ML20086P313
| ML20086P313 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 11/30/1983 |
| From: | Edwards N, Steinert L NUTECH ENGINEERS, INC. |
| To: | |
| Shared Package | |
| ML20086P296 | List: |
| References | |
| DET-04-028-1, DET-04-028-1-R01, DET-4-28-1, DET-4-28-1-R1, NUDOCS 8402270096 | |
| Download: ML20086P313 (10) | |
Text
DET-04-028-1 Revision 1 November 1983 ENRICO FERMI ATOMIC POWER PLANT UNIT 2 PLANT UNIOUE ANALYSIS REPORT VOLUME 1 GENERAL CRITERIA AND LOADS METHODOLOGY Prepared for:
Detroit Edison Company O
Prepared by:
NUTECH Engineers, Inc.
4 Approved by:
Issued by:
l 0
Dr.
N.
W.
- Edwards, P.E.
L.
D.
Steinert President Project Manager NUTECH Engineers, Inc.
O wanws nutenb A
REVISION CONTROL SHEET N
(Continuation)
TITLE: ENRICO FERMI ATOMIC POWER REPORT NUMBER: DET-04-028-1 PLANT, UNIT 2 Revision 1 PLANT UNIQUE ANALYSIS REPORT VOLUME 1 PRE-ACCURACY CRITERIA PRE-ACCURACY CRITERIA E
REV REV PARED CHECK CHECK PARED CHECK CHECK PAGE(S)
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V ABSTRACT The primary containment for the Enrico Fermi Atomic Power Plant, Unit 2, was designed, erected, pressure-tested, and ASME Code N-stamped during the early 1970's for the Detroit Edison Company by the Chicago Bridge and Iron Company.
Since that time new requirements, defined in the Nuclear Regulatory Commission's Safety Evaluation Report NUREG-0661, which affect the design and operation of the primary containment system have evolved.
The requirements to be addressed include an assessment of additional containment design loads postulated to occur during a
loss-of-coolant accident or a safety relief valve discharge event, as well as an assessment of the effects that these postu-lated events have on the operational characteristics of the containment system.
This plant unique analysis report documents the efforts under-I taken to address and resolve each of the applicable NUREG-0661 V
requirements, and demonstrates, in accordance with NUREG-0661 acceptance criteria, that the design of the primary containment system is adequate and that original design safety margins have been restored.
The report is composed of five volumes which are:
o Volume 1
- GENERAL CRITERIA AND LOADS METHODOLOGY SUPPRESSION CHAMBER ANALYSIS o Volume 2 o Volume 3 VENT SYSTEM ANALYSIS o Volume 4 INTERNAL STRUCTURES ANALYSIS SAFETY RELIEF VALVE PIPING ANALYSIS o Volume 5 This particular volume, Volume 1,
provides introductory and background information regarding the re-evaluation of the suppression chamber design.
This includes a description of the Fermi 2 pressure suppression containment system, a description of DET-04-028-1 Revision 0 1-iv nutggb
the structural and mechanical acceptance
- criteria, and the hydrodynamic loads development methodology
'ised in the analysis.
This document has been prepared by NUTECH Engineers, Incorporated (NUTECH), acting as an agent responsible to the Detroit Edison Company.
The volume number precedes each number assigned to
- pages, sections, subsections, tables, and figures within a given volume.
An Appendix A has been added (as part of Volume 1) to provide responses to NRC requests for additional information.
Appendix A is in a question-and-answer format and addresses topics in each of the five volumes of the report.
