ML20216C361
| ML20216C361 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/23/1987 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8706300269 | |
| Download: ML20216C361 (33) | |
Text
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DUKE POWER GOMPANY P.O. HOX 33180 CRARLOTTE, N.O. 28242
'HALU.TUCKEH
' TELEPHONE vece renament (704) 373-4531 muas4m emanconow June 23, 1987 U. S. Nuclear Regulatory Commission Attention: Document Control Desk-Washington, D. C. 20555
Subject:
Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 l
i Deletion of Upper Head' Injection System
Dear Sir:
My letter of June 12, 1987 transmitted a proposed Technical Specification amendment
- regarding the deletion of the Upper Head Injection System for Catawba ' Attachment 4 to that submittal provided marked-up FSAR pages. Attached herein are marked-up
'FSAR pages for the non-LC0A transients which were inadvertently omitted from the.
June 12, 1987-letter.
This letter supplements the previous amendment. request to Catawba's Technical y
Specifications. Accordingly, a check for $150.00 is not required.
Pursuant'to 10 CFR 50.91 (b) (1) the appropriate South Carolina State Offfcial is being provided a copy of this Supplement to our amendment request.
Very truly yours, p
/} /
g l
-Hal B. Tucker RWO/78/sbn Attachment f0 [
B706300269 B70623 l
0 PDR ADOCK 05000413 i
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U.:S..Nuclacr Reguictcry Cosmiccion
'J9ne 23,'1987'
,yp Page.Two-p'
~
F Exc t. Dr.;J. Nelson Grace,;Regiona1' Administrator-
' U. S. Nuclear _ Regulatory. Commission.
' Region II,
~101'Marietta' Street,:NW, Suite'2900L Atlanta,; Georgia 30323 a-
.Mr. Heyward Shealy, Chief?
Bureau of Radiological. Health South' Carolina Department of Health &.
. Environmental Control-2600l Bu11' Stree t Columbia, South Carolina 29201' L
1American. Nuclear Insurers
,c/o~Dottie:Sherman, ANI Library
- The Exchange, Suite 245 270 Farmington Avenue Farmington,~CT 06032 M&M Nuclear Consultants 1221 Avenue of the Americas New York,-New York 10020 INPO Records Center-Suite 1500>
.1100 Circle 75 Parkway
" Atlanta, Georgia. 30339 Mr. P. K.' Van Doorn NRC Resident. Inspector Catawba Nuclear Station i
j j
1 l
c e _..._.
CNS The ca ated sequence of events' for the excessive load increase i nt is shown on Ta 5.1.2-1.
15.1.3.3-Environ al Consequences 1
1 There will be no radiological c uences sociated with this event and
(
activity is contained within the fu di and reactor coolant' system within i
design limits.
1 15.1.3.4 Conc 1 s
l The anal resented above shows that for a ten percent ste ad increase, the remains above the limit value thereby precluding fuel or damage.
The plant reaches a s'tabilized condition rapidly following the load increase.,
15.1.4 INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE d
15.1.4.1 Identification of Causes and Accident Description J
The most severe core conditions resulting from an accidental depressurization of the Main Steam System are associated with an inadvertent opening of a single steam dump, relief, or safety valve. The analyses performed assuming 1
O a rupture of a main steam line are given in Section 15.1.5.
The steam release as a consequence of this accident results in an initial increase in steam flow which decreases during the accident as the steam pres-sure falls.
The energy removal from the RCS causes a reduction of coolant temperature and pressure.
In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity.
The analysis is performed to demonstrate that the following criterion is satisfied.
j 1
Assuming a stuck rod cluster control assembly, with offsite power i
available, and assuming a. single failure in the Engineered Safety Features System, there will be no consequential damage to the core or l
Reactor Coolant System after reactor trip for a stXeam release equivalent i
to the spurious opening, with failure to close, of the largest of any single steam dump, relief, or safety valve.
Accidental depressurization of the secondary system is classified as an ANS Condition II event.
See Section 15.0.1 for a discussion of Condition 11 events.
The following systems provide the necessary protection against an accidental depressurization of the Main Steam System.
1.
Safety Injection System actuation from any of the following:
a.
Two-out-of-three low steamline pressure signals in any one loop 15.1-8 Rev. 14
j m
1 3,
-(.
b.
Two-out-of-four low pressurizer pressure signals.
Two-out-of-three high containment pressure signals.
1 c.
2.
