View source for B13242, Forwards App 2C of Probabilistic Safety Study,Updated Through Rev 2.Topics Discussed include,thermal-hydraulic Bases for Success Criteria,Loss of Normal Power Transient & Loss of Feedwater Transient

{{Adams
| number = ML20247F562
| issue date = 05/19/1989
| title = Forwards App 2C of Probabilistic Safety Study,Updated Through Rev 2.Topics Discussed include,thermal-hydraulic Bases for Success Criteria,Loss of Normal Power Transient & Loss of Feedwater Transient
| author name = Mroczka E
| author affiliation = NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000245
| license number = 
| contact person = 
| document report number = B13242, NUDOCS 8905300106
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 80
}}

=Text=
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1 m UTILITIES                                               cer.erei civice seiden street. Beriin. connecticut v       c HARTFORD, CONNECTICUT 06141-027o L    t T J '%Z' O",""O''Z",",
(203) ess-sooo May 19, 1989 Docket No. 50-245 B13242 Re:     Integrated Safety Assessment Program U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:
Millstone Nuclear Power Station, Unit No.1 Probabilistic Safety Study In a {{letter dated|date=July 10, 1985|text=letter dated July 10, 1985}},II) Northeast Nuclear Energy Company (NNECO) submit-ted. the Millstone Unit No.1 Probabilistic Safety Study (PSS).                        Subsequent Revi-sions 1 and-2 were sent to th tals dated February 11,1987,guclear           Regulatory and February      Commi(spion 10, 1989,                    (Nr.C)' St:ff in submit-res'gectively.
NNEC0'has been requested by the NRC Staff to resubmit Appendix 2C, updated through Revision 2. Enclosed herein, please find toe requested document. Revision material for Appendix 2C previously sent to the NRC may be discarded, as the amended Appendix 2C is attached.
        'Should you require additional information, please contact us.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY                             ,
At% ku E. J/ Mr'oczkqy Senior Vice President
         .cc:     W. T. Russell, Region I Administrator M. L. Boyle, NRC Project Manager, Millstone Unit No. 1
: W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 (1)   J. F. Opeka letter to J. A. Zwolinski, " Millstone Unit No.1 Probabilis-tic Safety Study--Results and Summary Report," dated July 10, 1985.
(2)   E. J. Mroczka letter to the U.S. Nuclear Regulatory Commission, "Probabi-listic Safety Study Update," dated February 11, 1987.
(3)   E. J. Mroczka letter to the U.S. Nuclear Regulatory Commission, "Probabi-listic Safety Study Update," dated February 10, 1989.                                                  p(
8905300106 890519 PDR      ADOCK 05000245
                                                                                                                       ' l\)\
i L_____ _ __ _ _ __ f PEGC

Docket No. 50-245 B13242 L
i l
Attachment 1 Millstone Nuclear Power Station, Unit No.1 Probabilistic Safety Study                           'l Appendix 2-C i
I I
l May 1989

APPDDIX 2-C 'DIERNAL HYDRAULIC BASES FOR SUCCESS CRITDtIA
  .p        .               This appendix doctznents the analysis which formed the bases for the success criteria used in the study.       The results are obtained from five categories of
                                                    ~
analysis. These four categories are:
o    Detailed plant specific transient analysis performed by General Electric for Millstone 1 . 1139 S g the SAFE and CHASTE computer codes.
                          'o    Simplified boil-off and clad heat calculations.
o    Generic ' analysis performed for BWR/3s by General Electric for the BWR Owners Group.                                                                                  l 1
o    Generic analysis performed by the National Laboratories under contract with the NRC.
o- Best Estimate Analysis performed using the MAAP and COMPACT computer codes.
The following sections st1rdnarize the methods used in the analyses, and the V                        results.
i
                                                                                                                               )
m (G
:                                                                 MILLSTONE UNIT 1 i-                                                                PROBABILISTIC SAFETY STUDY

       .y
  -- Q Table of' Contents Page 2C.1   Method and Assumptions                                     2C-3 20.2   Loss of Normal Power Transient-                            2C            20 3   Station AC Blackout with Failure of IC Transient           20-10 2C.4   Loss of Feedwater Transient                                2C-18
(^)     2C.5   Stuck Open Safety / Relief Valve Transient                 2C-21
   %J 2C.6  -Isolation Condenser Related Times and Flow Rate            2C-30 2C.7  .Long Term Cooling Success Criteria                         2C-31 2C.8   Large Break LOCA Analysis                                  2C-35 2C.9   3 mall Break LOCA Analysis                                 2C-39 2C.10  References                                                 2C-46 1
2C-2 MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

2C.1'              Method'and Asnauption p                                        The success criteria used in this study are based on the results of Nj                                         the best ' estimate - (i.e. realistic) thermal hydraulic analyses.              In addition to . computer analysis with Millstone Unit 1 specific input (Reference -1), a number of simplified boiluoff and clad heat-up calculations (hand calculations) were performed.
The computer analysis utilized G.E.'s approved licensing evaluation
                           ~
models (SAFE and CHASTE). The SAFE code is used to track the vessel inwntory and to model ECCS performance. The CHASTE code is used to-calculate the fuel cladding heat up rate, peak cladding temperature and cladding oxidation. To obtain realistic results, aaveral inputs were changed from the values G.E. normally uses for licensing analyses. The following are the most significant realistic input changes:
o    Best Estimate Core Decay Heat Source
      -( m                                           The analysis utilized 1979 ANS decay heat curve with a multiplier U                                             of x 1.0.
o    Realistic Treatment of CCFL The SAFE code provides a good overall prediction of the system mass and energy balance. While the effects of Counter Current Flow Limitation (CCFL) are not modeled, the overall results remain conservative. CCFL at the top of the core tends to limit the downflow of LCC3 water delivered to the core from above.
CCFL at the bottom of the core, however, tends to limit the rate at which water drains from the core and facilitates refilling the core with water even before the lower plenum refills. While the combined effects are not equivalent, the lack of CCFL modeling in SAFE tends to produce a conservative estimate of core inventory.              l In addition, integral system tests demonstrate that CCFL at the top of the core breaks down rapidly during core spray injection G
(V AMENDMENT 2                                        2C-3 MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

[5 i
resulting in. r:pid draining of water from above the' core, producing rapid core reflooding.      Blerefore, neglecting CCFL.
     ;- .                 effects does not cause any non-conservatism in the prediction of         l overall system response.
O     Steam Cooling 1
The analysis took credit for steam cooling in the region of fuel above the two phase level (i.e. uncovered portion of the core).
: o. Measured System Performance Data Whenever available, the analysis used the measured system performance data instead of design minimum values.
Boil-off and heat-up calculations using the MAAP code utilized the          l 1979 ANS decay heat curve with a multiplier of x          1.0. The ' core boiloff rates for various scenarios where calculated by assuming the initial RPV inventory at saturated conditions at 1050 psia.
Initial RPV inventory at 100% inventory at 100% power is 4.06x10 5
                  . lbs, which corresponds to a level of 512.5" from the bottom of the RPV. Table 2C-1 stanmarizes water volumes for various elevations in the RPV.
I
    .r3()
em2                               M MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY i
                                                                                                 .l

Table-2C    sf q -                                  ~ RPV WATER VOLUMES VS. ELEVATIONS
    #   t                                                                                                                                             >
V-                               4 Set Points        Elevations     Water Volumes              Notes High level FW -        530 1/2"        Not Needed for Trib Set Point-                        Calculation Normal Level    -
512-1/2"        Not Needed for       Initial Inventory =
5 Calculations         4.06 x 10 Volume of Initial Inventory = 8855 ft 3                                                   .]
(Sat. Water 81050 psig)
Low Level -            482-1/16"       8420 ft3 Setpoint                                                                                                                             l
                                                                                                                                                      \
3
([             Low Level -            435-1/16"       7200 ft Setpoint Top of Active -        355-1/16"       6200 ft 3 Fuel                                                                                                                          l Bottom of'      -      211-1/16"       3210 ft 3 Active Fuel                                                                                                                          !
(
3 Ik>ttom of RPV -       0"              0 Ft 0
2C-5                                                                                        I MILLSTONE UNIT 1 PROBABILISTIC SAFEIT STUDY
-_--___-___-__:-_--_-_---.           ._    __                                                                                                         1

