ML20236N880
| ML20236N880 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 07/10/1998 |
| From: | Rainsberry J SOUTHERN CALIFORNIA EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, GL-96-6, NUDOCS 9807160037 | |
| Download: ML20236N880 (60) | |
Text
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+j SOUHt1RN CAUTORNIA EDISON
!d;"s"Cu I
1)lSON IN11.RNATIOMI.* Company July 10,1998 V. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C.
20555 Gentlemen:
Subject:
Docket Nos. 50-361 and 50-362 Generic Letter 96-06: " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions,"
Additional Information San Onofre Nuclear Generating Station Units 2 and 3
References:
1)
Letter from J. L. Rainsberry (SCE) to the Document Control Desk (NRC), dated February 3,1997,
Subject:
120 Day Response to Generic Letter 96-06: " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," San Onofre Nuclear Generating Station Units 2 and 3 2)
Letter from James W. Clifford (NRC) to Harold B. Ray (SCE), dated May 6,1998,
Subject:
Request for additional information regarding response to Generic Letter 96-06:'" Assurance of Equipment Operability and
)
Containment Integrity During Design-Basis Accident 1
Conditlons," San Onofre Nuclear Generating Station 330j8 Units 2 and 3
-This letter provides additional information to supplement the Southern
\\
California Edison Company's (SCE's) response to Generic Letter (GL) 96-06 (Reference 1). This additional information was requested by a May 6,1998 yi
. letter from Mr. James W. Clifford (Reference 2).
A I
Additionally, this letter corrects information provided in the SCE response to GL96-06(Reference 1). On page 18 of Reference 1 SCE stated that, with the exception of'the Reactor Coclant System (RCS) drain lines from the cold legs to the RCS drain tank, the piping systems evaluated were not insulated. SCE has now reanalyzed two penetrations with insulation and performed calculations San Onofre Nudear Generating Station P. O. Box 128 l
San Clemente. CA 92674-0128
't 9807160037 980710 P
i (DR ADOCK 05000361 D P&
Document Control Desk which confirmed the acceptability of these two penetrations at the maximum pressure and. temperature (see Section 4.2.4, Note (3), pages 10 and 11 of 15 intheenclosure).
If you have any questions or would like additional information, please let me
- know.
Sincerely, D.M/.taxsare For J. L. Rainsberry Enclosure cc:
E. W. Merschoff, Regional Administrator, NRC Region IV K. E. Perkins, Jr., Director, Walnut Creek Field Office, NRC Region IV J. A. Sloan, NRC Senior Resident Inspector, San Onofre Units 2 & 3 J. W. Clifford, NRC Project Manager, San Onofre Units 2 and 3 1
i t
_J
Enclosure Generic Letter 96-06 Request for Additional Information San Onofre Units 2 and 3
' Question (1)
For the four pipe runs where the insulation was modified, describe the requirements for the qualification of the insulation.
In addition, provide the following information for these four pipe lines and the other nine pipe runs qualified by detailed analysis.
Provide the applicable design criteria for the piping and the valves.
Include the required load combinations.
Provide a drawing of the piping run between the isolation valves.
Include the lengths and thicknesses of the piping segments and the type and thickness of the insulation.
Provide the maximum-calculated temperature and pressure for the pipe run.
Describe, in detail, the method used to calculate these pressure and temperature values.
This should include a discussion of the heat transfer model used in the analysis and the basis for the heat transfer coefficients used in the analysis.
Response to Question (1) 1.
Properties of PiDe Fiberalass Insulation Insulation properties per ASTM C-547 95, " Standard Specification for Mineral Fiber Pipe Insulation," Type II, were specified.
The purpose of modifying, or adding, the insulation is to reduce the heat flux from the containment atmosphere to the water trapped inside the pipe following the accident.
The heat flux depends on the thickness of the insulation, the properties of the insulation material, and the length of the pipe segment with insulation.
