ML17328A728
| ML17328A728 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/30/1989 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17328A716 | List: |
| References | |
| WCAP-11902-S01, WCAP-11902-S1, NUDOCS 9005250201 | |
| Download: ML17328A728 (50) | |
Text
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%CAP-11902 Supplement 1
'riESTiNGciCUSc. CLASS ii.-'ERATED POMER ANO REVISEO TEMPERATURE ANO PRESSURE OPERATION FOR DONALD C.
COOK NUCLEAR PLANT UNITS 1
& 2 LICENSING REPORT September 1989 WESTINGHOUSE ELECTRIC CORPORATION Energy Sys'tems Business Unit P.O.
BOX 355 Pittsburgh, Pennsylvania 15230 9 t 440:1(j/092189 c
soosa5oaoi poo514 PDR ADQCN, 05000316 P
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It TABLE S-2.1-1 COOK NUCLEAR PLANT UNITS 1
ANO 2 OESIGN POWER CAPABILITY PARAMETERS FOR RERATING PROGRAM Parameter (Unit 1, Original)
Case 1
(Unit 2, Current)
Case 2
- Power, MWt 3250 Core Power, MWt 3250 RCS Flow,(gpm/loop)~
88,500 Minimum Measured Flow, (total gpm)""
366,400 3423 3411 364,960 RCS Temperatures,
'F Core Outlet Vessel Outlet Core Average Vessel Average Vessel/Core Inlet Steam Generator Outlet Zero Load RCS Pressure, psia Steam Pressure,6psia Steam Flow, (10 lb/hr.tot.)
Feedwater Temperature,
'F
'A SG Tube Plugging Flow Oefinitions:
602.0 599,3 570.5 567.8 536.3 536.3 547.0 2250 758
- 14. 12 434.8 575.5 574.1
'547.0 2250 794'. 4 14.6 423.4 1PA avg./
15K -peak
- RCS Flow (Thermal Oesign Flow) - The conservatively low flow used for thermal/hydraulic design.
The design parameters listed above are based on this flow.
- ~Minimum Measured Flow - The flow specified in the Tech.
- Specs, which must be confirmed or exceeded by the flow measurements obtained during plant startup and is the.flow used in reactor core ONB analyses for plants applying the Improved Thermal Oesign Procedure,
"~Flow values supplied in FSAR6for Unit 2 are 141.3 x 10 lb/hr for vessel 6
coolant flow, and 134.9 x 10 lb/hr for active core flow.'ote:
Oashes in Case 2 indicate information which was not contained in the
- FSAR, and is therefore information which is unavailable to Westinghouse.
9144e:1d/092189 S-2.1-5
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TABLE S-2. 1-1 (Con t')
COOK NUCLEAR PLANT UNITS I ANO 2 OESIGN POWER CAPABILITY PARAMETERS FOR RERATING PROGRAM Parameter NSSS
- Power, MMt Core Power, MMt RCS Flow, (gpm/loop)"
Minimum Measured
- Flow, (total gpm} "
(Revised)
(Revised)
(Revised)
Case 3
Case 4
Case 5
3262 3425 3425
. 3250 3413 3413 88,500 88,500
, 88,500 366,400 366,400 366,400 (Revised)
Case 6
3425 3413 88,500 366,400 RCS Temperatures,
'F Core Outlet Vessel Outlet Core Average Vessel Average Vessel/Core Inlet Steam Generator Outlet Zero Load RCS Pressure, psia
~ Steam Pressure,6psia Steam Flow, (10 lb/hr.tot.)
Feedwater Temperature,
'F.
610.1 607.5 579.2 576.3 545.2 545.0 547.0 2250 or 2100 807
'14.20 434.8 583.6 614.0 613.9 580.7 611.2 611.2 549.7 581.8 581.8 547.0 578.7 578.7 513.3 546.2 546.2 513.1 546.0 546.0 547.0 547.0 547.0 2250 2250 2250 or or or 2100 2100 2100 603 820 806 14.98 15.07 15.06 442.0 442.0 442.0
/
SG Tube Plugging (average) 15 10 10 15 Flow Oefinitions:
"RCS Flow (Thermal Oesign Flow) - The conservatively low flow used for thermal/hydraulic design.
The design parameters listed above are based on thi s f1 ow.
""Minimum Measured Flow - The flow specified in the Tech.
Specs.
