ML20073Q701
| ML20073Q701 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 05/15/1991 |
| From: | Tuckman M DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9106030038 | |
| Download: ML20073Q701 (20) | |
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i DUKE POWER May 15, 1991 U.
S.
Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.
C.
20555
Subject:
McGuire Nuclear Station Docket Numbers 50-369 and -370 Catawba Nuclear Station Docket Numbers 50-413 and -414 Supplementary Information Relative to Topical Report BAW-10173; Boron Dilution Analysis By letter dated February 20, 1991, the NRC staff transmitted the Safety Evaluation Report (SER) for Topical Report BAW-10173.
The SER imposed 5 conditions on the use of the Topical, and requested that Duke respond to these conditions.
By letters of March 14, 1991 and April 25, 1991 responses were provided for these conditions.
It has since been determined that calculations relative to the response for condition number 4 (boron dilution recriticality times) contained incorrect assumptions regarding initial reactor coolant system volumes. The calculations have been redone and the revised response to Condition 4 is attached.
We regret any inconvenience this may have caused.
If there are any questions, please call Scott Gewehr at (704) 373-7581.
Very truly yours, M.
S. Tuckman I
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bordilut/ sag l
9106030038 910515 l
FDR ADOCV. 05000369 Q()l P
Nuclear Regulatory Commission May 15, 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 Atlatata, Georgia 30323 Mr. Heyward Shealy, Chief Bureau of Radiological Health South Carolina Department of Health &
Environmental Control 2600 Bull Street Columbia, South Carolina 29201 American Nuclear Insurers c/o Dottie Sherman, ANI Library The Exchange, Suite 245 270 Farmington Avenue Farmington, CT 06032 M&M Nuclear Consultants 1221 Avenue of the Americas New-York, New York 10020 INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30339 Mr.
W.
T. Orders Senior Resident Inspector Catawba Nuclear Station Mr.
P.
K. Van Doorn Senior Resident Inspector McGuire Nuclear Station
Catawba Units 1 and 2 Boron Dilution Accident Reevaluation Safety Evaluation 1.
Introduction The boron dilution accident has been reanalyzed to support recent Catawba reload cores. These flows are currently being administratively controlled until a change to Technical Specification 3/4.3.3.12 is applied for and received.
Catawba Units 1 and 2 are equipped with a Boron Dilution Mitigation System (BDMS) which serves to detect uncontrolled dilution events in Modes 3-6 of plant operation and secure possible dilution flowpaths by automatic 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 one or both train (s) of the BDMS is (are) 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 to change the Technical Specification flowrates to the following values:
Mode 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.
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 Baron 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.
1 A reanalysis of. the dilution events.with the-BDMS inoperable was performed to demonstrate that, given boundingLassumptions made on flowrates from:the Reactor Makeup Water System, conservative temperature i
differences between the diluted water. source and the Reactor-Coolant
- 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 specific evaluations.
The flowrates used in the evaluations with the BDMS inoperable are chosen to ensure that the operator will have 15 minutes-in which to terminate the dilution prior to criticality. The Technical Specification flowrates are conservatively chosen to ensure that the flowrate values used in the evaluations are not exceeded.
Results and Conclusions The evaluation of the boron dilution accident shows that the operator 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.
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Results The results following the startuo of an idle pumo with the above listed assumo-tions are shown in Figures 15.4.4-1 through 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 DNBR during the transient is considecably greater than :ne limit value.
Reactivity addition for the inactive 1000 startuo accident is due to the decrease in core water temperature.
During the transient, this cecrease 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, comouted with consideration of both moderator and Doppler reactivity feedback effects, is shown on Figure 15.4.4-1.
The calculated sequence of events for this accident is shown on Table 15.4.1-1.
The transient results illustratea in Figures 15.4.4-1 througn 15.4.4-5 indicate that a stabilized plant condition, with the reactor tripped, is approached rap-idly.
Plant cooldown may subsequently be achieved by following normal shutdown procedures.
15.4.4.3 Environmental Consecuences There would be minimal radiological consequences associated with startup of an inactive reactor coolant loop at an incorrect temperature.
Therefore, this event is not limiting.
The reactor trip causes a turbine trip and heat may be removea from the secondary system through the steam generator power relief salves or safety valves.