2C.2               Loss cf Normal Power Transient Following an LNP event, the reactor scams and the feedwater and CRD
  -- Q >
;  V                            pumps                  trip.             At ' this pointi all . injection' sources capable of maintaining the RPV inventory are lost. -                                               The imediate operator actions are to close the MSIVs and initiate the'IC (see ONP 503 B, Appendix 2B). In this analysis it is assumed tnat the operator closes the MSIVs. However, since ~the purpose of this analysis is to t
determine the mimimum time available for the operator. to restore RPV level, it is arbitrally assumed that the operator's immediate action of initiating                            the IC   fails. The RPV pressure rises to the Safety / Relief . the valve setpoint (1095 psig) and valve with lowest setpoint begins to cycle (See Section 2.4.1 for number of valves which will cycle). The RPV inventory depletes and the RPV level begins to drop. Table 2C-2 lists some of the pertinent data for this analysis.
E0P 571 (Appendix 2B) instructs the operator to depressurize the RPV to 900 psig- by manually opening S/R valves.                                                                                        The purpose of depressurization is to stop S/R valves from cycling and thus prevent
[q                           possible valve failure. The RPV pressure is maintained at 900 psig by
  ~V                            manually keeping S/R valve (s) open. As the RPV is depressurized, RPV inventory flashes. - As listed in Table 2C-2, total' loss of inventory 4
due to flashing is estimated to be 1.45 x 10 1bsm.
At ten minutes following the LNP, the RPV level drops to the low-low                                                                                            [
setpoint.                           The RPV inventory at that time has been depleted from an initial value of 4.06 x 10 5
lbs. to 3 39 x 10 5 lbs. If no additional operator action were taken and if FWCI and IC failed to initiate automatically on low-low level, the boil-off would continue and level would drop to the top of the active fuel (TAF) in about                                                                                               l 4
twenty-two minutes (corresponding RPV inventory lost is 9.95 x 10 lbs).
For this analysis, it is assumed that at ten minutes when the level is at the low-low level setpoint, the operator takes action to restore                                                                                             l    ,
RPV water level. The operator has at least 3 methods of restoring RPV s
NM 2                                                                               MILLSTONE UNIT 1 PROBABILISTIC SAFETl STUDY 4
                                                                                                    - _ _ _ _ . _ . _     _.-______.__._._______m_                          _ _ _ . _ . - _ _

f level and any one of the ' three methods is sufficicnt' as discussed below:
fG                            .
                      . Level Restored by FWCI o
The operator starts 1 Feedwater train. Since FWCI can inject                                                                   l water at full system pressure, no depressurization is needed.
FWCI will begin to increase RPV level, thus preventing core uncovery, o     Level Stablized and Recovered by IC:
The operator places the IC in service. The IC condenses the core boil-off and returns the condensate to the RPV. Since the IC can condense more steam than generated due to core boil-off at ten The S/R valves, minutes, the RPV will begin to depressurize.
which were manually opened to maintain RPV pressure at 900 psig are closed. Even though the RPV is not losing any inventory, the water level decreases due to increases in density of the remaining inventory as RPV pressure drops.      It is estimated that h
RPV pressure drops to 200 psig in about 1 hour, and RPV water level remains a few inches above the top of the active fuel                                                                    l (TAF). For this analysis, it is assumed that the operator restores RPV level within one hour by starting a pump (for example, a CRD pump) and thus prevents core uncovery.
o     Level Restored by Low Pressure Paes and The operator starts one low pressure pump- (LPCI or CS)
As               l manually opens 2 S/R valves to rapidly depressurize the RPV.
the pressure falls, the RPV inventory deceases due to flashing and core boil-off. The RPV collapsed level not only decreases due to a decrease in RPV inventory, but also due to the increase                                                               l in density of the remaining inventory. When the RPV pressure falls to 235 psig, the operating low pressure pump begins to inject. The pump injection rate increases as the RPV pressure 2C-7 AMENDMENT 2                                           HILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

w-I' drops. It 19 cstimated th:t ct 200 psig, th2 pump flow rata exceeds crN boil-off and flashing rate and therefore the RPV level begins to increase. The minimum RPV level achieved during the depressurization is estimated to drop below TAF. However, due to steam cooling, significant cladding damage is not expected.
1 1                The MAAP. Code (version 3 0) has been utilized to evaluate the minimm l
l       -        intervention time for each of the following operator actions following a Loss of Feedwater (bounds LNP case) which are considered to restore the RPV water level.
l
: 1. The operator recovers the feedwater within 25 minutes after the accident initiation time.
The operator manually opens the S/R valves within 20 minutes to               .I 2.
rapidly depressurize the RPV and initiate the low pressure coolant injection pumps (LPCI pumps) or the low pressure core spray pumps (LPCS).
h
(_/            A base case was conducted in which neither the RHR, nor any operator action to restore the RPV level was credited. Note that in all the cases of Loss of Feedwater which are considered in this section, no credit is given to the CRD flow. Results of various key events for the cases studied are summarized in Table 2C-2.2. Note that LPCI Flow started after about 21.7 minutes for the case where the operator manually opened the S/R valves within 20 minutes after the accident initiation.
1 I
2C-8 AMEDME T 2                                         MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

                                         .,                                                 Table 2C-2.1 O(                                                                     Pertinent Values for an LNP Event Description                             Value Initial RPV Inventory                             4.06 x 105 2bs.
Water Volume below Low-Low level Setpoint         7200 ft 3 Water Volume below TAF                            6200 ft 3 4
Inventory lost due to flashing in                 1.45 x 10 lbs.
depressurizing RPV to 900 psig Mass of water in RPV when collapsed level         3 39 x 105 lbs.
is at Low-Low level Setpoint (saturated water at 900 psig) 4 Inventory available for core boil-off             5 24 x 10 lbs.
above Low-Low level Setpoint Time at which collapsed level drops to            10 minutes
[q                                                 Low-Low level Setpoint L)
Mass of Water in RPV when collapsed               2 92 x 105 lbs.
level is at TAF (saturated water at 900 psig) 4 Inventory available for core boil-off             9 95 x 10 lbs.
above TAF Time at which collapsed level drops to            22 minutes TAF m
k.
AMENDMENT 2                                2C-9 MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

Table 2C-2.2 Summary of Key Events for the cases Studied j'    '
[ Accident Initiator: Loss of FW at t:03 l
Timing of Key Events                  i Type of                Operator        Core       Core     Fuel     Vessel
                          . Operator Action           Intervention     Uncovery Recovery Damage       Failure-  i No Operator Action 8 No IC                                      ~21                ~1.1      ~2.05
* No RHR                                   minutes              hours     hours
* NO CRD Flow FW recovered
* No IC                  FW recovered        ~21       ~25.6 8 No CRD Flow            within 25 min. minutes      minutes from accident
      /                                               initiation Open S/R Valves
* No IC                  S/R valves          ~21      ~22.2
* No CRD Flow            manually          minutes   minutes opened within 20 minutes from accident initiation I
i AMENDMENT 2                                 2C-9a MILLSTONE UNIT 1 I
PROBABILISTIC SAFETY STUDY

m l
u, h           2C.3    Station AC R1mnkw t With Fdilura of IC Transient L
   ,n The.' Station AC Blackout transient - is similar to the LNP transient
(,l              described in Section 2C.2. Following the blackout, the operator will close. the MSIVs and try to initiate the IC (See ONP 503C, Appendix                                        l 2B). Since the purpose of this analysis is to determine the minimum time avaiable for the operator to restore AC power and RPV level to                                              '
prevent core melt,     it is arbitrarily assumed that the IC is                                           l unavailable. The RPV pressure increases to the S/R valve set point and the S/R valves begin to cycle.
The only pump independent of station AC power is the diesel-driven fire pump (see Section 3 2.15). Use of this pump for RPV injection requires running fire hoses to a FW heater drain line in the Turbine                                              !
Building. It is estimated that it will take an operator about an hour                                      l      l to line-up the fire pump for injection to the RPV. Therefore, unless                                              j AC is restored earlier, all injection sources are lost for at least one hour. Since the IC has been assumed to be - failed, the RPV inventory will continue to be depleted and the RPV level will fall.
In a station-blackout scenario, since the operator. is aware of loss of                                   l all injection resources, it is likely that the operator will not depressurize the reactor to 900 psig to stop the S/R values from                                          l cycling. By not depressurizing the RPV, the operator prevents loss of inventory associated with flashing. This inventory becomes available for decay heat removal (i.e. , core boil-off) and thus delays core                                                ]
uncovery.                                                                                                         1 Table 2C-3 lists some of the pertinent values used in this analysis.
As the boil-off continues, the core will begin to uncover at . about                                              >
twenty-five minutes. As the RPV level drops to fifty feet above the                                               ;
bottom of active fuel (BAF), the peak cladding temperature (PCT) will be about 1700 F. In this analysis, it is conservatively assured that any further drop in RPV level will result in the onset of core                                                i melt.
l O                                                                                                                                 I AMENDMENT 2                             2C-10                                                                               '
MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