which must be confirmed or exceeded by the flow measurements obtained during plant startup and is the flow used in reactor core ONB analyses for plants applying the Improved Thermal Oesign Procedure, 9144e:1d/091BB9 S-2.1-6
J
~
s TABLE S-2.1-1 (Cont'd)
COOK NUCLEAR PLANT UNITS 1
AND 2 DESIGN POWER CAPABILITY PARAMETERS FOR RERATING PROGRAM Parameter (Revised)
(Revised)
(Revised)
(Revised)
Case 7
Case 8
Case 9
Case 10
- Power, MWt Core Power, MWt RCS Flow, (gpm/loop)"
Minimum Measured
- Flow, (total gpm)""
3600 3600 3600 3588 3588 3588 88,500 88,500 88,500 366,400 366,400 366,400 3600 3588 88,500 366,400 RCS Temperatures,
'F Core Outlet Vessel Outlet Core Average Vessel Average Vessel/Core Inlet Steam Generator Outlet Zero Load a
RCS Pressure, psia Steam Pr essure,6psia S team Fl ow, (10 1 b/hr. tot. )
Feedwater Temperature,
'F 585. 4 582.3 549.9 547.0 511.7 511.4 547.0 2250 or 2100 587 15.90 449.0 618.0 585.4 618.1 615.2 582.3 615.2 584.6 549.9
.584.7 581.3 547.0 581.3 547.3 511.7 547.4 547.1 511.4 547.2 547.0 547.0 547.0 2250 2250 2250 ol or or 2100 2100 2100 820 576 806 16.00 15.89 15.99 449.0 449.0 449.0 5
SG Tube Plugging (average) 10 10 15 15 Flow Definitions:
~
~RCS Flow (Thermal Design Flow) - The conservatively low flow used for thermal/hydraulic design.
The design parameters listed above are based on this flow.
Minimum Measured Flow - The flow specified in the Tech.
Specs.
which must be
. confirmed or exceeded by the flow measurements obtained during plant startup and is the flow used in reactor core DNB analyses for plants applying the Improved Thermal Design Procedure.
9144e:1d/091889 S-2.1-7
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S-3.3 NON-LOCA SAFETY EVALUATION S-3.3. 1 Introduction This section evaluates the effects of the Cook Nuclear Plant Rerat1ng Program on the non-LOCA transients;
-The non-LOCA safety evaluation provided within is applicable only for Unit 1, with the exception of the steamline break mass/energy releases (inside and outside containment).
The effort performed is to support Unit 1 operation with an uprated core power of 3413 MWt in the range of reactor vessel average temperatures between 547'F and 578.7'F at primary pressure values of 2100 psia or 2250 psia.
Table S-2. 1-1 (Cases 4
and 5) presents the range of conditions possible for the rerating of Unit 1.
The steamline break mass/energy release analyses are performed to support the potential future Unit 1 rerating as well as to bound a potential rerating of Unit 2.
Table S-2.1-1 (Cases 7 and 8) presents the range of conditions possible for the future rerating of Unit 2.
In addition, the evaluation performed is to support a maximum average steam generator.tube plugging level of 10K, with a peak steam generator tube plugging level of 15%.
The following non-LOCA safety evaluation also supports the change and/or relaxation of certain plant parameters to provide Unit 1 with increased I
operating margin and flexibility, Included in the non-LOCA safety evaluation are:
Increased Most Negative Moderator Temperature Coefficient (MTC)
(Tech Spec 3.1.1.4b)
Oegraded ECCS Charging Pump Flow (Tech Spec 4.5.2f)
Increased Main Steam)inc Isolation Valve (MSIV) Closure Time (Tech Spec 4.7.1.5b and Tech Spec Table 3.3-5 items 5h, 6h,
& 7c)
The evaluation conservatively assumes 0 ppm boron concentration in the 6oron Injection Tank (SIT).
9144e:1d/091889 S-3.3-1
The evaluation also supports a change to the steam generator water level program.
The existing level program is a ramp function from 33M narrow range span (NRS) to 44K NRS from 05 power to 20% power and a constant level at 44'A NRS between 20'A power and 100'A power.
The proposed steam generator water level program is a constant level at 44K NRS between 0/
power and 10'ower.
'i+
S-3.3.4.1 Steamline Break Hass/Energy Releases This section will discuss -the analyses of the steamline break event to determine the mass and energy releases inside containment and the superheated mass and energy releases outside containment for the Cook Rerating Program.
The analyses were performed to support the range of conditions possible for the rerating of Unit 1 as well as to position Unit 2 for a potential rerating.
The analyses also consider the relaxation of certain plant parameters (Section S-3.3-1).
914e:1d/091889 S-3.3-6
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2'
Steamline Hreak Mass/Energy Releases Inside Containment The current mass/energy 1 eleases for the inside containment analysis is based on work performed for Unit 2, which is applicable for Unit 1.
The calculation of the mass/energy release following a steamline break is described in the Cook Unit 2 FSAR Section
- 14. 1.5.
The steamline break mass/energy releases N
were recalculated to address the rerating of both Units and the relaxation of the plant parameters described in Section S-3.3. 1.
Steamline ruptures occurring inside a reactor containment structure may result in significant releases of high energy fluid to the containment environment, possibly resulting in high containment temperatures and pressures.
The quantitative nature of the releases following a steamline rupture is dependent upon the many possible configurations of the plant steam system and containment designs as well as the plant operating conditions and the size of the rupture.
These variations make it difficult to reasonably determine the single "worst case" for both containment pressure and temperature evaluations following a steambreak.
The FSAR'nalysis determined that the limiting scenario of the steambreak cases analyzed for the containment response evaluation were a break size of 0.942 ft occurring at 30'4 power for the 2
split rupture scenario and a break size of 4.6 ft occurring at full power 2
for the double-ended rupture scenario.
(The 30'A power split break case was slightly more limiting.)