The overpower reactor trips (neutron flux and AT) and the reactor trip occurring in conjunction with receipt of the safety injection signal.
3.
Redundant isolation of the main feedwater lines-a Sustained high feedwater flow would cause additional cooldown.. Therefore,.
q in addi',fon to the normal control action which will close the main feed-1 440.66 water valves following reactor trip, a feedwater isolation signal will rapidly close all feedwater control valves and back up feedwater isola-3 tion valves, trip the main feedwater pumps, and close the feedwater. pump discharge valves.
j 4.
Trip of the fast-acting steam line stop valves (designed to close in less- -
than 5 seconds) on:
1 Two-out-of-three.1ow steamline pressure signals in any one loop.
a.
.1 b.
Two-out-of-four high-high containment pressure signals.
Two-out-of-three high negative steamline pressure rate signals in c.
any one loop (used only.during cooldown and heatup operations).
k
' Plant systems.and equipment which are available to mitigate the effects'of the accident are also discussed in Section 15.0.8 and listed in Table 15.0.8-1.
15.1.4.2 Analysis of Effects and Consequences Method of Analysis The following analyses of a secondary system steam release are performed for this section.
1.
A full plant digital computer simulation using the LOFTRAN Code (Reference 1) to determine RCS temperature and pressure during cooldown, and the effect of safety injection.
l 2.
Analyses to determine that there is no damage to the core or reactor coolant system.
The following conditions are assumed to exist at the time of a secondary steam 1
system release:
1.
End-of-life shutdown margin at no-load, equilibrium xenon conditions, and with the most reactive rod cluster control assembly stuck in its fully withdrawn position. Operation of rod cluster control assembly banks during core burnup is restricted in such a way that addition of positive t
15.1-9 Rev. 14 Carryover
L ;[
u a
' reactivity in'a secondary system steam release accident will.not lead to j
a more adverse condition than the case ~ analyzed.
2.
'A negative moderator coefficient corresponding to the end-of-life' rodded core with the most reactive rod cluster control assembly.in the' fully withdrawn position..The variation of the coefficient with temperature i.
and pressure is included.
The K,ff versus temperature at 1000 psi q
corresponding to'the negative moderator temperature. coefficient used is j
shown in Figure 15.1.4-1.
l 3.
Minimum capability for injection of high concentration boric acid
' solution corresponding to the most restrictive single. failure in the Safety Injection System.
This corresponds to the flow delivered by one charging pump delivering its full contents to the cold leg header. No
.L credit is taken for the low concentration boric acid which must be swept from the. safety injection lines downstream of the refueling water storage -
tank prior _to the delivery of concentrated boric acid (4,990 ppm from the refueling water storage tank) to the reactor coolant loops. \\ goo 4.
The_ case studied is a steam ~ flow of 270 pounds per second at 1200 pounds l
per square. inch' absolute (psia) with offsite power available. This is the maximum capacity-of any single steam dump, relief, or safety valve.
. Initial hot shutdown conditions at time zero are assumed since this represents the most conservative initial condition.
Should the reactor be just critical or operating at power at the time of i
a steam release,:the reactor'will be tripped by the normal overpower protection when power level reaches a trip point.
Following a trip at power, the RCS contains more stored energy than at no-load, the average coolant temperature is higher than at no-load and there.is appreciable energy stored in the fuel.
Thus, the additional stored energy is removed via the cooldown caused by the steam release before the no-load condi-tions of RCS temperature and shutdown margin assumed in the analyses are
~
. reached. After the additional-stored energy has been removed, the cool-~
down and reactivity insertions proceed in the same manner as in the j'
, analysis which assumes no-load condition at time zero. However,.since the ' initial steam generator water inventory is greatest at no-load, the
' magnitude and duration of the RCS cooldown are less for steam line release occurring at power.
5.
In computing the steam flow, the Moody Curve (Reference 3) for f(L/0) =
O is used.
6.
Perfect moisture separation in the steam generator is assumed.
i Results The calculated time sequence of events for this accident is listed in Table 15.1.2-1.
l 15.1-10 Rev. 14 Carryover l
. 7 l
............r..........
l
U
~ 7;. p-CNS k
The results presented are a conservative indication of the events which would occur assuming a secondary system steam release since it is postulated that
{
all of the conditions described above occur' simultaneously.
