ML20064G752
| ML20064G752 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/31/1982 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML20064G750 | List: |
| References | |
| NUDOCS 8301110627 | |
| Download: ML20064G752 (95) | |
Text
{{#Wiki_filter:.-. . . 7: l 1 t 4 3 JOSEPH M. FARLEY NUCLEAR PLANT
; UNIT 1 AND UNIT 2 BORON CONCENTRATION i
REDUCTION IN THE BIT i i 4 ,
. j r
3 i i f i i l I Alabama Power Company December 1982 830111 %27 821230 PDR ADOCK 05000340 PDR P
TABLE OF CONTENTS Page I. Introduction 1 II. System Description 2 2 III. Steam Line Break Analysis 3
- A. Credible Steam Line Break
- l. Methods of Analysis
- 2. Results
- 3. Conclusions B. Hypothetical. Steam Line Break t
- 1. Methods of Analysis ,
- 2. Results
- 3. Conclusions I
- IV. Containment Analysis ?
13 A. Methods of Analysis B. Results i C. Conclusions ! J ; V. Plant Modifications 16 I
- VI. Technical Specifications 17
! l i 1 l 2
I I. Introduction The purpose of this report is to provide information for support of an application to amend the Technical Specifications associated with a reduction of the boron concentration requirements for the boron injection tank (BIT) and the associated recirculation system at the Joseph M. Earley Nuclear Plant - Units 1 and 2. The information provided herein includes a discussion of the BIT and associated systems, a steam line break analysis, a containment analysis, plant modifications, and proposed Technical Specification changes. The system description (section II) provides an overview of the boron injection system and associated components. Piping and instrument diagrams (P& ids) are included to enhance the understanding of the functions and interactions of the boron injection system components. Analytical models and calculational techniques are used in the steam line break analysis (section III) to show that elimination of the boron concentration requirements for the BIT can be accommodated within existing plant design criteria. This analysis addresses credible and hypothetical steam line breaks through a discussion of methodology, results, and conclusions. The effect of the BIT boron concentration reduction on the O' containment analysis (section IV) is evaluated along with an assessment of equipment qualification parameters to verify that all applicable design criteria are met. This evaluation consists of discussions of methodology, results, and conclusions. Section V addresses changes to the system that will be made as a result of the BIT boron concentration reduction. The proposed changes to the Technical Specifications are addressed in section VI. In conclusion, this report documents the analyses and evaluations necessary to justify elimination of boron concentration requirements for the BIT. The results of the analyses demonstrate that all plant safety design criteria are met and are in compliance with current NRC regulations. It is therefore concluded that the proposed Technical Specification changes, described in section VI, are justified. O V 1 i
t
~
System Description
[d. II. The BIT is a component of the boron injection system, which is a part of the safety injection system (SIS). The purpose of ; the boron injection system is to provide concentrated boric l acid to the reactor coolant system to mitigate the consequences , of postulated steam line break accidents. The principle components associated with the boron injection system are the BIT, the boron injection surge tank, and the boron injection recirculation pumps. The main purpose of the BIT is to store boric acid solution for injection into the reactor coolant system upon actuation of the ; safety injection system. The boric acid solution is 12 percent by weight and must be heat-traced in order to ensure the temperature of the solution in the tank (pes not drop below 135*F, the solubility limit of the solution. . Incorporated in the BIT is a sparger type inlet which distributes the incoming solution in a 360 degree fan in order to prevent channeling and ensures a radially homogeneous solution. The boron injection surge tank provides surge capacity for the BIT recirculation loop. During normal operation, the boric acid solution in the boron injection surge tank remains at approximately the same concentration as that in the BIT. O Recirculation of the boric acid solution through the boron injection tank, the boron injection surge tank, and associated piping is performed by the boron injection recirculation pumps. This prevents cold spots and stratification of the boric acid solution in the BIT during normal operation. Each unit , contains two of these pumps. One is in continuous operation while the other is a backup for maintenance and safety procedures. Additional information concerning the manner in which the BIT and its associated system interact with the safety injection system and the chemical and volume control system (CVCS) is provided in the Final Safety Analysis Report (FSAR) subsections '
- 6.3.2 and 9.3.4, respectively. The P& ids of the SIS and CVCS, which include the boron injection system, are provided in figures II-l through II-14.