Since no fuel camage is postulatec to occur from tnis transient, the radiological consecuences associated with this event would be less severe nan the steam line creak event analyzec in Section 15.1.5.
15.4.4.4 Conclusions The transient results show that the core is not adversely affected.
There is consideraole margin to the limiting DNBR.
Thus, no fuel or clad damage is pre-dicted.
15.4.5 A VALFUNCT:0N OR FAILURE OF THE FLOW CONTROLLER IN A SWR LCOP THAT RESULTS IN AN INCREASED REACTOR COOLANT FLOW RATE (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 Descriotion Reactivity can be added to the core by feeding primary grade water into the Re-actor Coolant System via the reactor makeuo portion of the Chemical and Volume Control System.
Boron dilution is a manual coeration uncer acministrative con-l 15.4-17
CNS trol witn proceoures calling for a limit on the rate and curation of dilution.
A boric acid blend system is provided to permit the operator to match the boron concentration of reactor coolant makeup water during normal charging to that in the Reactor Coolant System.
The Chemical and Volume Control System is designea to limit, even under various postulated failure mooes, the potential rate of dilution to a value wnich, after indication througn alarms and instrumentation, provides the operator sufficient time to correct the situation in a safe and orderly manner.
The opening of the reactor water makeup control valve provides makeuo to the Reactor Coolant System wnich can dilute the reactor coolant.
Inadvertent cilu-tion from this source can be readily terminated by closing the control valve.
In order for makeup water to be added to the Reactor Coolant System at pressure, at least one charging pump must be running in addition to a reactor makeup water pump.
The rate of addition of unborated makeup water to the Reactor Coolant System when it is not at pressure,is limited by acministratively limiting the output of the reactor makeup water pumps.
Normally, only one reactor makeup water pump is operating while the other is on stanoby.
With the RCS at pressure,
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the maximum delivery rate is limited by the control valve.
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.
In order to dilute, two separate operations are required:
1.
The operator must switch from the automatic makeuo moce to the dilute mode.
2.
The start button must te depressed.
Jmitting either step would prevent dilution.
Information on the status of the reactor coolant makeuo is continuousiv avail-able to the operator.
Lights are provided on the control board to inalcate the operating condition of the pumps in the Chemical and Volume Control System.
Alarms are actuated to warn the operator if boric acid or demineralized water flow rates deviate from preset values as a result of system malfunction.
-A boron dilution ; 5 ;.555ified 2; cr ^MS Conditian !! event, a fault of ocar'ta
' eeuency. - see Section 15.0.1 for ; aiscussica of C:navbien !! ever's The Boron Oilution Mitigation System (80MS) uses two source range detectors to monitor the suocritical multiolication of the reactor core.
An alarm setooint is continually calculated as 4 times the lowest measured count rate, including compensation for oackground and the statistical variation in the count rate.
Once the alarm setpoint is exceeded, each train of the BOMS 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 pumps from the VCT.
Since these functions are automatically actuated by the BDMS, no operator action is necessary to terminate the dilution event and recover the shutdown margin.
Because of the averaging scheme used by the 80MS 1
1 15.4-18
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to determine the count rate, there is a time delay or lag between the calculated output and the actual count rate.
This time celay 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 1 cps is assumea.
A boron dilution 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 Analvsis of Effects and Consecuences Method of Analysis To cover all phases of the plant operation, boron dilution during refueling, cold shutdown, hot shutdown, hot standby, startup, and power operation are con-sidered in this analysis. % imer& from a #a e /r e</ g es ilution Ourino Refuelino (Mode 6)
An u ontrolled boron dilution accident cannot occur during refueling as a result a reactor coolant makeuo system malfunction.
This accident is k
prevented administrative controls which isolate the Reactor Coolant System from the pot tial source of unborated water.
Valve NV230 in th CVCS will be locked closed during refueling opepations.
Thisvalvewillbloc(lantsystem.the flow paths which could allow unborated makeup to reach the reactor coo Any makeup which is requirecr'during refueling will be borate'dswater supplied from the refueling ter storage tank.
The most limiting alternate sour of uncontrolled bor,an dilution would be the inadvertent opening of a valve - the Baron Th mal Regeneration System (STRS).
