ML20207J106
| ML20207J106 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 08/24/1988 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20207J102 | List: |
| References | |
| NUDOCS 8808300229 | |
| Download: ML20207J106 (30) | |
Text
{{#Wiki_filter:. . . , e.,. Attachment 1 Proposed Technical Specification Chango i 8808300229 G80824 PDR ADOCK 05000413 P PNU i l
' ' ' .i '
PLANT SYSTEMS STEAM GENERATOR POWER OPERATED RELIEF VALVES LIMITING CONDITION FOR OPERATION . 3.7.1.6 Three steam generator power-operated relief valves (PORVs) and associated remote manual controls, including the safety-related gas supply systems, shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4.* ACTION:
- a. With one less than the required steam generator PORVs OPERABLE, restore the inoperable steam generator PORV to OPERABLE status within 30 days; or be in at least HOT STANDBY within the next-12 hours.
- b. With two less than the required steam generator PORVs 0PERABLE, restore at least one of the inoperable steam generator PORVs to ,
OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS l 4.7.1.6 Each steam generator PORV and associated remote manual controls including the safety-related gas supply systems shall be demonstrated OPERABLE:
- a. At least once per 24 hours by verifying that at least one of the two nitrogen bottles associated with each PORV has a pressure greater than or :
equal to 2100 psig, and
- b. At 1 cast once per 18 months and prior to startup ,
following any refueling shutdown by verifying that all steam generator PORVs will operate through one cycle of full travel using remote manual controls and safety-related gas supply. i l
*When steam generators are being used for decay heat removal.
CATAWBA - Units 1 & 2 3/4 7-9a
?
l 1 1 PLANT 5YSTEMS SASES 3/4.7.1.7 AUXILI RY FEEDWATER SYSTEM The OPERAaILITY of the Auxiliary Feedwater System ensures tnat the Reactor- .l l Coolant..Systas can ne cooled down to less than 350*F from normal operating l
. conditione irt the event of a foematerline-greek accident wittr a'werst case '
A *9' R **'.I'* W" l i' ' . . .
'.Ns ? .
.' k .
The Auaitiary Feedwater Systas is caoatier oil deHvering a total feedwater fiew orf at least 492. gas at a pressure of 1210 psig ta the entrance of at least two of the steam generators. This capacity is sufficient to ensure tnat aceouate feedwater flow is available to remove decay heet and reduce the Reector Coolant Systas tamperature ta less taan 350*P when the Residual Meet .Resevel ~ - Sysstas.sey
~' '
~
. s ue.placed- inta .coaration.-
3/4.T.1.3 SPECTFIC ACTTVTTY The limitations on Secondary Coolant System specific activity ensure that, the resultant offsita radiation dose will be 11mitad to a san 11 fraction of tr 10 CTR Part 100 dose guideline values in the* event of e- stems line rupture. t This dose ,also includes tne effects of a coincident 1 gpe reactar to secondary tune leek in. the steam generator of the affected staas line. These values are consistent with the assumptions used in the safety analyses. . 3/4.7.1.4 MAIN STEAM LINE ISOLATION VALVES y The QPBASILITY of the sein staas line isolation valves ensures that no g wre than one staan generstar will blow down in the event of a steen line N rupture. This restriction is required tar (1) sinimize the positive reac-tivity effects of the Reactor Coolant Systas coeldown assoctatet wita the b blowdown, and (2) limit the pressure rise within containment in tae event the x staas line rupture occurs within containment. The OPERA 8ILITY of the sein staas isolation valves within the closure times of tne Surveillance Require- l i ments are consistant wita the assumptions used in the safety analyses, . . 3/4.7.1.5 CONCEMSATT STORAGE SYSTEM The 07ERASILITY of the Condensate Starage Systas with the einimum water g volume ensures that sufficient water is available to maintain the Reactor Coolant systas at HOT STANCBY concitions for 2 hours followed ey accroximately 5 nours cooldown with staan disenerge to the ataesphers concurrent witn total The contained water volume limit includes an allowance h loss-of-offsignt power.for water not usuaale because of tant, disenarge line focation or aractaristics, 3/4.7.2 STEAM GENERATOR ARESSURE/ TEMP DATURE t. IMITATION The limitation on steam generator pressure and'tencerature ensures that the pressure-induced stresses in the steam generators do not exceed tne maximum allowaale fracture tougoness stress limits. The limitations of.70*F and 200 psig are cased on a steam generator RTNOT of 60*F and are sufficient to prevent brittle fracture.