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"" * ^"#"' UNIT 1 AlahamaPOWCf UNIT 1 AND UNIT 2 FIGURE II-l i
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.e.=. 2 7 D-2050s CHEF 1ICAL AND VOLUtiE NH E F ARLEY NUCLE AR PLANT CONTROL SYSTEM UNIT 2 Alabama Power UNIT 1 AND UNIT 2 FIGURE II-14
A 'III. Steam Line_ Break Analysis \sI - The " credible" steam line break (failure by-opening of'a single steam generator relief,~ safety, or turbine bypass valve).and the " hypothetical" steam line break (double ended rupture of a main steam line) serve _as the Farley steam line break licensing
-basis,.and define the existing requirements on the minimum BIT boron concentration. To show that a high concentration of boric acid solution is not essential to the-design criteria or operation of the plant, the following four cases were analyzed assuming a concentration of 0 ppm in the BIT: -Two " credible" steam line break cases with offsite power;available (uniform and nonuniform breaks) and the " hypothetical" steam line break,.
with and'without offsite power available. Only the limiting
" credible"_and " hypothetical" break cases are presented. The
" credible" steam line break and the " hypothetical" steam line break cases analyzed are discussed below.
III.A. Credible Steam Line Break ' The most severe core conditions resulting from an accidental-depressurization of the main steam system, which is classified as an ANS Condition II event, result from an inadvertent opening of a single steam, dump, relief, or. safety valve. The steam release as a consequence of this accident results in (~} ( ,f an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the reactor coolant system causes a reduction in coolant temperature and pressure. In the-presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity.
, For.the "crediblo" steam line break, i.e., the spurious opening of a single generator relief, safety, or turbine bypass valve, the radiation releases must remain within the requirements of 10 CFR Part 20. This is the ANSI N18.2 criterion for Condition II events, " Faults of Moderate Frequency." Although the plant may return to criticality, the above limit can be met by showing that there is no consequential fuel damage, i.e., that the departure from nucleate boiling (DNB) design basis is met.
Therefore, the analysis is performed to demonstrate that the following criterion is satisfied: Assuming the worst rod cluster control assembly stuck, with offsite power available, and assuming a single failure in the engineered safety features, there will be no consequential damage to the core or reactor coolant system-after reactor trip for a steam release equivalent to the spurious opening, with failure to close, of l the largest of any single steam dump, relief, or safety valve. i 3
- /
(,) III.A.1 Methods of Analysis Analyses of a secondary system steam release were performed-to determine the following: A. The reactor coolant system temperature and pressure, during cooldown, and the effect of safety injection with a full plant digital computer simulation using the LOFTRAN(a) code. This code was used in place of MARVEL,tb> which was used for the FSAR analysis. The LOFTRAN code includes models of the reactor core, steam generators, pressurizer, primary piping, protection systems and engineered safeguards systems. LOFTRAN also has the capability of modeling the entrance and exit temperatures of the cold and hot legs, for both faulted and unfaulted loops, respectively. The reactivity effects which occur during a steam line break cre computed as a function of power, coolant density, boron concentration, etc., by the code. B. That there is no damage to the reactor core or reactor coolant system. The following conditions are assumed to exist at the time of a secondary steam system release: '~' A. End-of-life shutdown margin at no-load, equilibrium xenon conditions, and with the most reactive rod cluster control assembly (RCCA) stuck in its fully withdrawn position. Operation of rod cluster control assembly banks during core burnup is restricted in such a way that addition of positive reactivity due to a secondary system. steam release accident will not lead to a more adverse condition than the case analyzed. B. A negative moderator coefficient corresponding to the end-of-life rodded core with the most reactive rod cluster control assembly in the fully withdrawn position. The variation of the coefficient with temperature and pressure is included. The keff versus temperature at 1000 lb/in.2 corresponding to the negative moderator temperature coefficient used is shown in figure III.A-1.
