ML19329A545
| ML19329A545 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Crane  |
| Issue date: | 10/30/1972 |
| From: | Desiree Davis US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Lainas G US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8001060060 | |
| Download: ML19329A545 (12) | |
Text
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[d' OCT 3 OTL l
Gb-A f'l nea W C. Lainas, Chief, Containment Systems Branch, Directorate of Licens THRU:
A. Schwencer, Chief, Pressurised Water Reactors Branch I g
CONTADcGMT PRAK PRESSURE AND INTERNAL PRESSURE DIFFERENTIAL In regard to Mr. R. C. DeYoung's technical assistance request (TAR) to Mr. J. M. Hendrie dated October 20, 1972, we understand that you will need additional information similar to that requested for the Kevaunes review to resolve the problems discussed in the TAR for Oconee and In an effort to obtain this information, requests Three Mile Island Unit 1.
for additional information concernha the containment peak pressure problem have been transmitted to the applicants for Oconee and Three Mile Island Unit 1.
Responses to these requests are expected by November 20, 1972 and will be forwarded to you upon receipt. For the second problem, containment j
internal pressure differentials, information needed for ocones is presently available since this information was supplied during the original review.
In regard A copy of the appropriate FSAR pages is enclosed for your review.
to Three Mile Island Unit 1, a request for additional information relative to this problem is in preparation and will be transnitted to the applicant.
Response to this request is also expected by November 20, 1972.
As was previously indicated in the October 25, 1972 notice ve are holding a aceting with BW on November 3,1972 to discuss several ECCS and containment topics. In particular, B&W wishes clarification of our requests concerning the contain=ent peak pressure problem. This would also be an excellent opportunity to discuss the containment internal pressure differentials problem relative to l
the enclosed Oconee FSAR naterial if your initini review of this caterial can be completed by this time.
Original Signed by
{
Donald K. Davis AW
/
E, Donald K. Davis h
A Pressurized Water Reactors Branch No. 4 jff#
Directorate of Licensing 8
l l
l Enclosure DISTRIBUTION Bubject File ces R. C. DeYoang PWR-4 Reading R. Tedesco RP Reading A. Schweneer DDavis B. Schierling Y. Pelt 4 c g/ 2,4 omet> J HR-4
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Dlav s:emp.
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Fons AEC-Ste (Rev.9 53) AECM 0240 t;. A covrasuort pacetzsc orricz into o..os.34.
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A breathing rate of 3.47 x 10-4 m3/s is assumed for the 2-hour exposure. For 3
the 30-day exposure, a breathing rate of 2.32 x 10-4 m /s is assumed.
The total integrated thyroid doses resulting-from this LOCA fission product release are 6.0 Rem for the 2-hour exposure at the 1 mile exclusion distance, and 5.5 Rem for the 30-day exp o ure at the 6 mile low population distance.
These doses The corresponding whole body doses are 0.014 Rem and 0.018 Rem.
are shown in Table 14-17.
The effect on the dose from an LOCA release was also investigated in the event LOCA occurred while the reactor building was being purged. Assuming that aall the fission products had been initially released 'to the reactor building, which was at a constant pressure of 59 psig, the flow from the purge vents would resemble flow through a long duct. With an isolation valve closing time of 5 seconds and with no credit taken for flow restrictions due to the closing valve, additional dose equivalent curies of 7 131 will be released. This re-lease corresponds to a 0.05 Rem additional '
roid dose at the site boundary.
14.2.2.3.8 Pressure Buildup In Primary And Secondary Due To Loss-Of-Coolant Accident Following the design basis accident y pe. rupture within either compartment (rg,
actor cavity _,_s, team generator), high-enthalpy water flows out of both ends of the pipe, flashing partly to steam. As the pressure builds up within the com-the steam-air-water mixture will flow through openings in the com-
- partment, partment into the main containment. The pressure built up in the compartment is dependent on the number and shape of vent areas leading into the main con-tainment, the volume of the compartment, and the blowdown rate from the broken pipe.
