ML20070L440
| ML20070L440 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 03/14/1991 |
| From: | Tuckman M DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-73771, TAC-73772, NUDOCS 9103200002 | |
| Download: ML20070L440 (19) | |
Text
i j1%n C6tnparg 4 1HDr n
- Nuclear 1%Inction Dept Oce 11rsident l'G lim Juni Nuckor Operutions s
' Charkorte. N C 282011907 i701)373 M1 -
DUKE POWER March 14, 1991 U. S. Nuclear Regulatory Commission l
ATTN: Document Control Desk i
Washington, D.
C.
-20555
Subject:
McGuire Nuclear Station Docket Numbers 50-369 and ~370 Catawba Nuclear Station Docket Numbers 50-413 and -414 Response to. Conditions Relativc tc the Use of Topical Report BAH-10173 (TAC Nos. 73771/73772)
By letter dated February 20, 1991, Mr. Robert Martin' transmitted the NRC staff's Safety Evaluation Report (SER) for topical report BAW-10173.
The SER found the Topical Report to be acceptable for-referencing in support of future reloads,-provided 5 conditions which the Staff requested be-addressed were met.
Attached please find responses to Conditions-1-4 as-specified-in the SER.
Note-that the response to Condition 5,
evaluation of radiological consequences of locked-rotor and single RCCA withdrawal events, has been provided previously to the staff in the Catawba Unit 1 Cycle 1
6 reload design submittal dated January-9, 1991.
1 If there are any questions,.please call Scott Gewehr atl(704) 373-75B1.
Very truly yours,
.)
h
-r M.
S. Tuckman topcond/ sag I f.
[
a
n Nuclear Regulatory Commission March 14, 1991 Page 2 cc:
Mr. T. A. Reed, Project Manager office of Nuclear Reactor Regulation U.
S. Nuclear Regulatory Commission Mail Stop 9H3, OWFN Washington, D.
C.
20555 Mr. R.E. Martin, Project Manager office of Nuclear Reactor Regulation-U.
S. Nuclear Regulatory Commission Mail Stop 9H3, OWFN Washington, D.
C.
20555 Mr.
S.
D.
Ebneter, Regional Administrator-U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street, NW - Suite-2900-Atlanta, Georgia. 30323 Mr. Frank McPhatter.
B&W Fuel-Company 3315 Old Forest-Road P.
O.
Box 10935 Lynchburg, VA-24506-0935 l
l l
L l s 1
Attachment I Responses to Conditions 1-3 for Use OF BAW-10173 Condition 1:
A benchmark analysis should be performed to verify appropriateness of the LYNXT modifications and modeling for the analysis of steam line break with loss of offsite power (SLB-LOOP).
The benchmark analysis should-be provided = for staff review to confirm the acceptability _ of the SLB-LOOP analysis results.
Response 1:
An analysis of the steamline break transient with-of fsite power was provided in Reference 1.
In response to a staff
- question, an analysis without offsite power _(LOOP) was q
provided in-References 2 and 3.
For_the case where offsite power-ir.~available, full reactor coolant-flow is maintained during the entire transient, and the and the analysis _was performed with the single-pass 5-channel 1/8 core LYNXT model (Reference 4),'which had been reviewed -previously as an acceptable LYNXT model.
For the LOOP-case, however, the core thermal-hydraulic. conditions are characterized by natural-temperature and power peaking conditions. _ Therefore,' for this-case a
modification-was-made to the LYNXT code by incorporating the implicit pressuro-velocity -(PV)' algorithm to ensure a convergent solution.
LIn addition, a nine-channel half-core model was used with the core modeled as three distinct regions; a cold region to. represent the faulted quadrant, a hot region to represent the combined intact
. quadrants, and a mix region to. provide an interface between the hot and cold core areas.
As discussed above, analysis of the SLB-LOOP. case required a larger, more detailed model than the full-flow case and an alternate solution scheme, the PV algorithm.
.In order-to provide a meaningful benchmark, the two modifications - were tested separately-by analyzing the=SLB transient with offsite power available. (the full-flow case), which-was the most-limiting SLB transient.
A statepoint analysis was used,-
consistent with the case analyzed for Reference 1.
In the first case
- analyzed, the full-flow SLB~ statepoint was reanalyzed using _ the same 5-channel model as before (Reference 1) but using the PV algorithm in place-of the standard solution technique (SCHEME).
In the-second case, the standard solution technique was used with the larger nine-channel half-core model described in Reference 2.