.QATAWSA - UNITS 1 ANO 2 __
8 3/4 7-2 -----.--_-____ __ __ - _ - __ _ _ _ _
- o. ,
INSERT FOR PAGE-B 3/4 7-2: The Surveillance Requirement for the Main Steam power-operated relief valves (PORV's) nitrogen supplies ensures that the PORVs will be' available to mitigate the consequencesaof a steam generator tube rupture accioent concurrent with loss of offsite power. This assunes that the , PORV o.' the ruptured steam generator is unavailable,-and that the other two are usci to cool the Reactor Coolant System inventory to less than the saturation temperature of the ruptured steam generator. Concurrent with the requirement that a specific number of PORVs be OPERABLE is the requirement that the associated PORV block valves upstream be open.or OPERABLE. Should an associated PORV block valvo be closed and inoperable, the PORV downstream of that block valve should also be considered inoperable and the applicable ACTION statement shall be entered until such time that the block valve is opened or returned to OPERABLE status. f
l Attachment 2 Steam Generator Tube Failuro Analysis i ( f l 1 i t f i i f I l
l CNS ,
)
of these postulated events, the classification ' system of ANSI-N18.2-1973 is ; utilized. The design of "systems important to safety" (including protection i systems) is consistent with IEEE Standard 379-1972 and Regulatory Guide 1.53 in I the application of the single failure criterion. All of the transients analyzed in Chapter 15 are analyzed assuming the most limiting single failure (e.g., loss of one protection signal or SI train failure). For the incidents of moderate frequency, including the complete loss of forced reactor coolant flow, the analysis shows that the DN8R remains above the limit value. Therefore, no fuel failure will occur. Table 15.0.8-2 lists'all the incidents of moderate frequency along with the worst single failure assumed and the effect such an assumption has on the results. For most accidents, the worst single failure has no effect at all, since the logic of the protection system (e.g., 2 out of 4) is designed to account for this. The basis for performing the Chapter 15 analysis includes allowing for the worst single failure in the protection system. For example, the electrical aspect'of this criterion, as stated in IEEE 279-1971, is met that "any single failure within the protection system shall not prevent proper protective action at the system level when required". The Catawba FSAR meets the worst single failure criterion. The SRP states that "an incident of moderate fre-quency in combination with any single active compocent failures, or single operator error, should not cause loss of function cf any barrier other than the fuel cladding". This criterion is met in the FSAR by the fulfillment of , 1 the design basis for incidents of moderate frequency. I In the analysis of the Chapter 15 events, control system action is considered only if that action results in more severe accident results. No credit is taken for control system operation if that operation mitigates the results of l an accident. For some accidents, the analysis is performed both with and N N33! C3 N 3) ! y pe opgration }g,dggermgg_the worst case. ] : y - :gj :: 15.0.8.2 Operability of Non-Safety Grade Equipment The transients analysis presented in Chapter 15 only assume non-safety. grade systems and equipment are operable in the following situations:
- 1) When operation of the system will cause the transient to be more severe.
If there is doubt about the system's effect, the transient is analyzed i and presented with and without the system available.
- 2) When a loss of a non-safety grade system initiates a transient by itself, it is not superimposed upon other transients unless there is a credible reason that one would cause the other. I The following non-safety grade systams and equipment are assumed operable in i some analyses presented in Chapter 15. I i
- 1) Automatic rod control 1
- 2) Pressurizer pressure control (power operated relief valves and spray) I 1
1 15.0-10 l i
.m v.-- g ,. m-,- -,,-- - , - , , e_-m,--.-,
i.., ... e {N,5 h [ressarie<u I-c.M b - V 3) Main Feedwater System AUTOMATIC R00 CONTROL 9 h The Automatic Rod Cont'rol System is assumed to be operable in the following
-4 transient analysis:
i 15.1.3 Excessive increase in secondary steam flow. 15.4.3 Rod Cluster Control Assembly Misoperation
+1 y v
15.6.1 Inadvertent opening of a pressurizer safety or relief valve.
~
Analyses of the excessive increase in secondary steam flow transient are done i , b with and without automatic rod control and are presented in Section 15.1.3. e Both cases meet acceptance criteria. c% 4 For the dropped rod (s) case of Section 15.4.3 the automatic rod control Oh ( system is assumed to be operable. Because of the power overshoot possibility N associated with automatic control, it is conservative to assume its function
-{j 4 " this accident, i
a p For the inadvertent opening of a pressurizer safety or relief valve transient fq it is conservative to assume the automatic rod control system is operable. E During this transient RCS ' pressure will be decreasing. Decreasing RCS pres-
. C, I sure will cause the reactor power to decrease due to moderator density feed-
- back. If the Automatic Rod Control System is operable it will function to 3 maintain power and average coolant temperature, thus causing a more severe transient. gg o -
PRESSURIZER PRESSURE CONTROL (Power operated Relief Valves [=d 0;;ny) E
.3 $ The Pressurizer Pressure Control System is assumed to be operable in the 4 following transients:
}d 15.2.3 Turbine trip h 15.4.2 Uncontrolled rod cluster control assembly bank withdrawal at. power D4 O
\5.4.1 M e m se.nas d w b b . < W e Analysis of the turbine trip event with and without pressurizer pressure i
o% h d control is presented in Section 15.2.3 of the FSAR. Both cases meet acceptance criteria. I For the rod bank withdrawal transient it is conservative to assume pressurizer 4 pressure control is operable. The limiting criteria for this transient is the , t. DN8 ratio. Maintaining RCS pressure low will result in lower DNB ratios. 4 4
$ MAIN FEE 0 WATER CONTROL SYSTEM
, .5 8 g The Main Feedwater Control System is assumed to function during the following transient analyses: i
- 15.1.1 Feedwater System malfunctions causing a reduction in feedwater d o temperature l h T.
