ML20070E602
| ML20070E602 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/14/1982 |
| From: | Kemper J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8212170315 | |
| Download: ML20070E602 (12) | |
Text
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PHILADELPHIA ELECTR1C COMPANY 23O1 M ARKET STREET P.O. BOX 8699 1881 -1981 PHILADELPHI A. PA.19101 ys3n DEC 14198 Mr. John F. Stolz, Chief Docket Nos. 50-277 Operating Reactors Branch #4 50-278 Division of Licensing U.S. Nuclear Regulatory Commission W::hington, DC 20555
Subject:
Peach Bottom Atomic Power Station - Urits 263 Additional Information for Mark I Containment Program Plant Unique Analysis Report
Dear Mr. Stolz:
This letter is in response to your request for additional information on the Mark I Plant Unique Analysis Report (PUAR) submitted for Peach Bottom Units 2 and 3 on April 30, 1982. As stated in Section 1 of the PUAR, the results of the analyses of all torus attached piping systems will be contained in an Addendum to the PUAR. This addendum will also address, in detail, questions related to safety relief valve (SRV) discharge piping. The addendum is now being prepared and will be submitted when complete.
The questions and our responses are as follows:
Question 1:
Provide a stamary of the analysis with regard to all piping systems attached to the torus which are not included in Reference 2, indicating whether all criteria requirements of Reference I have been met.
Response
The criteria requirements given in Reference I for the analysis of torus attached piping systems have been met.
Summary of analyses performed will be provided in Reference 6.
i Question 2:
Provide a summary of the analysis with regard to the safety relief valve piping systems, indicating whether all the criteria requirements of Reference 1 have been met.
Response
The criteria requirements given in Reference 1 for the analysis of Safety Relief Valve (SRV) piping systems have
]
been met.
Summary of analyses performed will be provided
/
in Reference 6.
8212170315 821214 PDR ADOCK 05000277 P
Mr. John F. Stolz, Chief Page 2 Question 3:
Provide the basis of demonstrating the operability of active components which do not meet Service Limits A or B (or more conservative limits) in the plant unique analysis.
Response
Operability of active components other than valves will be addressed in Reference 6.
End loads from piping on the valves and valve operator acceleration are determined for seismic and Mark I loading combinations.
Operability of active valves will be addressed in Reference 6.
Question 4:
Indicate whether the containment vacuum breaker valves mounted on the vent internal to the torus or on piping associated with the torus are considered Class 2 components as required by the criteria of Reference 1.
Response
In Peach Bottom Atomic Power Station, the containment vacuum breaker valves are mounted on the vent line internal to the torus. The vacuum breaker is considered Class 2 as required by the criteria of Reference 1.
Question 5:
With reference to Table 1 of Appendix B (attached),
indicate whether all loads have been considered in the analysis and/or provide justification if any load has been neglected.
Response
All the loads in Table 1 of Appendix B have been considered except environmental temperature on SRV piping (Item 6.7).
The justification for not including this load is as follows:
The SRV piping in the wetwell was analyzed for maximum temperature of 400*F.
This is much higher than any environmental temperature that is expected in the wetwell.
Therefore, these loads were not considered in the analysis of SRV piping.
Question 6:
Indicate whether all linear types of component support meet the criteria requirements as specified in Section 4.4 of Reference 1.
Response
All safety related linear type component supports meet the criteria requirements as specified in Section 4.4 of the Plant Unique Analysis Application Guide (PUAAG)
(Reference 1).
Supports for non-safety related structures such as the catwalk platform meet the original criteria of construction, the AISC code.
Mr. John F. Stolz, Chief Page 3 Question 7:
Provide justification for determining the load combinations indicated in Tables 5-1 to 5-5 of Reference 2 to be the governing load combinations.
Response
A brief description for determining the load combinations indicated in Tables 5-1 through 5-5 is given in Section 5.4 of the PBAPS PUAR (Reference 2).
Additional justification is as follows:
1.
There are no condensation oscillation (CO) loads during SBA.