Accordingly, the values of these parameters were determined based on meeting the l
applicable acceptance criteria explained below.
2.
Accentance Criteria A Design Basis Accident due to a Loss of Coolant Accident (LOCA) is classified as a faulted plant condition in the San Onofre 2/3 Updated Final Safety Analysis
' Report (UFSAR).
Following a postulated LOCA the piping evaluated in this calculation must be capable of maintaining structural integrity under the I
effects of pressure due to the thermal expansion of the trapped water, weight, I
and Design Basis Earthquake (DBE) loadings. Under this condition the ASME Code stress limit designated as a Level D Service limit was used as the acceptance criteria. The Level D service limits are intended to ensure that violation of the pressure retaining boundary will not occur.
I' b_---________-
Enclosure GL 96-06 Request for Additional Information
- The 1989 Edition of the ASME Code was used in this evaluation as allowed per Paragraph NA-1140 of the original code of construction (1974 Edition through 1974 Summer Addenda).which permits the use of specific provisions within later Code editions provided that all related requirements are met. The 1989 Code edition has more design criteria, categories, and definitions, which cannot be found in the original Code, available to address Level D limits, resulting in a
-more comprehensive design Code. A Code reconciliation of the differences between the original Code edition and the later 1989 Code edition was performed to demonstrate the fulfillment of the requirements for use of the later Code edition.
Based on the above, this evaluation was performed per the design criteria guidelines given in the 1989 Code edition, Subsection NC-3217 and Table NC-3217-1 (Class 2 vessel components).which states that when a complete-analysis is performed the stress limits of Appendix F may be used for the evaluation of loads with Level D service limits.
The stress limits per Appendix F, Article F-1331 are as follows:
The general primary membrane stress intensity,' P,, is the lesser of 2.4S, and 0.7S, where S, is the design stress intensity and S is the tensile u
u strength.
The primary membrane (general or local) plus primary bending stress is the lesser of 3.6S, (3.0S, was conservatively used intensity Pt+Pb in this evaluation) and 1.05S.
o Potential for unstable crack growth shall also be considered.
The finite element method was used to perform a thermal-structural evaluation of
)
the analyzed piping, between the two isolation valves, at the containment penetrations.
Finite element models were generated using the general purpose finite element program ANSYS.
Each penetration model included the analyzed piping, the penetration, the isolation valves, and part of the adjacent containment wall, including the steel lining. Sample lines were modeled as a
. pipe exposed to containment atmosphere. Stainless steel sample lines were conservatively modeled without insulation, and thermal insulation was inc'uded in the models for carbon steel sample lines (attached to penetrations No.17 f
and44).
'L---____-_________-________--
Enclosure GL 96-06 Request for Additional Information A thermal transient analysis was performed to calculate the temperature time history under-accident conditions. A Loss of Coolant Accident {LOCA) and Main Steam Line Break (MSLB) were considered, and results showed that a LOCA is more limiting due to the longer duration of elevated temperature.
Results of the thermal transient analysis were used to perform a structural analysis to calculate the trapped water pressure and the stresses' in the pipe wall as a-
. function of time.
The maximum stress values were used to perform an evaluation per the ASME Code.
2.1 General Membrane Stress Intensity (Py The general membrane stress intensity in the pipe wall due to internal pressure, P, is acceptable if i
P, s 2.4S, or 0.7S, (whichever is less) 2.2 Local + Bendina Stress Intensity (Pt + Pd ASME Code, Subsection NC-3652.2, Equation 9 was used as a guideline to calculate the combined pressure + deadweight +DBE, a
B P,,x D /2 t + B 3
o 2 M,/Z s S, I
where P,x
= maximum pressure. The maximum pressure in this calculation j
is taken as'the maximum pressure due to LOCA j
D
= pipe outside diameter o
t
= nominal wall thickness M
= resultant moment due to weight and sustained moment plus DBE i
Z
= section modulus B3 and B2
= stress indices The allowable stress, S., for Equation 9 was conservatively taken equal to j.