- a. WCAP-7907, T.W.T. Burnett, et. al., "LOFTRAN Code Description," October 1972.
- b. WCAP-7909, H. G. Hargrove, " MARVEL - A Digital Computer
\- Code for Transient Analysis of a Multiloop PWR System," October 1972. 4
]
m ._ (~N C. Minimum capability for injection of concentrated boric - ( ,/ acid solution corresponding to the most restrictive single failure in tho_ safety injection system is assumed. This corresponds to the flow delivered by one charging pump delivering its full contents to the cold leg header. No credit is taken for the low concentration boric acid which must be swept from the safety injection lines downstream of the refueling water storage tank prior to the delivery of corscentrated boric acid (2000 ppm from the refueling water storage tank) to the reactor coolant loops. D. The case at 1020 studied paia consists ofgenerator from one-steam a steam flow with of 229 lb/s offsite power available. This in the maximum capacity of any single steam dump, relief, or safety valve. Initial hot standby conditions with minimum required shutdown margin at the no load Tavn is assumed since this represents the most conservative initial condition. E. Should the reactor be just critical or operating at power at the time of a steam release,_the reactor will be tripped by the normal overpower protection when power level reaches a trip point. Following a trip at power, the reactor coolant system contains more stored rg energy than at no load, the average coolant ( ,) temperature is higher than at no load, and there is appreciable energy stored in the fuel. Thus, the additional stored energy is removed via the cooldown caused by the steam release before the no load conditions of reactor coolant system temperature and shutdown margin assumed in the analysir are reached. After the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis which assumes no load condition at time zero. However, since the initial steam generator water inventory is greatest at no load, the magnitude and duration of the reactor coolant system cooldown are less for steam line releases occuring at power. F. In computing the steam flow, the Moody Curve for fL/D = 0 is used. G. Perfect moisture separation in the steam generator is assumed. III.A.2 Results Figures III.A-2 through III.A-5 show the transient results for a steam flow of 229 lb/s at 1020 paia from one steam generator. The assumed steam release is that for the capacity of any (-] single steam dump, relief, or safety valve. LJ S
. . . a
4 ('N Safety injection (SI) is initiated automatically by. low
\s-) . pressurizer pressure. Operation.of one pump is assumed. Boron solution at 2000 ppm enters the reactor coolant system from the refueling water storage tank (RWST) providing sufficient negative reactivity to prevent core damage. The calculated transient is quite conservative with respect to cooldown, since no credit is taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the other steam generators. Since the core transient occurs over a period of about 5 minutes, the neglected stored energy will have sufficient effect in slowing the cooldown.
Following blowdown of the faulted steam generator, the plant can be brought to a stabilized hot standby condition through . control of auxiliary feedwater flow and safety injection flow, as described by plant operating procedures. The operating procedures would call for operator action to limit reactor coolant system pressure and pressurizer level by terminating safety injection flow, and to control steam generator level and reactor coolant system coolant temperature using the auxiliary feedwater system. Any action required of the operator to maintain the plant in a stabilized condition will be in a time frame in excess of 10 minutes following safety injection actuation. III.A.3 Conclusions The analysis shows that the criteria stated earlier in this section are satisfied. For an accidental depressurization of the main steam system, where the boron concentration in the BIT is O ppm, the minimum departure from nucleate boiling ratio (DNBR) remains well above the limiting value and no system design limits are exceeded. III.B. Hypothetical Steam Line Break The steam release arising from a rupture of a main steam line would result in an initial increase in steam flow which decreases during the accident as the steam pressure decreases. The energy removal from the reactor coolant system causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity. If the most reactive rod cluster control assembly is assumed stuck in its fully withdrawn position after reactor trip, there is an increased possibility that the core will become critical and return to power. The core is ultimately shut down by the boric acid delivered by the safety injection system. For the " hypothetical" steam line break, i.e., double ended rupture of a main steam line, the radiation releases must remain within the requirements of 10 CFR Part 100. This is the , (~} ANSI N18.2 criteria for Condition IV events, " Limiting Faults". (,,/- Alabama Power Company conservatively meets this for the Farley 6
9 (~') Units by demonstrating that the DNB design basis is mot, the 'u/ criterion typically used for Condition II events. The analysis of a main steam line rupture is. performed to demonstrate that the following criteria are satisfied:
- 1. Assuming a stuck RCCA, with or without offsite power, and assuming a single failure in the engineered safety features, the core remains in place and intact.