The results, summarized in the table below, show that the difference in results between the two solution schemes is less than 3% in predicted DNBR and there is no practical-difference between l
i I
the 1/8 core and 1/2 core models (actual calculated-difference 0.2%)
Full-flow Steamline Break DNBR Analysis Results QLgg Minimum DNBR Base with 1/8 core model 1.57 Base with PV algorithm 1.60-j Base with 1/2 core model 1,57 Condition 2:
Since the transition mixed core DNBR penalty is accomodated by setting the OFA enthalpy rise peaking limits at a level equal.
to 96 percent of the Mark-BW peaking limits without increasing the design DNBR limit, the mixed core penalty should be-assessed to-the OFA assemblies if a reload design is such that the enthalpy rise factor of the OFA fuel is higher-than 96 percent of that of the Mark-BN fuel, kesponse 2:
For Catawba Unit 1 Cycle 6 operation, no transition core penalty is necessary because the OFA enthalpy rise peaking factor limit is set at a level equal to 96 percent of that of the Mark-BW fuel.
-For reload designs other than Catawba 1 Cycle 6, transition mixed core penalties will be accomodated by assigning retained DNBR margin.
Condition 3:
Evaluations should be performed for each reload cycle to confirm that the values of the key reactivity parametsrs are within the bound of those specified. in-Tables A.1 anct A.2 of BAW-10173.
If any cf the cycle-specific value (sic) is-not bounded, new analyses of those transients affected should be performed to confirm the acceptance criteria are met, or a--
reload design change should be made to ensure the parameters are bounded by the values in the tables. - In addition, a RCCA misalignment analysis should be made for each reload cycle'to verify that the DNBR margin loss is less'than the available peaking margin at full power.
Response 3:
The reload report submittal for-Catawba 1 Cycle 6 (Reference 5) contains the results of the physics cycle-specific evaluation.
Tables 7-1 and 7-2 show that the key physics i
parameters are within the bounds specified in Tables A.1 and
)
A.2 in BAW-10173, Respectively.
The results of the RCCA misalignment analysis were acceptable and are also documented in Reference 5.
For reload designs other than catawba 1 Cycle 6, the values of key safety analysis physics parameters will be confirmed to be bounded by values assumed in the safety analysis applicable to each reload.
Otherwise, a new analysis or reload design change will be performed to ensure that the safety analyses remain valid, Condition 4:
Since no analysis was provided for the baron dilution accident, an analysis should be provided in the first reload analysis report with the Mark-BW fuel.
Response 4:
See Attachment II,
" Catawba Units
- and 2 Boron Dilution Accident Reevaluation".
'\\
References
- 1. Letter transmitting Topical Report BAW-10173, Revision 0, H. B.
Tucker to U.
S. Nuclear Regulatory Commission, March 30, 1989.
- 2. Letter transmitting Topical Report BAW-10173, Revision 1, H. B.
Tucker to U. S. Nuclear Regulatory Commission,- October 22, 1990,
- 3. Letter transmitting Topical Report BAW-10173, Revision 2, M. S.
Tuckman to U. S. Nuclear Regulatory Commission, November 28, 1990.
- 4. BAW-10156-A, "LYNXT - Core Transient Thermal-Hydraulic Program,"
February, 1986.
- 5. Letter transmitting Catawba Unit 1 Cycle _6 Releed Technical Specification Amendment, H.
B. Tucker to U.
S.
Faclear Regulatory Commissica, January 9, 1991.
1 s
Attachment II Catawba Units 1 and 2 Boron Dilution Accident Reevaluation 4
P'
1 l
Catawba Units 1 and 2 Boron Dilution Accident Reevaluation Safety Evaluation 1.
Introductior.
The boron dilution accident has been reanalyzed to support recent Catawba reload cores. These flows are currently being administrative 1y controlled until a change to the TS is applied for and approved.
Catawba Units I and 2 are equipped with~a Boron Dilution Mitigation Systeni (BDMS) which serves to detect uncontrolled dilution events in Modes 3-6 of plant operation and secure poasible dilution flowpaths by automati:-valve operation. The evaluation of dilution events in Modes 3-6 must demonstrate that the dilution will be terminated, either by the BDMS or by the operator, before criticality occurs, in the event that ono train of the BDMS is inoperable in these modes, the flowrate of the Reactor Makeup Water System is limited to values which have been shown to allow adequate operator action time to terminate the dilution before criticality occurs.