15.0-11
. - -- . _ . - - _- . - = - - . - - -
C N,$, l
. l 15.1.2 Feedwater System malfunction causing an increase in feedwater flow 15.1.3 Excessive increase in secondary steam flow 15.1.5 Steam system piping failure 15.4.2 Uncontrolled rod cluster control assembly bank withdrawal at power 15.4.4 Startup of an inactive loop -
15.5.1 Inadvertent operation of Emergency Core Cooling System during power operation 15.6.1 Inadvertent opening of a pressurizer safety or relief valve. Loss of main feedwater flow is a Condition II occurrence by itself and is ! analyzed in Section 15.2.7. There is no credible' reason for any of the Condition II events listed above to cause a loss of feedwater flow. There-fore, a loss of feedwater is not considered coincidently with those occurances listed above which are C1bndition II. , l , For the steamline break transient it is conservative to assume main feedwater is available. This maximizes the amount of steam generator. inventory avail-able to be blowdown and prolongs the transient. i Iv.56AT The respons'e_ times and discharge rates for some important plant valves and - g i umps are listed in Table 15.0.8-3. 15.0.9 FISSION PRODUCT INVENTORIES 15.0.9.1 Inventory in the Core The fission product radiation sources considered to be released from the fuel to.the containment following a maxi um credible accident are based on the assumptions stated in TID-14844 (Reference 1), namely 100 percent of the noble gases, 50 percent of the halogens and a core power level of 3565 MWt.
, The time-dependent fission product inventories in the reactor core are calculated by the ORIGEN Code (Reference 2) using a data library based on ENOF/B-IV (Reference 3). The core inventories are shown in Table 15.0.9-1.
i The Equilibrium Appearance Rate of Iodines in the RCS due to conservative and realistic fuel defects are shown in Table 15.0.9-2. 15.0.9.2 Inventory in the Fuel Pellet Clad Gap l i The radiation sources associated with accidents which may cause more than 1 percent failed fuel (loss of coolant accident, rod cluster control assembly ejection, and fuel handing accidents) are based on the assumption that the fission oroducts in the gap between the fuel pellets and the cladding of the damaged 'uel rods are released as a result of cladding failure. The gap activities were determined using the model suggested in Regulatory Guide 1.25. Specifically, 10 percent of the iodine and noble gas activity 15.0-12 -
1 csc a.T voe. F seR- Pese l 5 . o - 12. 1 I l l pre %utI%ER L EVEL Co4~rkOL
% 9eessaris<r Lass ( dnhal Sy.sk i.r asswned A k eprae in de. analysis el sl cam gene <A-dule ruf f ure in Sech.n i.s. 4 3 rrhe ol.&fW/
m<k<.up Sem de cAayiy pwys b /ny.c reacA,- h9 -J -A4ece-Ere inaww pruucy A s< w Jary kaby a. l l l l l 4 l
. - , , - . . - - . . ,_._,-,_.s-, . - - - . - - - - . . . , _ - - - - - - , _ . - . . - . . - , -
CNS uses six delayed neutron groups and contains a , detailed multi-region fuel-clad-coolant heat transfer model for calculating pointwise Doppler and moderator feed-back effects. The code handles up to 2000 spatial points, and performs its own steady state initialization. Aside from basic cross section data and
. thermal-hydraulic parameters, the code accepts as input basic driving functions such as inlet temperature, pressure, flow, boron concentration, control rod motion, and others. Various edits are provided, e.g. channelwise power, axial offset, enthalpy, volumetric surge, pointwise power, and fuel temperatures.
The TVINKLE Code is used to predict the kinetic behavior of a re% tor for tran-sients which cause a major perturbation in the spatial neutron flux distribution. TVINKLE is further described in Reference 10. 15.0.11.4 THINC The THINC Code is described in Section 4.4. N 15.0.12 ENVIRONMENTAL CONSEQUENCES A summary of the offsite doses is presented in Table 15.0.12-1. A description of each accident analysis is given in the appropriate section.