Radiation doses do not exceed the guidelines of 10 CFR
. 100.
- 2. Although DNB and possible cladding perforation following a steam pipe rupture are not necessarily unacceptable, the following analysis, in fact, shows that the DNB design basis is met f or any rupture assuming the most reactive RCCA stuck in its fully withdrawn position.
The rupture of a major steam line, which is classified as an ANS Condition IV event, is the most limiting cooldown transient and, thus, is analyzed at zero power with no decay heat. Decay heat would retard the cooldown thereby reducing the return to power. A detailed analysis of this transient with the most limiting break size, a double ended rupture, is presented here. I'T III.B.1 Method of Analysis - \_) The analysis of the main steam line rupture has been performed to determine: A. The core heat flux and reactor coolant system temperature and pressure resulting from the cooldown following the steam line break. As for the " credible" break, the LOFTRAN Code has been used instead of . MARVEL. B. The thermal and hydraulic behavior of the core following a steam line break. A detailed thermal and hydraulic digital-computer code, THINC, has been used to determine if DNB occurs for the core conditions
, computed in A. above.
The following conditions were assumed to exist at the time of a main steam line rupture: A. End-of-life shutdown margin at no load, equilibrium xenon conditions, and the most reactive RCCA stuck in its fully withdrawn position. Operation of the control rod banks during core burnup is restricted in such a way that addition of positive reactivity in a steam line break accident will not lead to a more (~} \.J adverse condition than the case analyzed. 7
I l
'[\_/)
B. A negative moderator coefficient corresponding to the f end-of-life rodded core with the most reactive RCCA in I the fully withdrawn position. The variation of the coefficient with temperature and pressure has been included. The keff versus average coolant temperature at 1000 lb/in.2 corresponding to the negative moderator temperature coefficient used is shown in figure III . A-1. The core properties associated with the sector nearest the affected steam generator and those associated with ' the remaining sectors were conservatively combined to obtain average core properties for reactivity feedback calculations. Further, it was conservatively assumed that the core power distribution was uniform. These two conditions cause underprediction of the reactivity feedback in the high power region near the stuck rod. To verify the conservatism of this method, the reactivity, as well as the power distribution, was checked for the limiting conditions for the cases analyzed. The core analysis considered the Doppler reactivity from the high fuel temperature near the stuck RCCA, moderator feedback from the high water enthalpy near the stuck RCCA, power redistribution and nonuniform core inlet temperature effects. For cases in which steam generation occurs in the high flux regions of the core, the effect of void formation was
\_ also included. It was determined that the reactivity employed in the kinetics analysis was always larger than the reactivity calculated including the above local effects for the conditions. These results verify conservatism, by an underprediction of negative reactivity feedback from power generation.