CO loads during IBA are included in chugging loads (Reference 3).
CO loads are, therefore, only during DBA.
This eliminates all the C0 loads during SBA and IBA, from Table 5-1 of the PUAAG (Reference 1).
2.
Specific Justification for Tables 5-1, 5-2, and 5-4 There are basically three sets of load combinations in Table 5-1 of the PUAAG (Reference 1).
- 1) loao combinations with DBA ii) load combinations with IBA/SBA iii) load combinations without LOCA.
Service Levels with and without IBA/SBA are the same because the Service Level A is the same as Service Level B for MC components and supports. This eliminates all the load combinations without LOCA.
3.
Specific Justification for Table 5-3 The load combinations are the same as given in Tables 5-1, 5-2, and 5-4 except one additional load combination was introduced due to Note 4 of Table 5-1 of PUAAG (Reference 1).
Question 8:
Indicate the number of fatigue cycles assumed for various components in order to determine the relevant fatigue usage factors and justify the assumptions. Also, indicate whether the fatigue usage factors for the SRV piping and the torus attached piping are sufficiently small so that a plant unique fatigue analysis is not warranted for piping. The hRC is expected to review the conclusions of a generic precentation (Reference 4) and determine whether it is sufficient for each plant unique analysis to establish that the expected usage factors for piping are small enough and do not warrant a plant unique fatigue analysis of the piping.
Mr. Jshn F. Stolz, Chief Page 4 The number of fatigue cycles during the plant life:
Response
1.
Fatigue cycles / events for Torus and Vent System (M-C Components)
A.
The minimum number of SRV discharges under normal operating conditions (NOC) assumed for fatigue was 800. Based on up-to-date information on Peach Bottom SRV actuations and similar information from other plants collected by General Electric Company, this is a conservative number. Also, since start-up, the set point-pressures on relief valves were' revised to increase the simmer margin. This has resulted in reducing the number of spurious actuations.
B.
There could be 5-0BE events. The stresses due to OBE are below the endurance limit; therefore fatigue usage factor is zero.
C.
There could be only one LOCA, any one of SBA, IBA or DBA, and during LOCA, there could be one OBE event. During SBA or IBA, there could be 5-SRV discharge events. All these SRV discharge events are assumed to be multiple valve.
Peak stresses of each cycle of LOCA, SRV discharge and OBE vere conservatively assumed to be exactly at the same time and were added by absolute sum method. It was also conservatively assumed that the maximum fatigue usage factor for SRV discharge (NOC) is for the same element as for LOCA.
2.
Fatigue cycles for SRV piping and the Torus Attached Piping For earthquake loads, 5 OBE events and 1 SSE event are used (10 stress cycles for each earthquake event).
For LOCA and SRV discharge loads, the equivalent number of cycles were determined by using the equivalent maximum stress cycle factors (R factors, Reference 4).
This procedure is similar to the approach used for Mark II plants and uses an exponent consistent with that specified for ASME Class 2 piping, cycle load analysis (para. NC-3611.2 of the code).
The stress cycles and usage factors are given as follows:
A.
Torus Attached Piping a)
Maximum usage factor for SRV (NOC) + DBA is 0.013 <<
l.0, for SRV (NOC) + IBA/SBA is 0.022 <<
l.0.
Mr. John F. Stolz, Chief Page 5 b)
Stress Cycles SRV(NOC) + DBA LOAD COMBINATION EFFECTIVE CYCLES 1-C0+0BE+MC+PR 1.000 2-C0+0BE 9.000 3-C0 80.000 4-CHUG +CSE 10.000 5-CHUG 22.000 6-PRECHUG 101.000 7-SRV(NOC)+MC+0BE+PR 40.000 8-SRV(N0C)+0BE+MC+PR 960.000 9-SRV(NOC) 4560.000 SRV(NOC) + IBA/SBA LOAD COMBINATION EFFECTIVE CYCLES 1-CHUG +SRV(IBA)+SSE+MC(IBA)+PR(IBA) 1.000 2-CHUG +SRV(IBA)+SSE 9.000 3-CHUG +SRV(IBA) 191.000 4-CHUG 321.000 5-C0(IBA) 3733.000 6-SRV(NOC)+MC+0BE+PR 50.000 7-SRV(NOC)+MC+PR 950.000 8-SRV(NOC) 4550.000 B.