" $0. U.S i W RITRAN-02 c1 y FITRA.N-02 a code capable of simulating most thermal-hydraulic transients of interest in both PWRs and BWRs. PITRA.N-02 has the flexibility to model any general fluid system by partitioning the system into a one-dimensional network of fluid volumes and connecting flowpaths or junctions. The mass, momentum, and energy equations are then solved by e= ploying a semi-implicit solution technique. The time step selection logic is based on algorithms that detect rapid changes in physical processes and limit time steps to ensure accuracy and stability. Although the equations describe homogeneous equilibrium fluid volumes, phase separation can be modeled by separated bubble-rise volumes and by a dynamic slip model. The pressuri:er and other volumes can be modeled as non-ewilibrium volu=es when I
such phenomena are present. Reactor power generation can be represented by either a point kinetics model or a one-dimensional kinetics model. Heat transfer across steam generator tubes and to or f rom structural ccmponents can i be modeled. Special component models for centriugual pumps, valves, trip logic, control systems, and other features useful for fluid system modeling are available. 15.0-15
, {
Table 15.0.3-1 Nuclear Steam Supply System Power Ratings t Transients Analyzed at 0 W t Hot zero thermal power 0 Reactor coolant pump thermal oower (assumed) 0 Transients Analyzed at 3427 Wt Guaranteed core thermal power 3411 Reactor coolant pump thermal power 16 Transients Analyzed at 3495 Wt i Guaranteed core thermal power (+2 percent) 3479 Reactor coolant pump thermal power 16 Transients Analyzed, at 3652 Wt Engineered safety features design rating (+2 percent) 3632 (maximum calculated turbine rating) Reactor coolant pump thermal power 20 Transients Analyzed at 3479 Wt l Guaranteed core thermal power (+2 percent) 3479 Reactor coolant pump thermal power 0 i c :
%mtenh AnQarcl. ed 3499 % "
guem6A we. bad p~u u 2. eu-Q 3ng Reu.br cooW4 pume Ocmal Pecer 1987 ' e
Table 15.0.3-3 Summary of Computer Codes and Methodologies Used in Accident Analyses j Computer Code Transientst Analyzed with that or Methodology Computer Code or Methodology , LOFTRAN 15.1.2*, 15.1.3*, 15.1.4, 15.1.5*, 15.2.3*, 15.2.6, 15.2.7, 15.2.8*, 15.3.1*, 15.3.2*, 15.3.3*, 15.4.2*, 15.4 + &- 15.4.4, 15.5.1, 15.6 ,- 4 4,4 FACTRAN 15.3.1, 15.3.2, 15.3.3, .4.1, 15.4.4, 15.4.8* THINC 15.3.1, 15.3.2, 15.4.1,~ 15.4.3*, 15.4.4 TWINKLE 15.4.1, 15.4.8* TURTLE 15.4.3*, 15.4.7*, 15.4.8* LEOPARD 15.4.3*, 15.4.7* C BASH 15.6.5 a g e
?
NOTRUMP 15.6.5 a-c <, SATAN-VI .5.6.5 d g WREFLOOD 15.6.5 d g POWLOCTA 15.6.5 d g LOCTA-IV 15.6.5* LOTIC-2 15.6.5 d g PAD 15.6.5 d g gcTAAV'N 15.6.3 + ITDP 15.1.2*, 15.1.3*, 15.2.3*, 15.3.1, 15.3.2, 15.4.2*, 15.4.3*, 15.4.4, 15.5.1, 15.6.1, WRB-1 15.1.2*, 15.1.3*, 15.2.3*,
. 15.3.1, 15.3.2, 15.4.1, 15.4.2*,
15.4.3, 15.4.4, 15.5.1, 15,6.1 W-3 15.1.4, 15.1.5* , TTransients are numbered according to the cases listed in Table 15.0.3-2. r I
*All cases of this transient used the computer code.
1987 Update 3
Table 15.0.3-4 (Page 3) Summary of Input Parameters for Accident Analyses Initial Moderator Moderator NSSS RCS Vessel Pzr Pzr Water Feedwater FSAR Temperature Density Output Flow T,yg Press. Volume Temperature Section (pcm/*F) (%ik/k/g/cc) Doppler (Mwt) (gpe) @ (psia) 3 (ft ) Q 15.4.4 NA 0.43 Minimum 2339 Note 12 580.7 2280 891 400 f 15.4.7 a NA NA NA 3427 NA NA NA NA NA 15.4.7 b NA NA NA 3427 NA NA NA NA NA 15.4.7 c NA NA NA 3427 NA NA NA NA NA 15.4.7 d NA NA NA 3427 NA NA NA NA NA 15.4.8 a Note 10 NA Note 5 3495 373,200 595.5 2208 NA NA 15.4.8 b Note 10 NA Note 5 0 171,672 557 2208 NA NA 15.4.8 c Note 10 NA Note 5 3495 373,200 595.5 2208 NA NA 15.4.8 d Note 10 NA Note 5 0 171,672. 557 2208 NA NA 15.5.1 NA 0.O Minimum 3427 387,600 590.8 2250 1100 440 15.6.1 +7 NA Minimum 3427 387,600 591.5 2250 1100 440 3491 594 /227 V31 15.6.3 NA NA NA M' 373,49& 595.5 2280 6 -444 15.6.5 a Note 11 NA Note 11 34791 377,000 590.8 2280 1172 440 15.6.5 b Note 11 NA Note 11 34797 377,000 590.8 2280 1172 440 15.6.5 c Note 11 NA Note 11 34791 377,000 590.8 2280 1172 440 t Pumps were assumed to trip at start. of transient. 1987 Update - _ - - _ - - _ - - - - . _ ---,--_----.-.-,------,-------_+--m,.___a - w -
, . . a .. CNS 15.6.3 STEAM GENERATOR TUBE FAILURE 15.6.3.1 Identification of Causes and Accident Description I The accident examined is the complete severance of a single steam generator tube. The accident is assumed to take place at power with the reactor coolant , contaminated with fission products corresponding to continuous operation with a . limited amount of defective fuel rods. The accident leads to an increase in . contamination of the secondary system due to leakage of radioactive coolant from the RCS. In the event of a coincident loss of offsite power or failure of the condenser steam dump system, discharge of activity to the atmosphere takes-place via the steam generator safety and power operated relief valves. In view of the fact that the steam generator tube material is highly