ML20006B888
| ML20006B888 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 01/26/1990 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20006B886 | List: |
| References | |
| NUDOCS 9002060056 | |
| Download: ML20006B888 (58) | |
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ITEM 1 INCREASE REFUELING BORON CONCENTRAT1011 A
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.O DEFINITIONS .
REACTOR OPERATING CONDITIONS (Continued)
Cold Shutdown Condition (Operating Mode 4)
. The nector coolant T eoid is less than 210*F and the mactor coolant is at shutdown boron concentration.
. t Refuelino shutdown Condition (Operating Mode 5)
The reactor coolant is at refueling boron concentration and Tcold is less than 210'F.
Refuelino Operation Any operation involving the shuffling, renoval, or replacement of nuclear fuel, CEA's, or startup sources.
,,..,, The Refuelino Boron Concentration @
A reactor coofant boron concentration of at least g which corre- l4_
sponds'to a shutdown margin of not less than 5% with all CEA's withdrawn.
Shutdown Boron Concentration The boron concentration required to make the reactor subcritical by the amount defined in paragraph 2.10.
Refuelino Outace or Refuelino Shutdown A plant outage or shutdown to perform refueling operations upon reaching the planned fuel depletica for a specific core.
Plant Operatino Cycle The time period from a Refueling Shutdown to the next Refueling Shutdown.
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AnendmentNo.24.72,#,$),/g/
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. + 2.0 LIMITING CONDITIONS FOR OPER Q 103 2.2 Chemical and Volume Control System (Continued) :
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- a. One of the operable charging pumps may be remove 6 from service provided two charging pumps are operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
e b. Both boric acid pumps may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l c i
- c. One concentrated boric acid tank may be out of service provided !
a minimum of 68 inches of 6 1/4 percent to 12 percent by weight E
boric acid solution at a temperature of at least 20'F above -
saturation temperature is contained in the operable tank and ;
provided that the tank is restored to operable status within 24 ;
L hours. !
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-d. Only one flow path from the concentrated boric acid tanks.to .
the reactor coolant system may be operable provided that either _
the other flow path from the concentrated boric acid tanks to L the. reactor. coolant system or the flow path from the SIRW tank i to the charging pumps is restored to operable status within 24 I hours. *
- e. One channel of_ heat tracing may be out of service provided it :
is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- f. One level instrument on each concentrated boric acid tank may be out of service for ?? hours.
Basis ;
' The chemical coolant systemand volume baron contro1gstem inventory. This provides control is normally of the reactor accomplished by using any one of the three charging pumps in series with one of the two boric acid pumps. An alternate method of boration will be to use the
- charging pumps directly from the SIRW storage tank. A third method will be to depressurize and use the safety injection pumps. There are two sources of borated water available for injection through three different paths.
(1) The boric acid pumps can deliver the concentrated boric acid tank l contents' (6-1/4 - 12 weight percent concentration of boric acid) to the charging pumps. The tanks are located above the charging pumps -
so that the boric acid will flow by gravity without being pumped.
(2) The safety injection pumps can take suction from the SIRW tank which maintains a boric acid concentration greater than the required refueling concentration.
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l- .: e 2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emeroency Core Coolino system Applicability Applies to the operating status of the emergency core cooling system.
Objective l
To assure operability of equipment required to remove decay heat from the core, ,
Specifications j
(1) Minimum Requirements i The reactor shall not be made critical unless all of the following [
conditions' are met:
- a. The SIRW tank contains not less than 283 100 gallons of water -
with a boron concentration of at lea {tG006 ppm at a temperature .
not less-than 50'F. Ng l-(
- b. One means of temperature indication (local) of the SIRW tank is operable. l
- c. All four safety injection tanks are operable and pressurized to '
at least 240 psig with a tank liquid of at least 116.2 inches
'Q (67%) and a maximum level of 128.1 inches (74%) with refueling
.Ud"' boron concentration. '
- d. One level and one pressure instrument is operable on each safety injection tank,
- e. One low pressure safety injection pump is operable on each bus. ;
- f. One high precsure safety injection pump is operable on each bus, i
- g. Both shutdown heat exchangers and three of four component cooling
- heat exchangers are operable. .
- h. Piping and valves shall be operable to provide two flow paths from the SIRW tank to the reactor coolant system.
1
- i. All valves, piping and interlocks associated with the above components and required to function during accident conditions are operable. HCV-2914, 2934, 2974, and 2954 shall have power ,
removed from the motor operators by locking open the circuit breakers in the power supply lines to the valve motor operators.
FCV-326 shall be locked open.
h 2-20 Amendment No. 17,32,43,IO3,II7,//
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, 2.0 .!MITING CONDITIONS FOR OPERATION '
w 2.3 ;meroency Core Coolina System (Continued) .
(3) Protection Against low Temperature Overpressurization The following limiting conditions shall be &pplied during scheduled 1 heatups and cooldowns. Disabling of the HPSI pumps need not be i required if the reactor vessel head, a pressurizer safety valve, or j a PORV is removed. l Whenever the reactor coolant system cold leg temperature is below l 320'F, at least one (1) HPSI pump shall be disabled. :
Whenever the reactor coolant system cold leg temperature is below 4 312'F, at least two (2) HPSI pumps shall be disabled. I i
Whenever the reactor coolant system cold leg temperature is below .
I 271'F, all three (3) HPSI pumps shall be disabled. ,
In the event that no charging pumps are operable, a single HPSI pump may be made operable and utilized for boric acid injection .
to the core. .
Basis
- The nomal procedure for starting the reactor is to first heat the reactor ;
coolant to near operating temperature by running the reactor coolant pumps.
.4 The reactor is then made critical by withdrawing CEA's and diluting boron in the reactor coolant. With this mode of start-up, the energy stored in the reactor coolant during the approach to criticality is substantially '
equal .to that during power operation snd therefore all engineered safety ,
features and auxiliary cooling systems are required to be fully operable. 1 During low power physics tests at low temperatures, there is a negligible amount of stored energy in the reactor coolant; therefore, an accident comparable in severity to the design basis accident is not possible and the engineered safeguards systems are not required.
Tlie SIRW tank contains a-migiipum of 283,000 gallons of usable water /D contain- '
ing at least1898-piim borontII. This is sufficient boron concentration to l#
provide a shutdown margin of 55, including allowances for uncertainties, with all control rods withdrawn and a'new core at a temperature of 60*F.(2) ,
The limits for the safety injection tank pressure and volume assure the required amount of water injection during an accident and are based on values used for the accident analyses. The minimum 116.2 inch level corresponds to a volume of 825 ft3 and the maximum 128.1 inch level corresponds to a volume of 895.5 ft3 Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked. Since the system is used for shutdown cooling, the valving will be changed and must be properly aligned prior to start-up of the g reactor.