SRV Discharge Line a)
Maximum usage factor for SRV(NOC) + DBAs0.046 (( 1.0 and for SRV(NOC) + IBA/SBAs0.102 ((
1.0.
b)
Stress cycles SRV(NOC) + DBA LOAD COMBINATION EFFECTIVE CYCLES 1-C0+0BE+MC+PR 1.000 2-C0+0BE 9.000 3-C0 80.000 4-CHUG +SSE 10.000 5-CHUG 22.100 6-PRECHUG 101.650 7-SRV(NOC)+MC+0BE+PR 40.000 8-SRV(NOC)+0BE+MC+PR 960.000 9-SRV(NOC) 3010.000 10-SRVTQF 4050.000
I Mr. John F. Stolz, Chief Page 6 SRV(NOC) + IBA/SBA LOAD COMBINATION EFFECTIVE CYCLES 1-CHUG +SRV(IBA)+SSE+MC(IBA)+PR(IBA) 1.000 2-CHUC+SRV(18A)+SSE 9.000 3-CHUG +SRV(IBA) 135.000 4-CHUG 321.000 5-C0(IBA) 3733.500 6-SRV(NOC)+MC+0BE+PR 50.000 7-SRV(NOC)+MC+PR 950.000 8-SRV(NOC) 3000.000 9-SRVRTQF 4050.000 The details of these calculations will be provided in Reference 6.
NOMENCLATURE Normal Operating Conditions NOC Dasign Break Accident DBA Intermediate Break Accident IBA Small Break Accident SBA Operating Basis Earthquake OBE Safe Shutdown Earthquake SSE Condensation Oscillations CO Chugging CHUG Thermal Loads MC Pressure Loads PR Safety Relief Valve Discharge SRV Quencher Discharge Drag Loads TQF Question 9:
With regard to damping values assumed for dynamic analysis of pipes less than 12 inches in diameter, indicate whether the assumed values are in accordance with the NRC Regulatory Guide 1.61 (Reference 5).
Response
The damping values assumed for the dynamic analysis of pipes less than 12 inches in diameter are in accordance with the NRC Regulatory Guide 1.61 (Reference 5).
Question 10:
Provide and justify the reasons for not considering a 180' segment if the torus in order to determine the effects of seismic and other nonsymmetric loads as required by the criteria of Reference 1.
Response
A 90' segment of the torus was used for the seismic analysis. By use of appropriate (symmetric and anti-symmetric) boundary conditions, this model is adequate for seismic analysis.
For other nonsymmetric loads, which are likely to give rise to lateral loads (i.e.,
loads on seismic ties), hand calculations were made to account for these loads. The stresses due to these loads were very small and were considered in the evaluation of seismic ties.
Mr. J:hn F. Stolz, Chicf Page 7 Question 11:
Indicate whether the torus nozzles have been analyzed for the effect of reactions produced by the attached piping using either Bijlaard formulas or finite element analysis, as required by the criteria of Reference 1.
Response
The torus nozzles have been analyzed for the ef fect of reactions produced by the attached piping using Bijlaard formulae, as required by the criteria of Reference 1.
Refer to Figure 6.4 of Reference 2 for photograph of modification.
Question 12:
Provide and justify the reasons for not considering a 180* segrent of the vent system in order to determine the ef fects of seismic and other nonsymmetric loads as required by the criteria of Reference 1.
Response
During the Mark I Short-Term Program and the early part of the Long-Term Program, several generic studies were performed to evaluate the response of the Peach Bottom vent system to asymmetric loads. The calculated stresses were very low.
Therefore, new detailed analyses of the 180' model for asymmetric loads were not performed for the Long-Term Program.