3.0 S,.
)
l
. 2.3
'Ev'aluation for Brittle Failure A fracture mechanics evaluation was performed to assess the potential for unstable crack growth in the' pipe segments subjected.to the overpressure i
conditions, as required by Appendix F.
The methodology for the brittle failure evaluation is based on calculating the stress intensity factor for a postulated j
' 1 l
_ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ = _ _ _ _ _ _ _
Enclosure GL 96-06 Request for Additional Information axial crack in the pipe wall. The crack depth is assumed to be one tenth of the pipe wall thickness.
Fracture mechanics correlations were obtained from Reference 1.
(a)
Thin Pipes (5s R /t s20) i A model representing an infinite pressurized circular cylinder with a postulated infinite part-through longitudinal crack. This model is conservative since an actual crack will have a finite length.
The stress intensity factor, K, is i
given by 2PR' R
2 2
g.g i
1 where
= 1.1 + A [4.951(a/t)o.+zs.092(a/t)']
2 F
1
= (0.125 R /t - 0.25 for 5 s R /t s10
= (0.2 R /t - 1).2s )or 10 s R /t A
i i
A f
s20 i
i a
= postulated crack depth t
= pipe wall thickness R,
= inside radius R,
= outside radius P
= internal pressure The above thin pipe fracture mechanics correlation was used for all piping I
except for the 3/4" sch 160 sample line. Attachment 1 provides a complete listing of the analyzed penetrations, including the size and material of the attached pipe.
(b)
Thick Pipes ( R /t s5) i The model described above is not applicable to'the 3/4" sch 160 pipe considered in this evaluation, since R /t does not fall within the range 5 s R /t s20. A i
i L
plate model with a single edge crack, of depth a, was used as an approximation.
The plate width is equal to the pipe wall thickness.
The stress intensity factor, K, is given by the following equation:
i a} f(")
K,
= I i
Enclosure GL 96-06 Request for Additional Information where o,,
= stress a
= crack depth t
pipe wall thickness 2 tan E f(S)
[0. 752 +2.02(E ) +0. 37 (1 -s i n E) 3]
t cos 3 2t Brittle fracture will not occur if the stress intensity factor, Kr, is less than the critical stress intensity factor, Kg, of the pipe material, i.e.,
K3sKIc The following values of KIc, representing lower-bound fracture toughness properties, were used based on Generic Letter 90-05 (Reference 2), dated June 15, 1990:
Kic = 35 ksi(in)% for carbon steel KIc=135ksi(in)4 for stainless steel The maximum K3 calculated for the carbon steel piping was 16.4 ksi(in)4, and 10.9 ksi(in)k for the stainless steel piping. Generic Letter 90-05 guidance is 1
tu appl a safety factor of 1.4 to the calculated K, which results in K3 = 23.0 ksl(in){ for carbon steel, and 15.3 ksi(in)%
i for stainless steel.
These values MKi are acceptable based on the values of KIc above.
?.4 Valves and Penetration Flued Heads Results of the finite element-based stress analysis showed that the maximum stresses occur in the piping attached to the penetration.
This result is expected since valve body walls are designed to have section properties larger than attached piping.
Furthermore, the limiting loads for the containment penetrations are the bending moments acting on the penetration flued head. The overpressure in the piping attached to the penetration has no impact on these 1
moments.
In addition, the penetration has a higher pressure capacity based on:,
a
Enclosure GL 96-06 Request for Additional Information (a)
The thickness of the penetration 2 the thickness of the pipe.
(b)
The flued head has a stiffening effect against internal pressure.
Therefore, the piping was considered limiting.
3.
Drawinas of the Analyzed Pinina Runs and Pinina Data The drawings of the analyzed pipe runs are included in Attachment 2.
4.1 Calculation of Maximum Pressure The time history of the temperature and pressure of the-trapped water were calculated using the finite element method.