2-22 Amendment No. J7,U,ff,f7,$f,7f,77,Jpp,jpg
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? 2. 0 ,!MITMNG CONDITIONS FOR OPERATION g .. 2.8 Itefue ina Doerations (Continuen) incident ~could occur during the refuelin I y a hazard to public health and safety (1)g Whenever made in s: ore geometry one flux monitor is sufficient. This permits operations that would changes result are not beingin
maintenance of the instrumentation. Continuous monitoring of radiation ;
levels and neutron flux provides immediate indication of an unsafe condi- ;
tion. The shutdown cooling pump is used to maintain a uniform boron -l concentration. . 1 The shutdown margin as indicated will keep the core suberitical even if !
all CEA's were withdrawn from the core. During refueling operations, the -
reactor refueling cavity is filled with approximately 250,000 gallons of_ /M l borated water. The borca concentratton of this water (at leasy1815. Opm I 4i boron) is sufficient to maint&in the reactor suberitical by moes-than 5%, '
including allowance for uncertainties, in the cold condition with all rods j
' withdrawn.(2) Periodic checks of refueling water boron concentration !
ensure the proper shutdown margin.
Communication requirements allow the !
control room operator to inform the refueling machine operator of any i impending unsafe condition detected from the main control board indicators during fuel movement.
In addition to the above engineered safety features, interlocks are utilized during refueling operations to ensure safe handling. An excess '
l weight interlock is provided on the lifting hoist to prevent movement of 1 more than one fuel assembly at a time. In addition, interlocks on the !
auxil,iary building crane will prevent the trolley from being moved over
,,, storage racks containing irradiated fuel, except as necessary for the w handling of fuel. The restriction of not moving fuel in the reactor for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the power has been removed from the core takes j advantage of the decay of the short half-life fission products and allows for any failed fuel to purge itself of fission gases, thus reducing the consequences of fuel handling accident.
The ventilation air for both the containment and the spent fuel pool area flows through absolute particulate filters and radiation monitors before discharge at.the ventilation-discharge duct. In the event the stack discharge should indicate a release in excess of the limits in the ,
i technical specifications, the containment ventilation flow paths will ba closed automatically and the auxiliary building ventilation flow paths -
will be closed manually. In addition, the exhaust ventilation ductwork from the spent fuel storage area is equipped with a charcoal filter which will be manually put into operation whenever irradiated fuel is being handled.(1)
Ref6tences
! (1) SAR, Section 9.5 (2) SAR, Section 9.5.1.2 -
l h 2-39 AmendmentNo.2A,75,Z82,///
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fa f 2.0 LIMITING CONDITIONS FOR OPERATIONS
! @ 2.14 Enoineered Safety Features System Initiation instrumentation W Settings (Continued)
(3) Containment High Radiation (Air Monitoring) (Continued) l l
4 The setpoints for the isolation function will be calculated in - ;
accordance with the ODCM. !
Each channel is supplied from a separate instrument A.C. bus i and each auxiliary relay requires power to operate. On failure
. of a single A.C. supply, the A and 8 matrices will assume a !
one-out of-two logic.
(4) Low Steam Generator Pressure >
A signal is provided upon sensing a low pressure in a steam generator to close the main steam isolation valves in order to 3 minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possible addition !
of reactivity. The setting of 500 psia includes a +22 psi uncertainty and was the setting used in the safety analysis.I3) ,
Closure of the MSIVs (and the bypass valves, along with main -3 feedwaterisolationandbypassvalves)isaccomplishedbythe i
steam generator isolation signal which is alpgica1.combTsation 3
g, of low steam generator pressure or high containment pressure, ,;
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As part of the AFW actuation logic, a separate signal is provided to tenninate flow to a steam generator upon sensing a ,
low pressure in that steam generator if the other steam generator 4 pressure is greater than the pressure setting. This is done to ;
minimize the temperature reduction in t'ne reactor coolant system in the event of a main steamline break. The setting of 466.7 psia includes a +31.7 psi uncertainty; therefore, a setting of 4'5 psia was used in the safety analysis.
(5) SIRW Tank Low Level, ,
Level switches are provided on the SIRW tank to actuate the .
valves in the safety injection pump suction lines in such a manner so as to switch the water supply from the SIRW tank to the containment sump for a recirculation mode of operation after a period of approximately 24 minutes following a safety injection signal. The switchover point of 16 inches above tank bottom is set to prevent the pumps from running d y during l the 10 seconds required to stroke the valves and ho d p / 9 C d, t- reserve approximately 28,000 gallons of at leas ' -ppm I borated water. The FSAR loss of coolant accide t1 nalysis(4) assumed the recirculation started when the minimum usable
- l. volume of 283.000 gallons had been pumped from the tank.
B 2-62 AmendmentNo.5,#,0.Q.M,J9/,[h
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- 4.0 XSIGN FEATURES .
! 4. 4 ruel 5torace 4.4.1 New Fuel Storace The new unirradiated fuel bundles will nomally be stored in the dry l new fuel storage rack with an effective multiplication factor of less ;
than 0.9. The open grating floor below the rack and the covers above ;
. the racks, along with generous provision for drainage, precludes i flooding of the new fuel storage rack. '
New fuel may also be stored in shipping containers or in the spent i l fuel pool racks which have a maximum effective multiplication factor l of 0.95 with Fort Calhoun Type C fuel and unborated water. 4 i
The new fuel storage racks are designed as a Class I structure. j J
4.4.2 Spent Fuel Storace ;
Irradiated fuel bundles will be stored prior to off-site shipment in the stainless steel lined spent fuel pool. The spent fuel pool is t' nomally filled with borated water with a concentration of at least 3 ppm.
gm ,
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.. The spent fuel racks are designed as a Class I structure. -
c#y.
Normally the spent fuel pool cooling system will maintain the bulk watcr temperature of the pool below 120'F. Under other conditions of fuel discharge, the fuel pool water temperature is maintained l below 140'F.
'J The spent fuel racks are designed and will be maintained such that the j calculated effective multiplication factor i's no greater than 0.95 (including all known uncertainties) assuming the pool is flooded with unborated water. The racks are divided into 2 regions. Region I racks !
are surrounded by Boraflext Region 2 racks have no poison. Acceptance !
criteria for fuel storage in Regions 1 and 2 are delineated in Section -
2.8 of these Technical Specifications.
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l 2 l 4-4 Amendment No. J),f),M,/I/
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j-SUSPEND CERTAIN SAMPLING WHEN ALL FUEL HAS BEEN REMOVED FROM THE REACTOR VESSEL i
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- 2.0 LE!ITI!!G CCI.'DITIONS FOR OPPRATIC;!
e 1 2.8 Refueling Overatiens ,
Applicability L ' Applies to operating limitations during refueling operations.
Objective To minimise the possibility of an accident occurring during )
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refueling operations that could affect public health and i safety.
g Scecifications k-
The following conditions shall be satisfied during any refuel- '
ing operations:
- .