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s*e TABLE OF CONTENTS (Concluded)
Page 1-4.2 Safety Relief Valve Discharge Loads 1-4.89 1-4.2.1 SEV Actuation Cases 1-4.94 1-4.2.2 SRV Discharge Line Clearing 1-4.100 Loads 1-4.2.3 SRV Loads on the Torus Shell 1-4.105 1-4.2.4 SRV Loads on Submerged 1-4.111 Structures 1-4.3 Event Sequence 1-4.116 1-4.3.1 Design Basis Accident 1-4.119 l-4.3.2 Intermediate Break Accident 1-4.125 1-4.3.3 Small Break Accident 1-4.127 1-5.0 SUPPRESSION POOL TEMPERATURE MONITORING SYSTEM 1-5.1 1-5.1 Suppression Pool Temperature Response to 1-5.2 SRV Transients 1-5.2 Suppression Pool Temperature Monitoring 1-5.9
System Design
s s_,/
1-6.0 LIST OF REFERENCES 1-6.1 APPENDIX A - RESPONSES TO NRC QUESTIONS ON THE FERMI 2 A-0 PLANT UNIOUE ANALYSIS REPORT DET-04-028-1 Revision 1 1-viii nutggb
LIST OF ACRONYMS ADS Automatic Depressurization System ASME American Society of Mechanical Engineers BWR Boiling Water Reactor CDF Cumulative Distribution Function CO Condensation Oscillaticia DBA Design Basis Accident DC/VH Downcomer/ Vent Header FSAR Final Safety Analysis Report FSI Fluid-Structure Interaction FSTF Full-Scale Test Facility HPCI High Pressure Coolant Injection IBA Intermediate Break Accident I&C Instrumentation & Control LDR Load Definition Report (Mark I Containment Program)
LOCA Loss-of-Coolant Accident LTP Long-Term Program MCF Modal Correction Factors NEP Non-Exceedance Probability NOC Normal Operating Conditions NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System PUAAG Plant Unique Analysis Application Guide PUA Plant Unique Analysis PUAR Plant Unique Analysis Report PULD Plant Unique Load Definition 9
DET-04-028-1 Revision 0 1-ix nutp_gf_1
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modifications before the final loads and load combinations were determined by the Mark I Owners Group.
In Amendment 12 to the Final Safety Analysis Report (FSAR), Article 3.8.2, Detroit Edison Company sub-mitted an interim LTP plant unique analysis (PUA).
Reference 17 to Article 3.8 of the FSAR described the program which was implemented by Detroit Edison Company to provide an early assessment of the Fermi 2 containment design for the original design loads and the newly defined suppression pool hydrodynamic loads.
The loads employed in the interim PUA were
)
established using available generic documents, with s
the objective of developing realistic design loads which would allow early plant modifications with a high probability of bounding the final loads.
Results of the interim PGA indicated that extensive l
modifications would be required to the suppression
- chamber, vent
- system, and suppression chamber internal piping and structures to re-establish the original design margins.
The nature and extent of l
[.
the modifications were discussed in the interim PUA l-I report (Reference 4).
Detroit Edison Company l
s l Q DET-04-028-1 1-1.3 I
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(
proceeded at risk to install extensive modifications in anticipation that they would be required to meet the LTP acceptance criteria.
The installation of selected m-difications was delayed until some of the specific Owners Group concerns about the NRC acceptance criteria were resolved.
These selected designs were re-evaluated in light of the resulting NRC criteria, and in some cases, the proposed modifications were redesigned.
The Fermi 2
containment modification status is provided in Table 1-1.0-1.
The configuration and geometry of the torus is discussed in Section 1-2.1.1.
This plant unique analysis report (PUAR) describes the final LTP PUA for the Fermi 2 containment.
The report documents the evaluation of the modified Fermi 2 suppression chamber and internals which was performed in accordance with the requirements of NUREG-0661.
The alternate criteria allowed by NUREG-0661, Appendix A, Article 2.13.9 was used in the evaluation of safety relief valve discharge loads.
As such, a series of in-plant tests will be performed after fuel load to confirm that the com-DET-04-028-1 1-1.4 Revision 1 nutp_qh
1 puted loadings and predicted structural responses for SRV discharges are conservative.
Reference 17 describes the planned SRV in-plant tests for Fermi 2.
The predicted response of the suppression chamber shell provided by this PUAR for each of the loads and load combinations is an essential input for evalu-ating the piping attached to the torus.