Reanalysis of the boron dilution event in Modes 3-5 shows a need tc change the Tet;inical Specification flowrates to the following values:
M_ ode Old Value New Value 3
200 150 4
80 150 5
80 75 2.
Accident Evaluation The following safety evaluation has been prepared to justify the revision in allowable flowrates during operation in Modes 3-5 with one train of the BDMS inoperable, t
CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant (Presented in FSAR Section 15.4.6)
This ANS Condition 11 event is analyzed to show that adequate time exists to terminate a dilution event prior to loss of shutdown margin.
Termination of a dilution event can result from actuation of the Boron Dilution Mitigation System (BDMS) in Modes 3-6 or by operator action following a high flux at shutdown alarm in case the BDMS is inoperable.
l l
t
- -._~. -. -
.- ~. -
e A reanalysis of the dilution events with the BDMS inoperable was performed to demonstrate that, given bounding assumptions made on flowrates from the Reactor Makeup Water System, conservative temperature differences between the diluted water source and the Reactor Cool at System, and conservative ratios of initial to critical boron concentrations, that the operator would have adequate time to terminate the dilution before criticality occurs.
The ratios of initial to critical boron concentrations assumed in the safety evaluation for the different modes of operation are confirmed to be bounded during cycle ryecific evaluations.
The flowrate used in the evaluations with the BDMS inoperable are Technical Specification flowrates chosen to ensure that the operator will have 15 minutes in which to terminate the dilution prior to criticality.
Results and Conclusions l
The evaluation of the boron dilution accident shows that the operator i
will be able to terminate a dilution event in Modes 3-5 prior to recriticality and that all Standard Review Plan acceptance criteria for the boron dilution event are satisfied.
FSAR markups based on the evaluation are attached.
4 a
5 i
. fO cs Osc e S ea 4hd pap CNS Results The results following the startup of an idle pump with the above listed assump-tions are shown in Figures 15.4.4-1 througn 15.4.4-5.
As shown in these curves, during the first part of the transient the increase in core flow with cooler water results in an increase in nuclear power and a decrease in core average temperature.
The minimum DN8R during the transient is considerably greater than the limit value.
Reactivity addition for the inactive loop startup accident is due to the decrease in core water temperature.
During the transient, this decrease is due both to the increase in reactor coolant flow and, as the inactive loop flow reverses, to the colder water entering the core from the hot leg side (colder temperature side prior to the start of the transient) of the steam generator in
,the inactive loop.
Thus, the reactivity insertion rate for this transient changes with time.
The resultant core nuclear power transient,, computed witn consideration of both moderator and Doppler reactivity feedback ef fects, is shown on Figure 15.4.4-1, 4
4 O
calculated sequence of events for this accident is shown on Table 15.4.11.
"y transient results illustrated in Figures 15.4.4 1 through 15.4.4-5 indicate t..dt a stabilized plant conoition, with the reactor tripped, is approached rap; idly.
Plant cooldown may subsequently be achieved by fcilowing normal shutdown procedures.
15.4.4.3 Environmental Consequences There would be minimal radiological consequences associated with startup of an inactive reactor coolant loop at an incorrect temperature.
Therefore, this l
event is not limiting.
The reactor trip causes a turbine trip and heat may be removed from the secondary system through the steam generator power relief valves or safety valves.
Since no fuel damage is postulated to occur from this transient, the radiologiul consequences associated with tnis event would be less severe than the steam line break event analyzed in Section 15.1.5.
15.4.4.4 Conclusions The transient results show that the core is not adversely affected.
There is considerable margin to the limiting DN8R.
Thus, no fuel or clad damage is pre-dicted.
l l
15.4.5 A MALFUNCTION OR FAILURE OF THE FLOW CONTROLLER IN A BWR LOOP l
THAT RESULTS IN AN INCREASED REACTOR COOLANT FLOW RATE j
(Not applicable to Catawba).
15.4.6 CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION THAT RESULTS IN A DECREASE IN BORON CONCENTRATION IN THE REACTOR COOLANT 15.4.6.1 Identification of Causes and Accident Description Reactivity can be added to the core by feeding primary grade water into the Re-u tor Coolant System via the reactor makeup portion of the Chemical and Volume Control System.
Boron dilution is a manual operation under administrative con-15.4-17
Ih0 OmegC 46 fW pCQC-i CNS 4
trol with procedures calling for a limit or the rate and duration of dilution.
l A boric acid blend system is provided to reamit the operator to match the boron r
concentration of reactor coolant makeup water during normal charging to that in the Reactor Coolant System.