For this case highly caratea CS water '
decleted of Doron as it casses through the BTRS and is returned ia t olume control tank.
The following conditions are assumed for an unca rolled Doron ailution during refueling:
1.
TechnicalSpecificationsredirethereactortobeboratedtoa concentration of 2000 pgat refueling. 'T{ie critical boron con-centration is conservatively estimated to b' 1739 ppm.
2.
Dilution flow is ssumed to be the design outpukof both reactor makeuo water putitos (300 gpm).
This is assumed alhqugn normally neither th eactor makeuo system nor the BTRS is ope ated at I
refuelin onditions.
3.
Mixi of the reactor coolant is accomplished by the operat n of one r-idual heat removal pump.
4 A minimum water volume (3588 fta) in the RCS is used.
This is the minimum volume of the RCS for residual heat removal system operation.
15.4-19
t Modes-3-6 are analyzed - witn two different methods for two different
- purposes.
First, with the BDMS assumed to be operable, the accident is analyzed to demonstrate that there is adequate time, without restrictions on the flow rates from potential dilution sources, for the 3DMS to terminatc the dilution prior to criticality.
This time consists of two components: 1) the period required to stroke the valves manipulated by the BDMS and 2) the period required, once the unborated water source has been isolated, to purge the remaining unborated water from the piping leadir.g to the RCS.
Second, with the BDMS assumed to be inoperable, the accident is analyzed to demonstrate that there is adequate time, possibly with restrictions on the flow rates from potential dilution sources, for the operator to terminate the dilution prior to criticality.
Since the BDMS is not used in Modes 1 and 2, the analysis of these modes is similar i
to the analysis of Modes 3-6 with the BDMS assumed to be inoperable, but without the restrictions on flow rates.
Alarm Function Which Initiatas Mitication Mitigation of a boron dilution accident is not assumed to begin until an alarm has warned of the-abnormal circumstances caused by the event.
For Modes 3-6 with the BDMS operable, the alarm function is provided by the measured source range count rate exceeding the BDMS setpoint.
For Modes 3-6 with the BDMS inoperable, the alarm function is provided by the source range high-flux-at-shutdown alarm exceeding its setpoint.
For Mode 2'and for manual rod control during Mode 1, the alarm function is provided by the earliest reactor trip setpoint reached.
Finally, for automatic rod control during Mode 1, the alarm function is provided by the alarm which occurs when the control rods reach their insertion limits.
Dilutien Volume A postulated dilution event procresses faster for smaller RCS water volumes. Therefore, the analysis considers the smallest RCS water volume in which the unborated water is actively mixed by f orced cir culation.
For Modes 1-3, the Technical Specifications require that at least one reactor coolant pump be operating. This forced circulaticn will mix the RCS inventory in the reactor vessel and each of the four reactor coolant loops.
The pressurizer and the pressurizer surge line are not included in the volume available f or dilution in Modes 1-3.
Por normal Operation in Mode 4,
forced circulation-is typically maintained, although the Technical Specifications do not require it.
The volume available for dilution in Mode 4 is therefore conservatively assumed to not include the upper head of the reactor vessel, a region which has reduced flow in the absence of forced circulation, or the pressurizer and the pressurizer surge line.
Since the Technical Specifications-do require operacility of all four steam generators during Mode 4,
all four of the reactor coolant _ loops, in addition to the remainder of the reactor vessel, are included in the RCS volume available for dilution.
For Modes 5 and 6, the reactor coolant water level may be drained to below the top of the main coolant loop piping, and at least one train of the Residual Heat Removal System (RHRS) is operating.
The volume available for dilution in these modes is - limited to the smaller volume RHRS train plus the portions of the reactor vessel and reactor coolant loop piping below the minimum water level (7.5 inches above the centerline of the hot and cold leg piping) and between the RHRS inlet and outlet connections.
The
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4 4
minimum water level used to calculate this volume is corrected for level instrument uncertainty.
Boron Concentrations The Technical Specifications require that the shutdown margin in-the various modes ~ be above a certain minimum value.
The difference in boron concentration, between the value at which the relevant alarm function is
. actuated and _the value at which the reactor is just critical,_ determines the_ time available to mitigate a dilution event.