(
14.2.2.3.8.1
. Pressure Buildup Within The Reactor Cavity i.<
that 9
The cavity model, as analyzed, consists of a 5520 cubic foot compartment, has a vent (always open) and concrete shield plugs which blowout during the accident. A value of six (6) square feet, representing the effective (always pen) vent area of tha cavity, was used. The total effective area of the vent 5*
and the shield plugs (af ter shield plugs blow out) is 75 square feet. The ei-fect of the increase in area with time, as the shield plugs blow, has been taken into account in the analysis by calculating the acceleration of the plugs due to the pressure buildup within the reactor cavity. The results of the analy-sis are presented in Figure 14-68.
The indicated rapid pressure drop results from the shield pi g blowout.
Largest Break Size Reactor Cavity Can Withstand An analysis of the preliminary design of the reactor cavity was initially made in June 1967 to determine _the largest break size that the cavity could with-stand. The documentation established a break size of 8 square feet with the associated peak pressure of 195 psi across the cavity walls. Reanalysis of the final design determined that the cavity could withstand a somewhat higher pres-sure (205 psi) associated with a larger break size (8.5 square feet).
Rev. 5.
5/25/70 14-64
.,ssible Within Rcactor C:tvity
} 3 f i % t g r ry)~ ~ %
/
Si;ilarl s tw ag -
,,is of the largest break size possible that could occur
~
-vity wan calculated to be 3.0 square feet which cou'.d
_within t e reactu 1 pressure across the cavity walls bf 120 psi.
This is
_prodoc eaints on the primary lines which limit hed by * -
the physical sepa-accoMP the_ ruptured lines.
The cavity, therefore, can safely ration { the c"d2
-- tes resulting from the largest break possible in the cool-withstag g p r ~
reactor cavity.
ant P P cInted with varinus breakA chart and tabulation of the re i
thin nizes cavity prennuri*.
is shown on Finnre 14-68.
2ssure llulldup Within The Steam f;enerator C g.2.2.L U 2 ompartments
-;enerator compartments designated East and West.
Th2re are two st f -
y -ed assumes two separate compartments.
The idea 1ized geomet r The East compart-e" v lume f 60,400 cubic feet and an " effective" flow gg g.. _
o
'otut has a The West compartment has an " effective" volume of est.
1000 sq*
an " effective" flow area of 1100 square feet.
"T*"
700 cubic fe* I
.cted for these calculations:
Only one dudownwas'in{](
the great,est energy release rate.the hot-leg rupture, since this case rePre-re d'
', f -tial across the East and West generator compartment walls is presented in Figures 14-69 and 14 70.
The prc tunction ~~
The values gssociated Vit~
differentials are 9.68 psi and 7.98 psi,'respectively.
f.-
11aximum and De' -
- ssures Within Steam Generator Compartments
~
The table belC' [ g x rator compartments:a summary of the maximum pressure and the design
.g sure in the 5'# "
STEAM GENERATOR COMPARTMENTS
.'X.
PRESS.
MAX./ DESIGN DESIGN PRESS.
g gigg
'ast West PRESS.
East West
't. 7 8.0 11.1 psi 14.0 S I*
- i E*i 87%
72%
/.g, ptions Were Incorporated In *.he Analysis -
gr thrughout the compartment are in thermal equilibrium at all times.
Duri 8 each ti",'._r,tep the blowdown mass expands isenthalpicly to the total com-The water present at that time could form more steam only e prcS6"#
The water is assumed to undergo no further change of.
ion.
pa o ),ynpr>r to the floor of the. compartment.
b halpic e P^) }'on to the partial pressure of the steam already in the co drop domone"cously with the air.
g,.,e-step the mass of water added is temporarily stored.
DurinS each If the-equilibrine eulatien indicates atomsphere superheating, then a sufficient 3
14-64a Rev. 5.
5/25/70 (New Page) i 1
_m
K (3)
.Kegg = Et i Orifice Coefficient Datn Available information on nominal orifice coefficients is summarized in Figure 14-73.