ductile ; Inconel-600, the assumption of a complete severance is considered somewhat conservative. The more probable mode of tube failure would be one or more Activi minor leaks of undetermined origin. System and is ansubject to continual accumulation of minorsurveillance,ty leaks i which exceeds the limit established in the Technical Specifications is not permitted during the unit operation. The operator is expected to determine that a steam generator tube rupture has occurred and to identify and isolate the affected steam generator on a re-stricted time scale in order to minimize contamination of the secondary system and ensure termination of radioactive release to the atmosphere from the affected unit. The recovery procedure can be carried out on a time scale which 1 ensures that break flow to the secondary system is terminated before water , level in the affected steam generator rises into the main steam piping (Refer- ! ence 14). Sufficient indications and controls are provided to enable the operator to carry out these functions satisfactorily. Immediately apparent symptoms of a tube rupture accident, such as falling pressurizer pressure and level and increased charging pump flow, are also , symptoms of small steam line breaks and loss of coolant accidents. It is ; therefore important for the operator to oetermine that the accident is a ! rupture of a steam generator tube, in order to carry out the correct recovery ! procedure. The accident under discussion can be identified by the following ' method. In the event of a complete tube rupture, the reactor coolant system - pressure decreases and the condenser air ejector radiation and/or steam genera-tor blowdown radiation monitors exhibit abnormally high readings. If the i containment pressure, containment radiation, and containment recirculation sump , level exhibit normal readings, then a steam generator rupture is diagnosed to ' have occurred. V Consideration of the indications provided at the control board, together with ; the magnitude of the break flow, leads to the conclusion that the accident ! diagnostics and isolation procedure can be completed within 30 minutes of reactor trip for the design basis event. Note that break sizes smaller than complete severance of a tube, with less break flow from primi.ry to secondary, exhibit a slower rise in steam generator j water level, and an increased time interval for actuation of the t' lowdown line 1 I l J
o.. . ..' CNS radiation monitor and the condenser air ejector radiation
- iherefore, more time may be available to the operator to diagnose the a w.cnt and take steps to isolate the ruptured steam generator.
If normal operation of the various plant control systems is assumed, the following events are initiated by a tube rupture:
- 1. Pressurizer low pressure and low level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the secondary side, steam flow /feedwater flow mismatch occurs as feedwater flow to the affected steam generator is reduced as a result of primary coolant break flow to that generator.
- 2. The decrease in RCS pressure, due to continued loss of reactor coolant inventory, leads to a reactor trip signal on low pressurizer pressure or overtemperature AT. The resultant plant cooldown following reactor trip leads to a rapid decrease in pressurizer level. A safety injection signal, initiated by low-low pressurizer pressure, follows soon after reactor trip. The safety injection signal automatically terminates normal feedwater supply and initiates auxiliary feedwater addition.
- 3. The steam generator blowdown liquid monitor and/or the condenser air ejector radiation monitor will alarm, indicating a sharp increase in radioactivity in the secondary system, and will automatically terminate steam generator blowdown.
- 4. The reactor trip automatically trips the turbine and, if offsite power is available, the steam dump valves open, permitting steam dump to the condenser. In the event of a coincident station blackout (loss of offsite power), as assumed in the transients presented in this section, the steam ,
dump valves autonatically close to protect the condenser. The steam i generator pressure rapidly increases resulting in steam discharge to the ; atmosphere through the steam generator safety and power operated relief valves. Steam flow as a function of time is constant initially until reactor trip. This is followed by turbine trip which results in a large , decrease in flow, but a rapid increase in steam pressure to the safety l valve setpoint. j i
- 5. Following reactor trip, the continued action of the auxiliary feedwater supply and borated safety injection flow (supplied from the RWST) provide a heat sink which absorbs the decay heat. j
- 6. Safety injection flow results in increasing pressurizer water level, the rate of increase depending upon the amount of auxiliary equipment operating. ,
A steam generator tube failure is classified as an ANS Condition IV event, a limiting fault. See Section 15.0.1 for a discussion of Condition IV events.