The evaluation of the Fermi 2 torus attached piping is described in Reference 18.
l l
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O Accordingly, with the submittal of this PUAR, Detroit Edison Company believes that the Formi 2 containment modification program has addressed the requirements of NUREG-0661 and the Fermi 2 Safety Evaluation Report (NUREG-0798 and NUREG-0798, Supplement No. 1).
l l
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i i
i 1-2.2 Operating Parameters
,j Plant operating parameters are used to determine f
many of the hydrodynamic loads utilized in the re-evaluation of the Fermi 2 suppression chamber design.
Table 1-2.2-1 is a summary of the operating parameters used to determine the Fermi 2 hydrodynamic l
loads.
t f
l.
DET-04-028-1 1-2.29 Revision 0 nutggh E
Table 1-2.2-1 SUPPRESSION CHAMBER OPERATING PARAMETERS CCMPONENTS CONDITION / ITEM VALUE FREE AIR VOLUME (1) 163,730 cu ft $
NORMAL OPERATING PRESSURE HIGH 1.8 psig LOW -0.5 psig NORMAL OPERATING TEMPERATURE NOMINAL BULK 1350F MAX BULK 150cF MIN BULK 105cF DRYWELL NORMAL OPERATING RELATIVE HIGH 90%
HUMIDITY RANGE LOW 0%
PRESSURE SCRAM INITIATION SETPOINT 2 psig i 0.2 psig DESIGN INTERNAL PRESSURE 56 psig DESIGN EXTERNAL PRESSURE MINUS 2 psid INTERNAL PRESSURE DESIGN TEMPERATURE 3400F POOL VOLUME MAX (HIGH WATER 3
LEVEL) 124,220 ft MIN (LOW WATER 3
LEVEL) 121,080 ft I2)
MIN (HIGH WATER FREE AIR VOLUME LEVEL) 127,760 f'.3 MAX (LOW WATER SUPPRESSION LEVEL) 130,900 ft LOCA VENT SYSTEM DOWNCOMER MIN (LOW WATER SUBMERGENCE (DISTANCE OF DOWNCOMER LEVEL) 3.00 ft DISCHARGE PLANE BELOW WATER LEVEL)
MAX (HIGH WATER LEVEL) 3.33 ft WATER LEVEL DISTANCE TO TORUS MAX (LOW WATER CENTERLINE LEVEL) 0.9166 ft MIN (HIGH WATER LEVEL) 0.5833 ft SUPPRESSION POOL SURFACE EXPOSED TO 10'788
'**'2 SUPPRESSION CHAMBER AIRSPACE NORMAL OPERATING PRESSURE RANGE HIGH 1.8 psig LOW -0. 5 psig O
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Table 1-4.1-3 DBA CONDENSATION OSCILLATION TORUS SHELL PRESSURE AMPLITUDES (Concluded)
A M RESSURE AMPLITUDE (psO FREQUENCY INTERVALS ALTERNATE ALTERNATE ALTERNATE ALTERNATE (Hz) 1 2
3 4
l 25-26 0.25 0.25 0.25 0.50 26-27 0.58 0.58 0.58 0.51 27-28 0.13 0.13 0.13 0.39 28-29 0.19 0.19 0.19 0.27 29-30 0.14 0.14 0.14 0.09 30-31 0.08 0.03 0.08 0.08 31-32 0.03 0.03 0.03 0.07 32-33 0.03 0.03 0.03 0.05
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33-34 0.03 0.03 0.03 0.04 x._ J 34-35 0.05 0.05 0.05 0.04 35-36 0.08 0.08 0.08 0.07 36-37 0.10 0.10 0.10 0.11 37-3d 0.07 0.07 0.07 0.06 38-39 0.06 0.06 0.06 0.05 39-40 0.09 O.09 0.09 0.03 l
l 40-41 0.33 0.33 0.33 0.08 41-42 0.33 0.33 0.33 0.19 42-43 0.33 0.33 0.33 0.19 43-44 0.33 0.33 0.33 0.13 44-45 0.33 0.33 0.33 0.18 45-46 0.33 0.33 0.33 0.30 46-47 0.33 0.33 0.33 0.18 l
l 47-48 0.33 0.33 0.33 0.19 l
48-49 0.33 0.33 0.33 0.17 49-50 0.33 0.33 0.33 0.21 l
DET-04-028-1 Md i
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Table 1-4.1-4 i
FSTF RESPONSE TO CONDENSATION OSCILLATION MAXIMUM MEASURED CALCULATED FSTF RESPONSE RESPONSE QUANTITY FSTF RESPONSE AT 84% NEP(1)
M8 MllB M12 BOTTOM DEAD CENTER 3.0 2.3 1.6 2.7 AXIAL STRESS (ksi)
BOTTOM DEAD CENTER HOOP STRESS (ksi) 3.7 2.6 1.4 2.9 BOTTOM DEAD CENTER 0.17 0.11 0.08 0.14 DISPLACEMENT (in.)