Mathematically, this time is a ' function of the ratio of these two concentrations, where a large ratio corresponds to a longer time.
During the reload safety analysis for each new. core, the above - concentrations are checked to ensure that the _- value of this ratio for each-mode is larger than.the corresponding ratio assumed _in - the accident analysis.
Each mode of operation covers a range of temperatures.
Therefore, within that mode, the temperature which minimizes this ratio is used for comparison with the. accident analysis ratio.
For accident initial conditions in which the control rods are withdrawn, it is conservatively assumed,- in
" calculating the critical boron ' concentration, that the most reactive RCCA does not fall into the cora at reactor trip.
This assumption _is also conservatively applied in Moce 3 when the' initial condition is hot zero power.
For. colder _ conditions in Modes 3-5, emergency procedures for la reactor trip.with a stuck RCCA require that, prior to the initiation of the cooldown, the boron concentration be increased by an amount which compensates for any RCCAs not completely inserted.
Dilution Flow Rate In the absence of flow rate restrictions, the dilution flow rate assumed to enter _ the RCS is ' greater than or equal to the design volumetric flow rate of both reactor makeup water pumps.
In a dilution event, these pumps are assumed to deliver unborated water to the suction of the centrifugal charging-pumps.
Since the water delivered by.these pumps..is typically colder than the RCS.. inventory, the unborated water expands within t h e _. R_CS, causing a - given volumetric flow rate measured at the colder temperature to - correspond to a larger volumetric dilution flow rate within the RCS.
This. density difference in'the_dil' tion flow rate u
is accounted for in the analysis.
The above assumption on flow rate is
- also conservatively-used for Mode 6,
even though valve NV-230 in the
' Chemical: and Volume Control System '(CVCS). is locked closed during refueling.
This valve blocks the flow paths which could allow unborated makeup to reach the RCS.
Any makeup.which is required during this mode
-is borated water supplied'from the refueling' water storage tank.
Results-The calculated sequence of avents is shown in Table 15.4.1-1.
Dilution Durino Modes in which the'BDMS is Recuired (Modes 3 6)
.During Mode 6'an inadvertent dilution from the Reactor Makeup Water o
System is prevented by administrative controls which isolate the RCS f rom -
potential sources of unborated makeup water.
The results presented in Table 15.4.1-1 f or this mode are for an assumed dilution event, for which
no mechanism-or-flow path has been identified.
For Modes 3-6-with the BDMS-operable, the results presented in. Table 15.4.1-1 show that there is adequate time to reach-the' BDMS alarm'setpoint, stroke closed the valves to isolate the source of unborated water,'and purge the unborated water already in the CVCS piping, before the shutdown margin is exhaust-ed.
For_ Modes 3-6 with the BDMS inoperable, the results presented in Table 15.4.1-1 show that, with-limitations on flow rates'from potential sources of unborated water, there is adequate time for the operator to determine the cause of the dilution, isolate the source of unborated water, and initiate reboration before the shutdown margin is exhausted.
In accordance with Reference _11, adequate time is judged to be at least 15 minutes for Modes 3-5 and _at least 30 minutes for Mode 6.
The results
- presented in Table 15.4.1-1.are for the dilution flow rates - which, assuming the boron concentration ratios are at the reload safety analysis limits, give exactly these operator response times.
Flow. rates are restricted, through Technical Specifications and administrative controls, to values which are less than these analy cd flow rates, thus in practice giving even longer operator response times.
Additional margin is provided by the fact there is typically margin between the assumed boron concentration ratio - for a given mode and the actual corresponding concentration ratio for the reload core.
l I
,4
CNS f
ilution During Cold Shutdown (Mode 5)
Con 'tions at cold shutdown require the reactor to be shut _down by at leas 1.0% a k.
The ratio of the 1.0% ak/k shutdown boron concentration to t.hs critica boron concentration is assumed to be the conservatively low value of 1.15.
Th following conditions are assumed for an uncontrolled borop' dilution during col hutdown:
1.
Oilu on flow is assumed to be the design output of th reactor makeup ater pumps (300 gpm).
2.
Mixing of he reactor coolant is accomplished b the operation of one residual he t removal pump.
3.
A minimum wate 3
volume (3588 ft ) in the RQ5 is used.