Figure 1 depicts a schematic of a typical model showing the modeled components and the thermal boundary conditions. Condensing and convective heat transfer coefficients were calculated as described in Section 4.2 below.
Figure 2 shows a computer plot of a typical finite element model generated using the general purpose finite element analysis program ANSYS', Revision 5.0A.
The finite element analysis included:
(a)
A thermal transient analysis to calculate the temperature time history throughout the accident based on the temperature inside containment following a LOCA, and using the calculated film coefficients (see Section 4.2 below).
1 J
(b)
Using the results of the thermal transient analysis, a structural analysis was performed to calculate the pressure time history of the water trapped between the two isolation valves.
The maximum pressure was ider.tified, 4
and the stresses were calculated in the pipe.
l Piping insulation was modified at penetrations 42 and 43 (Component Cooling Water (CCW)~non-critical loop), and penetrations 45 and 46 (containment normal chilled water). Table 1 gives the summary of the results obtained for the l
piping at these two penetrations.
The finite element analysis results were verified by hand calculations using a model geometry representing a pipe filled with water.
Temperature was raised from 70 to 450*F, and the pressure was calculated.
The difference between the finite element-calculated value and the hand-calculated value was less than 0.25%. i J
Enclosure-GL 96-06 Request for Additional Information 4.2 Calculation of the Heat Transfer Coefficient Per Reference 3 (Topical Report BN-TOP-3, Revision 4, Bechtel Power Corporation, 1983), the boundary condition for surfaces exposed to the containment atmosphere during an accident is j
Q
=(h - h )(T, - T,)(TEST) + h,(T, - T, )
c y
where Q
= the heat flux j
h
= condensing heat transfer coefficient e
h,
= convective heat transfer coefficient T,
= saturation temperature (dew point)
T,
= component surface temperature T,
= containment vapor temperature TEST = 1.0 i f. T, > T, TEST = 0. 0 i f T, < T, Conservatively, the containment vapor temperature was used throughout the analysis (T, 2T,).
The methodology of calculating the heat transfer coefficients h and h, is described below.
c 4.2.1 Condensina Heat Transfer Coefficients The condensing heat transfer coefficients, h, were calculated as follows:
e Condensing heat transfer coefficients were based on the Tagami and Uchida correlations (References 3 and 5). -These coefficients were calculated by l:
the computer program COPATTA, which was used to perform the accident analysis.
During the initial blowdown period of a LOCA, heat transfer to
.the structural heat sinks from the containment atmosphere region is determined in the COPATTA program by a " modified Tagami" condensing heat transfer coefficient function i
h-
.=h (t/t*)
c m
.where t
= time following the LOCA up to the time of initial peak containment pressure during the Reactor Coolant System (RCS) i blowdown period (seconds) 1 r
l 7
Enclosure GL 96-06 Request for
{
Additional Information l
t*-
= time to initial peak containment pressure during the RCS l
= 72.5 [Q/(Vt*)] d{ seconds) blowdown period h,,,
Q
=-integrated blowdown energy release out through time t* (BTV) 3 V
= containment net free volume (f t )
l i
l For time greater than t*, the heat transfer coefficient is the greater of either a Uchida condensing heat transfer coefficient based on the mass ratio of non-condensible gas (air) to steam, or a value based on an exponential decay from the maximum Tagami value to the Uchida value given p
by the equation h
' = h,,x[e '8530 ~ ")].
c l
l When the magnitude of the heat transfer coefficient calculated by this procedure becomes less than the Uchida value, the condensing heat transfer coefficient used is based on the Uchida data,.and the Uchida correlation continues to be used for the balance of the LOCA containment response analysis.
During the blowdown phase of the accident, a value of 4xTagami value was
]
l used.