However, it is difficult to conclude if these
- FSAP, cases remain bounding for the range of conditions possible for the reratings of both Units.
Adding to the difficulty in determining the effect of the rerating conditions are the plant parameters changes incorporated into the Cook Rerating Program.
The potential changes of certain plant parameters (i,e, relaxed most negative MTC limit, degraded ECCS performance, increased MSIV closure
- time, and 0 ppm BIT boron concentration requirement) are penalties in the calculation of mass/energy releases.
'It is not readily apparent as to the total impact of the combination of these changes.
As such, a series of steamline
- breaks, consistent with the cases presented in the FSAR, were analyzed to determine the containment response to a variety of postulated pipe breaks encompassing wide variations in plant operation, safety system performance, and break sizes.
91446:1d/091889 S-3.3-7
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Method of Anal sis The LOFTRAN computer code (Reference
- 2) was used to calculate the break flows and enthalpies of the release through the steambreak.
Hlowdown mass/energy releases determined using LOFTRAN include the effects of core power generation, main and'uxiliary'feedwater additions, engineered safeguards
- systems, reactor-coolant thick metal heat storage, and reverse steam generator heat transfer.
A bounding analysis was performed to address the range of conditions possible for the potential Unit 1 rerating and the potential Unit 2 rerating.
The assumptions on the initial conditions are taken to maximize the mass and total energy released.
The higher primary temperatures along with the higher uprated power level associated with the Unit 2 rerating parameters are conservative for the mass/energy release calculations.
The upper bound
'emperature of Table S-2. 1-1, Case 8 was used, Since the mass blowdown rate is dependent on steam pressure and the steam pressure is less for the lower bound temperature
- case, the steam pressure of the upper bound temperature case is limiting for the range of operating conditions possible for the reratings of Unit 1 and Unit 2.
The functions which actuate safety injection and steamline isolation during a
steamline rupture event are commonly referred to as the Steamline Hreak Protection System.
A plant's steamline break protection system design can have a large effect on steaml,ine break results.
The steamline break protection system designs for Unit 1 and Unit 2 are different.
Unit 1's design is referred to as an "OLO" steamline break protection system design.
Unit 2's design is referred to as a
"HYHRID" steamline break protection system design.
The two systems have the following characteristics:
9144e:1 d/091889 S-3.3-8
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- 1. -
pLD" Steamline Break Protectio~
Safety InJection Signals 1.
High"high steam flow coincident with low steamline pressure (two out of four lines) 2.
High-high steam flow coincident with low-low Tavg (two out of four lines) 3.
Two out of three differential pressure signals between a steam line and the remaining steam lines 4.
Two out of three low pressurizer pressure signals 5.
Two out of three hi containment pressure signals b
Steaml inc Isolation Signal s 1.
High-high steam flow coincident with low steamline pressure (two out of four lines) 2.
High-high steam flow coincident with low-low Tayg
( two out of four lines) 3.
Two out of four hi-hi containment pressure signals Unit 2 - "HYBRID" Steamline Break Protection
, Safety Injecti on S igna1 s 1.
Low steamline pressure (two out of four lines) 2.
Two out of three differential pressure signals between a steam line and the remaining steam lines 9144e:1d/091889 S-3.3"9
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C 3.
Two out of three low pressurizer pressure signals 4.
Two out of three hi containment pressure signals Steamline Isolation Signals 1.
Low steamline pressure (two out of four lines) 2.
High-high steam flow coincident with low-low Tavg
( two out of four lines) 3.
Two out of four hi-ni containment pressure signals The only differences between the Unit 1 and Unit 2 designs is the actuations from a high-high steam flow and low-low Tavg signal and the logic associated with the low steamline pressure signal required to actuate safety injection and steamline isolation.
For Unit 1, a high-high steam flow coincident with low-low Tavg signal actuates both safety injection and steamline isolation, For Unit 2, a high-high steam flow coincident with low-low Tavg signal actuates only steamline isolation.
- However, the difference is not significant for the calculation of the mass/energy releases since the analysis does not take credit for any ESF actuations on a high-high steam flow coincident with low-low Tavg signal.
Unit 1's design requires a coincidence between the low steamline pressure and high-high steam flow for protection actuation.
Unit 2's design only requires the low steamline pressure signal for protection actuation; no coincidence with steam flow is required.
The coincidence logic required for safety injection.initiation and steamline isolation on high-high steam flow and low steam pressure for Unit 1 is more limiting for the calculation of mass/energy releases inside containment than Unit 2's design.
Actuation of safety injection and steamline isolation wi 11 limit the mass/energy released to the containment.
Oelaying the safeguards initiation will result in a conservative calculation of the mass/energy 9144e:id/091889 S-3.3-10
C releases for the containment pressure and temperature evaluation.
The coincidence requirement for high-high steam flow with low steam pressure of the Unit 1 design increases the likelihood that safeguards initiation might be delayed compared to Unit 2's design where only a low steam pressure signal is required.
In the case where the coincidence logic prohibits safety injection and steamline isolation on high-high steam flow with low steam pressure, one of the other signals must be received before the safeguards are initiated.