INSIDE COLUMN 184 93 68 109 FORCE (kips)
OUTSIDE COLUMN 208 110 81 141 FORCE (kips)
NOTE:
(1) USING CO LOAD ALTERNATES 1, 2 AND 3.
DET-04-028-1 Revision 1 1-4.48 nutgch
this manner for Fermi 2 results in a probability that the force will be exceeded once per LOCA as a
function of the number of downcomers chugging.
The resulting exceedance probabilities for various cases of multiple downcomers chugging are presented in Table 1-4.1-14.
For fatigue evaluation of the downcomers, the required stress reversals at the downcomer/ vent header junction are obtained from the FSTF, RSEL reversal histograms.
The plant unique junction stress reversals are obtained by scaling the FSTF, RSEL reversals by the ratio of the chugging duration
(
specified for Fermi 2 to that of the FSTF.
Chugging
(
durations for the DBA, IBA, and SBA are specified in Table 1-4.1-12.
F l
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Table 1-4.1-14 PROBABILITY OF EXCEEDANCE FOR MULTIPLE DOWNCOMERS CHUGGING NUMBER OF PROBABILITY OF DOWNCOMERS EXCEEDANCE 2.91 x 10-3 5
10 1.45 x 10-3 20 7.27 x 10-4 40 3.64 x 10-4 80 1.82 x 10-4 NOTE:
1.
SZE THE RESPONSE TO NRC QUESTION 4 IN APPENDIX A FOR ADDITIONAL INFOlu1ATION ON THE MULTIPLE DOWNCCMER CHUGGING LATERAL LOAD.
O DET-04-028-1 Revision 1 1-4.84 nuta_c__h_
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1-4.2 Safety Relief Valve Discharge Loads This section discusses the procedures used to determine loads created when one or more SRV's is actuated.
When an SRV actuates, pressure and thrust loads are exerted on the SRVDL piping and the T quencher discharge device.
In addition, the expulsion of water followed by air into the suppression pool through the T-quencher results in pressure loads on the submerged portion of the torus shell and drag loads on submerged structures.
The T quencher utilized in the Fermi 2 plant is a plant unique version of the Mark I
T-quencher described in the LDR.
The Fermi 2 T-quencher has 20" diameter arms which are mitered at the connection to the ramshead portion.
This is accomplished to provide symmetrical torus shell loads upon SRV actuation, since the T-quenchers are installed on the torus ring girders at the miter joints.
Figure 1-4.2-1 illustrates this mitering concept and connection to the SRVDL.
t Q
DET-04-028-1 1-4.89 Revision 0 nutagh
To minimize torus shell pressure loads and to ensure adequate steam condensation performance, the Fermi 2 T-quencher utilizes the same hole diameter and minimum spacing as the Mark I T-quencher.
The hole pattern distribution along the arms was modified to accommodate Fermi 2 unique supports without sacrifice to the extended water clearing duration concept of the quencher.
The details of the hole distribution along the arm are illustrated in Figures 1-4.2-2 and 1-4.2-3.
Analytical predictions of torus shell pressures for Fermi 2
T-quencher discharges indicate improved performance over the standard Mark I T-quencher.
The torus shell loads are predicted utilizing the Fermi 2 T-quencher geometry and the hydrodynamic modeling techniques and analytical models used in the development of the Mark I T-quencher as contained in the Mark I LDR.
As allowed in Section 2.13.9 of Appendix A
of NUREG-0661, plant unique SRV testing at Fermi 2 will be performed to confirm that the computed loadings and predicte6 structural responses for SRV discharges are conservative.