(1)- The equipment hatch and one door in the air lock shall be properly-closed. In addition, all automatic contain- l ment isolation valves shall be operable or at least one valve in each line shall be closed.
(2) The five containment atmosphere and plant ventilation duct re.diatica monitors ths.t initiate closure of the f containment pressure relief, air sample, and purge
<(. . system valves shall be tested and verified to be oper-able im=ediately prior to refueling operations. The five menitors shall employ one-out-of-five logic from .
separate contact outputs for VIAS.
(3) Radiation levels in the containment and spent fuel storage areas shall be monitored continuously.
(h) ^ Whenever core geometry is being changed, neutron flux 1 shall be continuously monitored by at least two source range neutron monitors, vith each monitor providing continuous visual indication in the control room. When t core geometry is not being changed, at least one source range neutron monitor shall be in service.
1 (5) At'least one shutdown cooling pump and heat exchanger shall.be in operation. However, the pump and heat ex- -
changer may be removed from operation for up to one hour M per B hour period during the performance of core alter-stions in the vicinity of the reactor coolant hot les loops or during manipulation of a source. I E .
(6) During reactor vessel head removal and while refuel
.I sb operations are being performed ing boron concentration in thened shall be mainta reactor, in the the reactrefuel g, 'M
'Al, k ecolant system and shall be checked by sampling on ene shift. ,.
Amendment.io./7,h 2-37 a .- . - . - . - .
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2 . '-) LIMITING CONDITIONS FOR OPERATION l ,
2.8 ite t ueling Operations (Continued) kt- L .J 77 Direct communication be tween personnel in~the con- 'I Y
trol room and at the refuelin.3 machine shall be available whenever changes in core geometry are takinj place.
I M1 When irradiated fuel is being handled in the auxi-l .
7 'liary building, the exhaust ventilation from the y spent fuel pool area will be diverted through the J
charcoal f11ter.
e (L g M7 Prior to initial core loading and prior to refuel- l ing operations, a complete check out, including a b load test, shall be conducted on fuel handling cranes that will be required during the refueling operation to handle apent fuel assemblies.
A minimum of 23 feet of water above the top of the I l- k ,L.1-0 7 core shall be maintnined whenever irradiated fuel I is being handled.
[g,M-i~f Storage in Region 1 and Region 2 of the spont fuel l racks shall be restricted to fuel assemblies having initial enrichment less than or equal to 4.0 weight percent of U-235.
Storage in Region 2 of the spent fuel racks shall .l p M-27
.he restricted to those ansemblics whose paraineters f all within the " acceptable" region of Figure 2-10.
'If any of the above conditions are not met, all refueling operations shall cease immediatuly, work shall be initiated to satisfy the required conditions, and no operations that may change the reactivity of the core shall be made. Ilow-ever, refueling operations may commence and continue with less than 5 containment atmosphere and plant ventilation duct . radiation monitors provided that gross, particulate and ' iodine nonitors are monitoring the stack of fluent.
These three plant ventilation duct radiation monitors will initiate closure of the containment pressure relief, air sa nplo and purge system valves and shall employ a one-out-of-three lojic for the initiation of VIAS.
Irradiated fuel movement shiill not be initiated before the
. reactor core has decayed for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the reactor.has been operated at power levels in excess of 2%
rated power.
hasis The equipment an:1 general procedures to be utilized during refueling operations are discunsed in the USAR. De ta iled instructions, the above speci f icat ions, and the design of t hei f uel handling equipment incorpor.iting built-in inter-locks and safety features provide issurance that no Amencment No. 3, 2A, 25, 41. M 2-38 k + _ .
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i-r TABLE '-4 (Continued) l
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MINIMUM FRE0VENCIES FOR SAMPLING TEST
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Type of Measurement- Sample and Analysis and Analysis Frecuency ,
I
- 1. : Reactor Coolant
[ (Continued) :
~(c) Cold Shutdown (1) Chloride 1 per 3 days I' (OperatingMode4) :
i (d).- Refueling Shutdown (1) Chloride 1 per 3 days (3) ;
(Operating Mode 5) (2) Boron Concentration 1 per 3 days (3) l L
L (e) Refueling Operation (1) Chloride 1 per 3 days (3) l (2) Boron Concentration 1 per shift (3) ,
- 2. SIRW Tank Boron Concentration l'per 31 days [
I 3. Concentrated Boric Acid Boron Concentration 1 per 31 days ,
F Tanks it ;
g 4. SI Tanks Boron Concentration 1 per 31 days 't
~5. Spent Fuel Pool; . Boron Concentration 1 per 31 days I ,
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-(l) Until-the radioactivity.of th'e reactor coolant is restortd'tu s l Ci/gm DOSE EQUIVALENT l-131. j r ,
o -(2) Sample to be taken after a minimum of 2 EFPD and 20 days of power operation f
- have elapsed since reactor was subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
{
.(3) Boron and Chloride sampling / analyses are not required when the core has 4
- been off loaded. Reinitate boron and chloride sampling / analyses one shift .
prior to reintroduction-of fuel into the cavity to assure adequate shutdown l margin is maintained.
3-19 1
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i ITEl4 3 REVISE STORAGE REQUIREMENTS FOR SPENT FUEL POOL REGI0ff 2 s.
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( SPENT FUEL POOL REGION 2 STORAGE CRITERIA j
Minimum Required fuel Assembly Exposure os o function of
-Initial Enrichment to Permit Storage in Region 2 i'
,. a0.0 g
[, S 37.5 S. -s
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1 35.0 r ,
k '32.5 Acceptable w/0 CEA P
in Region 2 o 30.0 L-a'
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J 27.5 l
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M* - 0 22.5 '
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/ Accept ble ,wlTH CEA in Region 2 u 15.0
.x "o 12.5 o.
m 10,0 /
[ j j 5.0 NOT occeptable in Region 2 2.5 0.0 2.00 2.25 2.50 2.75 3.00 3.25 3.50 3.75 4.00 4.25 4.50 Assembly Initial Enrichment, w/o U-235
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SPENT FUEL POOL REGION 2 OMAHA PUBLIC POWER DISTRICT FIGURE
- STORAGE CRITERIA FORT CALHOUN STATION-UNIT No.1 2-10 ,
Amendment No. I
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'> f ALLOW DIRECT TRANSFER OF SPENT FUEL i p; FROM THE REACTOR CORE TO THE t
_. SPENT FUEL POOL REGION 2 t 9: ;
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[ 2.8 ite t ueli no Operations (Continued) ,
(7) Direct communication between personnel in the con-i trol room and at the refueling machine shall be available whenever changes in core geometry are
[
taking place.
(B). When irradiated fuel is being handled in the auxi-
liary building, the exhaust ventilation from the L spent f uel pool area will be diverted through the charcoal filter.