CNS i 15.6.3.2 Analysis of Effects and Consequences t
; Method of Analysis -
Detailed thermal-hydraulic calculations are performed to determine ~ primary to. , secondary mass release and to determine the amount of steam vented from each of the steam generators, using RETRAN-02 (Reference 15) as described in Reference ! 16. } l In estimating the mass transfer from the RCS.through the broken tube, the following assumptions are made* e
- 1. Reactor trip occurs _ manually at thirty (30) minutes. Loss of offsite power occurs at reactor trip.
- 2. Following the initiation of the safety injection signal, two high-head
, safety injection pumps are aligned to the safety injection flowpath and two intermediate-head safety injection pumps are actuated. These pumps i continue to deliver flow until safety injection is manually terminated by l l the operator. t I
- l j
- 3. After reactor trip, break flow reaches an equilibrium when it is balanced byincomingsafetyinjectionflowas.showninFigure15.6.31. The resultant break flow continues at approximately the same value from plant .I 3 trip until pressures are equalized. Operator actions are modeled to ,
i terminate break flow. - The above assumptions, extremely conservative for the design basis tube rup- ! j ture, are made to maximize doses and do not model expected operator actions for recovery. Plant characteristics and initial conditions are discussed in ; j Section 15.0.3. Results i The results are shown in the following figures: Figure 15.6.3-1 Break Flow i i
; Figure 15.6.3-2 Reactor Coolant System Pressure l l Figure 15.6.3-3 Reactor Coolant System Temperatures (For Ruptured loop) l j Figure 15.6.3-4 Reactor Coolant System Temperatures (For Intact Loops) i Figure 15.6.3-5 Pressurizer Water level j i Figure 15.6.3-6 Steam Line Pressure Figure 15.6.3-7 Steam Generator Water Levels
} The sequence of events is presented in Table 15.6.3-1, 15.6.3.3 Environmental Consequences
] The postulated accidents involving release of steam from the secondary system i
do not result in a release of radioactivity unless there is leakage from the RCS to the secondary system in the steam generators. A conservative analysis
; of the postulated steam generator tube rupture assumes the loss of offsite power. Thi:, causes the loss of main steam dump capabilities and the subsequent
{ .E
CNS venting of steam from the secondary system to the atmosphere. A conservative -
. analysis of the potential offsite doses resulting from this accident is pre-1 sented assuming primary to secondary leakage. This analysis incorporates ,
assumptions of 1 percent defective fuel and steam generator leakage of 1 gpm prior to the postulated accident for a time sufficient to establish equilibrium specific activities in the secondary system. Three postulated cases are i analynd: i Case 1: Normal equilibrium Technical Specification iodine concentrations l exist at the tiu.e of the accident.. Case 2: There is a pre-existing iodine' spike at the. time the accident. occurs. The reactor coolant concentrations are _the maximum i permitted for full power operation. l l Case 3: There is a coincident iodine spike at the time the accident occurs. The iodine concentrations are found by increasing the equilibrium appearance rate in the coolant by 500. l The primary and secondary coolant activities prior to the accident correspond to limits set by Technical Specifications. The following assumptions and parameters-are used to calculate the activity ! 4 release and offsite dose for the postulated steam generator tube rupture:
- 1. Prior to the accident, an equilibrium activity of fission products exists r in the primary and secondary system caused by primary to secondary leakage ,
in the steam generators. r J
- 2. The accident is initiated by the rupture of a steam generator tube, which results in the transfer of approximately 173,655 pounds of reactor coolant into the shell side of the defective steam generator.
- 3. Offsite power is lost, i i ;
; 4. The primary to secondary leakage is 1.0 gal / min in the nondefective ! steam generators. ]
- 5. The steam release fron, the defective steam generator terminates in 3890.3 j seconds. The release from the nondefective steam generators terminates in !
8 hours. l
- 6. All noble gases which leak to the secondary side are released.
- 7. The s'.eam generator iodine partition factor is 0.01 during the accident, except during the period frcm 30-45 minutes when tube uncovery is possible, i
During that period a steam generator iodine partition factor of 1.0 is used i for flashing flow and a partition factor of 0.01 is used for non-flashing j flow. l l l Y
CNS-
- 8. For Case 1, the primary coolant concentration is at the equilibrium Technical Specification limit.
- 9. For Case 2, the primary coolant concentration is at the maximum permitted for_ full power operation.
- 10. For Case, 3, the iodine spike occurs at the onset of the accident and continues for the duration of the accident. The-iodine concentrations are determined by increasing the equilibrium appearance rate by 500,
- 11. Other assumptions are listed in Table 15.6.3-2.
Based on the foregoing model, the thyroid and whole body doses'are calculated at the exclusion area boundary and the low population zone. The results are presented in Table 15.6.3-2. .The doses at these distances are below the regulatory limits of the 25 rem whole body and 300 rem thyroid doses estab- ' lished in 10 CFR 100. l l l l I 1
CNS 1 REFERENCES FOR SECTION 15.6 l
- 1. Burnett, T. W. T., et, al., "LOFTRAN Code Description", WCAr-7907, June 1972.