Th] RPV wat3r 1;v31 will drop to 50" tbove. BAF in about forty-eight minutcs, assuming that st am being g:ncratsd in ths lower portion of the core does not entrain any water droplets.       Therefore, the steam                 )
        /")             superheats as it rises through the uncovered fuel. However, if it is                     l assumed that the steam does entrain water droplets which boil in the                  l I upper portion of the core, the RPV water level is calculated to drop to 50" above BAF in about forty-three minutes. (For the later calculation, enough droplets were assumed to be entrained to keep steam saturated in the core). This analysis assumed the time when the
                -       RPV level drops to 50" above BAF to be forty-five minutes (an average value).
Justifications of PCT = 1700 When Level is at 50" + BAF Figures 2C-1 through 2C-4 provide the transient results of an isolatist event. Specifically,the scenario is as follows:
i I
o     Reactor trips from 100% power, FW flow drops to zero and o     IC Fails to initiate either autceatically or manually.                            j o     1 CRD pump (77 gpm) continues to run.
I             o    RPV Jater level drops to 50" above BAF in about 1 hour.
O o    A . a hour, the operator depressurizes the RPV by manually opening 2 S/R valves and recovers the level by injecting with one LPCI pump.
1 The PCT for this scenario is calculated to be 1520 F (Fig 2C-4).
The transient analysis results presented in Figures 2C-1 through 2C-4 can be used in estimating the PCT for the station blackout scent >rio under consideration. This is because the PCT in a transient is determined by three variables, which are:
l o    Maximum Core Uncovery (i.e. Minim]m level) and l                       o    Duration of Core Uncovery and o    Time after reactor trip of MAXIMUM core uncovery.
Q)
AMENDMENT 2 2C 11 MILISTONE UNIT 1 PROBABILISTIC SAFETY STUDY

L.
                    -In th3 two cas:s - being compared, thm minimum 1sval cchieved ' is the same (50" + BAF). Also, the duration of core uncovery is comparable, in both cases the core will .be- completely recovered in about forty h-
   ' (/              minutes 'arter uncovery starts.
l                    The time of maximum core uncovery is one hour in Figure 2C-3 versus 45 minutes   in the Station Blackout case.                                                         The higher decay heat associated uith earlier time of minimum level results in higher PCT.
Such an increase in the PCT can be estimated by taking ratios of decay heat as shown below:
PCT (@ 45 min level)= Decay Heat Fraction at 45 min
* PCT for 1 hr Decay Heat Fraction at 1 hour
                                            =   1.0% X 1520
                                            =   1700 F D)
G.
O-AMENDMENT 2 2C-12 MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

1, L                                                                                                   i l                                                                                                 -1
(                                                                                                   l Table 2C-3 i
PERTINENT VALUES FOR A STATION BLACK 0lrf TRANSIErr l
Initial RPV Inventory                         =     4.06 X 105 lbsm.
!                         Water Volume below TAF                        =     6200 ft 3 Water Volume below 50" + BAF                  =     4903 ft 3 Mass of wate~ in RFV when collapsed level is at TAF 5
(at 1100 psia saturated)                      =     2.92 X 10 1bm.
Inventory Available for boil-off before RPV level drops to TAF                 =     1.14 X 10 51bm.
Time at which collapsed level will drop to TAF                                   =     25 minutes Mass of Water in RPV when collapsed level' drops to BAF + 50"                                     5
                                                                        =     2.33 X 10 1bs Inventory available for boil-off when RPV level' drops to BAF + 50             =      1.73 X 1051bs" Assuming no entrainment of water droplets in steam rising through uncovered core, the RPV water level drops to 50" + BAF                            =     48 minutes Assuming rising steam through the core remains saturated due to entrained water droplets, the core level drops to 50" + BAF                                  =     43 minutes Average time when the core level drops to 50" + BAF                            =     45 minutes
(               -
FC-13 MILLSTONE UNIT 1 PROBABILISTIC SAFEIT bTUDY

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(
i 2C.4      Loss of Feedwater Transient rs                      The Loss of Feedwater (LOF) transient is similar to the Loss of Normal i
V                       Power (LNP) transient described in the Section 2C.2, except that the reactor trip signal is generated on low RPV water level rather than at normal RPV level as in the LNP transient.                                    i 1
MAAP (Modular Accident Analysis Program) Code (BWR version 3 0) has
                ,          been utilized to simulate the loss of Feedwater scenario. Operator intervention to recover the reactor coolant system (RCS) inventory has been credited in this study.
The analysis of this scenario, using the MAAP Code, showed that RPV level recovery could be achieved by either one of the following two manual operator actions:
1.) Recovery of feedwater (FW) after 25 minutes from the accident initiation time (t=0).
or*
n                              2.) Manually opening the safety relief valve (SR/V) after U                                   twenty minutes to depressurize the primary system and initiate either the low pressure coolant injection system (LPCI) or the or the low pressure core spray system (LPCS).
l                          The results of both scenarios are summarized in Tables 2.C.4.1 and 2.C.4.2.
Figuras 2.C.4.1 through 2.C.4.14 show the trends of key parameters during these scenarios.
Note:
l                         A parameter study on the recovery action time showed that recovering 1
Feedwater after fifty minutes from the initiation time has resulted in a maximum core temperature of about 2331 F (after 50 minutes) and that the reactor core was uncovered for about 30.5 minutes.
b) v 2C-18 NDM 2                                                   MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

w                                              ,
Alttrnativaly, whtn thD SRV's wers opened aft 2r 45 minut:s from ths initiation time, the maxima core temperature was about 1684 0 F
(-<                   (after 45 minutes) and the reactor core was uncovered for about 26.5 i                 minutes.
The followina assumptions are made during the loss of Feedwater Scenario:
       +
1.) Feedwater is lost at t=0.
  +
2.) Isolation condenser is inoperable.
1 3.) No CRD flow.
4.) No ADS.
    '\
C AMENDMENT 2                               2C-19 MILBTONE UNIT 1 FROBABILISTIC SAFETY STUDY E____________________         _ _ _ _ _ _ _ _ _

q l
Table 2C-4.1
(~J                     W_w4 of Events for loss of Feedwater Transient                                                            l v                                                                                                                                  ,
Time 0.0 secs.                          Trip of all FW pumps initiated                                            .,
                                                                                                                                    ^
5.0 secs.                          Feedwater flow drops to zero.
14.0 secs.                          RPV water level drops to Low Level Setpoint 38.0 secs.                          Reactor Trip signal is generated.
RPV water level drops to Low-Low Level Setpoint HSIVs begin to close.
Recirculation pumps are tripped.
IC initiation signal is generated.
   ']~
43 0 secs.                          Ms1Ys fully close and RPV pressure increases to the S/R valve setpoint.
8.00 min.                          RPV level drops to TAF + 3', if IC Auto Initiation fails.
15 00 min.                          RPV level drops to TAF, if IC auto initiations fails and the operator actions to recover level fail.                                              1 3
l 1
8 water volume below TAF + 3 ft is 7140 ft b
AMENDMENT 2                             2C-20 MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