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E (9) Prior to initial core loading and prior to refuel- I ing-operations, a complete check out, including a
[ load test, shall be conducted on' fuel handling crancs that will be required during the ref ueling operation to handic spent fuel assemblies.
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(10) A minimum of 23 feet of water above the top of the core shall be maintained whenever irradiate 1 fuel is being handled.
(11) Storage in Region 1 and Region 2 of the spent fuel racks shall be restricted to fuel assemblies having ,
. initial enrichment less than or equal to 4.0 weight
.. percent of U-235.
(12) Ator69e in' Region 2 of the spent fuel racks shall ;
he restricted to those ansemblies whose paraineters f.all.within the " acceptable" region of Figure 2-10. l if any of the above conditions are not met, all-refueling operations shall cease immediately, work shall be initiated to satisfy the required conditions, and no operations that ,
may change the reactivity of the core shall be made. Ilow- ,
ever, ref ueling operations may commence and continue with less than 5 containment atmo9phere and plant ventilation duct radiation monitors provided that gross, particulate and -iodine nonitors are monitoring the stack ef fluent.
These three plant ventilation duct radiation monitors will initiate closure of the containment pressure relief, air E - - -
sample and purge system valves and shall employ a one-out- ;
I ot-three lojic for the initiation of VIAS. i
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Irradiated fuel movement shsil not be initiated be fore the reactor core has decayed for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the reactor has been operated at power levels in excess of 21 rate 1 power.
T:a s i s The equi;. ment ani general procedures to be utilized during refueling operations are discunsed in the USAR. De t a iled instructions, the above specifications, and the design of the fuel handling equipment incorpora ting built-in inter-locku and safety features provide issurance that no AgencmentNo.5.24,25,di.M 2-33
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The spent fuel assembly may be transferred directly from the reactor t; ore to the spent fuel pool Region 2 provided the independent verification of assembly burnups as defined in Special Procedure SP-BURNUP.1 has been completed and the I assembly burnup meets the acceptance criteria identified in Technical l Specification Figure 2-10. l
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' DELETE REQUIREMENT FOR FUEL r M T _ ,' " - PERFORMANCE REPORT AT END N, ,
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. l 5.9.3 Special Reports ,
Special reports shall be submitted to the Regional Administrator of !
) the appropriate NPC Regional Office within the time period specified I for each report.- These reports shall be submitted covering the Lctivities identified below pursuant to the requirements of the applicable reference specification where appropriate: ;
L a. in service irspection repert, reference 3.3.
-b. Tendon surveillance, reference 3.5.
- c. Centainment structural tests, reference 3.0, i
- d. Special naintenance reports.
- e. Containment leaf rate tests, reference 3.5. ,
- f. Radioactive effluer.t releases, reference 2.0
- g. Paterials radiation surveillance specimens reports, reference 3.3.
Se-i- grf ewnence-ftMoving-eech-refveLing-evtage - C - l h.
t 13, % Fire protecticn equipment outage, reference 2.19.
$,6 Post-accident monitoring irstrumentation, reference 2.21. t 5.9.4 Unioue Peporting Requirements
- a. Radioactive Effluent Release Peport A report covering the operation of the Fort Calhoun Station during the previous six months shc11 be submitted within C0 days after January 1 and July 1 of each year per the requirements of 10 CFR 50.36a.
The radioactive effluent release report shall include a summary of the quantities of radioactive licuid and gaseous effluents and solid waste released from the plant as outlined'in Regulatory Guide 1.21 Revision 1.
The radioactive effluent release report shall include a sumary cf the meteorological conditient concurrent with the release of gasecus effluents during each ouarter as outlined in Pegulatory ;
Guide 1.21 Revision 1.
The radioactive effluent release report shall include an assessrent >
of radiaticn doses from the radioactive liquid and gasecus effluents released from the unit during each calendar cuarter as outlined in r egulatory Guide 1.21, Revision 1. In addition, the unrestricted crea boundary maximum noble gas gamma air and beta air doses shall be evaluated. The meteorological ecnditions concurrent with the 5-15 Amendment No. f, 7E, #F,E6, 110,
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INCREASE REFUELING BORON CONCENTRATION i-i.?
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QISCUSSION OF CHANGL The proposed amendment to the Technical Specification increases the reactor vessel refueling boron concentration from 1800 to 1900 ppm. The increased concentration is required to maintain adequate shutdown margin for Cycle 13.
L The refueling boron concentration is defined to provide adequate shutdown
- margin and provide adequate operation response time in the event of an
! inadvertent dilution event. The shutdown margin and boron concentration
[' requirements are determined using OPPD methodology outlined.in the OPPD Topical Report OPPD-NA-8303 Rev. 02 (Reference 1). The boron dilution event analysis ensures a dilution to critical time of not less than 15 to 30 minutes (mode e dependent) assuming the maximum credible influx of unborated water.
~
The requirements'of the refueling boron concentration are to provide:
- 1) a dilution time to critical not less than 30 minutes, and
- 2) a shutdown margin of not less than 5% assuming a freshly loaded core with all CEAs withdrawn, in accordance with the Technical L Specification definition of refueling boron concentration.
For Cycle 13, the current Technical Specification refueling boron concentration of 1800 ppm meets neither of the above criteria.. A revised refueling boron concentration of 1900 ppm has been shown to be adequate to fulfill both of the
- above requirements. The attached table summarizes the results of the boron l
dilution analyses for Cycles 12 and 13 for the Refueling Mode (Mode 5).
-JUSTIFICATION Increasing'the reactor vessel refueling boron concentration to 1900 ppm will maintain an adequate shutdown margin for Cycle 13.
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TABLE 1 REFUELING MODE INPUTS AND RESULTS OF THE BORON DILUTION EVENT I Parameter Cycle 12
'Cvele 13 Critical Boron Concentrations.
All Rods Out, Zero Xenon (ppm) 1400 1454 Inverse Boron Worth Assumed in Dilution to critical calculation (ppm / Zap)
-55 -55 Time to Loose Prescribed shutdown Margin (min.) 31.2 33.3 Refueling Boron Concentration 1800 1900 Actual Shutdown Margin (2) 5.0 5.2 i
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i NO SIGNIFICANT HA7ARDS l
l The ~ proposed amendment to the Technical Specifications will increase the ;
reactor vessel refueling boron concentration from 1800 to 1900 ppm. !
The Technical Specification document changes required are found in Sections '
2.2, 2.3, 2.8,-2.14'and 4.4 on pages 2, 2 18, 2-20, 2-22, 2-39,.2-62, and 4.4.
The refueling boron concentration for Cycle 13 is defined to provide adequate' !
shutdown margin and provide adequate operation response time in the event of an :
inadvertent dilution event. The shutdown margin and boron concentration requirement is determined using OPPD methodology outlined in the OPPD Topical .