The Chemical and Volume Control System is designed to limit, even under various postulated failure modes, the potential rate of dilution to a value which, after indication through alarms and instrumentation, i
provides the operator sufficient time to correct the situation in a safe and orderly manner.
i Tria opening of the reactor water makeup control valve provides makeup to the Reactor Coolant System which can dilute the reactor coolant.
Inadvertent dilu-tion from this source can be readily. terminated by closing the control valve.
In order for makeup water to be added to the Rea.ctor Coolant System at pressure, at least one charging pump must be running in addition to a reactor makeup water pump.
- ihe rate of addition of unborated makeup water to the Reactor Coolant System when it is not at pressure is limited by administratively limiting the output i
of the reactor makeup water pumps.
Normally, only one reactor makeup water pump is operating while the other is on standby.
With the RCS at pressure, i
the maximum delivery rate is limited by the control valve.
1 The boric acid from the boric acid tank is blended with primary grade water in '
{
the blender and the composition is determined by the preset flow rates of boric acid and primary grade water on the control board, j
In order to dilute, two separate operations are required d
1.
The operator must switch from the automatic makeup mode to the dilute mode.
r 2.
The start button must be depressed, t
Omitting either step would prevent dilution.
fr Information on the status of the reactor coolant makeup is continuously avail-l able to the operator.
Lights are provided on the control board to indicate the operating condition of the pumps in the Chemical and Volume Control System.
Alarms are actuated to warn the operator if boric acid or dominerelized water i
flow rates deviate from preset values as a result of system malfunction, j
A boron dilution is classified as an ANS Condition-II~ event, a fault of moderate f
frequency. See Section 15.0.1 for a discussion of Condition II events, j
The Baron Dilution Mitigation System (BDMS) uses two source range detectors to moniter the subcritical multiplication of the reactor core.
An Alarm setpoint i
is continually cc.lculated as 4 times the lowest measured count rate, including compensation for background and the statistical variation in the count rate..
Once the alarm setpoint is exceeded, each train of the 80MS will automatically shut off both reactor makeup water pumps, align the suction of the charging pumps to highly borated water from the RWST, and isolate. flow to the charging i
pumps from the VCT.
Since these functions are automatically actuated by the BDMS, no operator action is ntcessary to terminate the dilution event and.
i recover the shutdown margin.
Because of the averaging scheme used by the BDMS 15.4-18
.-1.wr
=.---
e.
-y-w,,ymo--,.-.,,~
wm..,
---.r--.
e,.,
yy--,,-.,.-,,_,
,wneww
--ey.-
+-y--,
n-us-+-
w~
ev e
+
9 CNS to determine the count rate, there is a time delay or lag between the calculated output and the actual count rate.
This time delay is a function of the initial, steady-state count rate.
In order to maximize this time delay, a lower bound on the initial count rate of I cps is assumed.
A boron di'ution is classified as an ANS Condition II event, a fault of moderate frequency.
See Section 15.0.1 for a discussion of Condition II events.
15.4.6.2 Analysis of Effects and Consequences
$thodofAnalysis To cover all phases of the plant operation, boron dilution during refueling, cold shutdown, hot shutdown hot standby, startup, and power operation are co.n-sidered in this analys.fs. I n ik e. coucW e n os au mca r'd + N. dMenm t i n the irmpomhre ch Ybe chh* cci W(^N & 9 ou r c & c, t nd 4% recicle,r Dilution During Refueling (Mode 6) c.colant Mcm n 5.cr \\ exhd a n e t % c.% u b,d i n 4 o c
on el i l on r d e..
An uncontrolled boron dilution accident cannot occur during refueling as a result of a reactor coolant makeup systLa malfunction.
This accident is prevented by administrative cor.trols Aich isolate the Reactor Coolant System from the potential source of unborated water.
l Valve NV230 in the CVCS will be locked closed during refueling operations.
This valve will block the flow paths which could allow unborated makeup to l
reach the reactor coolant system.
Any makeup which is required during refueling will be borated water supplied from the refueling water storage l
tank.
The most limiting alternate source of uncontrolled boron dilution would be the inadvertent opening of a valve in the Boron Thermal Regeneration System (BTRS).
For this case highly borated RCS water is depleted of boron as it passes through the BTRS and is returned via the volume control tank.
The following conditions are assumed for an uncontrolled boron dilution during refueling:
i 1.