Table 2C-4.2
 '/     .
Imsa of Feedwater with Operator Intervention                                J U                                  (FW recovered after 25 minutes)                                       l Key Event                                 Time Reactor Trip                              ~17 0 secs.
Core Uncovery                              21.1 min.
(RPV Lr   A drops to TAF)
Core Recovery                              25.6 min.
Maximum Core Temperature                   ~672 F after 25 min.
Table 2C-4 3 Imss of Feedwater With Operator Intervention (SRV opened after 20 minutes)
Key Event                                  Time Reactor Trip                               ~17.0 secs.
Core Uncovery                              20.0 min.
i (RPV Level drops to TAF)
Core Recovery                             22.2 min.
Maximum Core Temperature                  ~593 F after 21.7 min.
LPCI Initiation                           21 7 min.
AMENDMENT 2                         2C-20a MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

l.(                                             FIG 3RE 2.C.4.1 S                                 CORE WATER HEIGHT                                                                                1 l'J/~N.
, i's                               REFERENCED TO. VESSEL BOTTOM MP1 LOFW2                                                                 ;qi CASE-STUDY'NO.2                                                                     j 27-JUN-88.10:35:20                                                                       J 1
LOSS OF FEEDWATER SCENARIO 50 -                  (WITHOPERATORINTERVENTION)-
q
_                                       \
4C-i Maximum Core Temperature s672*F-after 25 minates 10                       h 4.5 mg l
30-                               \                                                          TOP 0F ACTIVE FUEL 20-                                           '
4 l
l       '
i      i   '
i            '  l  '
i  '
l '  I Os             0.0 01     .l   0.2     0.3           0.4       0.5           0.6                         0.7 0.0     0.9   1.0 m2                                         TIME IN HOURS

FIGJRE 2.C.4.1                        HEIGHT OF WATER IN THE SHROUD 1/N                                 BEFERENCED'TO' VESSEL B0TTOM                                                                                                            -l V   '
MP1 LOFW2 CASE STUDY NO.2
                                            '27-JUN-88.10:35:20 LOSS OF FEEDWATER SCENARIO 60-               (WITHOPERATORINTERVENTION) 50-f3 V                 -
d u.
40-30-
                            /
                                                                                               ~
Feedwater Recovery after 25 minutes L
s                  20 ,   ,   , ,
0.0   0.1   0.2    0.3    0.4      0.5            0.6           0.7             0.8               0.9                   1.0 TIME IN HOURS AMENDMENT 2 i.
__.-_u   _  -a._ _ _ _       __._m._u___m          ._.u-___,__.____..-__-.2_         _ _ _ _ _ _ . _ _ _ _ -

L                                                                                                                                          MGURE'2.C.4.1-3 0'             -
MAXIMUM TEMPERATURE IN CORE MP1 LOFW2
           ' (n-     <)                                                                                                                        CASE STUDY NO.2 27-JUN-88.10:35:20 LOSS OF FEEDWATER SCENARIO 670 -
660 -
ss 672*F after i-                                                                                          650 .                                                                                                                                           25 minutes                                                      l 1,,
640-630-620-610-u.
UJ M.600-S C
590 580-570-560-
                                                                                                   ~                                                                                                                     l 550-    j 540-                                                                                                                                  %
530-(
i          i                           i     .        -
i             .
i   .
i .
i 0.0        0.1             0.2               0.3                                0.4                                              0.5    0.6   0.7     0.8       0.9     1.0 TIME IN HOURS AMENDMENT 2

FIGURE 2.C.4.1-4
  .- o o
        . -- m AVERAGE CORE TEMPERATURE
         .I          4                                                                                       NP1 LOFW2'
        'V                                                                                               CASE STUDY NO.2 27-JUN-88.10:35:20 LOSS OF FEEDWATER SCENARIO 580   -
(WITHOPERATORINTERVENTION)
H L                                   -                                                                                                                  .
578 F after 25 minutes i
570 -
560-
                                                          - u.
W 8
u                                 ,      5 e
                                                                           ,550             l f
i l
1 540-530-     ,    ...        .     .
                                                                                                                 .i.i                                       i                       i.i                                            i 0.0     0.1 0.2     0.3   0.4    0.5                            0.6    0.7                  0.8                      0.9                      1.0 1IME IN HOURS AMENDMENT 2

FIGURE 2.C.4.1-5 i :'                                                                                                                    i i
g                                                  WATER MASS IN CORE-
; ?bj' -                                                        MP1 LOFWE CASE STUOY NO.2 27-JUN-88.10:35:20 LOSS OF FEEDWATER SCENARIO-130000    -
(WITHOPERATORINTERVENTION) l-120000    --
f i
110000   -
I 100000 -
90000-
                                           . b f~ '
g   80000-E 8
7'.000-1 60000-50000-40000-30000-Feedwater Recovery
[ after 25 minutes 20000-r)                               i     -
: i.     ...          .,.,,,,,,,,,
     ;                               0.0 0.1       0.2  0.3   0.4      0.5 0.6  0.7   0.8   0.9       1.0 AMENDMENT 2                           TIME IN HOURS

_ _ _ - _ _ _ _ _ _ _ _ - - .                             . . ~ . - - _ _            _        ._ ____ _ __ _-_.                   _ _ _     __    ___ _ -
e-         ,
FIGURE 2.C.4.1-6 CORE INLET FLOW RATE                                                                                           ~
MP1 LOFW2 t       '
                                                                                                          . CASE STUDY NO.-2 1
27-JUN-88.10:35:20-LOSS OF FEEDWATER SCENARIO l
      ;                                                 8.0E+07          ,                            (WITHOPERATORINTERVENTION) l l                                                       7.OE+07 -
1
: 6. 0E +07 -
: 5. 0E + 07 -
: 4. 0E+07 -
E                                                                                                                                                                    I E                                                                                                                                           '
                                                      - 3. 0E + 07 -
_N
: 2. 0E4 07 -
: 1. 0 E + 0 7 .-
o
: 0. 0E+00 -                                 -
                                                                                                                                     \      Feedwater Recovery                                                        .
after 25' minutes
                                                    - 1. 0 E + 0 7 -
g                                                              4 l                 '          '
i  '4    '
4 i '
i'i'4
   .. id -                                                         0.0       0.1             ' 02
                                                                                               .i      03 .l      0.4                   0.5   0. 6    0.70.8                 0.9              1.0 TIME IN HOURS AMENDMENT 2

1 FIGURE 2.C.4.1-7'                                           ,,
PRESSURE IN: PRIMARY SYSTEM
-- (~T                                                              MP1 LOFW2
~ (J                                                       CASE STUOY NO.2 27-JUN-88.10:35:20 LOSS OF FEEDWATER SCENARIO.
1140-                             (WITHOPERATORINTERVENTION) 1130-1120-1110-
                                                                !l l
Feedwater Recovery after 25 minutes 1100-              j i
1090-                       :
1080-L               I           <
1070-                  !                        ,
1060-                                   j h                  4 G
1050-                         f      l,'          I
                   ' 1040-                      l                 I 1030-                                                                    {
1020-1010-1000-990-980-l 960-950 ,   .   ,   .          ,         ,      .,.,...,.,.                   ,    ,
-{                        0.0    0.1         0.2         0.3         0.4   0.5    0.6  0.7 0.8 0.9   1.0 TIME IN HOURS AMENDMENT 2

FIGURE 2.C.4.1-8 nV CORE WATER HEIGHT-REFEFiENCEO TO VESSEL BOTTOM MP1 LOFW3-CASE STUDY N0.3 27-JUN-88.13:21:24 t                                                  L'0SS OF FEEDWATER SCENARIO 50-                       (WITHOPERATORINTERVENTION)
                                    'I                                     h     f
                                     ~
                                                                                                                     ~~
40-                                          }
t                                                   .2 min.
t b     30-TOP OF AC_TJVE_EVEL _
u.
Maximum Core Temperature 593*F after about 21.7 minutes.
20-B
__..____....__...i __ __ __ _ -._OTTOM                    _ . _ _ _ . _ 0F  _ _ _ACTIVE
_ _ _ _ . _ .FUEL LPCI Started after 21.7 minutes SRV opened after 20 minutes i .
10                                                        '         '        '            '
i    '                         '                      i          l            l         i                      '
i 0.0        0 .l     .i 1 '02   ' 03
                                                         .i         0 .l4   ' 0 .l 5 0.6        0.7           0.8       0.9                          1.0 AMENDMENT 2                                             TIME IN HOURS l
l'
: l.                                                                                                       ._ -             __ __ ___ - _ _ _ _ __         a

l FIGURE 2.C.4.1-9
    #                                                    HEIGHT.0F WATER IN THE SHROUD
                                                                         . REFERENCED TO VESSEL BOTTOM-MP1 LOFW3 CASE STUDY NO.3-27-JUN-88.13:21:24 LOSS OF FEEDWATER SCENARIO 70 -                '(WITHOPJRATORINTERVENTION)
SRV opened after 20 minutes 60-50-O                                                 -
s                                                4              i o%
                                                                                                                                                                  ,o r
40-                                                                          a
                                                                                                                                                           'e   N 4
30-                                                                                                                                               !
                                                               )
LPCI Started after 21.7 minutes 20 .     .     .      .
                                                             .       i     .      .   .
i                   .
i     .
r- . i
                  ~
0.0 0.1        0.2    0.3     0.4                     0.5       0.6                         0.7         0.8       0.9   1.0 TINE IN HOURS AMENDMENT 2

m    ,
i                                                                                                                    i
 ~
FIGURE 2.C.4.1-10 Q                                            MAXIMUM TEMPERATURE IN CORE MP1 LOFW3-CASE STUDY N0.3 27-JUN-88.13:21:24 LOSS OF FEEDWATER SCENARIO 600 -                 (WITHOPERATORINTERVENTION) 593*F after 21.7 minutes 1
SRV opened after 20 minutes
                                              .500-O V                                  m Ls.J .
p 400-2C 300-O                                         2oo ,
0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 AMENDMENT 2                                                              TINE IN HOURS
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _              _                                                                           i