This is the minimum volume h the RCS for residual hea't removal system operation.
Dilution Ouring Hot Shutdown Mode 4)
Conditions at hot shutdown rea e the reacto to be shut down by at least 1.3% ak/k.
The ratio of the 1.3% ak/k shutddwn boron concentration to the l
critical boron concentration is as med tcybe the conservatively low value of 1.15.
The following conditions are ssu d for an uncontrolled boron dilution during hot shutdown:
1.
Oilution flow is assumed t,o be the design output of both reactor makeup water pumps (300 m).
2.
Mixing of the reactor / coolant is a omplished by the operation of one residual heat removai pump.
3 3.
A minimum water
- ume (3588 ft ) in th RCS is used.
This is the minimum volume f the RCS for residual h t removal system coeration.
Oilution Durino Hot 5 ndbv (Mode 3)
Conditions at hot s ak/k shutdown marp(tandby require the reactor to have av ilable at least 1.30%
n.
the most reactiver rod cluster control assemoly (RCCA) stuckM ut of the core.
The stuck rod c/se is assumed to occur immediately after a r' ctor trip and
!s therefore phalyzed at no-load conditions.
The case with no stuck rod is analyzed at J50 F which is conservative since this is the lowes permissible temperatury'in this mode.
For oath cases analyzed, the ratio of he 1.3% ak/k shutdown oron concentration to the critical boron concentration i assumed to be the nservatively low value of 1.15.
The following conditions e assumed l
I in eac' case for a continuous boron dilution during hot standby:
i 1.
Oilution flow is assumed to be the design output of both react r l
makeup water pumps (300 gpm).
3 2.
A minimum water volume (9029 ft ) in the Reactor Coolant System i L
used.
This corresponds to the active volume of the Reactor Coolan System while on natural circulation, i.e., the reactor vessel upper head and the pressurizer are not_ included.
\\
15.4-20
-~
. ~
- NS ilution Durina 5tartuo (Moce 2)
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St&qtuo is a transitory mode of oceration.
In this mode the plant is being f
taker ( from one long term mode of oceration, hot stanoby, to another, power f
operation.
The plant is maintainea in the startup moae only for the ourgese of startuo\\testingatthebeginningofeacncycle.
During this mode of ope' ration, the plan is in manual control, i.e., Tavg/ rod control is in manual.,All normal act ns required to enange cower level, either uo or down, re6uire operatorinigiation, The Tecnnical Soecifications reauire a shutdodn margin of 1.3% Ak/k and four reactor coolant pumps operating.
Additionaf conditions assumed are:
1.
Dilution flow rate is a conservatively high char gi g flow rate (300 s
gpm) cons stent with Reactor Coolant System oper on at 2250 psia and 557 F.
2.
A minimum RCS olume of 9800 ft.
This is a conservative estimate of 3
(
the active RCS olume, minus the pressurize. volume.
1 3.
The HZP, ARI, N-1\\ critical baron concentration is assumea to be the caservatively higN value of 1350 ppm /, 4ith a very conservative l
cone. ant boron wortrt of 15.0 pcm/ ppm.
Dilution Durina Pt.wer Operation Mode 1)
With the unit at power and the Re ctor Coa ant System at pressure, the dilution rate is limited by the caca'(ity/of the charging pumps (analysis is performed assuming all charging pumos Wre in operation for manual roa control, 300 gal / min,althougnonlyoneisnors.allyinoperation)
For automatic rod controlatpower,aflowcapacityOf'300Nal/minisalsoassumed, The effec-
-tive reactivity accition rate is p' functiogof the reactor coolant temperature and boron concentration.
Additt4nal concit.gns assumea are:
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1.
A minimum RCS volume of 9800 ft3 TM s is a conservative estimate o*
the active RCS vg(ume, minus ;ne cress'ttrizer volume.
I 2.
The reactivity / nsertion rate calculated is based on a conservatively high value fjfr the expected boron concentration at power at which
[
shutdown m rgin is lost (1150 ppm).
The operator is a) rted to an uncontrolled reactivity inseqtion by an over temoerature AT t/io or by the roa ir.sertion alarms cecending on whether the plant is in mandal or automatic roc control.
Results The calc, ated sequence of events is shown in Table 15.4.1-1.