C. Minimum capability for injection of boric acid
- solution (2000 ppm from the RWST) corresponding to the '
most restrictive single failure in the safety injection system. The flow corresponds to that delivered by one pump delivering full flow to the cold leg header. No credit has been taken for the low concentration borated water which must be swept from the lines downstream of the refueling water storage tank prior to the delivery of concentrated boric acid to the reactor coolant loops. The calculation assumes the boric acid is mixed with and diluted by the water flowing in the reactor j coolant system prior to entering the reactor core. The concentration after mixing depends upon the relative flow rates in the reactor coolant system and in the SIS. The variation of mass flowrate in the reactor coolant system due to water density changes is (~T included in the calculations, as is the variation of
-s / flow rate in the SIS due to changes in the reactor 8
( ) coolant system pressure. The SIS flow calculation includes the line losses in the system as well as the SI pump head curve. The boric acid solution from the safety injection system is assumed to be uniformly delivered to the three reactor coolant loops. The boron in the loops is then delivered to the inlet plenum where the coolant (and boron) from each loop is mixed and delivered to the core. The stuck RCCA is conservatively assumed to be located in the core sector near the broken steam generator. Because the cold leg pressure is lowest in the broken loop due to larger loop flow and a larger loop pressure drop, more ' boron would actually be delivered to the core sector where the power is being generated, enhancing the
- effect of the boric acid on the transient. No credit was taken for this in the analysis.
For the cases where offsite power is assumed, the sequence of events in the SIS is the following. 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 SI pump starts. In 12 seconds, the valves are assumed to be in their final ( ') position and the pump is assumed to be at full speed. The volume containing the O ppm borated water is swept into the core, before the 2000 ppm borated water from the RWST reaches the core. This delay, described above, is inherently included in the modeling. In cases where offsite power is not available, a 12 second delay to start the standby diesel generators in addition to the time necessary to start the safety , injection equipment is included. D. Design value of the steam generator heat transfer coefficient including allowance for fouling factor. E. Since the steam generators are provided with integral flow restrictors with a 1.061 fta throat area, any .' rupture with a break area greater than 1.061 ft 2 regardless of location, would have the same effect on . j the nuclear steam supply system (NSSS) as the 1.061 : fta break. The following cases have been considered in determining the core power and reactor coolant ' system transients: !
- i. Complete severance of the pipe, with the plant initially at no load conditions, full reactor coolant flow with offsite power available. This
()/ s is the most limiting case. i 9
~ x' l' ) 11. Case i with_ loss =of offsite_ power simultaneous
~
;with the initiation of the safety-injection signal. Loss of offsite power results in reactor coolant pump coastdown.
F. Power peaking factors corresponding to one stuck-RCCA and nonuniform core inlet coolant temperatures are. determined at end of core life. The coldest core inlet temperatures are assumed to occur in the sector with the stuck rod. The power peaking factors account for the effect of the local void in the region of the stuck control assembly during the return-to power phase following the steam line break. This void, in conjunction with the large negative moderator coefficient, partially offsets the effect of the stuck assembly. The power peaking factors depend upon the core power, temperature, pressure, and flow and thus, are different for each case studied. The core parameters used for each of the two cases correspond to values determined from the respective transient analysis. Both the cases above assume initial hot shutdown at time zero since this represents the most pessimistic initial condition. Should the reactor be just I~}
\- critical or operating at power at the time of a steam line break, the reactor will be tripped by the normal overpower protection system when power level reaches a trip point. Following a trip at power, the reactor coolant ayatem contains more stored energy than at no-load, the average coolant temperature is higher than at no load and there is appreciable energy.storod in the fuel. Thus, the additional stored energy is removed via the cooldown caused by the steam line break before'the no load conditions of reactor coolant system temperature and shutdown margin assumed in the analysis are reached. After the additional stored energy has been removed, the cooldown and reactivity insertion proceed in the same manner as in the analysis, which assumes no load condition at time zero.
G. In computing the steam flow during a steam line break, the Moody curve for fL/D = 0 is uned. H. Perfect moisture separation in the steam generator is assumed. I. Feedwater additionally aggravates ;ooldown accidents like the steam line rupture. Therefore, the maximum feedwater flow is assumed. All the main and auxiliary (~} feedwater pumps are assumed to be operating at full x_- 10
~
I % x ; l j
) . capacity When the rupture-occurs,Jeven'though~the plant-is assumed to'belin-a hot standby (condition.