As
- such, the Unit 1 steamline break protection system design was assumed in this bounding analysis for the calculation of the mass/energy releases inside containment.
Assumptions A series of steamline breaks were analyzed to determine the most severe break condition for the containment temperature and pressure response.
The following assumptions were used in the analysis:
a.
Oouble-ended pipe breaks were assumed to occur at the nozzle of one steam generator and also downstream of the flow restrictor.
Split ruptures were assumed to occur at the nozzle of one steam generator.
b.
The blowdown is assumed to be dry saturated steam.
c.
As discussed
- above, the Un.it 1 steamline break protection system design is assumed.
However, credit was not taken for safeguards actuation on high steam line differential pressure or high-high steam flow coincident with low-low Tavg.
d.
Steamliae isolation is assumed complete 11 seconds after the setpoint is reached for either high-high steam flow coincident with low steam pressure or hi-hi containment pressure.
The isolation time allows 8 seconds for valve closure plus 3 seconds. for electronic delays and signal processing.
The total delay time for steamline isolation of 11 seconds is assumed to support the relaxation of the main steam isolation valve (MSIV) closure time.
9144e: Id/092189 S-3.3-11
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4.6 ft2 and 1.4 ft2 double-ended pipe breaks were evaluated at
- 102, 70, 30, and zero percent power levels f.
Split pipe ruptures were evaluated at 0.86 ft, 102'A power; 2
2 0.908 ft, 7'ower; 0.942 ft, 371. power; and 0.4 ft, hot 2
shutdown.
These split break sizes for each power level were modeled because they reflect the largest breaks for which ESF actuations (i.e., steamline isolation, feedwater isolation, and safety injection) must be generated by high containment pressure trips.
The high"high steam flow coincident with low steam pressure is not reached for these break sizes or smaller break sizes.
(Reference 5) g.
Failure of a main steam isolation valve, failure of a feedwater isolation valve or main feed pump trip, and fai lure of auxiliary feedwater runout control were considered.
Two cases for each break
~ size and power level scenario were evaluated with one case modeling the HSIV failure and the other case modeling the AF'H runout control failure.
Each case assumed conservative main feedwater addition to bound the feedwater isolation valve or main feed pump trip fai lure.
h.
The auxiliary feedwater system is manually re-aligned by the operator after 10 minutes.
i.
A shutdown margin of 1.3'A 4k/k is assumed.
This assumption includes added conservatism with respect to the Unit 1 end-of-life shutdown margin requirement of 1.6X dk/k at no load, equilibrium xenon conditions, and the most reactive RCCA stuck in its fully withdrawn position.
The Unit 1 end-of-life shutdown margin requirement was used as the basis for this assumption since it is more limiting than the existing Unit 2 shutdown margin requirement.
j.
A moderator density coefficient of 0.54 hk/gm/cc is assumed to support the relaxation of the most negative moderator temperature coefficient limit.
91440:1 d/091889 S-3.3-12
k, Hinimum capability for injection of boric acid (2400 ppm) solution corresponding to the most restrictive single failure in the safety injection system.
The Emergency Core Cooling System (ECCS) consists of the,following systems:
- 1) the passive accumulators,
- 2) the low head safety injection (residual heat removal)
- system,
- 3) the high head (intermediate head)'afety-injection
- system, and 4) the charging safety injection system.
Only the charging safety injection system and the passive accumulators are modeled for the steam line break accident analysis.
The modeling of the safety injection system in LOFTRAN is described in Reference 2.
Figure 3.3-52 of NCAP-11902 presents the safety injection flow rates as a function of RCS pressure assumed in the analysis.
The flow corresponds to that delivered by one charging pump delivering its full flow to the cold legs.
The safety injection flows assumed in this analysis take into account the degradation of the ECCS charging pump performance.
No credit has been taken for any borated water that might exists in the injection lines, which must be swept from the lines downstream of the boron injection tank isolation valves prior to the delivery of boric acid to the reactor coolant loops.
For this analysis, a boron concentration of 0 ppm for the boron injection tank is assumed.
After the generation of the safety injection signal (appropriate delays for instrumentation, logic, and signal transport included),
the appropriate valves begin to operate and the safety injection charging pump starts.
Ln 27 seconds, the valves are assumed to be in their final position (YCT charging pump suction valve has closed following opening of RHST charging pump suction valve) and the pump is assumed to be at full speed and to draw suction from the RHST.
The volume containing the low concentration borated water is swept into the core before the 2400 ppm bor ated water reaches the core.
This delay, described
- above, is inherently included in the modeling.
1.
For the at-power cases, reactor trip is available by safety injection
- signal, overpower protection signal (high neutron flux reactor trip or OPLT reactor trip), and low pressurizer pressure reactor trip signal.
9144a:1d/091889 S-3.3-13
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m.
For reactor coolant pump (RCP) operation, offsite power is assumed available.
Continued oper ation of the reactor coolant pumps maximizes the energy.transferred from the reactor coolant system to the steam generators.
I n.
No steam generator tube plugging is assumed to maximize the heat transfer characteristics.
Single Failure Effects a.