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Dilution Durina Refuelina (Mode 6) 0 ng refueling, an inadvertent dilution from the Reactor Makeup Water
$stemispreventedbyadministrativecontrolswhichisolatetheRCSfromt\\
~
potential source of unborated makeuo water.
I 15.4-21 1
2S kihe most limiting conditions for an inaavertent dilution f rom either the STRS oq the Reactor Makeup Water System occur with the RCS drained to 26" above tne bo'ttom ID of the reactor vessel inlet nozzles.
/
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The results for Mode 6 indicate that there are 1.95 minutes available between the timb the BDMS output exceeds the alarm setooint and the shutdown marfjin s
is exhausted.
A conservative response time of 25 seconds is assumea f4r tne valves actuated by the BDMS to open or close.
Therefore, this analysis
~
demonstrates \\that there is suf ficient time availaDie (s1.5 minuteH for any remaining diluted water to be flushed f rom the charging lines arpf borated water from the WST to be injected into the RCS prior to a lost of snutdown margin.
Dilution Durino Cold hutdown (Mode 5)
While in cold shutdown, the RCS thermal conditions are/
maintained while operating on the Residua N eat Removal System (RHRS)fwith the RCS drained to H
26" above the bottom 10 of\\the reactor vessel inlet nozzles.
N
/
The results for Mode 5 indicate that there are/.95 minutes available oetween 1
the time the BDMS output exceeds the alarm setooint and the shutdown margin is exhausted.
A conservative rebponse time'of 25 ::econds is assumed for the valves actuated by the BDMS to\\open pf close.
Therefore, this analysis demonstrates that there is sufficierft t,tme available (~1.5 minutes) for any remainingdilutedwatertobeflushed>(romthecharginglinesandborated water from the RWST to be injected jhtoNthe RCS prior to a loss of shutdown
- margin, j/
Dilution Durina Hot Shutdown (M e 4)
\\
The results for Mode 4 indic e that there are 5 minutes available between the time the BDMS output e>(ceeds the alarm setooilig and the shutdown margin is exnausted.
A conservptive response time of 25 seqonds is assumea for the valves actuated cyftne BOMS to coen or close.
Thbrefore this analysis l
demonstrates that tnere is suf ficient time available (4,5 minutes) for any remaining dilutea wat'er to oe flusned from the charging i'ines and borated water from the RWST/ o be injectea into the RCS prior to a loss of snutdown t
margin.
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Dilution Durine/
'x Hot Standby (Mode 3)
The results or Mode 3 indicate tnat there are 6.13 minutes availabie between the time tfie BDMS output exceeos tne alarm setooint and the snutdown INrgin is exhauited.
A conservative response time of 25 seconds is assumeo foh the vanes actuated by the BDMS to open or close.
Therefore, this analysis demops'trates that there is sufficient time available (s5.7 minutes) for an remaining diluted water to be flushed from the charging lines and borated w der from the RWST to oe injected into the RCS prior to a loss of shutdown fargin.
y Dilution Durina Startuo (Mode 2) l This mooe of operation is a transitory made to go to power and is the operational mode in wnicn tne operator intentionally dilutes and withdraws 15.4-22
9 CNS 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 hours.
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 be 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 suf ficient 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 mode, 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 it least 15.2 i nutet for operator action prior to a loss of shutdown margi to terminate the dilution.
. ade@e +;m e Ca& leas + 15 mhdes per Refer'nte II)
Dilution Durina Full Power Ooeration (Mode 11 1.
With the reactor in automatic control, the power anc 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 (cf the crccr
^f 63.9 -inutes) to determine the cause of dilution, isolaththeprimarygradewatersource,andinitiatereborationbefore the( otal shutdown margin is lost due to dilution.
% at leas + IS minu tes per R e le.re nc e //
2.
With the reactor in manual control and if no operator action is taken, the power and temperature rise will cause tne reacter to reach the over-temperature ST trip setpuint.
The bornn nilutic" accident 4^
thi; ca;c-
_is essent4elly ic:ntical to red cluster-contrc! men!y =ithemwe'r s c4 dent..