This value was conservatively used during the transition phase instead of the exponential decay allowed by Stan(ard Review Plan -
6.2.1.5, Reference 6.
l-l During the long-term post-blowdown phase, a valt.e of 1.2xUchida was used i
per Standard Review Plan - 6.2.1.5, Reference 6.
4.2.2 Forced Convection Heat Transfer Coefficients The_ reference for the forced convection film coefficients is Kreith, F.,
" Principles of Heat Transfer," Third Edition, Intext Educational Publishers,
)
1973 (Reference 4).
The Nusselt number for flow across a cylinder is C,Re
Nu
=
1 4
Enclosure GL 96-06 Request for Additional Information I
where-l Re (Reynolds number)
= V D, p/p U
= blow down velocity D,
= pipe outside diameter p
= density
= dynamic viscosity The values of the constants C and m depend on the range of.the values of Re.
3 3
The forced convection heat transfer coefficient, h,, is h,
= k Nu/0, where k is the thermal conductivity.
4.2.3 Liauid-to Pine Heat transfer Coefficient l
.PerSouthernCaliforniaEdison's(SCE's)ContainmentPressure/ Temperature (P/T)
Analysis for a Design Basis LOCA, the containment emergency sump pump, penetration, No. 54, is required-to operate up to the recirculation mode, which starts 2,284 seconds from the onset of a design basis LOCA.
The temperature.of g
the trapped water in piping at penetration No. 54 between the two butterfly l
valves will-increase due to the heat transferred from the hot sump water. Part of the line is submerged under the containment sump water during the design basis accident.
l The heat transfer coefficient between the sump water and the outside surface of the elbow is calculated using the correlation from Reference 4.
Nu
= 0.53(Gr Pr)2" I
and:
o h
= k Nu/D, where Nusselt number-Nu
=
Grashof number Gr
=
3 2
p'gp(T - T.)D /
.)
=
density I
p
=
u g
= gravity acceleration p
coefficient of volume expansivity
=
T
= pipe surface temperature j
T.
= temperature at a point sufficiently removed from the pipe i
D,
= pipe outside diameter
_g.
i
Enclosure GL 96-06 Request for Additional Information
= dynamic viscosity Pr
= Prandtl number h
= heat transfer coefficient k
= thermal conductivity The properties of water are calculated at the mean film temperature.
4.2.4 Calculated Maximum Pressure and Temperature Table 1 provides a summary of the results of the finite element analysis. The table gives the maximum pressure, P,, calculated for the analyzed penetrations.
The trapped water temperature was calculated internally by the program, and was used as input to the structural analysis step.
The temperature time history was not included in the output since the purpose of the analysis was to calculate the maximum pressure, which was used in the ASME Code evaluation.
Table 1 Maximum Pressure Results Component P,
Component P,
Description (ksi)
Description (ksi) 2" sch 160 lines to 6.8 6" sch 80 line at See Note penetrations No. 2 and penetrations No. 36 (1) 8.
and 37.
3" sch 10s line at 2.1 3/4" sch 160 sample and 26.1 j
penetration No. 11.
drain lines connected to See Note penetrations: 1,4, 7 and (2) 12.
10" sch 40 lines at 1.4 3/4" sch 160 sample and 12.2 penetrations No. 42 and drain lines connected to See Note 43, penetrations: 17 and 44.
(3) 24" sch 10s lines at 0.38 8" sch 40 lines at 3.2 penetrations No. 54 and penetrations No. 45 and l
55.
46.
2* sch 40 line at 5.3 penetration No. 6. - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Enclosure GL 96-06 Request for Additional Information Notes:
(1)-
The piping temperature is normally high. Results show that the temperature of the trapped water remains higher than the temperature inside containment.
Therefore, it was concluded that no over-pressurization will occur.
(2)
Stainless' steel sample lines were conservatively analyzed as un-insulated piping. Actual piping is insulated.
(3)
Carbon steel sample lines were analyzed as insulated piping (2.5" insulation thickness).