- 2. Chelemer, H., Boman, L. H., Sharp, D. R., "Improved Thermal Design Proce-dures", WCAP-8587, July 1975.
- 3. "Acceptance Criteria for Emergency Core Cooling System for Light Water Cooled Nuclear Power Reactors , 10 CFR 50.46 and Appendix X of 10 CFR 50.
Federal Register, Volume 39, Number 3 January 4, 1974.
- 4. Bordelon, F. M., H. W. Massie, and T. A. Borden, "Westinghouse ECCS <
Evaluation Model-Summary", WCAP-8339, (Non-Proprietary) July 1974. !
- 5. Bordelon, F. M. , et. al. , "SATAN-VI Proaram: Comprehensive Space Time Dependent Analysis of Loss of Coolant",'WCAP-8302, (Proprietary) June l 1974, and WCAP-8303 (Non-Proprietary) June 1974.
l 6. Kelly, R. O. , et. al. , "Calculated Model for Core Reflooding Af ter a Loss of Coolant Accident (WREFLOOD Code)", WCAP-8170 (Proprietary) and , i WCAP-8171 (Non-Proprietary), June 19-74. 4
- 7. Hsieh, T. , and Ray'mond, M. , "Long Term Ice Condenser Containment LOTIC .
i Code Supplement l , WCAP-8355 Supplement 1, May 1975, WCAP-8354 (Proprie-ta y), July 1974.
- 8. Bordelon, F. M., et. al., "LOCTA-IV Program: Loss of Coolant Transient Analysis", WCAP-8301, (Proprietary) and WCAP-8305, (Non-Proprietary) June 1974.
1 i 9. Rahe, E. P., Westinghouse letter to Thomas, C. O., U.S.N.R.C., Letter Number NS-EPR-2673, October 27, 1982,
Subject:
"Westinghouse Revised PAD l Code Thermal Safety Model", WCAP-8720, Addendum 2 (Proprietary).
i 10. Westinghouse ECCS Evaluation Model, 1981 Version", WCAP-9220-P-A, Rev. 1 j (Proprietary), WCAP-9221-A, Rev.1 (Non-Proprietary), February,1982,
- 11. Lee, H. Tauche, W. D., Schwarz, W. R., "Westinghousa Small Break ECCS Evaluation Model Using the NOTRUMP Code", WCAP-10081-A, August 1985, i 12. Salvatori, R. , "Westinghouse Emergency Core Cooling System - Plant Sensi-l tivity Studies", WCAP-8340, (Proprietary) July 1974, i 13. Young, M., et al., "BART-1A: A Computer Code for the Best Estimate Analyzed Reflood Transients", WCAP-9561-P-A,1984 (Westinghouse
} Proprietary). i 14. i Attachment 5 to December 7,1987 letter from Duke Power (H. B. Tucker) to NRC, "Overfill Evaluation for Steam Generator Tube Rupture". l 1 s
CE
- 15. J. H. McFadden, et. al., "RETRAN A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems", NP-1850-CCH-A, Revision 3, June 1987.
- 16. "Thermal-Hydraulic Transient Analysis Methodology", OPC-NE-3000, Duke Power Co., July 1987.
1
,.6, ....- CNS Table 15.6.3-1 Steam Generator Tube Rupture Sequence of Events - Time Event (minutes) Tube rupture occurs. 0.0-Manual SI actuation, reactor trip, turbine trip, loss of offsite
. power 30.0 Steam'line PORVs and' lowest steam line safety. valves open 30.1 Charging flow isolated 30.5' Auxiliary feedwater flow delivered to steam generators 31.0 All steam relief valves except failed PORV have reseated 31.4 Operator closes block valve on failed PORV 39.2 Intact stcam line PORVs cycle to relieve pressure 47.4-59.9 Operator diagnoses accident and closes MSIV on ruptured steam generator' 56.6 Lowest safety valve on ruptured steam gAnerator cycles to relieve pressure 57.7-64.8 Operator begins cooldown 61.6 Cooldown completed 64.8 Operator begins depressurization 66.3 Break flow terminated as pressures equalize 67.6 I
l 1 l 1
1 g . y . . . ,. A s
~ CNS Table _15.6.3-2 (Page 1)
Parameters for Postulated Steam Generator-Tube Rupture Accident Analysis
- 1. Data and assumptions used to estimate radioactive source from postulated accidents'
- a. Power level (MWt) 3565.
- b. Percent of fuel defected 1.
- c. Steam Generator tube leak rate prior to and during accident (gpm) 1.
- d. Offsite power Not available
- 2. Data and assumptions used to estimate activity released .