Technical Specifications require the reactor to be borated to a concentration of 2000 ppa at refueling.
The critical boron con-centration is conservatively estimated to be 1739 ppm.
i 2.
Dilution flow is assumed to be the design output of both reactor i
makeup water pumps (300 gpm).
This is assumed although normally neither the reactor makeup system nor the BTRS is operated at refueling conditions.
3.
Mixing of the reactor coolant is accomplished by the operation of one residual heat removal pump.
4.
A minimum water volume (3588 ft ) in the RCS is used.
This is the 3
minimum volume of the RCS for residual heat removal system operation.
15.4-19
l CNS Dilution During Cold Shutdown (Mode 5) l Conditions at cold shutdown require the reactor to be shut down by at least 1.0% ak/k.
The ratio of the 1.0% ak/k Shutdown boron concentration to the critical boron concentration is assumed to be the conservatively low value of 1.15.
The following conditions are assumed for an uncontrolled boron dilution during cold shutdown:
l 1.
Dilution flow !s assumed to be the design output of both reactor makeup water pumps (300 gpm).
i 2.
Mixing of the reactor coolant is accomplished by the operation of one residual heat removel pump.
3.
A minimum water volume (3588 ft ) in the RCS is used.
This is the 8
minimum volume of the RCS for residual heat removal system operation.
(4 c\\c\\ h Dilution During Hot Shutdown (Mode 4) 1 l
Conditions at hot shutdown require the reactor to be shut down by at least
- 1. 3% ak/k.
The ratio of the 1.3% ak/k shutdown boron concentration to the critical boron concentration is assumed to be the conservatively low value of 1.15.
The following conditions are assumed for an uncontrolled boron dilution during hot shutdown:
l 1.
Oilution flow is assumed to be the design output of both reactor makeup water pumps (300 gpm).
2.
Mixing of the reactor coolant is accomplished by the operation of one residual heat removal pump.
3.
A minimum water volume (3588 ft ) in the RCS is used.
This is the 3
minitrum volume of the RCS for residual heat removal system operation.
011ution Ouring Hot Standby (Mode 3) b Conditions at hot standby require the reactor to have available at least 1.30%
ak/k shutdown margin.
This mode of operation is analyzed both with and without l
the most reactive rod cluster control assembly (RCCA) stuck out of the core.
The stuck rod case is assumed to occur immediately after a reactor trip and is therefore analyzed at no-load conditions.
The case with no stuck rod is analyzed at 350'F which is conservative since this is the lowest permissible temperature in this mode.
For both cases analyzed, the ratio of the 1.3% ak/k shutdown boron concentration to the critical boron concentration is assumed to be the conservatively low value of 1.15.
The following conditions are assumed in each case for a continuous boron dilution during hot standby:
1.
Dilution flow is assumed to be the design output of both reactor makeup water pumps (300 gpm).
2.
A minimum water volume (9C29 ft ) in the Reactor Coolant System is 3
used.
This corresponds to the active volume of the Reactor Coolant System while on natural circulation, i.e., the reactor vessel upper head and the pressurizer are not included.
Ac\\ck C
15.4-20
i e
(A) In case the BDMS is inoperable, the operator is alerted to a dilution event by the high-flux at shutdown alarm. A flowrate of 80 gpm and a boron ratio of 1.15 are chosen to demonstrato adequate operator action time.
(B) in case the BDMS is inoperable, the operator is alerted to a dilution event by the high-flux at shutdown alarm.
A flowrate of 186 gpm and a boron ratio of 1.15 are chosen to demonstrate adequate operator action time.
(C) In case the BDMS is inoperabic, the operator is nierted to a dilution event by the high-flux at shutdown alarm. A flowrate of 155 gpm and a boron ratio of 1.15 are chosen to demonstrate adequato operator action time.
l i
.,.{
I
i (do Cha ne % k t' W3 pac 6 "'
CNS Dilution During Startup (Mode 2)
Startup is a transitory mode of operation.
In this mode the plant is being taken from one long term mode of operation, hot standby, to another, power operation.
The ; dant is maintained in the startup mode only for the purpose of startup testing at the beginning of each cycle.
During this mode of operation, g
the plant is in manual control, i.e., Tavg/ rod control is in manual.
All normal actions required to change power level, either up or down, require operator initiation.
The Technical Specifications require a shutdown margin of 1.3% Ak/k and four reactor coolant pumps operating.
Additional conditions assumed are:
1.