The normal containment atmosphere temperature at San Onofre is 120*F. The maximum calculated temperature during a LOCA is 295.4*F at 60 seconds from the-onset of the accident. The temperature decreases slowly to 120*F at'1,000,000 seconds..
Question (2)
In its submittal, the licensee indicated that spring loaded, diaphragm actuated isolation valves' associated with penetrations 13 and 26 provide inherent relief to prevent over pressure.
In addition, the licensee stated that a solenoid valve would relieve to prevent over pressure for one line inside containment associated with the Reactor Vessel Head Vent and Pressurizer Vent lines.
Provide the following information for these pipelines:
Describe the applicable design criteria for the piping and valves. Include the required load combinations.
Provide a drawing of the. valve.
Provide the pressure at which the valve was determined to lift off its seat or leak and describe the method used to estimate this pressure. Discuss any sources of uncertainty associated with the estimated lift off or leakage pressure.
l Provide the maximum-calculated stress in the piping run based on the l,
estimated lif t off or leakage ' pressure.
l>('
' Response to Question (2)
L 1.-
Piping stresses are verified to satisfy AS",E Code Level D criteria for pressure, weight, and DBE loads with the valves leaking to relieve pressure.
The faulted allowable stress used is 2.4S, where Sn is the o
allowable stress of the pipe material at a temperature of 300oF.
l L_
Enclosure GL 96-06 Request for Additional Information 2.
The calculated lift off pressure for both valves HV-5804 and HV-7513 is 729 psi. The lift off pressure acting on the port area is calculated equal to the actuator spring load plus the packing friction force. Where, per the valve vendor (Fisher), the actuator spring load is the product of the valve's Initial Low Bench Set Pressure and Actuator Diaphragm Effective Area, and the packing friction force is proportional to the size of the valve stem.
It was assumed that the packing friction force of actuator diaphragm valves is the same as for the motor operated valves with the same stem configuration.
Per the San Onofre Design Standard Generic Letter 89-10, Motor Operated Valve Program, the packing friction force is 1,000 lbs per inch of stem diameter.
Therefore, the packing frictionforceforthe3/4"valvestemofthesetwovalvesis750lbs.
For calculation purposes, a lift off pressure of 800 psi (instead of the calculated 729 psi) was conservatively used to account for any uncertainties in the calculation of the packing friction force.
Valve drawings are attached as Items 21 and 22 in Attachment 2.
3.
The maximum calculated Level D stress of 22,608 psi is within the allowable ctress of 39,840 psi at 300*F.
References 1.
Anderson, T.
L., " Fracture Mechanics Fundamentals and Applications," CRC Press, 1991.
2.
Generic letter 90-05, June 15,1990.
Subject:
Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 piping."
3.
Topical Report BN-TOP-3. Revision 4
" Performance and Sizing of Dry Pressure Containments," Bechtel Power Corporation, March 1983.
4.
- Kreith, F., " Principles of Heat Transfer," Third Edition, Intext Educational Publishers,1973.
5.
References for Taaami and Uchida Correlations:
(a)
Tagami, Takashi, " Interim Report on Safety. Assessment and Facilities Establishment (SAFE) Project,"' February 28, 1966, Hitachi Ltd, Tokyo, Japan, f
I I i
i Enclosure l
GL 96-06 Request for Additional Information l
References (Continued)-
L (b)
Tagami,- T., et aZ, " Studies for Safety Analysis of Loss-of-Coolant-l
- Accidents in Light Water-Power Reactors," NSJ-Tr-223, March 1968,
. Japan Atomic Energy Research Institute, Japan.
(c)
Uchida, H., et al, " Evaluation of Post-Incident Cooling Systems of l
Light Water Power Reactors," Third International Conference on the Peaceful Uses of Atomic Energy, Volume 13, Session 3.9, New York, 1965.
(d)
D.C. Slaughterbeck, " Review of Heat Transfer Coefficients for Condensing Steam in a Containment Building Following a Loss-of-Coolant Accident, IN-1388, September, 1970.