Tne maximum reacti. ity inscrtior rate for Mrnn dilutinn ic approx mate!y 1.El pcm/;cc and is within tce range of incertier "Stes i
-a r.c l y z e d.- Prior to the overtemperature aT trip, an overtemperature ST alarm and turbine runback would be actuated.
There is adequate time available (cf the order af
.0 -i nutes) after a reactor trip for the s
operator to determine the cause of dilution, isolate the primary grade water sources, and initiate reDoration before the reactor can return to criticality, s +- leas + 15 m m des per Re f'erence ll 15.4.6.3 Environmental Consecuences 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 reacto-trip causes a turbine trip, and heat may be removec from the secondary system througn the steam generator power relief valves or safety valves.
Since no fuel damage occurs from this transient, 15.4-23
- - - = - -
- ~
^
~
La rht W1 S is hof'vuvis be ic e.
, L ey s.ca rc e
- unk qainsh - n dildie c
eveni-M Aede s 3 %ro C is
,ve7 aired
+o b e p a sume,(,Q+r r suh eperdie n a ctio n isyer r e.y e ns e Hine s are rnswl aegens, when 4hese joy er resy> e ns e
}imes a re.
conside red, if hs ac e ssa ry 4 <es}r;e} phe % w rafe, from, po, f
benJM dildjon s wces LNS y
the radiological consequences associated with t.his 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 initia e boration to recover the shutdown margin.
l 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 results indicate that the BDMS will automatically terminate a dilution event in Modes 3 through 6 prior to a loss of shutdown marging 15.4.7 INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN IMPROPER POSITON 15.4.7.1 Identific4 tion of Causes and Accident Description l
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 burn ble poison rods into a new core without burnable poison rods.
l l
Any error in enrichment, beyond tb? normal manufacturing tolerances, can cause l
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 Conditin I and Condition II 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 fu 1 assembly is marked with an identification number and loaded in accordance with a core loading dia-I l
gram.
Before core loading, the fuel assemblies in the Spent Fuel Pool, desig-l l
nated for the next fuel cycle, will have the fuel assembly identification l
numbers and insert identification numbers checked.
Following core loading, the I
fuel assembly idereification numbers are again checked as final assurance that the core has been loaded properly.
~
The power distortion due to any combination of misplaced fuel assemblies would significantly raise peaking f actors and would be readily observable with incore i
l 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 enthalpy rise.
Incore flux measurements are taken during the startup subsequent to every refueling operation.
l 15.4-24 1988 Update
CNS REFERENCES FOR SECTION 15.4 1.
Risher, D. H., Jr. and Barry, R. F., " TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Coce", WCAP-7979-A (Proprietary) and WCAP-8028-A (Non-Proprietary), January 1975.
2.
Hargrove, H. G., "FACTRAN - A Fortran-IV Code for Thermal Transients in a UO Fuel Rod", WCAP-7908, June 1972.
2 3.
Chelemer, H., Boman, L. H., Sharp, D. R., " Improved Thermal Design Pro-cedures", WCAP-8567, July 1975.
4.
Burnett, T. W. T., et al., "LOFTRAN Code Description", WCAP-7907-P-A (Proprietary) and WCAP-7907-A (Non-Proprietary), April,1984.
5.
" Westinghouse Anticipated Transients Without Trip Analysis", WCAP-8330, August 1974.
6.
Morita, T., et al., " Dropped Rod Methodology for Negative Flux Rate Trip Plar.ts", WCAP-10297-A (Proprietary ) and WCAP-10298-A (Non-Proprietary),
June 1983.
7.
Taxelius, T. G. (Ed), " Annual Report - Spert Project, October, 1968, September,1969", Idaho Nuclear Corporation IN-1370, June 1970.
8.
Liimataninen, R. C. and Testa, F. J., " Studies in TREAT of Zircaloy-2-Clad, U0 -Core Simulated Fuel Elements", ANL-7225, January - June 1966, 2
- p. 177, November 1966.
9.
Bishop, A. A., Sanburg, R. O. and Tong, L. S., " Forced Convection Heat Transfer at High Pressure Af ter the Critical Heat Flux", ASME 65-HT-31, August 1965.
10.