J. -The effect'of heat transferred from-' thick metal in the pressurizer and. reactor vessel upper head is~not included in the cases analyzed. Studies previously. .. = performed have shown that the heat transferred to the 1 coolant from these latent-sourcesLis ainet benefit in
'DNB and reactor coolant system' energy when the effect ' of the extra heat on reactivity and peak power is considered. -
III.B.2 Results Figures III.B-1 through III.B-4 show the reactor coolant system
- transients and core heat flux following a-main steam line '
i rupture at initial no load conditions. Offsite power is assumed to be available so that-full reactor coolant flow
. exists. The transient'shown~ assumes an uncontrolled steam release from only.one steam generator.. Should: the core be critical at near zero power when the rupture occursi the;
- - initiation of safety injection by high steam flow coincident with low steam zline pressure or ~ 1ow Tavg will' trip the reactor.
- Steam release from more than one steam generator will be 1
prevented by automatic trip of the fast acting isolation valves in the steam lines by high-high containment pressurefsignals or by high steam flow coincident with low-low Tavg signal or low steam line pressure. Even with the failure of one valve, l release is limited to no more than 10 seconds for the other i steam generators while the one generator blows down. The steam ; line stop valves are designed to be fully closed in less than 5 seconds from receipt of a closure signal. h As shown in figure III.B-4, the core attains criticality with the RCCAs inserted ~(with the design shutdown assuming one stuck RCCA) before boron solution at 2000 ppm enters.the reactor 4 coolant system. A peak core power less than the nominal full . power value is attained.
- t The core and systems responses were also calculated for the
!. case' discussed above with additional loss of offsite power at the time the safety injection signal is generated. For the - ! DNBR evaluation,' power feedback and power shape analysis i , consistent with the fluid conditions was used. i
- Following blowdown of the faulted steam generator, the plant can be brought to a stablized hot standby condition through ,
control of the auxiliary feedwater flow and safety injection J flow'as described by plant operating procedures. The operating
- procedures would call for operator action ~to limit reactor coolant system pressure and pressurizer level by terminating i safety injection flow and to control steam generator level and j reactor coolant system coolant temperature using the auxiliary llO feedwater system. Any action required of the operator to 11
ys
') maintain the plant in a stabilized condition will be'in a time N-frame in excess of 10 minutes following safety injection actuation.
A DNB analysis was performed for.both of these cases. It was found that the DNB design basis was met'for both cases, assuming 0 ppm in the BIT. III.B.3 ' Conclusions The analysis has shown that the criteria stated earlier in the accidental depressurization"of the secondary system section are satisfied. Although preventing clad damage is not necessary for Condition IV events, the results show that the DNB design basis is met. The dose evaluation, which was performed in the FSAR assuming 1 percent failed fuel, therefore continues to demonstrate that the Condition IV accident criteria are satisfied. O 4 4 12 -
.O ZERO POWER 1000 psia END OF LIFE RODDED CORE WITH ONE RCCA STUCK IN THE FULL OUT POSITION AND ZERO 1.06 -
BORON CONCENTRATION 1.04 --
, 1.03 -
I x 8
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, JOSEPH M. FARLEY 1.061 FT DOUBLE ENDED RUPTURE ggg gg,g NUCLEAR PLANT UNIT 1 AND UNIT 2 BIT CONCENTRATION O PPM FIGURE III.B-1 44401
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. 4440-1
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STEA'iLINE BREAK TRANSIENT JOSEPH M. FARLEY 1.061 FT DOUDLE ENDED RUPTURE I AlabamaPower d NWAR PLANT BIT C NCEMTRATION O M UNIT 1 AND UNIT 2 1 FIGURE III.B-4 44441 - t
( IV. Containment Analysis V) Steam line ruptures occurring inside the containment structure may result in significant pressure and temperature transients. The reduction of the boron concentration in the BIT will effect the containment analysis for a main steam line break due to a change in the mass / energy release. A complete spectrum of breaks has been analyzed to determine the impact of the worst breaks with respect to pressure and temperature. A. Methods of Analysis The containment pressure analyses were performed using , the Bechtel computer program, COPATTA.'A' The COPATTA model calculates both the pressure and temperature within the containment regions and the temperatures within the containment structures. Blowdown studies have been performed by Westinghouse to determine the mass and energy input rates for the containment analysis, as described below. The COPATTA program uses a three region containment model consisting of the containment atmosphere (vapor region), the sump (liquid region), and the water contained in the reactor vessel. Mass and energy are transferred between the liquid and vapor regions by boiling, condensation, or liquid dropout. Evaporation O~ may be considered by specifying a convective heat transfer coefficient between the sump liquid and the atmosphere vapor regions. However, since any heat transfer in this mode is small, a conservative coefficient of zero is assumed. Each region is assumed to be homogeneous, but a temperature difference can exist between regions. Any moisture condensed in the vapor region during a time increment is assumed to fall immediately into the sump region. Noncondensible gases are included in the vapor region. The COPATTA model also includes such engineered safety features as containment sprays and air coolers. The structural heat sinks can also remove energy from the atmosphere.