Reference 17 describes the planned SRV in-plant tests for Fermi 2.
DET-04-028-1 1-4.90 Revision 1 nutggb
=,
a-l l-6.0.
LIST OF RF.FERENCES 1.
" Mark I Containment Program Load Definition Report," General Electric _ Company, NEDO-21888, Devision 2, December 1981.
2.
" Mark I' Containment Program Structural Accep-tance Criteria Plant-Unique Analysis Applica-tions Guide,'" Task Number 3.1.3, Mark I Owners Group,. General Electric Company, NEDO-24583,
- Revision 1, July 1979.
3.
" Mark I Containment Long-Term Program," Safety Evaluation
- Report, USNRC, NUREG-0661, July 1980.
4.
"Enrico Fermi 2 Atomic Power Plant, Unit 2,
Interim Plant Unique Analysis Report," NUTECH, DET-01-074,'May 1978.
i.
5.
"The General Electric Pressure Suppression Containment Analytical Model," General Electric Company, NEDO-10320, April, 1971; Supplement 1, May, 1971'; Supplement 2, Jaruary 1973.
6.
" Mark I Containment Program - Plant Unique Load O
Definition," Enrico Fermi Atomic Power Plant:
Unit 2,' General Electric Company, NEDO-24568, Revision 1, June 1981.
~~
7.
" Mark I Containment Program Quarter-Scale Plant Unique Tests, Task Number 5.5.3, Series 2,"
l General Electric Company, NEDE-21944-P, Volumes 1-4, April 1979.
l 8.
- Patton, K.T.,
" Tables of Hydrodynamic" Mass Factors for Translational
- Motion, ASME
- Manuscript, Chicago, November 7-11, 1965.
9.
-Miller, R.R.,
"The Effects of Frequency and Amplitude of Oscillation on the Hydrodynamic Masses of Irregularly-Shaped Bodies, MS Thesis, Univerity of Rhode
- Island, Kingston, R.I.,
1965.
10.
Fitzsimmons, G.
W.
et al.,
" Mark I Containment Program Full-Scale Test Program Final Report, Task Number 5.11,"
General Electric Company, NEDE-24539-P, April 1979.
f-DET-04-028-1 6-1 l
Revision 0 N
f l
l
_..., _.. _.. _ _ _. - _ _ _ _ _. _, _ _ _ _._ _, _.-__ _ _ _.. ~ _. _..
?1.
" Mark I
Containment Program Letter Reports MI-LR-81-01 and MI-LR-81-01-P, Supplemental Full-Scale Condensation Test Results and Load Confirmation-Proprietary and Nonproprietary Information," General Electric Company, May 6, 1981.
Full-Scale Test 12.
" Mark I containment Program Evaluation of Supplemental Tests,"
Program General Electric Company, NEDO-24539, Supple-ment 1, July 1981.
13.
- Hsiao, W.
T.
and Valandani, P.,
" Mark I
Containment Program Analytical Model for Computing Air Bubble and Boundary Pressures Resulting from an SRV Discharge Through a
T-Quencher Device," General Electric Company, NEDE-21878-P, August 1979.
14.
Letter from T. A.
Ippolito (NRC) to J.
F. Quirk (GE) dated October 16, 1981.
15.
" Suppression Pool Temperature Limits for BWR Containment," USNRC, NUREG-0783,
- Draft, July 1981.
16.
" Operating Problems with Target Rock Safety-Relief Valves at BWR's,"
- USNRC, Office of Inspection and Enforcement, IE Bulletin No. 80-25, December 19, 1980.
l 17.
Letter EF2-59,029 from Harry Tauber (Detroit Edison) to B.
J. Youngblood (NRC), " Submittal of SRV In-Plant Test Plan," dated August 18,
)
1982.
1 18.
"Enrico Fermi Atomic Power Flant Unit 2 Plant Unique Analysis Report for Torus Attached l
Piping and Suppression Chamber Penetrations,"
l NUTECH, DET-19-076-6, Revision 0, June 1983.
l DET-04-028-1 1-6.2 l
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