Dilution flow rate is a conservatively high charging flow rate (300 gps) consistent with Reactor Coolant System operation at 2250 psia ar.d 557'F.
2.
A minimum RCS volues of 9800 ft.
This is a conservative estimate of 3
the active RCS volume, minus the pressurizer volume.
3.
The HIP, ARI, N-1, critical boron concentration is assumed to be the conservatively high value of 1350 ppe, with a very conservative constant boron worth of 15.0 pcm/ ppm.
Dilution During Power Operation (Mode 1)
With the unit at power and the Reactor Coolant System at pressure, the dilution rate is limited by the capacity of the charging pumps (analysis is performed assuming all charging pumps are in operation for manual rod control.
300 gal / min, although only one is normally in operation).
For automatic rod control at power, a flow capacity of 300 gal / min is also assumed.
The effec-tive reactivity addition rate is a function of the reactor coolant temperature and boron concentration.
Additional conditions assumed are:
1.
A minimum RCS volume of 9800 ft.
This is a conservatiw estimate of 3
the active RCS volume, minus the pressurizer volume.
2.
The reactivity insertion rate calculated is based on a conservatively high value for the expected boron concentration at power at which chutdown margin is lost (1150 ppm).
The operator is alerted to an uncontrolled reactivity insertion by an over temperature AT trip or by the rod insertion alarms depending on whether the plant is in manual or automatic rod control.
Results The calculated sequence of events is shown in Table 15.4.1-1.
Oilution Durina Refuelino (Mode 6)
During refueling, an inadvertent dilution from the Reactor Makeup Water System is prevented by administrative controls which isolate the RCS from the potential source of unborated makeup water.
15.4-21 k
I
" ' ' " ' ' ' ~
l 1
i C,g The most 1 @iting conditions for an inadvertent dilution from either the BTRS
.l or the Reactor Makeup Water System occur with the RCS drained to 26" above the bottom ID of the reactor vessel inlet nozzles.
7 '> L The results for Mode 6 indicate that there are 1.95 minutes available between the time the BDMS output exceeds the alarm setpoint and the shutdown margin t
!{t k-(o is exhausted.
A conservative response time of 25 seconds is assumed for the valves actuated by the BOMS to open or close.
Therefore, this analysis 4 -
- t] 6.C[a demonstrates that thers is sufficient time available (s1.5 minutes) for any remaining diluted water to be flushed from the charging lines and borated S d ~o water from the RWST to be injected into the RCS prior to a loss of shutdown maroin. W A N Soms inoperabic h optred w win be a k r4 fd b
't Dilution Durina C&id Shutdo$n (WX -M-SNddown crWm 'nynute4Ac dAdion,
ac Con b g %. Web and w i n ha v c.
_._. C 33 o ter Mode 5) M nwM 9 +d N.
~
1 prior 4 o et Act 1
- ( " $ 1, While in cold shutdown, the RCS thermal conditions are maintaintd while L
t-
- /
operating on the Residual Heat Removal System (RHRS) with the RCS drained to o3 26" above the bottom 10 of the reactor vessel inlet nozzles.
W( y $
- g)
The results for Mode 5 indicate that there are 1.95 minutes available between the time the BOMS output exceeds the alarm setpoint and the shutdown margin
- ! a+ 4 t9 is exhausted.
A conservative response time of 25 seconds is assumed for s
0 Y f:
the valves actuated by the 80M3 to open or close.
Therefore, this analysis d } g{
demonstrates that there is sufficient time available (*1.5 minutes) for any remaining diluted water to be flushed from the charging lines and borated
\\+v e water from the RWST to be injected into the RCS prior to a loss of shutdown y,
margin.
i
+
}
i.
~ DO Oilution Ouring Hot Shutdown (Mode 4) n d 2&
, bjA The results for Mode 4 indicate that there are 1.95 minutes available between the time the BOMS output exceeds the alarm setpoint and the shutdown margin g -Q o
is exhausted.
A conservative response time of 25 seconds is assumed for c'3 the valves actuated by the 80MS to open or close.
Taerefore, this analysis q
demonstrates that there is sufficient time available (s1.5 minutes) for any i nq1. remaining diluted water to be flushed from the charging lines and borated L ' o C ;[ water from the RWST to be injected into the RCS prior to a loss of shutcown
]4 $ 9.t. margin, k
.y
~ d 0ilution Durina Hot Standby (Mode 3) t.t 4 p a,
- t - b d't The results for Mode-3 indicate that there are 6.13 minutes available between
.C " the time-the BOMS output exceeds the alarm setpoint and the shutdown margin ed d is exhausted.