Report OPPD NA-8303, Rev. 02 (Reference 1). The boron dilution event' analysis ensures a dilution to critical time of not less than 15 to 30 minutes (mode r
-dependent) assuming the maximum credible influx of unborated water. . ,
.r BASIS FOR NO SIGNIFICANT HA7ARDS DETERMINATION This proposed amendment does not involve a significant hazards consideration because the operation of Fort Calhoun Station in accordance with this amendment
- would not:
- 1) Involve an increase in the probability or consequences of an accident previously evaluated. The revised refueling boron concentration ensures the existence of both a 5% Acor greater shutdown margin with all CEAs ,
withdrawn from the core and a dilution time to critical which is greater or '
equal to 30 minutes. Thereforo, this change does not increase the probability or consequences of a previously evaluated accident.
- 2) Create the possibility of a new or different kind of accident from any accident previously evaluated. It has been determined that a new or different kind of accident is not created because no new or different modes of operation are proposed for the plant. The continued use of the existing
. Technical Specification controls prevents the possibility of a new or different kind of accident. +
- 3) Involve a reduction in the margin of safety. Specifications involving the minimum refueling shutdown margin are maintained above the minimum required margin and the dilution time to critical conforms to current plant conditions and, therefore, preserves the margin of safety. Increasing the boron concentration ensures that the minimum shutdown margin is maintained and, therefore, will not reduce the margin of safety.
Based on the above considerations, OPPD does not believe that this amendment involves a significant hazards consideration.
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W ITEli 2 SUSPEND CERTAIN SAfiPLING WHEN ALL FUEL HAS BEEN REliOVED FR014 THE REACTOR VESSEL
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DISCUSSION OF CHANGE The proposed amendment to the Technical Specification allows for suspension of boron sampling in the reactor vessel when all fuel has been removed. Sampling will be reinitiated prior to reintroduction of the fuel into the reactor vessel 3
to insure adequate shutdown margin.
Suspension of boron sampling of the reactor vessel coolant when all fuel is removed will not affect the plant safety since no fuel is present, n
'The reactor vessel coolant boron concentration requirement is based on the need
< for adequate shutdown margin when fuel is present. When all the fuel is removed, the need for boron is eliminated and hence the need for sampling is eliminated. Elimination of the sampling requirement-for the reactor vessel-head removal will not adversely impact the safe operation since the shutdown i margin calculations'do not credit the CEA's. The intent of the reactor vessel i < head removal would be to ensure CEA's were not inadvertently withdrawn causing
- a criticality excursion, however since the refueling shutdown calculations include an all rods out assumption, then the deletion of the boron shift sampling requirements will not change the safety analyses. A sampling
- frequency of once per.3 days would be consistent with refueling shutdown conditions.
The deletion of the chloride sampling will not adversely impact the fuel since the purpose of maintaining the chloride chemistry level is to meet warranty obligations-of the fuel vendor and reduce the possibility of intergranular stress corrosion cracking in the fuel assembly material. The chloride chemistry level is established to prevent any potential degradation of the fuel mechanical design properties or RCS piping. When fuel is not present in the reactor cavity a sampling frequency of once per 3 days is consistent with refueling shutdown conditions. The chloride chemistry level of the fuel assemblies is met by sampling of the Spent Fuel Pool.
JUSTIFICATION The suspension of reactor vessel coolant boron sampling or chloride sampling l when all fuel is removed from the vessel does not compromise or affect the safety of the plant operation, i
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Il NO SIGNIFICANT HAZARDS i- The proposed amendment to the Technical Specification allows for suspension of boron and chloride sampling in the reactor vessel when all fuel has been removed.
'The Technical Specification document changes required are contained in pages 2-37, 2 38, and 3-19 of Sections 2.8 and 3.2.
s- The reactor vessel coolant boron concentration requirement is based on the need for adequate shutdown margin when fuel is present. When all the fuel is removed, the need for boron is eliminated and hence the need for sampling is eliminated. Elimination of the sampling requirement for the reactor vessel head removal will not adversely impact the safe operation since the shutdown margin calculations do not credit the CEAs. The intent of the reactor vessel head removal would be to ensure CEAs were not inadvertently withdrawn causing a criticality excursion, however since the refueling shutdown calculations include an all rods out assumption, then the deletion of the boron shift sampling requirements will not change the safety analyses. A sampling frequency of once per 3 days would be consistent with refueling shutdown conditions.
The deletion of the chloride sampling will not adversely impact the fuel since the purpose of maintaining the chloride chemistry level is to meet warranty obligations of the fuel vendor and reduce the possibility of intergranular i stress corrosion cracking in the fuel assembly material. The chloride chemistry level is established to prevent any potential degradation of the fuel mechanical design properties or RCS piping. When fuel is not present in the reactor cavity a sampling frequency of once per 3 days is consistent with l refueling shutdown conditions. The chloride chemistry level of the fuel assemblies is met by sampling of the Spent fuel Pool.
i BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION This proposed amendment does not involve a significant hazards consideration because the operation of Fort Calhoun Station in accordance with this amendment would not:
- 1) Involve a significant increase in the probability or consequences of an )
accident previously evaluated. This change allows for the suspension of boron and chloride sampling during the time the fuel is removed from the reactor vessel with no changes in specifications. Since the fuel source is i removed, shutdown margin in the reactor vessel is not required and hence !
boron sampling is not required and the mechanical design properties of the !
fuel or RCS piping are not subject to potential degradation due to !
intergranular stress corrosion cracking potentially induced by a high chloride level . Therefore, this change does not increase the probability or consequences of a previously evaluated accident.
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- 2) Create the.. possibility of a new or different kind of accident from any accident previously evaluated. It has been determined that a new or t different kind of accident is not created because no new or different modes of operation are proposed for the plant. ~ The use.of the proposed revised Technical Specification controls will not result in the possibility of a l
new or different kind of accident.
- 3) Involve a significant reduction in a margin of safety. Specifications E ' involving the boron sampling ensure that the shutdown' margin. conforms to L current plant conditions and, therefore, preserves the margin of safety.
P Since the fuel source is removed, shutdown margin in the reactor vessel is not required and hence boron sampling is not required. The fuel
[1 manufacturer's chloride chemistry requirements are met. by sampling of the
[ Spent Fuel Pool during the period the core is offloaded. This maintains-L the mechanical design properties of the fuel. Consolidation of all the L . boron and chloride sampling requirements in one location in the Technical
' Specifications ensures compliance of sampling requirements and, therefore, p will-not reduce-the margin of safety.
F Based on the above considerations, OPPD does not believe that this amendment
- involves .a significant hazards ~ consideration.
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.; ITEM'3 jk"y J" ' REVISE STORAGE REQUIREMENTS FOR
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SPENT FUEL POOL REGION 2 I
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a DISCUSSION OF' CHANGE The proposed amendment to the Technical- Specifications will allow spent fuel assemblies with limited exposure to be stored in Region 2 of the spent fuel pool, if a full length Control Element Assembly (CEA) Lis inserted in the fuel assembly prior to the move to Region 2.