E
. Figures 15.1.4-2 and 15.1.4-3 show the transient results for a steam flow of i
270 lb/sec at 1200 psia.
l
'The' assumed steam release is typical of the capacity of any single steam dump, relief, or safety valve. Safety injection is initiated automatically by low pressurizer pressure. Operation of one centrifugal charging pump is assumed.
Boron solution at 2,04; ppm enters the RCS from the refueling storage water tank (RWST) providinglw (fficient negative reactivity to prevent core damage, su 900 The transient is quite conservative with respect to'cooldown, since'no credit is. taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the other steam generators.
Since the tran -
- sient occurs over a period of about 5 minutes, the neglected stored energy is likely to.have a significant effect in slowing the cooldown.
15.1.4.3 Environmental Consequences The inadvertent opening of a single steam dump relief or safety valve can result in' steam release from the secondary system.
If steam generator leakage exists coincident with the failed fuel conditions, some activity will be released.
IC 15.1.4.4 Conclusions The analysis shows that the criteria stated' earlier in this section are satisfied.
For an accidenta) depressurization of the main steam system, the DNB design basis as' stated in Section 4.4 is met and system design limits are exceeded.
15.1.5 STEAM SYSTEM PIPING FAILURE 15.1.5.1 Identification of C"auses and Accident Description
'The steam release arising from a rupture of a main steam line would result in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure.
In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of pos,tive d
reactivity.
If the most reactive rod cluster control assembly (RCCA) is as-sumed stuck in its fully withdrawn position after reactor trip, there is an increased possibility that the core will become critical and retura to power.
A return to power following a steam line rupture is a potential r,roblem mainly because of the high power peaking factors which exist assuming the most rsac-tive RCCA to be stuck in its fully withdrawn position.
The core is ultimately shut down by the boric acid injection delivered by the Safety Injection System.
i 15.1-11 Rev. 14 Carryover h
._ g s
4-CNS The' analysis of a main steam line rupture.is performed to demonstrate that the
,following criteria are satisfied:
Assuming a stuck RCCA with or without offsite power, and assuming a.
single failure in the engineered safety features, the core remains in place and intact. Radiation doses do not exceed the guidelines of l
110CFR100.
Although DNB and possible clad perforation following a' steam pipe rupture I
are not necessarily unacceptable, the following analysis, in fact, shows that no DNB occurs for any rupture assuming the most reactive assembly stuck in its fully withdrawn position.
The DNBR design basis is discussed in Section 4.4.
A' major steam line rupture-is classified as_an ANS Condition IV event.
See Section 15.0.1 for a discussion of Condition IV events.
i
-Effects of minor secondary system pipe breaks are bounded by the analysis l
presented in this section. Minor secondary system pipe breaks are classified as Condition III events, as described in Section 15.0.1.3.
i The major rupture of a steam line is the most limiting cooldown transient and j
is analyzed at zero power with no decay heat.' Decay heat would retard the cooldown thereby reducing the return to power. A detailed analysis of thisL transient with the most limiting break size, a double ended rupture, is presented here.
The following functions provide the protection for a steam line rupture:
1.
Safety Injection System. actuation from any of the following:
Two-out-of-three low steamline pressure signals in any one_ loop.
c a.
b.
Two-out-of-four low pressurizer pressure signals.
Two-out-of-four high Containment pressure signals.
c.
2.
The overpower reactor trips (neutron' flux and AT) and the reactor trip occurring in conjunction with receipt of the safety injection signal.
3.
Redundant-isolation of the main feedwater lines.
Sustained high feedwater f?ow would cause additional cooldown.
Therefore, in addition to the normal control action which will close Q
l the main feedwater valves a feedwater isolation signal will rapidly 440.66 close all' feedwater control valves and back up feedwater isolation valves, trip the main feedwater pumps, and close the feedwater pump discharge valves.
4.
Trip of the fast acting steam line stop valves (designed to close in less than 5 seconds) on:
15.1-12 Rev. 14 Carryover i
),
s CNS 4
Two-out-of-three low steam line pressure signals in any one loop.
'a.
i b.3 Two-out-of-three hign-high containment pressure signals.
.l
.Two-out-of-three high negative steam line pressure rate signals in j
c.
any one loop (used only during cooldown and heatup operations.
Fast-acting isolation valves are provided in each steam line; these valves 1
will fully.close within 10 seconds of a large break in the steam line. For breaks downsteam of the isolation valves, closure of all valves would com.
1 plately terminate the blowdown.
For any break, in any location, no more than
' one steam generator would experience an uncontrolled blowdown even if one.of the isolation valves fails to close. A description of steam line isolation is included in Chapter 10.
