ML18078B116
| ML18078B116 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/29/1979 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Parr O Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904020085 | |
| Download: ML18078B116 (9) | |
Text
Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201 /430-7000 March 29, l979 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
Mr. Olan D. Parr, Chief Light Water Reactors Branch 3 Division of Project Management Gentlemen:
RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-3ll Public Service Electric and Gas Company hereby transmits 60 copies of revised pages to its report "Evaluation of the Reactor Coolant System Considering Subcompartment Pressuri-zation Following a LOCA for Salem Units No. l and 2 11 which was submitted on March 6, l979.
This revision involves the following:
l)
Replacement pages for Sections 6.l, 6.2, 6.4 and Table 9
- 2)
Replace Table 8 with Tables 8A and BB
- 3)
Insert additional pages, Table ll and Figure 54 Should you have any ~uestions, please do not hesitate to con-tact us.
Enclosure The Energy People R. L. Mittl General Manager -
Licensing and Environment Engineering and Construction 7 904 02 oo 8'S 95-2001 (400M) 9-77
.e
. 6.0 RCS EQUIPMENT 6.1 Steam Generator and Reactor Coolant Pump Supports The supports were analyzed for the effects of asymmetric pressure loads combined directly with LOCA loop depressurization loads using the computer program WESAN.
These total LOCA effect loads were combined with seismic loads by the SRSS method.
Tables BA and 88 give maximum member stresses and the ratios of the combined stresses to allowables for critical members of the steam generator and reac-tor coolant pump supports.
Included in the Tables are material types and yield stresses. Of the break cases considered, only the governing (maximum) member stress is given.
Reference 4 includes mathematical models and drawings for the steam generator and reactor coolant pump supports.
As shown in the Tables, the permissible stresses, are greater than the calculated stresses for all members.
the permissible stresses are defined in the ASME Boiler and.Pressure Vessel Code,Section III, Subsection NF, with faulted condition increase factors.
Member 31 on the reactor coolant pump support had a stress ratio of 99% which indicates that the actual stress in the member is below the allowable. Conservatisms in the calculation* of stresses, determination of allowable stresses, redundant design of the supports, calculational methods for LOCA and asymmetric pressure loads, and combination of loads provide a large margin between allowable stresses and stresses required to induce failure. The factor that the member stress is less than allowable is sufficient to insure that the member is adequate for all loadings considered.
6.2 Steam Generator Upper Lateral Supports The steam generator upper lateral supports were analyzed for the effects of asymmetric pressure loads combined with LOCA loacts due to breaks in the reactor coolant loop and the main steam line. The total loads from the effect of postulated pipe breaks were combined with normal loads and seismic loads.
l.
Loop Break The analysis shows that for the breaks in the primary ~oop, the maximum stress in the support ring band is 43% of allowable and maximum load on the snubbers is 20% of rated capacity.
- 2.
Main Steam Line Break The analysis shows that the maximum stress in the upper support rina is less than 81% of allowable.
The maximum snubber load (occurring in the compression side inubbers) was 98% of rated capacity.
The rated capacity is conservative since the actual snubber capacity is greater than the listed rating. The snubber loads are less than rated capacity; therefore, the snubbers are adequate.
Since the allowable stress limits, as defined in ASME Boiler and Pressure Vessel Code Section III Subsection NF in the ring band, are all greater than the actual stresses, the steam generator upper lateral support is considered adequate for all loadings considered.
6.4 Reactor Vessel Supports The reactor vessel supports were analyzed using the finite element code WECAN with the maximum loads obtained from *the reactor pressure vessel LOCA analysis (Ref. A, Table 2-8) described in Section 5.
The normal loads were combined with the SRSS of LOCA and seismic loads.
The three dimensional mathematical model using beam and plate elements is shown in Figure 54.
Points of load application are as indicated on the Figure.. The vertical applied load was uni-formly distributed in the downward direction, while the horizontal load was applied in shear at the shoe.
Maximum stresses and the stress ratios for each component of the reactor vessel support are listed in Table 11, as are the respective yield strengths at tem-perature. Regions of maximum stress are indicated by the numbers on the Figure.
All components of the reactor vessel supports had stresses below the limits defined in the ASME Boiler and Pressure Vessel Code,Section III, Subsection NF, Article XVII-2000; there-fore the reactor vessel supports are adequate for all loadings considered.
TABLE BA I. Reactor Coolant Pump Lower Supports A.
Columns Break Member Ratio Stresses (ks1)
Material Axial Sh'ear Y Shear Z Bending v Bending Z Yield Hot Leg 41 26.3%
3.82 0
1.37 8.83 0
42 ASTM 441 Cold Leg 41 57.6%
7.19
.124
.996 20.2
.552 42 ASTM 441 e
B. Frame Hot Leg 40 49.8%
9.21
.330 0
.00435 10.4 42 ASTM 441 Cold Leg 31 99%
33.3
.856
.0679 0
6.55 42 ASTM 441
A.
Columns Break Member Hot Leg 72 Cold Leg 79 B.
Frame Hot Leg 9
Cold Leg 30 Ratio 14.4%
18.8%
95%
89.6%
TABLE BB II. Steam Generator Lower Supports Axial 4.22 4.17 1.17 1.03 Shear Y
.00495
.00495
.287.
.501 Stresses Shear Z
.129
.140 21.3
.134 (ksi)
Bending Y 1.21 2.06 17.9
. 919 Bending Z
.146
.145 1.92 2.36 Yield 42 42 42 42 Material "
ASTM 441 ASTM 441 ASTM 441 ASTM 441
TABLE 9 REACTOR COOLANT LOOP PIPING STRESS
SUMMARY
Maximum Stress Due to Maximum Stress Due to Asymmetric Pressure Other Faulted Condition Allowable Location (ksi}
Loads (ks i} l,2 Str2ss (ksi)
Hot Leg 2.6 28.5 Crossover Leg 2.8 23.4 Cold Leg 0.53 26.1 (1)
The "Other Faulted Condition Loads" are: deadweight, pressure, Design Basis Earthquake, LOCA loop hydraulic forces and reactor vessel motion.
Seismic and LOCA stresses are combined by SRSS.
(2) These stresses represent the maximum RCL piping stresses resulting from postulated breaks at the reactor vessel inlet, reactor vessel outlet and reactor coolant pump outlet nozzles (Ref. 4, Table 3-1).
50.1
- 50. l 50.1
TABLE 11 Reactor Vessel Supports Location Stress (ksi)
Ratio Yield Stress (ksi)
Material Shoe (1) 30.7 41%
74.5 A508, Cl. 4 Shear Panels (2) 19.. 2 55%
35.2 A588 Horiz. E of Box (3) 22.8 65%
35.2 ASTM 441 Vert. E of Box (4) 28.2 79%
35.2 ASTM 441 Pins (5) 61.3 82%
130.0 ASTM SA540 Gr.
824 Anchor Bolts 125.0 AB4140 Shear Pins (6) 9.0 23%
69 AISI 4340
I I I I
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I I I I I -I I I I -I I I I I I I I I I I I I I
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FIGURE 54.
Reactor Vessel Support Model