A conservative response time of 25 seconds is assumed for IO!the valves actuated by the BDMS to open or close.
Therefore this analysis N
3 + s 4 dem nstrates that there is sufficient time available (s5.7 mlnutes) for any remaining diluted water to be flushed from the charging lines and. borated g
water from the RWST to be injected into the RCS prior to a -loss of shutdown
{ margin.
Oilution During Startup (Mode 2)
Thi. mode of operation is a transitory mode to go to power and is the l
operational mode in which the operator intentionally dilutes and withdraws 15,4-22 1
jVo C h0( n t es d c) i
-4 hb pay.
CNS l
control rods to take the plant critical.
During this mode, the plant is in manual control with the operator required to maintain a very high awareness of the plant status.
For a normal approach to criticality the operator must manually initiate a limited dilution and subsequently manually withdraw the control rods, a process that takes several nours.
The plant Technical Specifications require that the operator determine the estimated critical position of the control rods prior to approaching criticality thus assuring that the reactor does not go critical with rods below the insertion limits.
Once critical, the power escalation must t'e sufficiently slow to allow the operator to manually block the Source Range reactor trip after receiving P-6 from the Intermediate Range (nominally at 105 cps).
To fast a power escala-tion (due to an unknown dilution) would result in reaching P-6 unexpectedly, leaving insufficient time to manually block the Source Range reactor trip.
Failure to perform this manual action results in a reactor trip and immediate shutdown of the reactor, allowing sufficient time prior to a loss of shutdown margin for the operator to terminate the dilution event.
However, in the event of an unplanned approach or dilution during power escalation while in the startup mooe, the plant status is such that minimal impact will result.
The plant will slowly escalate in power to a reactor trip on the Power Range Neutron Flux Low Setpoint (nominally 25% RTP).
After reactor trip, there is at least 15.2 minutes for operator ection prior to a loss of shutdown margin to terminate the dilution.
Dilution During Full Power Operation (Mode 1) 1.
With the reactor in automatic control, the power 6iid temperature increase from boron dilution results in insertion of the rod cluster control as-semblies and a decrease in the shutdown margin.
The rod insertion limit alarms (low and low-low settings) provide the operator with adequate time (of the order of 64.9 minutes) to determine the cause of dilution, isolate the primary grade water source, and initiate reboration before the total shutdown margin is lost due to dilution.
2.
With the reactor in manual control and if no operator action is taken, the power and temperature rise will cause the reactor to reach the over-temperature AT trip setpoint.
The boron dilution accident in this care is essentially identical to rod cluster control assembly withdrawal accident.
The maximum reactivity insertion rate for boron dilution is approximately 1.64 pcm/sec and is within the range of insertion rates analyzed.
Prior to the overtemperature AT trio, an overtemperature aT alarm and turbine runback would be actuated.
There is adequate time available (of the order of 47.0 minutes) af ter a reactor trip for the operator to determine the cause of dilution, isolate the primary grade water sources, and initiate reboration before the reactor can return to l
criticality.
i 15.4.6.3 Environmental Consequences There would be minimal radiological consequences associated with a Chemical and Volume Control System malfunction that results in a decrease in boron concen-tration in the reactor coolant.
The reactor trip causes a turbine trip, and heat may be removed from the secondary system through the steam generator power relief valves or safety valves.
Since no fuel damage occurs from this transient, 15.4-23 s
... - ~,
- - - ~.
o e
CNS the radiological consequences associated with this event are less severe than the steam line break event analyzed in Section 15.1.5.
15.4.6.4 Conclusions For Modes 1 and 2, the results presented above show that there is adequate time for the operator to manually terminate the source of dilution flow.
Following termination of the dilution flow, the reactor will be In a stable condition.
The operator can then initiate boration to recover the shutdown margin.
For Modes 3 through 6, the BDMS, as described in Section 7.6.2.4, is the primary source of protection against a dilution event. Even considering the conservative delays assumed in this analysis, the preceding resuits indicate that the BDMS will automatically terminate a dilution event in Modes 3 through 6 prior to a loss of shutdown margin. L mot +W 60m5 is hoper @Je w_.
hgh ~S hu-ch Sbddocon ctiorm M\\cwo a6eguode. h rne. 4ec opercdor' 15.4.7 INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN celon ho IMPROPER POSITON
-4cc m'md e %A Clerd prior %
15.4.7.1 leentification of Causes and Accident-Description eqcep Fuel and core loading errors such as can arise from the inadvertent loading of one or more fuel assemblies into improper positions.-loading a fuel rod during manufacture with one or more pellets of the wrong enrichment, or the loading of a full fuel assembly during manufacture with pellets of the wrong enrichment will lead to increased heat fluxes if the error results in placing fuel in core positions calling for fuel of lesser enrichment.