- a. Iodine partition factor for initial 0.01 Non-flashing flow steam release from defective steam 1. 0 Flashing flow generator
- b. Iodine partition factor for steam 0.01 release from nondefective steam generators
- c. Iodine partition factor for air ejector 0.12
- d. Percent of flashing steam 0.16
- e. Steam release from defective steam 73,679 generator (1b)
- f. Steam release from three nondefective steam generators (lbs)
(0-2 hr) 339,192 (2-8 hr) 1,300,960
- g. Reactor coolant released to 173,655 defective steam generator (1bs)
- 3. Dispersion data l
- a. Distance to exclusion area boundary (m) 762.
- b. Distance to low population zone (m) 6096, t
,.g.... 1 i CNS l Table 15.6.3-2 -(Page 2) Parameters for Postulated Steam Generator . Tub'e Rupture Accident Analysis
- c. X/Q at exclusion area boundary (sec/m3) 5.5E-04 (0-2 hr)
- d. X/Q at low population zone (sec/m3) 1.8E-05 (0-8 hr)
- 4. Dose data
- a. Method of dose calculations Regulatory Guide 1.4
- b. Dose conversion assumptions Regulatory Guides 1.4, 1.109
- c. Doses (Rem) ,
Case.1 (No iodine spike) Exclusion Area boundary i Whole Body 3.1 E-1 Thyroid 1.5 1.ow population zone Whole Body 1.0 E-2 Thyroid 5.0 E-2 Case 2 (With pre-existing iodine spike) Exclusion area boundary Whole body 3.1 E-1 Thyroid 9.1 E+1 Low population zone Whole body 1.0 E-2 Thyroid 3.0 Case 3 (With coincident iodine spike) Exclusion area boundary Whole body 3.1 E-1 Thyroid 2.6 E+1 Low population zone Whole body 1.0 E-2 Thyroid 9.0 E-1
6 5" x N B R E ' -- E a
* ~
N[% N [am 0 W G P M iee e~pi'e v i , i . .. , ,, , ,,,,, ,, , ,,, ,,,, , ,,,,,,,,, ,,,,, ,,, e ie 29 M 40 64 6e 7e TIE (MICEst l STEAM GENERATOR TUBE RUPTURE um CATAWBA NUCLE.AR STATION Figure 15.6.3-1 i
I 2300 R 2100 P 1900 x R E S S 1788 y U R E 1548 P S I O 3ee i 11W ,, ,,,,,, e is as as o se s. 7. f!E (MIMJTES) l I i l STEAM GENERATOR TUBE RUPTURE Cmf CATMBA Figure 15.6.3-2 NUCLEAR STATICN l l e -- ,,-g.-----g, -- , ., , _ . - - , - .
,- - -,,,--m-- x - , - *, , , -
e
')
R U P t s4. 4 U R l E D { sie L 0 0 P 5"
\
E N E 558 h a u R SM E s D " E 4# g'' ' ' ' '' ''' e is a m 4e se 58 Tile (MIMJtESI 1 I 1 STEAM GENERATOR TUBE RUPTURE j mamm CATAWBA NUCLEAR STATICN < Figure 15.6.3-3
)
4
i-
- sy I
H T g4g A C l T t sie 0 0 P [" \
- N l
R 558
\ ~
\
i U R N %. E SM s D E ~ o dos .... . . ... .. . ... .....
, e is a se 4e se se re ;
f!tt (MIMJTES) i I l g STEAM GENERATOR TUBE RUPTURE fCATMANUCLEARSTATION Figure 15.6.3-4 l l
.o :e .,,
re P 69 R E S S 50 U R I 48 Z E R 3e L E Y 20 E L
\ I g 18 N e-.
L
....... . ..... f.. . . .... .........
e 19 29 30 44 58 68 79 TIME (MIMJTES) i STEAM GENERATOR TUBE RUPTURE
" ' t *" CATAWBA NUC1 EAR.STATICt1 Figure 15.6.3-5 i
1 el
l l
.s > .
l l S i2ee i E A i,
')*. P L I N
E 1000 P y :, a : E t s See 0 8 s . U S R .,,-- E g 4 P S I G 700 ~ I' ii., ,,,,,,,,, .. .. ,,,,,, ,,,,, y T , g 3 P I E 3 g y , , ,ig y
* 'e ,, , ,e se e, ,,
TIT (MINUTES) RUPTIAED INTACT ------ 1 l l l l l
@, "" W % NUCIIAR STEAM STATION GENERATOR TUBE RUPTUR Figure 15.6.3-6 i
l l
o
- Dg.,- t, l
t ies e s G H R 60 L 'o,,." E ,___________ ___________ ____________ ..,-s; W ,' 40 - '
- E L I x
2e .,' a a-il . i ill ,, . O - ,,,,,,,,,, ,,,,,,,,, ,,,,,,,,, ,,,,,, ,, ,,,,,,,,, ,, ,,,,,, ......... 0 10 29 30 48 58 69 7e TIME (NINUTES) RUPTURED I NT ACT ---- 1 1 1 i 1 l STEAM GEf1ERATOR TUBE RUPTURE
! CATAWBA NUCLA STATION Figure 15,6.3-7 1
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