Question 13:
Indicate whether the vent penetration in the drywell has been analyzed and the calculated stresses are within the allowables specified in the criteria of Reference 1.
Response
Analysis of the vent system showed that the stresses in the vicinity of vent penetration in the drywell were low.
(See Tables 7-2 and 7-3 of Reference 2.)
Even af ter considering stress intensification factors, these stresses were well within the allowable limits. Therefore, no separate analysis for this penetration was performed.
Question 14:
Provide more details on the calculated stresses and fatigue evaluation of the intersection region boceeen the vent header and the downcomer identified on page 7-18 of the PUA report of Reference 2 and confirm whether the calculated stress indicated in Table 7-2 of Reference 2 corresponds to the maximum value in this region.
Response
1.
Details of Calculated Stresses in Table 7-2 A.
In the Peach Bottom vent system, the ventheader/
downcomer intersections are the most highly stressed regions. Also, the stresses in the venthender at these regions are higher than the stresses in the adjacent downcomers.
This is true because the downcomer is thicker than the ventheader.
Mr. J:hn F. Stolz, Chief Page 8 Re ponse:
A.(Continued) 1 The computer output for each load case is in the 1
form of hoop, longitudinal and shear stresses.
i Membrane and membrane plus bending stress intensities were calculated using special purpose computer program per the method outlined on page 6-23 of the PUAR (Reference 2).
At each of the intersections, twenty critical elements were chosen for further evaluations. Sixteen elements out of these twenty belonged to the ventheader region. Stress intensities reported in Table 7-2 correspond to the maximum in each category of stress intensity. Note that in calculating the stress intensities, the following conservatisms are included:
a)
Instead of adding the stresses due to various loads first and then calculating the stress intensities, the stress intensities due to each load were calculated first and then added.
b)
Stress intensities due to all the loads were added by the absolute sum method.
2.
Additional Details for Fatigue Evaluation A.
Fatigue evaluation for chugging loads was carried
)
out in accordance with LDR methodology (Reference 3). A brief description is as follows:
a)
A finite element analysis of the vent system was performed with unit static loads at the l
bottom of the downcomers. One unit load was l
applied in a direction parallel to the ventheader i
and the other was perpendicular to the first
(
one. The stresses and stress intensities were calculated.
b)
A finite element vibration analysis of the vent i
system was performed to calculate the downcomer frequency.
This frequency was used to calculate the fatigue chugging loads. Stress intensities calculated in (a) above were multiplied by the fatigue chugging loads, c)
The peak stress intensities for the fatigue l
evaluation were calculated by multiplying the stress intensities calculated above by the j
stress concentration factor, d)
Fatigue usage factor for stress histogram (Reference 3) was calculated from Table 1.9.1 of the ASME Code. The maximum fatigue usage factor for chugging during IBA/SBA is 0.039.
_ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _. _ ~. _ _ _ _ _ _. _ _,
Mr. John F. Stolz, Chiof P:g2 9 B.
A brief description of fatigue evaluation for CO and SRV discharge loads follows:
a)
A dynamic analysis was performed for CO loads using NASTRAN computer program. The computer output was in the form of hoop, longitudinal and shear stresses. Membrane plus bending stress intensities were calculated by the method given on page 6-23 of Reference 2.
The peak stress intensities were obtained by multiplying the above stress intensitias by stress concentration factor.
The maximum stress intensity for C0 loads is at 8 Hz frequency.
It was conservatively assumed that CO load would remain at 8 Hz for the entire duration of the C0 event. The maximum fatigue usage factor for DBA CO load, calculated according to Table 1.9.1 of ASME Code is 0.113.
b)
A dynamic time history analysis was performed for SRV diocharge loads using NASTRAN computer program. Range of stress difference and alternative stress intensities (S,) were calculated from stresses based on the method given in Subsection NE of the ASME Code.
Peak stress intensities were obtained by multiplying these alternative stress intensities by a stress concentration factor.
It was conservatively assumed that the maximum SRV discharge force is applied to the same downcomer during all the SRV discharge events throughout the life of the plant.