Failure of a main steam isolation valve (HSIV) increases the volume of steam piping which is not isolated from the break, When all valves
- operate, the piping volume capable of blowing down is located between the steam generator and the first isolation valve.
If this valve fails, the volume between the break and the isolation valves in the other steamlines, including safety and relief valve headers and other connecting lines, will feed the break.
For the cases which modeled a
failure of a HSIV, the steamline volumes associated with Unit 2 were assumed since the volume available for blowdown for this scenario is greater than Unit 1.
For the cases which did not model a failure of a MSIV, the steamline volumes associated with Unit 1 were assumed since the volume available for blowdown for this scenario is greater than Unit 2.
g b.
Failure of a diesel generator would result in the loss of one containment safeguards train resulting in minimum heat removal capability.
,c.
Failure of a feedwater isolation valve would result in additional inventory in the feedwater line which would not be isolated from the steam generator The mass in this volume can flash into steam and exit through the break.
For consistency with the FSAR steamline break mass/energy release analysis, all cases conservatively assumed failure
. of the feedwater isolation valve, which resulted in the additional inventory available for release through the steambreak and in higher than normal main feedwater flows.
9144e:1d/091889 S-3.3-14
Fai]ure of the auxiliary feedwater runout control equiPment would result in higher auxiliary feedwater flows entering the steam generator prior to re-alignment of the AFM system.
For cases where the runout control operates proper ly, a bounding constant AFM flow of 670 gpm to the faulted steam generator was assumed.
This value was increased 'to 1325 gpm to simulate a'failure of the runout control.
Results The steamline break mass/energy releases inside containment were calculated to account for the r ange of condi tions possible for the potential reratings of Unit 1 and Unit 2 and for the relaxation of certain plant parameters.
One set of mass/energy releases were calculated to bound the reratings for both Units incorporating the limiting steamline break protection design of Unit 1.
The analysis assumptions support relaxation of the most negative moderator temperature coefficient limit, degradation of the charging pump performance of the Elhergency Core Cooling System, extension of the main steam isolation valve closure
- time, and relaxation of the minimum 8IT boron concentration requirement.
Section S-3.4.2.1 presents the containment integrity evaluation for a main steamline break using the mass/energy releases calculated here.
As discussed in Section S-3.4.2.1, the limiting scenarios of the steambreak cases analyzed for the containment response evaluation were a break 'size of 4.6 ft occurring at 102% power with a main steamline isolation fai lure for the double-ended rupture scenario and a break size of 0.86 ft occurring at 102/
power with an auxiliary feedwater runout protection failure for the split rupture scenario.
Table S-3.3-4 presents the mass/energy releases for these limiting steambreak cases of the containment response evaluation.
4
S-3.
3.6 REFERENCES
l.
Augustine,
- 0. B., and Cecchett,
- 0. L., "Reduced Temperature and Pressure Operation for Donald C.
Cook Nuclear Plant Unit 1 Licensing Report,"
MCAP-11902, October 1988.
2.
- Burnett, T, M.. T., et al.,
"LOFTRAN Code Description,"
MCAP-7907-A, April 1, 1984.
3.
Butler, J. C.,
and Love, 0.
S, "Steamline Break Mass/Energy Releases for Equipment Qualification Outside Containment,"
MCAP-10961, Rev.
1 (proprietary) and MCAP-11.184 (nonproprietary),
- October, 1985.
4.
Hollingsworth, S. 0.,
and Mood, 0. C.,
"Reactor Core Response To Excessive Secondary Steam Releases,"
MCAP-9227, January 1978.
5.
- Land, R. E.,
"Mass and Energy Releases Following a Steam Line Rupture;"
MCAP-8860, September 1976.
6.
"American Electric Power Service Corporation Donald C.
Cook Nuclear Plant Unit 1:
Safety Evaluation for Including Uncertainty Oue to Operator Readability of Pressurizer Pressure Instrumentation,"
AEP-89-216, Letter from J, C. Hoebel (M) to R.