/~T a. Bechtel Topical Report BN-TOP-3, Revision 4, " Performance
(-) and Sizing of Dry Pressure Containment." 13
'r
[ ' The following main steam line breaks were analyzed:
%))
Case 1: Full DE Rupture at 102% Power Case 2: 0.7 ft2 DE Rupture at 102% Power Case 3: 0.6 fta DE Rupture at 102% Power Case 4: 0.645 ft2 Split Rupture at 102% Power Case 5: Full DE Rupture at 70% Power Case 6: 0.6 ft2 DE Rupture at 70% Power Case 7: 0.5 ft2 DE Rupture at 70% Power Case 8: 0.681 ft2 Split Rupture at 70% Power Case 9: Full DE Rupture at 30% Power Case 10: 0.5 ft2 .DE Rupture at 30% Power ' Case 11: 0.4 ft2 DE Rupture at 30% Power Case 12: 0.7065 ft2 Split Rupture at 30% Power r Case 13: Full DE Rupture at Hot Shutdown Case 14: 0.2 ft2 DE Rupture at Hot Shutdown Case 15: 0.1 fta DE Rupture at Hot Shutdown Case 16: 0.3 ft2 Split Rupture at Hot Shutdown This represents a complete spectrum of the worst case main steam line breaks. The mass and energy release data used in this analysis ! [_s was generated using the methods given in Appendix A of ; the Safety Analysis Standard 12.2, revision 2, of 1 WCAP-8822 a3 and WCAP-8860. The blowdown is used t in conjunction with a degraded containment spray and e one air cooler to allow for a single failure of one ' diesel generator. B. Results
.z. t Figures IV-1 through IV-32 show the containment i pressure and temperature results for this analysis. !
The peak pressure obtained was 46.1 psig for case 16, a 0.3 ft2 split rupture at hot shutdown. The highest t temperature reached was 378*F for case 3 and case 4, a ' O.6 ft2 double ended rupture at 102% power and a 0.645 l ft2 split rupture at 102% power, respectively. ' i i
- a. WCAP-8822 (Proprietary), R. E. Land, " Mass and Energy ;
Release Following A Steam Line Rupture," September 1976.