Fatigue usage factor is calculated based on Table 1.9.1 of the ASME Code. The maximum fatigue usage factor for SRV discharge events is 0.032.
3.
Stress intensities reported in Table 7-2 (Reference
- 2) correspond to the maximum in each category of stress intensity.
Question 15:
Indicate the calculated maximum upward and downward loads at a typical torus support and also show that the actual stress in the anchor bolt is less than the allowable.
Response
Calculated maximum upward and downward loads at a typical torus support are as below:
(ABSOLUTE SUM)
SERVICE LEVEL A
C Upload (KIPS) 248 324 Download (KIPS) 908 984
-A
,m,-
+-
A nm.
m.Gw u n i
mL--m----
- ~
m 4
-,-1
-A
-.---+,4:--
-w A
-+++--2 i
Mr. J:hn F. Stolz, Chief Page 10 4
The stresses in the anchor bolts are 50% of the allowable l
for Service Level A and 49% of the allowable for Service i
Level C.
Refer to Figure 6.2 of Reference 2 for a photograph of the modification.
Should you require any further information on this subject, please do not hesitate to contact us.
4 Very truly yours, J J.~ b -
i; l
DM0/dec 13/6 i
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REFERENCES:
1.
NEDO-24583-1 " Mark I Containment Program Structural Acceptance
]
Criteria - Plant Unique Analysis Application Guide", General Electric Company. San Jose, CA, October, 1979.
2.
Peach Bottom Atomic Power Station Units 2 and 3 " Plant Unique Analysis Report. Mark I Containment Program", Philadelphia Electric Company, Revision 0, April, 1982.
3.
NEDO-21888 Revision 2, " Mark I Containment Program Load Definition Report", Cencral Electric Company, San Jose, CA, Novembe r, 1981.
4.
P. M. Kasik, " Mark I Piping Fatigue".
Presentation at the NRC meeting, Bethesda, MD, September 10, 1982.
5.
US NRC Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants", October, 1973.
6.
Addendum to Peach Bottom Atomic Power Station Units 2 and 3
" Plant Unique Analysis Report, Mark I Containment Program", under preparation.
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- 1. Containment Pressure and Temperaturo X
X X
X X
X X
X
'X
- 2. Vont Systom Thrust Loads X
X X
- 3. Pool Swoll 3.1 Torus Not Vertical Loads X
X 3.2 Torus Shell Prossuro Histories X
X 3.3 Vont System impact and Drag X
X X
3.4 Impact and Drag on Other Structures X
X X
l 3.5 Frothlmoingomont X
X X
X X
3.6 Pool Fallback X
X X
X X
3.7 LOCA Jot 3.8 LOCA Bubble Drag X
X X
- 4. Condonsation Oscillation 4.1 Torus Sholl Loads X
X y
4.2 Load on Submerged Structures X
X X
l 4.3 Lateral Loads on Downcomers X
X 4.4 Vont System Loa'Js X
X i
- 5. Chugging O
5.1 Torus Shell Loads X
X l
5.2 Loads on Submerged Structuros X
X X
t 5.3 Latoral Loads on Downcomers
,, X X
1 5.4 Vont Systom Loads X
X f
- 6. T-Quenchor Loads 6.1 Olschargo Lino Cloaring X
j 6.2 Torus Shell Prossures X
X 6.4 Jet Loads on Submerged Structures X
X X
X 6.5 Air Bubblo Drag X
X X
X l
6.6 Thrust Loads on T-Quenchor Arms X
6.7 S/RVOL EnvironmentalTomporaturo X
j
- 7. Ramshead Loads 7.1 Discharge Line Clearing fx]
7.2 Torus Shell Pressuros.
l'Xl @
Z 7.4 Jot Loads on Submorged Structures X1 X'
X' X'
X
,X, X,
7.5 Air Bubblo Drag X
i
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7.6 S/RVOL Environmental Temperature 8
5 Loads required by NUREG-0661 and included in PUA report.
X Not applicable.
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