The spent fuel storage racks at the Fort Calhoun Station are organized into two regions. Either new fuel or. discharged fuel may be stored in Region 1, but only spent fuel meeting the minimum exposure requirements of Technical Specifications Figure 2-10 may be stored in Region 2.
The computer code selection, qualification and analysis were performed by Pickard, Lowe and Garrick.
The computer codes-used were thoroughly qualified by comparison with measured critical experiments and other published data. These codes and the qualification performed are described below.
LEOPARD The LEOPARD code determines fast and thermal spectra and cell averaged cross-sections, using only geometric, a.aterial composition and' temperature data. The code models the fission product cross-sections for depleted fuel after removal from the reactor, and provides a direct interface to PDQ and CINDER.
BLACKCYL The BLACkCYL code uses LEOPARD-generated microscopic cross-sections and geometric data to produce four-group cell-averaged cross-sections for the CEAs for PDQ.
CINDER The CINDER code is a zero-dimensional depletion code used to determine the behavior of fission products as a function of time both during the power phase and after shutdown. After shutdown, the decay of fission products causes the average fission product absorption cross-section to vary, falling initially but rising monotonically after approximately half of a year until the end of the time for which calculations were performed (40 years). The stable isotopes accumulated after 40 years precludes the possibility of a decrease below the minimum value even after 40 years. The primary quantities provided by CINDER for spent- fuel rack criticality analysis are the average absorption cross-sections for the decaying fission products when these are a minimum, since this maximizes the reactivity of the fuel.
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I A' preprocessor'(INCIND) and'a post processor (CORR 34) are used with CINDER. i INCIND converts a LEOPARD output file into a CINDER input file. CORR 34 determines-the. ratio of the fission product-absorption cross-sections generated !
by CINDER to those produced _by LEOPARD at'the same fuel exposure.
PDQ The PDQ code is a diffusion theory code used to' determine the kao of the rack using cross-sections generated directly or indirectly from LEOPARD or BLACKCYL.
PDQ can be used for resolution of one-dimensional problems.
2.1.2 Oualification 1 The LEOPARD, PDQ, and BLACKCYL codes have been- qualified by the following extensive series of benchmarks against measured criticals:
- s Fourteen Westinghouse UO2 Zircalloy-4
~
experiments with 2.719 w/o U-235 fuel. glad cylindrical core critical s Five Battelle criticals with 2.35 w/o U-235 fuel.2 ,
- s. Five Saxton criticals with mixed oxide fuel.3 e Six Esada criticals with mixed oxide fuel.4 m- Ten Gattelle criticals with 4.31 w/o U-23E fuel.5 a Five Battis criticals with cylindrical absorbers.6 The first five sets of criticals were calculated using LEOPARD and PDQ only, while the final set required the use of BLACKCYL as well. The mean calculated k of these 45 measured criticals is 0.9947 and the standard deviation is
-0.0041. This results in a bias in the results of 0.0053 and a 95/05 uncertaintyLof 0.0086.
LEOPARD calculated results were compared to measured values of fuel composition as a function of burnup for the Yankee Core I and 11 and Saxton Core II spent fuel. LEOPARD accurately traced the actual fuel composition as a function of l exposure. CINDER and LEOPARD results were shown to be in close agreement.
CINDER calculated values of fission product concentrations were in good agreement with the activities for an average end-of-life core given in the WASH-1400 Reactor Safety Study.
Base Case The criticality analysis of the spent fuel pool storage rack was performed by:
- 1. Using LEOPARD to calculate the condition of the depleted fuel with enrichments of 3.25 w/o &-235 at several different exposure levels.
- 2. Using INCIND, CINDER and CORR 34 to compute the minimum fission product absorption cross-sections at selected exposures for each enrichment.
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- " 3.- Using LEOPARD, with input from the depletion runs and the fission
,- 4 product absorption cross-sections, to compute the cell-averaged cross.
sections for the fuel.
'4. Using LEOPARD to compute the cross-sections for the water and stainicis steel present in'the rack.
- 5. Using LEOPARD and.BLACKCYL to compute the cell-averaged cross-sections for the control element assembly cells. ..
- 6. .Using PDQ along with geometric information and the cross-sections generated above to calculate-the k,for the rack, with the CEAs in place,-for the enrichments and exposures used above. ,
l This analysis yields the k ,of the rack as a function.of enrichment and .
j Texposure,;before biases,' uncertainties, and accident conditions are considered.
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Calculation Results. q I
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The calculations performed for this analysis included the determination of.the 4-minimum amount of Byo. remaining in the CEAs after their exposure in the core and the determination of the minimum allowable fuel exposure as a function of '
initial enrichment for an assembly with an used CEA in place to be allowed into Region 2 of'the spent. fuel pool.
- The density of.B10 in the control element assembly is large enough that it essentially' depletes from the outside to the inside. This is so because the thermal disadvantage factor for a fresh control element at hot conditions is more than 50, and the absorption mean free path of thermal neutrons in the.
control element r.ssembly is 0.01 cm. Thus, as the B 10 is burned out of the i control element, it is assumed that the radius of the absorber decreases, while- i che number density of: Byo in the remaining central portion remains constant.
,_ A series of LEOPARD, BLACKCYL, and PDQ calculations were performed to determine the rate at which neutrons were absorbed in the control element as a function of the amount of B io remaining. Separate calculations vere performed for the ,
four outer rods of the CEA and the single center rod. Since virtually all of-the absorption is by the B10, this gives a differential equation for the amount of B10 as a function of exposure. This differential equation may be solved approximately by assuming that the rate of absorption varies in a piecewise linear fashion between the data points, which is reasonable.in light of the actual rates, as shown in Figure 1. The solution of a differential equation is shown graphically.in Figure 2.
The evaluation conservatively assumed full CEA insertion even though the rods have limited use for flux shaping, z.e. limited rod insertion. The CEAs were in place for cycles 1 through 10, which had a combined availability of 3,266 full power-days, which means that the CEAs were inserted for no more than 663.0
.i full power days. The minimum amount of Big remaining in the CEAs at 653.0 full power days is 85% as shown in Figure 2.