Also included among possible core loading errors is the inadvertent loading of one or more fuel assemblies requiring burnable poison rods into a new core without burnable poison rods.
Any error in enrichment, beyond the normal manufacturing tolerances, can cause power shapes which are more peaked than those calculated with the correct enrichments.
There is a 5 percent uncertainty margin included in the design value of power peaking factor assumed in the analysis of Condition I and Condition 11 transients.
The-incore system of moveable flux detectors, which is used to verify power shapes at the beginning of cycle, is capable of revealing any assembly enrichment error or loading error which causes power shapes to be peaked in excess of the design value.
To reduce the probability of core loading errors, each fuel assembly is marked with an identification number and loaded in accordance with a core loading dia-gram.
Before core loading, the fuel assemblies in the Spent Fuel Pool, desig-I nated for the next-fuel cycle, will have the fuel assembly identification numbers and insert identification numbers checked.
Following core loading, the fuel assembly identification numbers are again checked as final assurance that l
the core has been loaded properly.
The power distortion due to any combination of misplaced fuel assemblies would significantly raise peaking factors and would be readily observable with'incore flux monitors.
In addition to the flux monitors, thermocouples are located at the outlet of about one third of the fuel assemblies in the core.
There is a high probability that these thermocouples would also indicate any abnormally high coolant entid py rise.
Incore flux measurements;are taken during the startup subsequent to every refueling operation.
15.4-24 1988 Update
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l d
i Table 15.4.1-1 (Page 2) f Time Seouence of Events for Incidents Waich Cause l
]
Reactivity and Power Distribution Anomalies j
t Accident Event Time (sec.)
l Startup of an Initiation of pump startup 1.0
[
Inactive Reactor i
Coolant Loop at Power reaches P-8 trip 13.4 an Incorrect setpoint i
Temperature Rods begin to drop 13.9 l
Minimum DNBR occurs 15.0 l
4 CVCS Malfunction that Results in a i
Decrease in the Boron Concentration 3
in the Reactor Md el boM Coolant c< ftac ha cI 1.
Dilution during Dilution begins 0
b 00tNb(O600D refueli BDMS setpoint is exceeded T37' I A -F i
Criticality occurs.
730-- -)SQ ]
r 2.
Dilution during Dilution begins 0
cold shutdown (Boms opembid BDMS setpoint is exceeded
-&33 (r I N 6--y Criticality occurs
.a w. 7 z.2 i
i l
3.
Dilution during Dilution begins 0-l hot shutdown-(60ms opemM BDMS setpoint is exceeded
$37 /357 C ----y Criticality occurs
-750-
/(oS/
i 4a.
Dilution during
-Dilution begins 0
hot standby
/
((60mS opemb)d BDMS setpoint is exceeded 1520-/I4',t w/o stuck rod Criticality occurs 4887' /4c,3 g4 4b.
Dilution during Dilution begins 0-hot standby (w/ stuck rod)
-1520 'llW
~
(600%operwble). BDMS setpoint is exceeded 1887-lg3 j
Criticality occurs Ed i
L t
?
?
i*
...a-.
.-.-.L..
s s
s
- 4. Dilution during refueling Dilution begins 0
(DDMS inoperable)
High-flux-at-shutdown 3400 alarm reached Criticality occurs
$280 O Dilution during cold shutdown Dilution begins 0
(BDMS inoperable)
High-flux-at-shutdown 1886 alarm reachod Criticality occurs 2786 C.. Dilution during hot shutdown Dilution begins 0
(BDMS inoperable)
High-flux-at-shutdown 2040 alarm reached Criticality occurs 2940 O Dilution during het standby Dilution begins 0
(w/o stuck rod)
(BDMS inoperable)
High-flux-at-shutdown 2444 alarm reached Criticality occurs 3344 E, Dilution during hot standby Dilution begins 0
(w/ stuck rod)
(BDMS inoperable)
High-flux-at-shutdown 2444 l
alarm reached Criticality occurs 3344 i
4
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