B. Bennett (AEPSC),
September
- 1989, 9144e:1d/091889 S-3.3-26
TABLE S-3 3-4
.STEAMLINE BREAK MASS/ENERGY RELEASES INSIDE CONTAINMENT 102%
POMER DER (4.6 FT')
BREAK FAILURE - MSIV TIME
~SEC 0.00 0.20 3.60 6.60 12.80 13.00 13.20 13.40 13.60
- 14. 00 14.40 14.80 15.00 15.20 15.60 15.80 16.00 16.60 17.20 17.60 17.80 18.40 18.60
'8.80 19.20 23.80 28.80 30.40 36.. 40
- 39. 20
- 50. 70 57,20 106.20 109.20 111.20 118.20 125.20 136.20 602.70 MASS
~LBM/SEC 0.00 10430.00 6552.00 5612.00 4978.00 4913.00 4847.00 4781.00 4716.00 4587.00 4458.00 4332.00 4269.00 4206.00 4083.00 4022.00 3961.00 3782.00 3606.00 3492.00 3435.00 3268.00 3213.00 3158.00 3050.00 1876.00 1623.00 1575.00 1461.00 1431.00 1369.00 1356.00 1331.00 1331.00 1184.00 308.70 188.10 98.97 93.24 ENERGY BTU x 10'/SEC 0.0
- 1. 250 7.883 6.748 5.974 5.895 5.816 5.737 5.660 5.504 5.350 5.198 5.123 5.047 4.899 4.826
- 4. A3 4.538 4.328 4.190 4.122 3.921 3.856 3.790 3.660 2.251 1.421 1.883 1.746 1.708 1.634 1.618 1.588 1.587 1.409 0.358 0.217 0.114 0.107 9144e:1d/092189 S-3.3-31
TABLE S-3.3-4 (Cont'd)
STEAMLINE BREAK MASS/ENERGY RELEASES INSIDE CONTAINMENT 102%
POWER SPLIT (0.86 FT')
BREAK FAILURE - AUXILIARY FEEDMATER RUNOUT PROTECTION TIME
~SEC 0.00 0.20 1;60 2.00 2.40 2.80 4.20 4.40 8.60 9.40 12.00 12.60 15.80
'8.00 21.40 22.60 23.60 23.80 25.00 32.00 32.20 33.80 42.00 42.60 43.20 43.80 44.40 55.20 67.20 80.20 82.20 96.20 98.70 118.20 124.20 282.70 285.20 290.20 292.70 297.70 302.70 320.20 MASS
~LBM/SEC 0.00 1394.00 1366.00 1358.00 1350.00 1342.00 1316.00 1312.00 1550.00 1575.00 1632.00 1638.00 1635.00 1618.00 1458.00 1400.00 1357.00 1349.00 1302.00 1103.00 1098.00 1064.00 928.70 920.80 913.10 905.70 898.40 799.10 732.60 691.30 686.60 662.50 659.50 645.70 643.60 633.20 633.10 615.00 579.70 556.60 490.40 304.70 ENERGY BTU x 10'/SEC 0.0000 1.6690 1.6370 1.6270 1.6170 1.6080 1.5770 1.5730 1.8540 1.8840 1.9500 1.9570 1.9530 1.9340 1.7460 1.6790 1.'6280 1.6180 1.5630 1.3260 1.3210 1.2810 1.1180 1.1090 1.1000 1.0910 1.0820 0.9625 0.8823 0.8325 0.8269 0.7977 0.7941 0.7775 0.7749 0.7623 0.7622 0.7402 0.6977 0.6695 0.5896 0.3643 9144@:1d/092189 S-3.3-32
TABLE S-3.3-4 (Cont'd)
STEAMLINE BREAK.
MASS/ENERGY RELEASES INSIDE CONTAINMENT 1025 POWER SPLIT (0.86 FT')
BREAK FAILURE - AUXILIARY FEEDWATER RUNOUT PROTECTION TIME
~SEC 330.20 340.20 352.70 525.20 535.20 600.20 605.20
'ASS
~LBM/SEC 238.70 206.50 190.20 181.90 182.00 182.10 190.70 ENERGY BTU x 10'/SEC 0.2845 0.2456 0.2259 0.2160 0.2160 0.2162 0.2258 9144e:1d/0921S9 S-3.3-33
S-3.4.2.1 Main Steamline Break (MSLB) Containment Integrity Introduction and Background An evaluation was performed to determine the impact of reduced temperature and pressure operation on the Donald C.
Cook Nuclear Plant Unit 1 Long-Term Hain Steamline Break Containment Integrity analysis.
This evaluation is documented 9144e:1d/091889 S-3.4-1
f I
- t. 1 C'1
< ~
~'0 ii.'<
44'
in Section 3 4 2 of MCAP-11902 "REDUCED TEMPERATURE AND PRESSURE OPERATION FOR OONALO C.
COOK NUCLEAR PLANT UNIT 1 LICENSING REPORT,"
and it was concluded that reduced temperature and pressure operation did not have an adverse impact on the analysis results and conclusions.
This Section documents the analysis performed for both Oonald C.
Cook Nuclear Plant Units 1 4 2 to determine the-impact of the rerated conditions described in Section S-2.1 on Containment Integrity following a Hain Steamline Break.
A series of main steamline sp'lit and double-ended breaks were analyzed as a
part of the original licensing basis for Oonald C.
Cook Nuclear Plan. Unit 2
,to determine the most severe break condition for containment temperature and pr essure response for this design basis event.
The analysis and evaluation are discussed in Reference 1.
These results documented in the FSAR show that the most limiting double-ended break was the 4.6 square foot break, occurring at 102'f. power with main steam isolation valve failure.
The most limiting split break was the 0.942 square foot break, occurring at 30'A power with the failure of auxiliary feedwater runout protection.
The calculated peak temperatures for these cases were 319.1'F and 328.1'F respectively.
Additional generic sensi tivities discussed in Reference 2, illustrate that other smaller breaks were not limiting.
Purpose The purpose of the analysis documented in the following paragraphs is to demonstrate that the peak containment temperature resulting from a design basis main steamline break will not exceed the equipment qualification temperature criterion for Oonald C.
Cook Nuclear Plants Units 1 and 2, at the rerated conditions.