- . Criticality Analysis The criticality' analysis was performed by determining the k,as a function of initial enrichment and exposure, while considering the effects of calculational biases and uncertainties, perturbations such as mechanical-Ltolerances, and postulated accident conditions to' determine the minimum
. allowable exposure, as a function of enrichment, for an assembly with'a CEA in
- place to be moved to Region 2. For additional conservatism, it was assumed that only 752 of the B10 calculated value remains in the CEAs to be used in
' Region 2.
l Fuel Burnup. Uncertainty The uncertainty caused by the fuel burnup is estimated to be no more than'52 of 4 the calculated reactivity loss from the fresh state to the required exposure, j LThis'is based on calculated versus experimental results for fuel lifetime, on :
the~. ability of LEOPARD to accurately track fuel depletion, and on the ability 1 of CINDER to predict radionuclide. concentration. 4 1
Perturbations I The perturbations considered in the original Fort Calhoun Nuclear Station spent fuel rack analysis. included variations in fuel b'ox dimensions, water channel 1
width, stainless steel box thickness, pellet density, and pool temperature, i The first four.have only a small effect and are considered to be of the same i magnitude as in the original analysis. The most reactive temperature was 200'F in the original case, so a case with an enrichment of 4.00 w/o and 15,000 MWD /MTU exposure was run at 200"F, resulting in an increase of 0.0120 in k , somewhat less than in the case with no CEA. This is due to the decrease'in the degree of overmoderation in the racks with the CEAs in place.
The results of the analysis are summarized in Tables 1 and 2. Table 1 shows -l the' base case k, values in_the Region-2 rack with CEAs inserted, and j
'lable 2 shows the biases and uncertainties for one of the cases run, 4.00 w/o i U-235 fuel at an exposure of 15,000 MWD /MTU and the adjusted.k, values in ;
, the' rack after biases and uncertainties are included. 1 The worst case rack k,value including all biases and uncertainties munt be maintained below 0.95. To allow for uncertainty in determining the actual exposure and interpolating between different enrichments, the exposure required ;
for a k,of 0.9450 will be' determined. For an initial enrichment of 3.25 I w/o this exposure is 4,900 MWD /MTU. For an initial enrichment of 4.00 w/o this exposure is 12,800 MWD /MTU. These results are shown graphically in in the proposed revision to Figure 2-10 of the Technical Specifications, e
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l .CEA Integrity ,
, . 45 CEAs are currently discharged in the Spent Fuel Pool due to the CEA fingers
- '" approachingLthe 12 mean unrecoverable circumferential' cladding strain design criteria. At the End of Cycle 9 (EOC 9) operation all full length assemblies g .were inspected by Combustion Engineering and a report submitted to the NRC as detailed in Reference 7. The NRC questions and the OPPD responses are ,
l . -contained in References 8 and 9. respectively. Reference 10 contains the NRC l: SER on.thelCEA inspection report. The inspection noted that all of the CEA fingersfin the 45 CEAs had maintained structural integrity through EOC 9. Ten
( Pb of 45 CEAs'were discharged at this time. Because the remaining 35 CEAs, which operate through EOC 10 before discharge, were verified to have maintained their l integrity at EOC 9 and calculated not to have exceeded the 12 mean l i '"
unrecoverable cladding strain limit at.EOC.10, it is reasonable to assume that these 135 discharged CEAs have not had a breach of cladding integrity. .Thus it is concluded that a31 45 of the discharged full length CEAs currently residing Lin the Spent Fuel Pool are considered to have maintained their integrity.
l l The CEAs,have an estimated 85% of the. absorber worth remaining;which will be p sufficient in the absence of a neutron flux to maintain the 0.95 k,
- l. requirements for-the design life of the plant.
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g Base Case and Adjusted km Values-in the ReRion 2 Rack with CEAs Inserted ~
l Initial Fuelf Enrichment Exposure . Base Case Adjusted
- w/o HWD/MTU Computed km km -
-3.25 0. 0.9599 0.9865 3.25. 5000, 0.9175 0.9440 3.25 10000. 0.8801- 0.9092 4.00 10000. 0.9354 0.9629 4.00- 15000. 0.9023 0.9304 4.00 20000. 0.8704 0.8991 j--
- Adjusted for biases and uncertainties v
TABLE 2 Summary of Reactivity Biases and Uncertainties ;
for Fort Calhoun Region 2 StoraRe Rack Reactivity Description- Effect, k km Basic Cell, 4.0 w/o U-235 6 15,000 MWD /MTU 0.9023 Calculational Biases s
LEOPARD /PDQ/BLACKCYL model +0.0053 Most Reactive Temperature +0.0120 Mesh Spacing Effect -0.005 TOTAL BIAS +0.0168 Basic Cell Including Biases 0.9191 Tolerances and Uncertainties
. LEOPARD /PDQ/BLACKCYL model (95/95) 0.0086
-Depleted Fuel Reactivity Uncertainties 0.0056 Minimum Assembly Pitch 0.0035 Stainless Steel Thickness 0.0026 Fuel Pellet Density (10.005 in) 0.0017 Fuel Pellet Diameter (10.0005 in) 0.0003 h TOTAL' STATISTICAL UNCERTAINTY 0.0113 NWximum value of km including all biases and uncertainties for basic cell with 4.0 w/o fuel at 15,000 MWD /MTU 0.9304
P N r La- . , t y ~ ,
!/'*' J2 .3 REFERENCES s'
- l. . V.E. Grob P.W. Davison, et al., ' Multi-Region Reactor Lattice Studies -
~Results of Critical Experiments in Loose Lattices of UO2 Rods in-H 2 0', WCAP-1412, Westinghouse Atomic Power Division, dated 1960 ;
t 2.c Battelle Pacific Northwest Laboratories, ' Critical Separation Between Subcritical Clusters of 2.35 Vt! U-235 Enriched UO2Rods in Water with Fixed Neutron Poisons", PNL-2438' o
c 3.: W.L. Orr et al., "Saxton Plutonium Program, Nuclear Design of the Saxton B Partial Plutonium Core', WCAP-3385-51, dated 12/65 2
- 4. R.D.'Leamer et al., 'Pu0 2 -UO2 Fueled Critical Experiments', y WCAP-3726-1, dated 7/67 ;
- 5. S.R.'Beirman, B.M. Durst, and E.D. Clayton, ' Critical Separation Between Subcritical Clusters of 4.29 w/o U-235 Enriched UO 2 Rods in. Water with .;
Fixed Neutron Poisons',-Battelle Pacific Northwest Laboratories, NUREG/CR-0073, dated 5/78 -
- 6. G.S. Hoovler et al., ' Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel *, Nuclear Technology, V51, pp l
- 217-237,. dated 12/80 P r
.7.- Letter LIC-86-044. R.L. Andrews (OPPD) to A. Thadani (NRC), 'CEA g Inspection Results Report', dated February 28, 1986 1
1,
Results', dated July 31, 1986
- 9. Letter, R.L. Andrews (OPPD) to D.E. Sells (NRC), 'CEA Inspection Results*, dated October 3, 1986 ' l r
4
- 10. Letter W.A. Paulson (NRC) to R.L. Andrews (OPPD), 'Special Report on End of Cycle 9 Control Element Assembly (CEA) Inspection Results and Impact on Cycle 10 Operation', dated March 3, 1987 u.
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The. inclusion of a-discharged full length CEA in an assembly is acceptable for- :
' storage in Region 2 of the spent fuel pool since the 0.95 k, criticality .
a : criteria is met'and assembly meets the burnup requirements of~ revised Technical !