Tienel Loss Coefficient Data Tle" re < psi reil exper imeni ni head losn coefficient dntn are reprenented in Fip,uren 74 and 14-75.
,lixuansion Factor The expansion factor used in the orifice flow relation is that applicable to a converging nozzle:
2 2
~k-1
~
A P
2 k
2 y
- 1. -
1-A1 k-1 P1 P1 y.
2 2
p2 E
A2 P
12 Al P1 p1
.J klere:
A1 = upstream area A2 = orifice area P1 = upstream pressure P2 = dosmstream pressure k = isentropic exponent The isentropic exponent (k) for these calculations is based on the mixture of air and steam.
This calculation is conservative becQpse for orifices which are not well rounded conpressibility effects increase the expansion factor slightly.
Rev. 5.
5/25/70-14-64c (New Page)
a;-
qu;ntity ofl this ' temporarily stored watcr is flashed into steam.such that the atmosphere is just saturated.
If there is not sufficient water ~in the tem-
.porary storage to saturate the atmosphere, then the atmosphere is allowed to superheat.
The flow out of the compartment into the main containment is calculated using the compressible flow equations-for subsonic and' choked flow. The orifice coefficient and-expansion factors used are discussed in Section 14.2.2.3.8.4.
The flow is assumed to become sonic at the critical pressure ratio defined by:
k P2 2
k-1 P1 k+1
.Where:
P2 = downstream pressure P = upstream pressure y
k = isentropic exponent (see Section 14.2.2.3.8.4)
The volume of the compartment is measured to the minimum cross sectional area of each of the flow openings leading out of the compartment.
-14.2.2.3.8.4 Calculation of Orifice Coefficient and Expansion Factor The orifice coefficients (C) used in the orifice flow relation are sensitive to Reynolds number, orifice size, and orifice shape. Orifice coefficients typically become independent of Reynolds number at high Reynolds numbers. Rey-nolds numbers through-orifices in the present area are greater than 10, so 6
the Reynolds number effect does not apply here.
Sufficient experimental information on orifice co'efficients for our geometry is not available. However, information is available on the head loss coefficient (K) defined by:
2 P=KV (1) 2 Where:
P = Pressure drop across orifice V== Velocity through orifice The relation between (C) and (K) is:
C=
1 (2)
/K The advantage.of the head loss coefficient is-that the total head loss for a
. complicated flow system can'be determined ~from an equivalent K obtained by adding K's for separate parts of the system (i.e. bends, expansions, contrac-
-tions, etc.) as-follows:
u,a Rev. 5.
5/25/70
=
720 E
7,-
200 I
1
- 180 a
f
\\
f 140 B
/
IE 33 Cm 100 s
/\\/
\\
/
N
\\
l0
/ / XXTN g
,0 X
/
XV \\\\
0 O.001 0.010 0.100 1.000 10.000 TIME AFTER RUPTURE, SECONDS CURVE BREAK SIZE
- DESCRIPTION A ' ""
8.5 sq. ft.
Corresponds to maximum break size reactor cavity con withstand B
3 sq. ft.
Correspo* ds to maxirr.um hot leg break n
size possible within the reactor cavity C
1 sq. ft.
1.0 square foot hot leg rupture D
0.4 sq. ft.
0.4 square foot hot leg rupture See Figures 14-71 and 14-72 for description of blowdown data used Figure 14-68 Pressure Transients for Range of Rupture Sizes Within Reactor Cavity.
i OCONEE NUCLEAR STATION Figura 14 68
[
-. _ ~
(New) Rw. 5 5/25/70
8 6
~.
6 E
/
. f-5
/
g 4-B w
5 E
2
/
0
'f
O.001 ',,,,,
0.010 0.100 1.000 10.000 TIME AFTER RUPTURE, SECONDS Pressure Differential Across West Compartment Walls Versus Time After Double-Ended Hot-Leg Rupture (Case 1-14 Square Feet).