1
' Steam flow is measured by monit' ring dynamic head in nozzles located in the o
throat'of the steam generator. The effective throat area of the nozzles is j
1.4 square feet, which is considerably less than the main steam pipe area;-
j
~ thus, the nozzles also serve to limit the maximum steam flow for a break at
.any location.
Table 15.1.5-1 lists the equipment required in the recovery from a high energy line rupture. - Not all equipment -is required for any one particular break, i
since the requirements will vary depending upon postulated break location and details of balance of plant design and pipe rupture criteria as discussed.
(~'
elsewhere in7this application. Design criteria and methods of protection of 1
safety-reF 'd. equipment from the dynamic effects of postulated' piping ruptures a.e provided in Section 3.6.
15.1.5.2
' Analysis of Effects and Conseauences 1
Method of Analysis The' analysis of the steam pipe rupture has been performed to determine:
1.
The core heat flux and RCS temperature and pressure resulting from the cooldown following the steam line break. The LOFTRAN CODE (Reference 1) has been used.
2.
ThethermalandhydraulicbehaviorofthecorefollowingastAamline break. A detailed thermal and hydraulic digital computer code, THINC, has been used to determine if DNB occurs for the core conditions a
computed in item 1 above.
Studies have been performad to determine the sensitivity of steamline break results to various assumptions (Reference 4).
Based upon this study, the
.following conditions were assumed to exist at the time of the main steam line break accident.
1.
End-of-life shut down margin at no-load, equilibrium xenon conditions, and the most reactive RCCA stuck in its fully withdrawn position.
I l
15.1-13 Rev. 14 Carryover l
l l
-.m,..
j'
._ b CNS i
Operation of.
the control rod banks during core burnup is restricted-in such a way that addition of positive reactivity in a steam line break accident will not lead to a more adverse' condition than the case analyzed.
2.
A' negative moderator coefficient corresponding to the end-of-life rodded core with the most reactive RCCA in the fully withdrawn position. The i
variation of the coefficient with temperature and pressure has been included. The K,ff versus temperature at 1000 psi corresponding to the j
negative moderator temperature coefficient used is shown in Figure
{
15.1.4-1.
The effect of power generation in the core on overall
~
. reactivity is shown in Figure 15.1.5-1.
q i
The core properties associated with the' sector nearest the affected steam.
l generator and those ' associated with the remaining sector were conserva-tively combined to obtain average core properties for reactivity feedback l
calculations.
Further, it was conservatively assumed that the core power
. distribution was 'unifom.
These two conditions cause underprediction of the reactivity feedback in the high power region near the stuck rod.
To
'l verify the conservatism of this method, the reactivity as well as the 1
,{
powcr distribution was checked for the limiting statepoints for the cases j
analyzed.
-j This core. analysis considered the Doppler reactivity from the high fuel
,.I temperature near the. stuck RCCA, moderator feedback from the high water
!j enthalpy'near the stuck RCCA, power redistribution and nonuniform core.
Li inlet temperature effects.
For cases in which steam generation occurs j
in the high flux regions of the core, the effect of void-formation was j
also. included.
It was determined that the reactivity employed in the j
' kinetics analysis was always larger than the reactivity calculated 1
including the above local effects for the statepoints.. These results verify conservatism; i.e., underprediction of negative reactivity feedback from power generation.
3.
Minimum capability for injection of boric acid C,000 ppm from the RWST) solution corresponding to the most restrictive single failure in the Safety Injection System.
The Emergency Core Cooling System consists of three systems:
- 1) the passive accumulators, 2) the Residual Heat Removal System, and 3) the Safety Injection System. Only the Safety Injection System and the accumulators are modeled for the steam line break accident analysis.
The actual modeling of the Safety Injection System in LOFTRAN is described in Reference 1.
The flow corresponds to that delivered by one charging pump delivering its full flow to the cold leg header. No credit has been taken for'the low concentration borated water, which j
must be swept from the lines downsteam of the RWST prior to the delivery of high concentration boric acid to the reactor coolant loops.
For the cases where offsite power is assumed, the sequence of events in the Safety Injection System is the following. After the generation of 15.1-14 Rev. 14
i a
y 30 i
the safety injection signal (appropriate delays for instrumen tion, logic, and signal transport included), the appropriate valves begin to t
operate and the high head safety injection pump starts.