FIGURE 2.C.4.1-ll ~
       .l                                                                                  AVERAGE CORE TEMPERATURE
_.                                                                                                               MP1 LOFW3 CASE STUDY NO.3-27-JUN-88.13:21:24 LOSS OF FEE 0 WATER SCENARIO 600-                   (WITHOPERATORINTERVENTION)
                                                                                           ')
                                                                                                                               $                                                                                                                                                  1 I                                                                                                                                                    i 500-
                                                                                                                                                                                                                                                                               . .i
        ._i '                                                                                                                                                                                                                                                                   '
(                                                     u.                                                                                                                                                                                                                  I w
p:400--
S o.
300-SRV opened after 20 minutes I
200 ,   ,,,,,,,,,,,,,,,,,,,,
l l                                                                                         0.0  0.1       0.2         0.3        0.4                                                      0.5 0.6         0.7 0.8                                        0.9             1.0 i
TIME IN HOURS                                                                                                                                       i AMENDMENT 2                                                                                                                                                                                                                                                       l e
i t_._.___.         -                    __                      ._ _ _ . _ . _ _ _ . _             _            _ _ _           . _ . _ _ _ _ . _ _ _ _ . _ . _ . _ _ . _ _ _                     . _ _ _     _ _ _ _ . _ _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _                  __

L
                                                 - FIGURE 2.C.4.1-12.
WATER MASS IN CORE r Q'                                                      MP1 LOFW3 CASE STUOY NO.3 27-JUN-88.13:21:24 LOSS OF FEEDWATER SCENARIO 160000-               (WITHOPERATORINTERVENTION) 150000-140000-130000-120000     ,   ,
110000-100000-90000-g     80000-E                                                                                 I 2     70000-60000-50000-40000-30000-20000-10000-LPCI Started i              after 21.7 minutes 0-O              -10000 ,    .     ,   ,
0.2 0.3 0.4 0.7 0.8 0.9 1.0 0.0     0.1                       0.5   0.6 TIME IN HOURS AMENDMENT 2 L--_____:_-_______-____________

         ~!
U                                      FIGURE 2.C.4.1-13 CORE INLET FLOW RATE
  . b~.'
MP1'LOFW3 CASE STUDY NO.3 27-JUN-88.13:21:24 LOSS OF FEEDWATER SCENARIO B.OE+07-     .                  (WITHOPERATORINTERVENTION) 7.0E407-                                                                                                                               ,
: 6. 0E + 07 -
: 5. 0E + 07 -                                                                  g
: 4. 0E+ 07 -
E.
k
: 3. 0E+07 -
i
: 2. 0E+ 07 -
1.0E+07-
                                                   \
0.0E+00-                   --
                                                                     =p LPCI Started after 21.7 minutes l
     <^
                             - 1. 0E + 07 -
I '
i '              '        '
i   '
i '  i  '
i '
i   '
0.0      0.1     02.i  ' 03
                                                                  .i      0 .l 4  0.5    0.6   0.7 0.8    0.9      10   .i TIME IN HOURS AMENDMENT 2.

FIGURE 2.C.4.1-14 n                  PRESSURE IN PRIMARY SYSTEM                                                                                           !
MP1 LOFW3 CASE STUOY NO.3-27-JUN-88.13:21:24 LOSS OF FEEDWATER SCENARIO 1200-                 (WITHOPERATORINTERVENTION) 1000-
                                                 ~
900-800-i 700-h    600-o_
500-400-300-2-                   200-SRV opened after 20 100-      minutes 1                                                                                        )
(                    0, 0.0 0.1 0.2 0.3    0.4 0.5 0.6 0.7 0.8 0.9 1.0 AMENDMENT 2                                 TIME IN HOURS                                                                          ,
                                                                                      - _ - . _ _     _            ____.______m__-____-- _m

i 2C.5    Stuck Open Safety / Relief Valve Transient b) v In a Stuck Open Safety / Relief Value (SORV) transient, the electric pressure regulator (EPR) or mechanical pressure regulator (MPR),
maintains the RPV pressure constant by partially closing the turbine control valve. Feedwater flow adjusts to maintain the RPV level.
Diversion of 10% of the steam (capacity of 1 S/R Valve) to the torus through the open valve, slowly heats the torus. However no automatic reactor trip signal is generated.           The Technical Specification requires the operator to initiate a manual scram when the torus heats                             l up to 110 F (See EOP 580 Appendix 2B).               For the purpose of calulating operator action times, it is arbitrarily assumed that the reactor trip is initiated at the same time when the S/R valve sticks open (i.e. t=0 sec). Tnis assumption has no eifect on the calculation as the RPV parameters (level & pressure) remain unchanged until reactor trip is initiated.
Following the reactor trip, the RPV pressure begins to drop.                        The MSIVs close on LOW pressure (825 psig) in about four minutes (Figure n
2C-5). Feedwater would restore and maintain the RPV level, therefore t) it was arbitrarily assumed that Feedwater is initially unavailable.
At this point, all injection sources capable of maintaining RPV level are   lost. The RPV level        drops due to core boil-off and depressurization associated with the stuck open S/R valve. The core begins to uncover in about fourteen minutes.
The following discussion shows that if the operator takes corrective action at twenty minutes, the PCT will be less than 2300 F and significant fuel damage will be avoided. The corrective action postulated is as follows:
At twenty minutes:          The operator opens 2 S/R values (in addition to   the valve stuck cpen) to rapidly depressurize the RPV and starts at least 1 LPCI pump.   (The operator may also use the IC O
AMENDMENT 2                             2C-21 MILLSTONE UNIT 1                                         l PROBABILISTIC SAFETY STUDY

to aid in d:pr:ssurization. Howev2r, it is not credited).
n U                                               As shown in Figures 2C-5 and 2C-6, at twenty minutes the RPV
                                               . depressurized to about 420 psia cnd the level drops to 25.2 ft (4.4 ft below the TAF). When the operator opens 2 S/R valves, the RPV rapidly depressurizes. When the RPV pressure falls to the shut-off head (235 psig) of the LPCI pumps, the pumps begin to inject.        The injection rate increases as the HPV pressure drops. It is estimated that at 200    !
psia, the pump flow rate exceeds the core boil-off and flashing rate (due to depressurization) and therefore, the RPV level begins to        ,
increase.
Table 2C-5 summarizes pertinent data for this transient.         The RPV minimum level (20 ft) occures at 1600 secs.         (20 min + 400 sec, duration of depressurization).
i Figures 2C-7 through 2C-9 provide RPV pressure, levels and clad          ,
temperature plots for a SORV case with 1 CRD pump and 1 LPC1 pump
,r3                                             operating. No operator action to open additional S/R valves (for U                                               faster depressurization) is credited.       The PCT for this case is
                                                ~2300 F. The results of this transient can be used in estimating the PCT for the SORV case being analyzed here (i.e. With operator action at 20 minutes, but without the CRD pump). Thus the parameters which mainly determine the PCT, can be compared between the two cases. Tne parameters are:
o     Maximum core uncovery (minimum level),
o     Duration of core uncovery and,                                     '
o     Time after RT of maximum core uncovery.
Comparing the transient results presented in Figures 2C-7 through 2C-9 with the SORV case being analyzed, the following can be concluded:
o     The minimum level shown in Figure 2C-8 is lower than the minimum level calculated (20 ft.).                                         l A                                                                                                                        I bj AMENDMENT 2                               2C-22 MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