Risner, D. H., Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods",
WCAP-7588, Revision 1-A, January 1975.
tj,
- u. G. N. R. C. 0 4fe'c e o f A!wlea r Reac fo r fy Llohn " Ma nda rd gevo es F'l.m ", AWKE G -eP00,
Revision I,
b y 1991, c e cko n ts.cl.L 15.4-37
)
i Table 15.4.1-1 (Page 2)
Time Seouence of Events for Incidents Which Cause Reactivity and Power Distribution Anomalies Accident Event Time (sec.)
Startup of an Initiation of pump startup 1.0 Inactive Reactor Coolant Loop at Power reaches P-8 trip 13.4 an Incorrect setpoint
-Temperature Rods begin to drop 13.9 Minimum DNBR occurs 15.0 CVCS Malfunction that Results in a Decrease in the l
a#4c4c2 fase
[ rep ace wdh M5ed 60'"
Boron Concentration in the Reactor Coolant h
h. Dilution during Dilution begins N fueling BDMS setpoint is exceeded 633 Criticality occurs 750 2.
Dilution duri Dilution begins 0
cold shutdown MS setpoint is excee0<fd 633 Cri 41ity occurs /
750 3.
Dilution during Dilution b i 0
hot shutdown BDMS setpoint is e ceeded 633 Criticality occurs 750
/
4a.
Dilution during
'lution begins 0
hot standby (w/o stuck rod)
BDMS setpoint is exceeded 1520 Criticality occurs 1887 j.
4b.
Dilutio during Dilution begins 0
l hot andby l
stuck rod)
BDMS setpoint is exceeded 1520 Criticality occurs 1887
--m --
m..-
+.
E.-
la.
Dilution during Dilution begins power operation (manual rod _ control)
Reactor trip setpoint reached
-0 Operator terminates dilution
-<1039 lb..
Dilution during Dilution begins-power operation (automatic rod Rod insertion limit alarm 0
control)
.setpoint reached Operator terminates dilution
<1618 2.
Dilution during Dilution begins startup-Reactor trip setpoint reached 0
Operator-terminates dilution
<1039
- 3a. -
Dilution during Dilution begins 0
-hot standby (BDMS operable)
.BDMS setpoint reached 1237 Dilution source isolated 1262 Borated water reaches core
<1523 3b..
Dilution during Di.lution begins.
O hot standby (BDMS inoperable)
High-flux-at-shutdown 1816 alarm setpoint reached Operator _ terminates dilution
<2716 4a.
Dilutioniduring Dilution begins 0
hot shutdown
-(BDMS operable)
BDMS setpoint reached 1359 i
Dilution source isolated 1384 Borated water reaches core
<1681 4b.
' Dilution during-Dilution begins-0
-hot shutdown
-(BDMS inoperable).
High-flux-at-shutdown 1816 alarm setpoint reached.
-Operator-terminates-dilution
<2716 Sa, Dilution during Dilution begins O'
cold shutdown (BDMS operable)
BDMS setpoint reached 739 Dilution source isolated 764 Borated water reaches core
<885
1 Table 15.4.1-1 (Page 3)
Time Seouence of Events for Incidents Which Cause ReactivityandPowerDistributionAnomalies Dilution during Power range low setpoint reactor startup trio due to dilution Criticality occurs (if dilution 912 continues after trip) 6.
Dilution during N
full power operation a.
Automatic OperatoNrecejves low-low rod in-0 reactor sertion lDrit alarm due to dilution control
/ 'N Shutdow,n margin lost (if dilution 3894 cont'inues after trip
[Dilutionbegins b.
Manual 0
g/
reactor control Reactor trip setpoint reached fo 1074 overtemperature AT
/
Shutdown margin is lost (if dilution 3894 continues after trip)
I L/
h ref ace ~,%
> n se H % d e inec< f ege l
Accident Event Timo (sec.)
I Sb.
Dilution during Dilution begins O
cold shutdown (BDMS inoperable)
High-flux-at-shutdown 1816 alarm setpoint reached Operator terminates dilution
<2716 6a.
Dilution during Dilution begins O
refueling (BDMS operable)
BDMS setpoint reached 1024 Dilution source isolated 1049 Borated water reaches core
<1267 6b.
Dilution during Dilution begins 0
refueling (BDMS inoperable)
High-flux-at-shutdown 3441 alarm setpoint reached Operator terminates dilution
<5241 i
i I
- - - -