In
- seconds, the valves are assumed to be in their final position and the pump is assumed to be at full speed.
The volume containing the low concentration j
borated water is swept before the 4r40& ppm water reaches the core. This delay, described above, is inherently included in the modeling.
'f 900 In cases.where offsite power is not available, an additional 10 'second delay is assumed to start the diesels and to load the necessary. safety injection equipment onto them.
4.
Design value of the' steam generator heat transfer coefficient including allowance for fouling factor.
5.
Since the steam generators are provided with integral flow restrictors l
with a 1.4 square foot throat area, any rupture with a break area greater than 1.4 square feet, regardless of location, would have the same effect on the NSSS as the 1.4 square foot break. The following cases have been considered in determining the core power and RCS transients:
Complete severance of a pipe, with the plant initially at no-load a.
. conditions, full reactor coolant flow with offsite power available.
'l b.
Case (a) with loss of offsite power simultaneous with the steam
./-
line' break and initiation of the safety injection signal.
Loss of offsite power results in reactor coolant pump coastdown.
i 6.
Power peaking factors corresponding to one stuck RCCA and non-uniform core inlet coolant temperatures are determined at end of core life. The coldest core inlet temperatures are assumed to occur in the sector with l
the stuck rod.
The power peaking factors account for the effect of the local void in the region of the stuck control assembly during the return to' power phase following the steam line break.
This void in conjunction with the large negative moderator coefficient partially offsets the effect of the stuck assembly.
The power peaking factors depend upon the j
core power, temperature, pressure, and flow, and, thus, are different for l
each case studied.
The core parameters used for each of the two cases correspond to values determined from the respective transient analysis.
h Both cases above assume initial hot standby conditions at time zero since this represents the most pessimistic initial condition.
Should the reactor be just critical or operating at power at the time of a steam line break, the reactor will be tripped by the normal overpower protec-
)
tion system when power level reaches a trip point.
Following a trip at power, the RCS contains more stored energy than at no-load, the average coolant temperature is higher.than at no-load and there is appreciable energy stored in the fuel. Thus, the additional stored energy is removed L
15.1-15 Rev. 14 Carryover
CNS via the cooldown caused by the steam line break before the no-load conditions of RCS temperature and shutdown margin assumed in the analyses are reached.
After the addit.ional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis which assumes no-load condition at time zero.
7.
In computing the steam flow during a steam line break, the Moody Curve (Reference 3) for f(L/D) = 0 is used.
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Results The calculated sequence of events for both cases analyzed is shown on Table 15.1.2-1.
The results presented are a conservative indication of the events which would l'conditionsdescribedaboveoccursimultaneously.
occur assuming a steamline rupture since it is postulated that all of the t
s Core Power and Reactor Coolant System Transient Figures 15.1.5-2 through 15.1.5-4 show the RCS transient and core heat flux following a main steamline rupture (complete severance of a pipe) at initial no-load condition (case a),.
Offsite power is assumed available so that full reactor coolant flow exists.
The transient shown assumes an uncontrolled steam release from only one steam
. generator.
Should the core be, critical at near zero power when the rupture occurs the initiation of safety' injection by low steam line pressure will shut down the reactor.
Steam release from more than one steam generator will be prevented by automatic trip of the fast acting isolation valves in the steam-j lines by low steamline pressure signals, high-high containment pressure i
signals, or high negative steamline pressure rate signals.
Even with the i
failure of one valve, release is limited to no more than 10 seconds for the other steam generators while the one generator b!ows down.
The steam line i
stop valves are designed to be fully closed in less than 5 seconds from receipt of a closure signal.
As shown in Figure 15.1.5-3 the core attains criticality with the RCCAs inserted (with the design shutdown assuming one stuck RCCA) shortly after boron solution at ppm enters the RCS.
The continued addition of boron results in a peak cor power significantly lower than the nominal full power value.
1900 15.1-16 Rev. 14
\\
J
i e
CNS The calculation assumes the boric acid is mixed with, and diluted by, the
' water flowing in the RCS prior to entering the reactor core.
The concentra-tion after mixing depends upon the relative flow rates in the RCS, ' : tt:
4lu, Land in the Safety Injection System.
The variation of mass flow rate in_.
I the RCS due to water density changes is included in the calculation as is the variation of flow rate in the Safety Injection System due to changes in the
. RCS pressure.- The-Safety Injection System flow calculation includes the line losses in the system as well as the pump head curve.