(
          )
o    Duration of the core uncovery in Figure 2C-8 is also longer.
im.                                            'Ihis is because, there is no rapid depressurization which allows a quick recovery (See Fig. 2C-7). Whereas, in the SORV case being analyzed here, the operator depressurizes the RPV at 20             l minutes, which allous a quick recovery of level by the LPCI-pumps.
d
                                        ,    o-   In the two cases being compared, the time of minimum level is comparable.
BaseJ on the above assessment, it is clear that the transient presented in Figures 2C-7 through 2C-9 is a more severe one.
Therefore, it can be concluded that for a SORV case if the operator takes' the corrective action, at 20 minutes, the resulting transient will have a PCT of 2300 F.
The previous' analysis examined the response of the In-vessel Key parameters such as the reactor pressure vessel (RPV) level, pressure, and inventory.
C'MPACT Code (Compartmentalized Analysis - of Containment Transients) .
has been utilized to study the ex-vessel key parameters related to the LPCI pump cavitation and the torus temperature during stuck open safety / relief valve scenarios with and without isolation condenser.
COMPACT Code has also been used to determine the conditions (time of addition, flow rate, and LPCI pump throttling) of water addition to the torus to prevent the LPCI pump cavitation.          The requirement of preventing pump cavitation can be represented by the following-i                                           inequality:
(                    >
SH) required H)available l~
2                                M MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

The analysis conducted by COMPACT Code is based on the following assumptions:
     /    )
L ,J 1 - Initial temperature of water in the torus is 75 F.
0 2 - Emergency Service Water temperature "ESW" is 70 F.
3 - Reactor tripped after seven minutes. The decay heat added to the the torus water reflects the reactor trip time.
4 - One LPCI loop is available (one LPCI ptmp and one LPCI heat exchanger).
5 - Two ESW pumps delivering a total of 5000 gpm to the LPCI heat                                                                i
                                       =xchanger.
Analysis of the results of COMPACT runs showed that:
n.
j    ()                           1- If the isolation condenser is operable (IC has heat removal
(                                       capability of 206 BTU /hr) and LPCI ptap throttling is credited, then no water addition to the torus is needed. Figure 2C-9A compares the NPSH available versus NPSH required during the
!                                       150,000 recond transient. As can be seen from Figure 2c-9A, the LPCI pump i,hrottled three times with the first throttling step started after approximately four hours.
l                                       Figure 2C-9B shows the torus water temperature during the transient. The maximum temperature of approximately 169 F was reached after about 32.5 hours.
2- This case considered the respor.se with the isolation condenser l                                       inoperable and no water addition to the torus.                                                              Figure 2C-9C l                                       compares NPSH available versus NPSH required of the LPCI pump.
As can be seen, pump cavitation occured after about 6.21 hours.
      -                                 Figure 2C-9D displays the torus temperature during the 150,000 1
V AMENDMENT 2 2C-23a MILLSTONE UNIT 1 PPOBABILISTIC SAFETY STUDY

second transient.      Tha. maximum temperature was about 211 F reached after about 16.81 hours.
p 3 - A final case considered the response with the isolation condenser-              i is inoperable and water addition to the torus is considered. In               ,
this case study, it is assumed that water is added to the torus               !
at 200 gpm flow rate. Water addition started after 1. hour'for a~             l duration ci about 14.41 hours (this corresponds to the addition
            .        of 4 feet of water to the torus).
Figure 2C-9E compares NPSH available versus NPSli required of the LPCI pump. As can be seen in Figure 2C-9E the LPCI pump has to be throttled three times (from 5000 gpm to 3500 gpm)' to avoid cavitation.
Figure 2C-9F shows the torus water temperature ouring the transient. Water addition started when the torus water temperature reached approximately 135 F. 'Ihe maximum torus temperature of approximately 186 F was _ reached after about
        ]             17 73 hours.
i 2                               LS MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY p
L ...                                                        -- __ ___ --_________ _ --_ _- _ _ _ _

p
 ' ' 7'                                                                            Table-2C-5 hJ -
               }                                                    Data for Stuck Open S/R-Valve Transient Descriptions                                             Value Core'Uncovery Starts-                                      ~14 minutes.
b                                         At~20 minutes: RPV Level                                    25.2 ft; RPV Pressure                   420 psia.
RPV Inventory                           5 2.74x10    lbs.
Depressurization from 420 to 200 psia:
Duration of depresurization (estimate)                      400 s'ecs.
Loss of Inventory due to flashing                           2.5x104 lbs N
Loss of Inventory due to core-boil off                      1.9x10 lbs RPV Inventory when the pressure is 200 psia                          5 2.30x10    lbs RPV Level when the pressure is 200 psia                    ~20. ft.#
I                            #This .value accounts for                        shrinkage    of the    RPV    inventory  due. to
                             <lepressurization.
i, 2C-24 MILLSTONE UNIT 1.
PROBABILISTIC SAFETY STUDY
__=_=__

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                             =         2                I O
I L. OWO      12WF- O[<J.O MqW1 5&*
LIll'

FIGURE 2C.9A
  -. f' .
    \                    COP ACT                   SORV CASE WITH THE ISOLATION CONDENSER OPEMBLE                                     9'02 32 19-A % 61-l 50                                         .
I i'                                            ,
t
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AMENDMENT 2 i
j  .-..... . .                      .

1 f3                                                          FIEJRil: 2C.G3
( )?-
C N ACI -             SORV CASE WITH THE ISOLATION CONDENSER OPERABLE 208.
I         I 175                                                                     +                                   -
                                                                                                                                      .l i50.    /   _
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4 e                                                   FIGURE 2C.90
. x) w
,                          COM ACT          SORV CASE WITHOUT 1      1 ADDITION AND                                                          E 10 55 30-MV-l THE ISOLATION CONDEh_ J IS INOPERAhLE 80.
I i
: 40.                                                                                                                       '
START OF. PUMP                       /
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                                                  / CAVITATION        7 REQUIR@          /
s
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                                                                /                    /
  >Os.
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     'n
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     ,                      - AMENDMENT 2

                                                                                                                                                                                                                                             )
l fN                                                                              FIGURE 2C.9D                                                                                                                                         1 k'.Y      '
C W ACT                       SORV CASE WITHOUT WATER ADDITION AND THJ                                                                    6'05:54 30 N E 65 ISOLATION CONDENSER IS INOP.ERABLE i,
l 200.                                              \
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S                    M5      ----        -  - - - - - -    ------      - - - - - -      - - - - -    ------                - - - - - -           - . - - .
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                                                                                                              -           ==              a-                                                                        m TIME (SECONDS) f                                    AMENDMENT 2

4
(                                         FIGURE 2C.9E COMPtCT 6:528dO.0i-Jd6-6E MPs SORV C ASE8 fl00% R.H.'1979ANS, TRIPS AT 7HIN.NO IC. ADD WATER 200 C8'M) j WESTINGHOUSE PROPRIEi ARY CL ASS 2                                            '
50                                AVAIL ABLE NPSH (FT) 40,
                                                             /
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                                       /
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p 20.
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              ~ 0.                                                                                                       I i
8            8       8          8         8          8 0             8                                                8          8 0           8 0          8        0          8
                                                                    -          -        -          c.
TIME (SECONDS) l l
L'                                                                                               _    ...._...___.__.__a

  'f e'(3-/.                                                FIGURE 2C.9F
           -COM*ACT                                                                                                                                                                                              8 51844 On-J06-6E MP1' SORY CASES (100% R.H.1979ANS. TRIPS AT THIN.NO IC, ADD WATER 200 GPM)
WESTINGHOUSE PROPRIETARY CLASS 2 TEMPERATURE       LOWER Rf 02sN  NODE 3 (DEC F)
                   .    .                            186 F 175                    ,                       \
i,          x 150.
                                /
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                                                                                                                                                                                                                      ~
TIME (SECONDS) g        AMENDMENT 2                                                                                                                                                                                                          ,
J                                                                                                                                                                                                                                .
1 J
i 1