The containment pressure response generated for the LOCA Containment Integrity analysis for the double-ended pump suction RCS break case (Reference
- 3) bounds the Main Steamline Break containment pressure
- response, and therefore is not a concern here.
This analysis assumes reduced safety injection flow, due to degradation of ECCS performance, closure of the RHR crosstie valves and the current containment heat sink information.
91440: 1 I/091889 S-3.4-2
, Analytical Assumptions The analysis performed for the Rerating Program is. consistent with the Reference 1 analysis except for assumptions directly related to the rerating parameters.
The analytical effort provides bounding system calculations for both Units 1
& 2 at the rerated plant conditions-described in Section S-2. 1.
A spectrum of split breaks is analyzed at 0.86 ft,
102$ power; 0.908 ft,
2 2
70/ power; 0.942 ft, 30'A power and 0.4.ft, hot shutdown.
Double-ended 2
2 breaks of 1.4 ft and 4.6 ft are analyzed at power levels of 102/,
- 70K, 2
2 30Ã and zero power levels.
The break sizes analyzed in the present analysis are based on the current FSAR analysis.
As in the FSAR analysis, loss of one containment safeguards train was also assumed for all the cases in addition to the single failure assumed in the mass and energy release calculations.
4 h
The following cases were analyzed for containment response:
h A.
Split break cases 1) 0.86 ft, 102% power, 2
2) 0.86 ft, 102% power, 2
3) 0.908 ft, 70li power, 2
4) 0.908 ft, 70'A power, 2
5) 0.942 ft, 30'A power, 2
6) 0.942 ft, 3Ã power, 2
7) 0.40 ft, hot shutdown, 2
8) 0.40 ft, hot shutdown, 2
MSIV failure AFRP failure MSIV failure AFRP failure MSIV failure AFRP failure MSIV failure AFRP failure Note:
MSIV - Main Steam Isolation Valve AFRP - Auxiliary'eedwater Runout Protection 9144e:1d/091889 S-3.4-3
8, Oouble-ended ru ture cases~
1) 4.6 ft, 102% power, 2
2) 4.6 ft, 102K power, 2
3) 4.6 ft, 70'A power, 2
4) 4.6 ft, 70A power, 2
5) 4.6 ft,.
30'A power, 2
6) 4.6 ft, hot shutdown, 2
7) 1.4 ft, 102% power, 2
8) 1.4 ft, 102K power, 2
- 9) 1.4'ft, 70li power, 2
- 10) 1.4 ft, 30'A power, 2
- 11) 1.4 ft, hot shutdown, 2
MS IV fai 1 ure AFRP failure MSIV failure AFRP failure MSIV failure MSIV failure MSIV failure AFRP failure MSIV failure HSIV failure MSIV failure Note:
"The limiting 4.6 ft double-ended failure cases (1024 and 70'A 2
power), with HSIV failure were analyzed with AFRP failure and found to be less limiting than the cor~esponding HSIV failure cases.
Therefore only the most limiting 1.4 ft (102'A power) was analyzed with AFRP failure.
The mass and energy releases to the containment as a result of the postulated accident are calculated using the LOFTRAN computer code (Reference 4).
The mass and energy releases are calculated using two different failures for each case
- namely,
- 1) failure of the auxiliary feedwater runout protection and
- 2) failure of the main steam isolation valve.
As in Reference 1,
no credit is taken for entrainment.
Section S-3.3.4.1 presents additional details regarding the calculation of the inside containment steamline break mass and energy releases.
The LOTIC-III computer code (Reference
- 5) is used to calculate the consequence of these releases, in particular the peak containment temperature.
The main steam line break containment integrity calculations are performed with an additional fai lure of one of the containment safeguards
- trains, which results in minimum spray flow (this includes a
1PA degradation in the spray pump flow).
Where applicable, input data consistent with that of the LOCA containment integrity analysis (Reference
- 3) is used.
9144e:1d/091889 S-3.4-4
6 The total initial ice mass assumed is 2.11 x 10 lbs.
The initial conditions in the containment are a temperature of 120'F in the lower and dead ended compartments, a temperature of 27'F in the ice condenser, and a temperature of 57'F in the upper compartment.
All volumes are at a
pressure of 0.3 psig and a relative humidity of 15'A, The refueling water storage tank (RHST) temperature is assumed to be 100'F.
A spray pump flow of 1900 gpm to the upper compartment and 900 gpm to the
lower compartment is assumed, at a temperature of 100'F.
The spray flow is initiated 45.0 seconds after the containment reaches the hi-hi pressure signal of 3.5 psig.
This setpoint includes instrument uncertainties.
Results The results of the analysis show that the maximum calculated containment temperature is 324.9'F for the 4.6 ft double ended rupture at 102% of the full power.
The mass and energy calculations for this case are based on the main steam isolation valve failure.
The maximum containment temperature calculated for the limiting small split break (0.86 ft at 102% of full power) is 324.4'F.
The auxiliary feedwater runout protection fai lure is assumed for this case.
Table S-3.4-1 and Figures S-3.4-1 through S-3.4-4 show the results for the two limiting cases.