- OCONEE NUCLEAR STATION
&catn
)-
Figure 14 - 60 it;..; n u,. o v ie ;.:; c
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10 9
8 7
e sZ Et:'. $
5
'O U9. 4 0
E
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3 2
I I
1
' ' ' '.010 '
' ' ' '.100
' ' ' '.0@
10.000 O
0 0
1 O.001 TIME AFTFR RUPTURE, SECONDS Pressure Differential Across East _ompartment Walls Versus Time After Double-Ended Hot-Leg Rupture (Case 1-14 Square Feet).
th%
outrata - OCONEE NUCLEAR STATION W
Figure 14 - 70 (New) Rev. 5 5/25/70
b.
Section 5.1.2.5 requires a minimum of 0.15 percent bonded steel rein-
' forcing (as stated above) for any location.
At the base of the cylin-der, the controlling design case requires 0.25 percent vertical rein-
' forcing. As a result of pursuing the recommendation of the AEC Staf f to further investigate current research on shear in concrete, several steps were taken:
1.
The work of Dr. Alan H. Mattock was reviewed and he was re-k tained as a consultant on the implementation of the current research being conducted under his direction.
The criteria
[
has been updated in accordance with his recommendations.
2.
Concurrently with reviewing Dr. Mattock's work, the firm of T. Y. Lin, Kulka, Yang and Associates was consulted to review the detailed design of the cylinder to slab connection.
It was their reconmendation to use approximately 0.5 percent re-inforcing rather than the 0.25 percent reinforcing indicated by the detailed design analysis for the vertical wall dowels.
This increase would assure that there was sufficient flexural steel to place the section within the lower limits of Mattock's test data (approximately 0.3 percent) to prevent flexural cracking from adversely affecting the shear capability of the ection.
5.1.5 INTERIOR STRUCTURE 5.1.5.1 Design Bases The Reactor Building interior structure (comprising all elements inside the Reactor Buil' ding shell) is a Seismic Class I structure and is designed on the following bases:
The stresses in any portion of the structure under the action of dead a.
load, live load and design seismic load will be below the allowable stresses given by either the ACI Building Code, ACI 318-1963_except as 1.
, h.
y ed,_in 5d 2.,6, AISC Manual of Steel Construction, 6th Edition. ' '
b* h b.
The stresses in any portion of the structure under the action of dead load, and thermal load will be below 133 percent of the allowable t
t stresses given in (a).
The capability to safely shut down the plant will be maintained under c.
the combined action of dead load, maximum seismic load, pressure and jet impingement load.
The latter two loads are based on' the rupture UT'one pipe in the primary loop. The deflections of structures and supports under these combined loads would be such that the functioning of engineered safeguards equipment would not be impaired.
The yield
~ 1oad equations in Appendix SA are adhered to except that local yielding is permitted for pipe, jet or missile barriers provided there is no general failure.
Rev. 1.
9/15/69 5-39
i U
5.1.5.2 Design. Loads and Materials The Reactor Building interior structure consists of (1) the reactor cavity, (2) two steam generator compartments, and (3) a refueling pool which is located between the steam generator compartments and above the reactor cavity.
The reactor cavity houses the reactor vessel and serves as a biological shield wall.
The reactor cavity is also designed to contain core flooding water up to the level of the reactor nozzles.
The primary functions of the steam generator compartment walls are to serve as secondary shield walls and to resist the pressure and jet loads described below.
The foundations for all NSSS equipment including the reactor vessel, the steam generators, and the pressurizer are designed to remain within the elastic range during rupture of any pipe combined with the " maximum earthquake."
The design pressure differential across walls and slabs of enclosed compart-ments in the internal structure are as follows:
Reactor Cavity
- 208 psi East Steam Generator Compartment - 11.1 psi West Steam Generator Compartment - 11.1 psi In addition to the peak pressure dif ferentials, the steam generator compart-ment walls are designed for simultaneous action of a single _ jet impin3 ament, load and the safe shutdown earthquake. Design of structures was done using C
A conventional' structural analytical techniques.
/
Pipe whipping restraints are provided for the main steam, feedwater and other high-pressure piping in accordance with criteria in Section 5.4.