L 2C.6       Isolation condenser Related Times and Flow Rate.
                        -1he Isolation Condenser (IC) removes core decay heat and depressurizes the RPV by condensing.the core boil-off and returning the condensate:
back to the RPV. The' shell ' side ' of the IC contains 2941 ft 3
                         .(22,000 gallons) - of water, out of which.' 2861 ft3 i is the . usable
                         ' inventory.- (The remaining inventory is below the tube' bundle)~. 'The
                        ' total heat removal capacity of this. initial usable-inventory is about'
* 8 1 99 x 10     BTU.
As a.part of the Technical. Specification surveillance, a heat balance was performed on the IC to determine its heat removal capacity. The 8
IC ' heat removal . rate has been calculated to be 2.73 x 10                         Btu /hr
                        ..(based on results of test data). Based on this heat renoval rate, the .I
                        . limited IC inventory will last for 44 minutes. It should be noted that this heat removal rate applies'when the RPV is at fbil operating pressure. As the RPV depressurizes, the IC heat removal rate drops.
Therefore, the IC inventory will actually last for more than 44 minutes.      At 44 minutes the core decay heat can . be removed by supplying 215 gpm.to the IC shell.
  ' h. .
                                                                                                                       )
i AMEND 11ENT 2                                2C-30                                     ,
PROBABILISTIC SAFETY STUDY

4  20.71    Imag Term Cooling !kaooess Criterio
     <       <            Any accident or transient can be ' divided into two parts; Short Term and -Long Term Cooling (LTC). During short term mitigation,- the operator is concerned about reactivity control. (Reactor Trip),
pressure. control- and RPV water ' level..'After the RPV level is recovered, the operator needs to establish . long term decay heat-removal.
This study considered four LTC systems, which are:
Main Condenser (MC)
Isolation Condenser (IC)
Shutdown Cooling (SDC)
Alternate Shutdown Cooling (ASDC)
To establish the requirements for long term cooling, two parameters need to be determined - These are:
                               .The minimum time'available for the operator to establish LTC and, L\
The heat removal capacity of the system needed to ensure LTC.
The most limiting scenario to determine these two parameter is ' the MSIV-closure event (excluding AWS initiated sequences).
In a transient caused by MSIV closure, the reactor -trip is generated--
either by the MSIV position or high dome pressure. The.RPV level can be recovered by Feedwater (W) or by low pressure (LPCI/CS) pumps. If the.RPV level' is being' maintained by the W system, long term cooling l
can be' delayed until W inventory is exhausted. However, if the RPV level is being maintained by the low pressure pumps, LTC needs.to be established before the torus heats up to a temperature where the pumps I
cavitate and are unable to maintain the RPV level.                                           The latter scenario provides a more restrictive (i.e. shorter) time requirement.
l-                                                                    2C-31 1
MILLSTONE UNIT 1                   1 PROBABILISTIC SAFETY STUDY ^
                                              -               - - - -    -  - - - - _ - - - - - __                         _       D

Tablo 2C-6 summarizes the datn used in calculating th3 minimum time        ;
available for the operator to establish LTC.         The initial torus temperature assumed in the analysis is 81 F. This value is the
   ''')
  ''      actual torus temperature measured during sumer of 1985. As torus heats, the LPCI pumps need to be throttled to maintain adequate Net        j Positive Suction Head (NPSH). The LPCI pumps can maintain a flowrate of >     3000 gpm (per pump) at a torus temperature of < 176 F. It should be noted that, this LPCI flow-torus temperature relationship
        . does not credit any increase in back pressure.      As the torus heats, the containment pressure begins to increase, unless the drywell is vented. The   higher  containment  pressure   increases      the  NPSH available. Therefore, the LPCI flowrate calculated in this analysis 0
is the minimum value. As the torus heats up above 176 F the LPCI flow rate drops rapidly and the pumps cavitate completely. Therefore, to ensure continuous operations of the LPCI pumps,              the torus 0
temperature needs to be maintained below 176 F.
During blowdown, the torus heats up from 81 F to 104 F as the RPV stored energy discharges into it. The torus continues to heat up due p''     to the addition of the core decay heat. If, no cooling is provided to      ,
the torus, it heats up to 176 F in 4.8 hours.
Therefore, the operator has at least 4 hours to initiate the torus cooling or establish a LTC system, which would stop further addition of heat to the torus. Selection of 4 hours as the maxinum operator action time ensures adequate time to place the LTC system in service.
6 The core decay heat at 4.8 hours is 65 x 10               BTU / hour. The Isolation Condenser, Main Condenser and the Shutdown Cooling systems remove the core decay heat by directly cooling for RPV. Thus further addition of heat to the torus is terminated.       Each of these three systems can match the decay heat well before 4.8 hours. It should be noted that only one loop of SDC system is needed. One loop of the SDC is defined as follows:
2C-32
[')
(Gt MILLSTONE UNIT "i PROBABILISTIC SAFLTY STUDY l                                                                                    !

                                            ._1 SDC pump E"
1 SDC Heat Exchanger b-                                   1 RBCCW Pump 2 RBCCW Heat Exchangers The Alternate SDC provides long Term Cooling by cooling the toms. At.
a torus temperature of 176 F, two trains of: the Alternate SDC 'can 6
remove well above 65 'x' 10 BTU /hr of heat from the torus. Clearly, the. torus water temperature will begin to drop before 4.8 hours as the torus heat removal rate is larger than the heat addition rate to it (i.e. the decay heat level). hio trains of Alternate SDC are required and defined as:
2 LPCI Containment Cooling Heat.Exenangers 1 LPCI Pump per Heat Exchanger 1 ESW Pump per Heat Exchanger The. COMPACT         Code       (Compartmentalized    Analysis               of    Containment
  .                      Transients) has been us'ed to develop the success criteria for toms                            i long - term cooling under the Loss of Normal Power (LNP) condition without crediting the Isolation Condenser (IC).
The following initial conditions were assumed in the analysis:
1 1 - Initial torus water temperature = 75 F (Best Estimate Value) 2 - Emergency service water temperature = 700 (Best Estimate Value) 3 - No Isolation Condenser 4 - Two emergency service water pumps (ESW pumps) delivering a total 1                             of 5000 gpm.
5 - One LPCI train with one LPCI pump in service and one LPCI heat exchanger.
c.
           . AMENDMENT 2                                           -33 MILLSTONE UNIT 1
:  _ _ _         _     _ - -   = _ _ _ _ . _ - - _ _ _ _ -

p         g U[                .
t                 '6 - The dtsign. flow rato of ths LPCI e sp is 5000 gpm.                                                                                                                                   1 e
        ;          7      .LPCI punp throttling is creditW. Up to four throttling steps, fA      -
with a flow reduction of 5000 gpm in each step, the LPCI pump is allowed to be throttled from 5000 gpm to 3000 gpm.
                                                                                                                                                                                                             -l
                  ;The analysis ' conducted by the. COMPAW Code included three scenarios .
for the LNP case; Case A No credit is given to the Alternate Shutdown Cooling System (ASDC) as a means of removing heat from the torus- As can be                                                                                                       .
seen in Figure 2C-9G, the LPCI pump throttled from 5000 spm to 3000 gpm and finally cavitated after 5.15 hours from the accident
              ,             initiation.
Figure 2C-9H shows that the torus water temperature r eached about-193 F after 5 98 hours.
O Case B Both LPCI ptap throttling and torus heat removal through the ASDC System are credited.                                                   Figure 2C-9I shows . that ' the LPCI pump throttled from 5000 gpm to 3000 gpm and finally ~ cavitated after 11.15 hours.
,                          Figure 2C-9J shows that the maximum torus water temperature is 195 F reached                  after about 24 31                                          hours from the accident-initiation time.                    At the maximum temperature, the torus heat removal through the LPCI heat exchanger matches the decay heay o                        delivered to the torus water.                                                     After the point of maximum temperature, the heat removal by the LPCI heat exchanger is greater than the decay heat and the torus water temperature-starts to decrease.
D i
MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

Q3e C
(-   :.
                         'With the stuck open safety relief valve (SORV) scenario, COM'ACT
    ''-                   Code analysis' showed that when the isolation condenser is credited as a means of removing heat from the RPV,. no kater addition to the                  torus 'is         required.           The maxixmum torus temperatu "ewas               169 F after 35.23 hours.              The LPCI pump'did not cavitate provided that pump throttling is credited.
Shce the stuck open safety relief . valve is a more limiting -
set.nario which bounds the~ Loss of Normal Power case, then it can be concluded that crediting the IC as well as LPCI planp throttling are sufficient conditions to prevent .LPCI pump.
cavitation given the same assumptions of initial conditions which were previously cited in the SORV. scenario.
I J.
o                                                                                                                                       :
AMENDMENT 2 2C-33b MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