'6.
Standard Review Plan - 6.2.1.5, Minimum Containment Pressure Analysis for Emergency Core Cooling System Performance Capability Studies, U.S. Nuclear l
Regulatory Commission Standard Review Plan, Office of Nuclear Reactor Regulation, NUREG - 0800 (Formerly NUREG - 75/087).
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Enclosure GL 96-06 Request for Additional Information Analyzed Piping Systems Data Summary l
f
i Enclosure l
GL 96-06 Request for Additional Information y
a.
Penetration No. 2 System description
- letdown line to letdown heat exchanger l
Penetration material
- SA-182 F316 Head diameter
- 10.75" Line No.
- 1208ML002 Pipe size
- 2" sch 160 Pipe material
- SA-376 TP316 b.
. Penetration No. 6 System description
- SIS drain & test line Penetration material
- SA-182 F304 Head diameter
- 10.75" Line No.
- 1204ML111 Pipe size
- 2" sch 40s Pipe material
- SA-376 TP304 c.
Penetration No. 8 System description
- charging line to regenerative heat exchanger Penetration material
- SA-182 F316 i
Head diameter
- 10.75" Line No.
- 1208ML104 Pipe size
- 2" sch 160 Pipe material
- SA-376 TP316 d.
Penetration No. 11 I
System description
- nuclear service water Penetration material
- neck extension SA-376 TP304 head fitting SA-403 WP304 Head diameter
- 10.75" Line No.
- 1415ML157 Pipe size
- 3" sch 10s Pipe material
- SA-376 TP304 l
)
t ---_------_-----_J
Enclosure GL 96-06 Request for Additional Information e.
Penetrations No. 36 and 37 System description
- steam generator secondary water blow down Penetration material
- SA-350 Gr LF2 Head diameter.
- 18" Line No.
- 1301MLO!5 Pipe size-
- 6" sch 80 Pipe material
- SA-333 Gr 6
-Insulation thickness
- 3.5"
' Insulation type
- AHC f.
Penetrations No. 42 and 43 System description
- component cooling water inlet and outlet Head diameter
- 24" Line No.
- 1203ML200 (penetration 42 - CCW inlet)
- 1203ML162 (penetration 43 - CCW outlet)
Pipe size
- 10" sch 40-Pipe material
- SA-106 Gr B
-Insulation thickness
- 3" (added) g.
Penetrations No. 45 and 46 System description
- Chilled water inlet (penetration 45 - chilled. water inlet)
- Chilledwateroutlet(penetration 46-chilledwater outlet)
Penetration material
- SA-106 Gr B Head diameter
- 18" Line No.
- '1513ML698 (penetration 45 - chilled water inlet)
- 1513MLO41 (penetration 46 - chilled water outlet) 1 Pipe size
- 8" sch 40
' Pipe material'
- SA-106 Gr B p
Insulation thickness
- 3" (modified)
L i
L,
I l
Enclosure GL 96-06 Request for Additional Information Attachment I h.
Penetrations No. 54 and 55 System description
- containment emergency sump recirculation Penetration material
- SA-182 F304 Head diameter
- 40" Line No.
- 1204ML004 (penetration 54)
- 1204ML003 (penetration 55)
Pipe size
- 24" sch 10s Operating Temperature
- 225"F Pipe material
- SA-358 Gr 304 Class 1 1.
3/4" sch 160 Samnle Line Penetrations Several lines, were listed below, evaluated.
I-1 Penetration No. 1 System description
- pressurizer steam ' space sample line Penetration material
- SA-182 F316 Head diameter
- 10.75" Line No.
- 1212MLO37 Pipe size
- 3/4"sch160
' Pipe material
- SA-376 TP316 I-2 Penetration No._1 System description
- RC sample from hot leg loop Penetration material
- SA-182 F316 Head diameter
- 10.75" Line No.