The materials used for the above structural elements are as Eo'15ows:
~
e Structural Steel -ASTM A36 Concrete
-f'
= 4000 psi at 28 days.
-f' = 5000 psi at 28 days (for steam generator bases, reactor foundation, and primary shield wall).
Reinforcing Bars -ASTM A615, Grade 40 for Bars #11 and under ASTM A615, Grade 60 for Bars large; than #11.
5.1.5.3 Missile Protection High-pressure reactor coolant system equipment which could be the source of missiles is suitably screened by the concrete shield wall enclosing the reactor coolant loops and by special missile shields to block any passage of missiles to the Reactor Building walls.
Potential missile sources are oriented so that the 5-40 a
acket 50-269, -270, and -287 FSAR Suppisment 8 Szptemb2r 14,1970 REQUEST 12 Discuss the analysis which shows that primary pipe whip will not cause failure of the secondary system.
RESPONSE
A detailed study of the primary loop was performed to determine potential pipe break locations which could possibly cause either fluid impingement or pipe impact forces on the secondary system.
The results of this evaluation indicated the most credible break locations which could cause either of these e'ffects are:
a guillotine break at the pump discharge in the cold leg 1.
pi ing; P
a longitudinal split in the vertical pump suction segment of 2.
the cold leg piping; or, a longitudinal split in the vertical segment of the hot leg 3.
piping.
All of.the above breaks could potentially af fect the generator The main steam lines, however, because of their prosimity to it.
are shielded f rom the ef fects of pipe breaks by the generator.
The primary piping and steam generator were analyzed for each 'of the above -breaks and supports provided to restrain the pipe from whipping In addition, the stresses in the generator shell into the generator.
forces were calculated and found to be due to the fluid impingement within acceptable limits.
~
The restraints on the prima'ry loop 'are shown in Figures 12-1 and The coolant pump is restrained by steel supports from the 12-2.
The hot leg piping is restrained by the con-primary shield wall.
crete support at. the primary cavity penetration, an intermediate steel support from the primary wall, and another steel support near The vertical segment of the cold the generator upper tube sheet.
leg piping is restrained by a steel support midway along its length,
~
which would spread any rupture load over a larger area of the generator shell.
To verify the location and size of the piping supports, the piping was analyzed for rupture loads occurring at the worst point along The rupture thrust force was assumed equal to Px A, its length.
wbere P is the coolant pressure and A the flow cross-sectional area The thrust was applied as an equivalent static force of the pipe.
Assuming the force to be a using a dynamic load factor of 2.0.
h point load acting at the midpoint of the span between supports, t e The supports are piping stresses were calculated using beam models.
- located so as to prevent the formation of plastic hinges in the piping, which would lead to an unstable linkage-type structure and possible impacting against the generator.
FSAR Supplement 8-22 i
eket 50-269, -270, and -287 r$AR Supplement 8 Szptember 14, 1970 To evaluate the effect of fluid jet impingement on the generator, an equivalent static pressure load on the shell was calculated. A 2 for the cold leg was break of 14 ft2 for the hot leg or 8.5 f t assumed. The maximum initial mass velocity was computed using the methods outlined in the report " Maximum Two-Phase Vessel Blowdown From Pipes, APED-4827," by F. J. Moody. It was assumed that the fluid Icaves the break in a direction normal to the pipe and that its velocity undergoes a 90* change in direction upon impinging on the OTSG. The resulting shell pressure loading was calculated to be 1300 psi.
A shell analysis was performed on the OTSG to determine the stress intensity due to the above loading. A B&W proprietary digital computer code, which considers two-dimensional shells with asymmetric loading, was utilized. The loading distribution and stress model are shown in Figures 12-3 and 12-4 The maximum stress intensity was computed
,e 38,6'0. psi. This is less than the allowable stress of 46,670 psi.
Based an these results for the 36" ID pipe break, it was concluded that the OTSG shell could 4
also withstand the reduced loading which would be generated by a 28" ID break.
FSAR Supplement 8-23
.