Table 2C-6
               /           (DATA USED IN CAIELLATION OF LONG TERM COOLING SUCCESS CRITERIA)
Parameter                               Value 3
Initial Torus Water Volume '(Minimum Value)            98,000 ft Initial Torus Water Temperature                             81 F (Nominal Value) 0 Maxinum Torus Temperature (Acceptable Value)               176 'F Initial RPV Inventory                                  4.06 x 105 lbs Initial RPV Pressure                                  ~1000 psia Final RPV Pressure (at the end of blowdown)                 20 psia (assumed)
RPV Stored Energy discharged to the Torus (during blowdown)               1,41 x 100 BTU Torus Temperature at the end of blowdown                 104 F                     j Energy Needed to Increase the torus temperature 0                                       8 from 104 F to 176 F                              4.39 x 10 BTU 1
Time when the integral core decay heat matches 8                                       4.8 hours 4.39x10 BTU 6
The core decay heat at 4.8 hours                        65 x 10 BTU /hr           l l
l l
                   'See text for discussion                                                               i t
(
l 2C-34 MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY
  - - _ - .-_-              -    __-                                 _       _           _    _     _ _ a

FIGURE 2C.9G r
   .(f
    .N
              -COHDACT-            MP-1 LOSS OF NORMAL POWER                                                        '12'06:34 Od-APR-BE (NO IC, NO SDC. AND WATER ADDITION) 50.
40 AVAILABL1: NPSH 30
                                        \                          Y
                                                                  /      '
            ,                                        REQUIRED NPSH
: Ong                                                                             START OF PUMP N \
: 20.                                                           CAVITATION 60.
i 0.
Y                         Y                 h                                $              $
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                                                           -                 -                                ~              ~
TIME (SECONDS)
AMENDMENT 2

r.
FIGURE 2C.9H r).
COMc AC T
                                              'MP-1 LOSS OF NORMAL POWER                     12 ' 07 4 55 Od- A*R- 6f (NO IC, NO SDC, AND NO TORUS WATER ADDITION) 200 Tmax = 193'F /
155 .
                                                                            /
150.                      .
O E              125          /
                                        .l M              100.     -
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l 3               75 l
               -l 50.
25.
1
: 0.                                                                                                  ,
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                                                       -                  -             ~                            g TIME (SECONDS) ry AMFRDMENT 2                                                                                          j

     ,n -
i f-FIGURE 2C.9I COM*ACT MP1 LOSS OF NORMAL POWER
                                 -(NO IC AND NO TORUS WATER ADDITION)-
50, 40.
(
REQUIRED NPSH J0.
n N
: 20.                      <          A s\
AVAILAELE                       ,
NPSH                           !
10.
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0      Y                     Y         Y
     /~Y                    8  8          8          8         8     8 8 8 O                       ~  -          ~
e         e     e ~
P.
TIME (SECONDS)
AMENDMENT 2                                                                  '
L

i'M
     -V                                                                                                              FIGURE 2C.5LT COPPACT MP1 LOSS OF NORMAL POWER 12'52 47 30-M M-6 (NO IC AND NO TORUS WATER ADDITION) 300.
Tmax = 15 5'F
                                                                               .                           [
ir5 f
150.
125 g,C v
           .R                                                            ico.
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O                                                               50.
: 25.                   _ _ _ _ _
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                                                                                                                                           -                          -     ~
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AMENDMENT 2 h-             - - _ - . _ - - - _ _ _ _ _ _ _ _ _ _ . _ . - _ _ _ . . _ _ _ _ _ _ _ _ _

      ;,~
lb
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FIGURE 2C.9K C0FFACT                                                  MP-1 LOSS OF NORMAL POWER                          6: 45: 40 3:-mar.-61 (NO IC AND WITH WATER ADDITION TO THE TORUS) so.                                                                                           .
40.
e AVAILAB' E NPSH                                 1 J0.                    A-                     #
E                                                  D cc                                                                   s\                                     .
U[                                                                         REQUIRED NPSH I
20 I
lo.                                                                 _
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                                                                                       -           n_        m_      C_        e TIME (SECONDS)
AMENDMENT 2

E.
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3 l                                                                                                                                                                                                                      1
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1,                                                              .C W Act MP-1 LOSS OF NORMAL POWER (NO IC AND WITH WATER ADDITION TO.THE TORUS) 200.
                                                                       .                                                                                                          I I
Tmax = 178'F 175.            C~                   -
                                                                                                                               ~' '
(
                                                                     '150
: g.        125-t                   /
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O                                           AMENDMENT 2

p  2C.8 Large Break IACA Analysis
  %)
This section summarizes the results of a large break LOCA analysis with orJ y 1 LPCI pump available.         The analysis assumed that the reactor is operating under normal full power condition. A guillotine fracture of the recirculation suction line occurs in one recirculation loop. A high drywell pressure signal of 2 psig occurs almost imediately and scrams the reactor.
The reactor water level and pressure drop rapidly uncovering the core. The FWCI, the LPCS, 4 LPCI, IC, end APR are all signaled to start. Of all these systems, only 1 LPCI pump is assumed to start.
The LPCI injection valve opens on low vessel pressure permissive and the LPCI pump injects water into the vessel at 15 seconds reflooding the core. No operator action is required.
Figures 2C-10 through 2C-12 provide the transient plots.                                 The peak clad temperature calculated is 2100 F.
O V
One Core Spray pump is more effective than one LPCI pump because the Core Spray provides spray cooling as well as reflooding capability even    though the flow rate capacity of the pumps is similar.
Therefore, large break LOCA with 1 CS pump will result in a PCT <,
2100 F.
i 2C-35                                                                                    l MILLSTONE UNIT 1                                                             l PROBABILISTIC SAFETY STUDY                                                   I l
,                                                                                                                               I

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2C-37

1    "    l C
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AW                                                _
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i-2C.9                                             Small Break LOCA Analysis py This section summaries results of a small break LOCA analysis with 1 LPCI pump and 2 S/R valves available. The analysis assumed that the reactor is operating under normal full power conditions.                                                                       A small 2
break ~ of 0.07 ft       which is the limiting break size occurs in the recirculation line of one of the recirculation loops. A high drywell pressure signal of 2 psig occurs which scrams the reactor.and signals the ECCS to initiate. Coincidently, a loss of feedwater is assumed.
The reactor water level and pressure start to drop. The water level drops to the trip setpoint of the MSIVs. The MSIVs close and the vessel pressure subsequently rises to the setpoint of the S/R valves.
The S/R valves cycle to limit vessel pressure. The APR timer is started.
Two minutes after receipt of the low-low water level signal, the APR timer times out and a pressure signal across the one running LPCI pump
: f.                                                                        confirms that the pump is running.                        All 4 valves of te APR are automatically signaled to open but only 2 of the 4 valves are assuraed to open. The reactor depressurizes until the LPCI injection valve opens allowing the 1 LPCI pump to inject water into the reactor
                                                                      .reflooding the core.
Figures 2C-13 through 2C-18 provide the transient plots. The peak clad temperature calculated is 1811 F. Similar to the discussion f
j                                                                         presented for the Large Breaks, the PCT with 1 core spray pump instead of 1 LPC1 pump is lower,1790 F.
2C-39 MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY

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2C-45

,q      2C.10         References
'\,,)
: 1.      AE-106-1283, Rev. 1,  " Realistic Evaluations of Degraded Accident Events for Millstone Unit 1", May 1984. (Proprietary)
: 2.       NEDO-24708A, " Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors Volumes 1 & 2" Rev. 1, December 1980.
3        NUREG/CR-4165 (EGG-2379), Severe Accident Sequence Analysis Program   -
Anticipated Transient Without Scram Simulations for Brown Ferry 1",
Idaho National Lab, February 1985.
4        NECC-22166 " Power Suppression and Boron Remixing Mechanism for General Electric BWR Emergency Procedure Guidelines", August 1983.               .
l
: 5.       NEDC-20850, " Anticipated Transient Without Scram:       Study For the Millstone Power Station Unit 1, Docket #50-245", March, 1975.
W I
2C-46 MILLSTONE UNIT 1 PROBABILISTIC SAFETY STUDY
                -- -       -      -               -                -                         J}}