Comparison of these results to the current FSAR results with respect to the peak containment temperature indicates that the FSAR result was more limiting.
This is due to the lower mass and energy releases inside containment, calculated for the present analysis.
The peak temperature shown in the FSAR for the limiting spl'it break case (0.86 ft at 102% of full 2
power, with auxiliary feedwater runout protection failure) is higher than the 91@le:1d/091889 S-3.4-5
I E
l J
M,4I p*
present case.
- However, the FSAR results for the limiting double-ended rupture case (4.6 ft at 102K power', with main steam isolation valve failure) is lower than the present double-ended results.
A detailed study of the results shows that even though the mass and energy releases within containment are lower in both the present
- cases, the double-ended break results in a higher temperature due. to reduced flows from the lower compartment into the ice-condenser.
The peak occurs very early in the transient (within the first ten seconds).
At this early time the only heat removal systems that exist are the containment wall heat sinks and the heat flow between the compartments.
In the present
- case, heat removal by the walls is better (due to more detailed modeling of the walls), but the heat flow from the lower compartment into the ice-condenser is'ower (due to the lower initial temperature assumed in the ice-condenser
'and the upper compartment, which affects the driving force through the ice-condenser).
Conclusions I
The main steamline break containment integrity analysis has been performed consistent with the current licensing basis analysis and Oonald C.
Cook Nuclear Plant Units 1
& 2 rerating program, considering the present plant operating conditions.
The results of this analysis are bounded by the current FSAR results.
This analysis therefore demonstrates that the containment heat removal systems function to rapidly reduce the containment pressure and temperature in the event of a main steamline break accident.
S-3.4.3 References 1.
Mestinghouse letter 0 NS-THA-1946, 9/20/78,
" American Electric Power Projects Donald C.
Cook Unit 2 (Docket 50-316)
Response
to Question 022.9".
2.
Mestinghouse letter fAEP-80-525, 3/10/80, "Response to NRC Question 022.17 - AMP's steamline break analysis".
3.
MCAP-11908," Containment Integrity Analysis for Donald C.
Cook Nuclear Plant Units 1 and 2", July 1988.
4.
MCAP-7907-P-A (Proprietary),
"LOFTRAN Code Description", April 1984.
5.
MCAP-8354-P-A (Proprietary),
Supplement 2,
"Long Term Ice Condenser Containment Code - LOTIC-3 Code",
February 1979.
9144e:1d/091889 S-3.4-7
TABLE S-3.4-1 MAIN STEAMLINE BREAKS Type of Break Break Size (FT
)
2 Oouble-Ended Rupture 4.6 Split Break 0.86 Type of Failure MSIV AFRP max
(
Time of Tmax (sec)
Px (psig)
Time of P
(sec) 324.9
- 6. 39 8.62 14.01 324.4 50.72 7.24 50.72 Note:
MSIV - Main Steam Isolation Valve AFRP - Auxiliary Feedwater Runout Protection 9144e:1d/092189 S-3.4-8
P t) 4 ~
neer Coeyer teen 240.
200m 180.
140.
120'pper COSPi I'SOflt 100.
o.
100.
1SO.
" ZOO.
ZSO.
S00.
SSO.
~.
CSO.
SOO.
SSO.
~C TINE (SEC)
CONPARTNENT TENPERATURE Figure S-3.4 4.6 ft Double-Ended
- Rupture, 102'A Power, MSIV Failure 2
S-3.4-9
t' f~
II
9.0
(,oaaea'mpl t~C 8.5 Coepertaent
$.0 T.S
~4 7.0 I@a CC an an Cg 6.5 CL b.0 5.5 5.0 4.5 4,0 0,
50,
- 100,
'150, 200.
250.
MO ~
350.
400'50
'IVE (SEC)
CONPARTMEHT PRESSURE F
~
2 igure S-3.4 4.6 ft Double-Ended
- Rupture, 102K Power, MSIV failure 914@a:1d/081689 S"3.4-10
a I I
250.
4 Cl UJ 200 0 I
175.
150.
125.
100.
wUgsl'~rraent 50.
0.
100.
500.
600.
700.
TINE (SEC)
I CONPARTNENT TENPERATURE Figure S-3.4-3 " 0.86 ft Split Break, 102K Power, AFRP Failure 9144e:1d/081889 S-3.4-11
C lh 1
Uppar CclparcROAC 7.
~i~r Coaparcecn~
d.
UJ ac S.
0.
100.
300.
400.
$00.
700.
TINE (SEC)
CONPARTNENT PRESSURE Figure S-3.4 0.86 ft Split Break, 102% Power, AFRP Failure 9144a:14/OB16B9 S-3.4-12
WESTINGHOUSE CLASS II
%CAP-11902 REDUCED TEMPERATURE AND PRESSURE OPERATION FOR DONALD C.
COOK NUCLEAR PLANT UNIT 1 LICENSING REPORT D. L. Cecchett D. B. Augustine October 1988 WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Business Unit P.O.
Box 355 Pittsburgh, PennsyIvania 15230 7980e:id/100588
2A Figure 3.3-52 Safety In)ection Fl~ Supplied by One Charging Puap 3.3-135