- 1212MLO31 Pipe size
- 3/4"sch160
. Pipe material
- SA-376 TP316 I-3 Penetration No. 7 l
System description
- RCP seal bleed-off Penetration material
- SA-182 F316 f
Head diameter
- 10.75" Line No.
- 1201ML052 Pipe size
- 3/4"sch160 Pipe material
- SA-376 TP316 t !
i
Enclosure GL 96-06 Request for Additional Information I-4
' Penetration No. 12~
. System description 1
- pressurizer surge line sample Penetration material
- SA-182 F316 Head diameter
- 10.75" Line No.
- 1212MLO43 Pipe size
- 3/4"sch160 Pipe material
- SA-376 TP316 1
1-5 Penetratio'n No. 17 i
System description
- steam generator secondary water sample Penetration material
- SA-350 Gr LF2 Head diameter
- 10.75" Line No.
- 1301ML014 Pipe size
- 3/4"sch160 Pipe material
- SA-333 Gr 6 1-6 Penetration No. 44 System description-
- steam generator secondary water sample Penetration material
- SA-350 Gr LF2 Head diameter
- 10.75"
'Line No.
- 1301ML018
-Pipe size
- 3/4"sch160 Pipe material'
- SA-333 Gr 6 l.
! c-
l Enclosure GL 96-06 Request for Additional Information I
l l
l I
i Drawings of Piping Segments Between Isolation Valves I
e i
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l l
l Enclosure GL 96-06 Request'for Additional Information 4
Analyzed Piping Systems Pinino Systems Drawinas
.The drawings, listed below, are attached.
These drawings pertain to the piping system where the insulation was modified. Also included are the drawings of valves HV-5804 and HV-7513.
Item No.
Description 1-Isometric drawing, piping at penetration No. 42, 2
Piping segment between isolation valves at penetration No. 42.
3.
I'sometric drawing, piping at penetration No. 43, 4
Piping segment between isolation valves at penetration No. 43.
5 Isometric drawing, piping at penetration No. 45.
6 Piping segment between isolation valves at penetration No. 45.
7 Isometric drawing, piping at penetration No. 46.
8 Piping segment between isolation valves at penetration No. 46.
9_
Isometric drawing, Piping at penetration No. 2,' sheet l',
'10 Isometric drawing, Piping at penetration No. 2, sheet 2.
11 Piping segment between isolation valves at penetration No. 2.
12 Isometric drawing, Piping at penetration No. 6.
13 Piping segment between isolation valves at penetration No. 6.
14 Isometric drawing, Piping at penetration No. 8.
15 Piping segment between isolation valves at penetration No. 8.
16 Isometric drawing, Piping at penetration No. 11.
17 Piping segment between isolation valves at penetration No. 11.
18 Isometric drawing, Piping at penetration No. 36.
19 Isometric drawing, Piping at penetration No. 54.
20 Piping segment at penetrations No. 54 and 55.
J
-21 Valve HV-5804 drawing.
22 Valve HV-7513 drawing.
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Lon eT-xa-2 ,os, ue T wow g35 ' AllFR osh I-17-20 fytyy estaEnos BECHTEL WERC0iff0R57'C" ]#'.'8 C L. A. POWER DIVISION - JOB NO.10079 S S3 87 mIyw.m. SOUTHERN CALIFORNIA EDISDN CO. Item-14 Isometric drawing, piping ~ d $'e73 SAN ONOFRE NUCLEAR GENERATING STATION 7 at penetration No. 8 NIT 2 02olr 6oo 57 exon:riuc oce; vare on s t. iece ~~~~ 9t- ?$!U u-mnau,,o.,, S ?-/208+n -/04 2 "C FEO(2IN8) m a gF.2 g;'",. A. y /.g DRamNo NuMsER REv. ~~ [tf ?BGA {'yf,} 52-120p-til.-/04 7 22A/2A*
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