- b. WCAP-8860 (Non-Proprietary), R. E. Land, " Mass and Energy i N
Release Following A Steam Line Rupture," September 1976. 14
I C. Conclusions
])
The above results show that the maximum pressure in the containment is below the containment design pressure of 54 psig. In addition, the components j covered by I.E. Bulletin 79-01B and NUREG 0588, and . required for safe shutdown and accident mitigation, maintain their environmental qualification for the resulting temperature and pressure profiles inside the containment as determined by the above analysis. t 4 S
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V. Plant' Modifications ([ No modifications are currently ~ anticipated other than changes to the BIT recirculation system and the BIT heat tracing as a-result of reducing the boron concentration in the BIT to O ppm. Since the boron concentration will be reduced to O ppm, the BIT recirculation system and the BIT heat tracing are no longer necessary to maintain the boron in solution and may be changed to facilitate operational efficiency. O I O 16
-[x-} VI. Technical Specifications The analyses presented'in sections-III and IV of this report concluded that removal of the BIT boron concentration requirements can be accommodated within existing plant design criteria and safety limits. In this regard, it is proposed to delete paragraphs 3.5.4.1, 4.5.4.1, and Bases paragraph 3/4.5.4, concerning BIT boron concentration requirements, from the plant Technical Specifications. It is also proposed to delete paragraphs 3.5.4.2 and 4.5.4.2, concerning the heat tracing requirements of the BIT system, from the plant Technical Specifications. Additionally, Bases paragraph 3/4.5.4, concerning the refueling water storage tank, was modified to include a discussion of the contribution of the borated water from the tank to counteract any positive increase in reactivity caused by reactor coolant system cooldown transients. These proposed changes to the Technical Specifications for Units 1 and 2 are presented in figures VI-l through VI-6.
I e 17
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1 i
') EMERGENCY C' ORE COOLING SYSTEMS (G
l 3/4.5.4 BORON INJECTION SYSTEM i BORON _ INJECTION TANK
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TilIS SPECIFICATION. DELETED O V FARLEY-UNIT 1 3/4 5-9 O JOSEPH M. F ARLEY NWLE AH PL ANr CIPICATION CHANGE AllllX10111 POWCf UNIT 1 AND UNIT 2 FIGURE VI-l 44401
EMERGENCY CORE COOLING SYSTEMS IIEAT TRACING TilIS SPECIPICATION DELETED 4 O FARLEY-UNIT 1 3/4 5-10 f O A .x>stPH M. F Antr y PROPOSED TECl!NICAL SPECIFICATION cirnNGEs Alal)a 11tilk)tyt'r A uUcteAnetANT UNIT I AND UNIT 2 f FIGURE VI-2 4440 3 l
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i [- ,') EMERGENCY CORE COOLING SYSTEMS V 3ASES _ The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance requirements for throttle valve
, position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA.
Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. 3/4.5.4 BORON INJECTION SYSTEM THIS SPECIFICATION DELETED 3/4.5.5 REFUELING WATER STORAGE TANK () The OPERABILITY of the Refueling Water-Storage Tank (RWST) as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture. The OPERABILITY of the RWST as part of the ECCS also ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain suberitical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. 4 FARLEY-UNIT 1 B 3/4 5-2 1 JOSEPH M, FARLEY
" = ^"'^"' SI'ECIFICMION CHANGES AlabamaPower A UNIT 1 ANO UNIT 2 FIGURE VI-3 l 444&i
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.x)steH M. rant ry PROPOSED TECIINICAL Altilxt:11tilbwer uoct 4n et4ur sPEcIPIcATron cliancEs UNIT 1 AND UNir 2 FIGURE VI-4 44403
( EMERCENCY CORE COOrdtJG SYSTEMS II_ EAT TRACING l TilIS SPECIPICATION DELETED 1 O l l FARLEY-UNIT 2 3/4 5-10 () PROPOSED TECilNICAL EIE7^"a$^^"a" 3"'C'"' ^T' " "^" "" AlabamaIbwer A UNIT 1 AND UNif 2 FIGURE VI-5 44408
_m v) EMERGENCY CORE COOLING SYSTEMS BASES The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assemptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that prbper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. 3/4.5.4 BORON INJECTION SYSTEM Tl!IS SPECIFICATION DELETED 3/4.5.5 REFUELING WATER STORAGE TANK bi ( ,/ The OPERABILITY of the Refueling Water Storage Tank (RWST) as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown. RCS cooldown en be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture. The OPERABILITY of the RWST as part of the ECCS also ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condition following mixing of the RWST and RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. FARLEY-UNIT 2 B 3/4 5-2 i A./' JOSEPH M' FARLEY
"""'^""*"I SPECIFICATION CIIANGES AlabaillaIDWCf UNIT 1 AND UNIT 2 FIGURE VI-6 44443
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