Figures 15.1.5-5 through 15.1.5-7 show the salient parameters for case b, which corresponds to the case discussed above with additional-loss of offsite
~
. power at the time the safety injection signal is generated.. 'S: EWi; h *;;;';,c, @;t = t?:; t'n ' :' d : M : ::d t: :t:-t th; di n:1 ' :dd'- D %
i t 4 -- + a 1n carande +n e+m,+ +he e da+y 4nfar+4nn n,
in w
+k.
op 7-Criticality is achieved later and the core power increase.7.eis slower than in the similar case with offsite power available..The ability of the emptying i
steam generator.to extract heat from the RCS is reduced by the decreased flow in the RCS.
The peak power remains well below the nominal. full power value.
-j l
It should be noted that following a steamline break only one steam generator i
blows down completely. Thus, the remaining steam generators are still avail-1 able.for dissipation of decay heat after the initial transient is over.
In j
the case of loss of offsite power this heat is removed to the atmosphere via r
l the steamline safety valves.
Marcin to Critical Heat Flux A.DNB analysis was performed for both of these cases.
It was found that both cases hee.a minimum DNBR greater than the limit value.
hew a.
15.1.5.3 Environmental Consequences The postulated accidents involving release of steam from the secondary system do not result in a release of radioactivity unless there is leakage from the RCS to the secondary system in the steam generators. A conservative analysis of the potential offsite doses resulting from this accident is presented con-sidering equilibrium operation based upon a 1 gpm steam generator leak rate prior to the postulated accident.
Two postulated cases are analyzed:
l Case 1:
There is a pre-existing iodine spike.
Case 2:
There is a coincident iodine spike.
.The primary and secondary coolant activities correspond to limits set by Technical Specifications prior to the accident.
The following assumptions and parameters are used to calculate the activity release and offsite dose for a postulated steam line break l
l 15.1-17 Rev. 14 4
,i
1
)
~
1
)
1 1
i i
i l
I e
INSERT A e
THe Safety Inf,ection System delay time includes 10 seconds to start the diesel and 30 seconds to ' allow proper alignment.of valves and for charging pumps to reach full speed.
1
.l l
O
\\
i f'
1 1
)
i O
b 1
i
.]
CNS 1.
Prior to the' accident, an equilibrium activity of fission products exists in the primary and secondary systems caused by a primary-to-secondary leakage in steam generators..
2.
The. total primary-to-secondary leak rate for the duration ~ of the accident is 1.0 gpm, with.0.347 gpm (500 gal / day) in the defective steam generator and 0.653.gpm in the nondefective steam generators.
3.
Offsite power'is lost.
4.
The initial steam release for the defective steam generators terminates-in 30 minutes, the steam release from the nondefective steam generators continues for eight hours..
- 5. - All noble gases which leak to the secondary side are released via the steam release.
6.
The iodine partition factor in the defective steam generator is 1.
4 7.
The iodine partition factor in the nondefective steam generator is' 0.01.
l 8.
For case 1, the iodine concentrations are assumed to be the maximum permitted for full' power operation..
l 9.
For Case 2,*the iodine spike occurs at the onset of.the accident and-
-(
continues for the duration of the accident.
Spike concentrations are-(
determined by increasing the equilibrium appearance rate in'the coolant l
by a factor of 500.
-j l 10. Other assumptions are listea nn Table 15.1.5-2.
Based on the foregoing model, the thyroid and whole body doses are calculated at the exclusion area boundary and the low population zone. The results are
,. l presented in Table 15.1.5-2.
{
15.1-10 Rev. 14 p
.i t
~
TABLE 15.1.2-1 (Page 2) l Time Sequence of-Event for Incidents Which.
i Cause an Increase in Heat Removal By The Secondary System Accident Event Time (sec.1 Stabilized conditions 50 reached (approximate time only)
.4.
Automatic Reactor 10% step load increase' O. 0 Control (Maximum moderator feedback)
Stabilized conditions 50 reached (approximate time only)
Inadvertent opening of Inadvertent opening of 0.0 a steam generator relief one main steam safety or safety valve or relief valve
~
Pressurizer empties
-WG.
16 1.
l f;988 ppm boron reaches core 4i+ 118
[
Steam system piping failure i
1.
Case a Steam line ruptures 0
Pressurizer empty
-it II Criticality attained
-l+
16.5 l
0,000 ppm boron reaches core 46-70 1100 2.
Case b Steam line ruptures O
Pressurizer empty te= 13 i
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