Specification Figure 2-10 as a function of the initial enrichment.
L w Administrative = controls, consioting of- footnoting the requirements that the CEA -
remain in'the fuel assembly while resident in Region 2, will assure compliance ,
with the Technical Specification requirements. A physical restraint system as shown in Figures l'through 6 will supplement the administrative controls. '
-Table 3 is lua installation procedure for the connection. clip to be utilized as i i a physical restraint. Figures 3 and 4 are sketches of the installation tool
.4 while Figures 5 and 6.are sketches of the connection clip and the clip installed on the CEA/ fuel assembly upper end fitting, respectively.
The' material for the' connecting clip was selected'for compatibility with the Spent Fuel Pool environment, the CEA and the fuel assembly.- The connecting Lclip will also be designed to support the loading of a fuel assembly and CEA assembly to facilitate any necessary movement of the pair.
Technical Data Book Figure I.B.1-3 is-provided as Figure 7 to indicate p.stential candidates for storage in Region 2.
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- '. TABLE 3 CONTROL ELEMENT ASSEMBLY IQ FUEL ASSEMBLY CONNECTION CLIP ,
INSTALLATION' PROCEDURE The Control Element Assemblies (CEA) will be attached to the :i
_ fuel assemblies through use of a connection clip. The clip and ,
tool to attach the clip are shown in Figures 1, 2, 3 and 4.' The ,
procedure'to_ accomplish this is outlined as follows:
-1. Attach _ top of. clip (Figure 1) to bottom of tool by inserting tool bottom locking tab (Figure 2) into slot in' top of clip (Figure 1) and rotating tool 90 0 to slot counterclockwise. This will t attach clip.to tool so that it can be lowered into pool.
- 2. Lower tool with clip into pool over CEA as shown. '
in Figure 4.
- 3. Press down on tool to engage rotator bar into top of slot of clip which will at the same time lower clip attachment feature below engagement bar.on fuel assembly top nozzle.
0
- 4. With' tool in engaged position rotate tool 90
' clockwise, then release downward load on tool.
'The clip.will now be engaged to fuel assembly top-nozzle.
- 5. Rotate tool another 90 0 clockwise to align clip locking tab on bottom of tool with slot in top of clip.
- 6. Lift tool out of pool for next CEA to fuel assembly clip connections.
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FIGURE 3 TOOL FOR CONNECTION Engagement position CLIP INSERTION Insertion position on fuel assembly into pool o
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- NO SIGNIFICANT HAZARDS LThis proposed amendment to the Technical Specifications will allow spent fuel assemblies with limited exposure to be stored in Region 2 of the spent fuel.
_ pool, if a full' length Control Element Assembly-(CEA) is inserted in the fuel assembly prior to the move to Region 2. 1
- The Technical . Specification -document changes are defined by the updated Figure 2-10 of Section 2.8.
Four postulated accident situations are of poten .al concern from a criticality standpoint. These are a boron dilution accident, 'a dropped fuel _ assembly
. . lodging sideways on top of the rack, a loss of pool water, and failure _to insert or inadve% dly remove a CEA from a low burnup spent fuel assembly.
The postulated borw dilution accident is analyzed as the base case since the l base case includes no soluble boron in the pool. j i .l Since the' rack' design ensures enough separation between the top of the fuel and i the top of the rack _(about 27") to effectively decouple the dropped assembly from the rack, a postulated dropped -assembly will have no' effect on reactivity.
Also, the effect of the dropped assembly is to reduce neutron leakage,- and no credit for axial leakage was taken in the base case. ;
A postulated loss of pool' water reduces the effective water density of the pool if the loss of water is due to the pool boiling, and can lead to an increase in reactivity if the pool. is initially overmoderated, as it is in Region 2. I However, since this is an accident, only the consideration of each single . !
failure--in isolation is required. This means that credit may be taken-for the ;
minimum soluble boron concentration of 1700 ppm when analyzing this accident. !
Note that- 1700 ppm is less than the proposed Cycle 13 value of 1900 ppm and i therefore is conservative. This level of boron more than compensates for the increase in k- due to a reduction in effective water density in the original i spent fuel rack storage analysis. The insertion of a CEA displaces water l reducing the initial level of overmoderation; thus,. the previous analysis ;
conservatively bounds this case.
l A low burnup spent fuel assembly requiring an inserted CEA for transport and storage -into Region 2 will be administratively controlled to preclude a criticality accident. This will be effected by attaching a clip to tie the CEA j.
and fuel assembly together in Region 1 prior to transfer to Region 2. The clip '
cannot be removed by the grapple on the fuel handling machine. The clip will :
prevent the inadvertent removal of the CEA from the fuel assembly.
BASIS FOR NO SIGNIFICANT HA7ARDS CONSIDERATION This proposed amendment does not involve a significant hazards consideration because the operation of Fort Calhoun Station in accordance with this amendment would not: j
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated. This change allows the storage of fuel not meeting the burnup of criteria of the current Technical Specification to be stored in Region 2 of the spent fuel pool provided a CEA is inserted and the assembly burnup meets the acceptance criteria of the proposed revised t
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l Technical, Specification Figure 2-10.. The' Region 2 racks _have been analyzed U
to ensure the minimum 0.95 K margin is maintained. Therefore, this c , change does,not increase the probability.or consequences of a previously i evaluated accident.
b -2) Create the possibility of a new or different kind of accident from any-o, accident previously evaluated. It has'been determined'that a new or
~different kind of. accident is not created because no new or.different modes ;
L of operation are proposed for the plant. The continued use of the same
- . Technical Specification controls prevents the possibility of a new or
- different kind of accident.
t 4 0 3) Involve a significant reduction in a margin' of safety. Specifications s ~ ' involving the storage of. spent fuel in Region 2 of the storage pool conform
'to ' current plant conditions'and, therefore, preserve the margin of safety.
Storage procedures contain strict administrative requirements and burnup requirements which are independently verified as acceptable and, therefore,-
will not reduce the margin of safety.
- l Based on the above considerations, OPPD does not believe that this amendment involves a significant hazards consideration.
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, ;v es ALLOW DIRECT TRANSFER OF SPENT-FUEL FROM THE REACTOR CORE TO THE
, SPENT: FUEL P0OL REGION 2 y ,
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DISCUSSION OF CHANGE
-The proposed amendment to the Technical Specifications will specifically allow W spent fuel transfer directly from the reactor core to the spent fuel pool Region-2.
Prior to the transfer of a fuel assembly during refueling operations, the Special Procedure SP-BVRNVP-l' must be completed to insure an independent review of the assembly _ burnup to verify all variables are within the acceptable ranges
! identified in Technical Specification Figure 2-10.
The direct transfer of fuel from the reactor core to the spent fuel pool Region
~
2 requires adequate procedures and independent checks to ensure that the spent fuel meets the acceptance criteria defined in Technical Specification Figure 2-10.& The existing procedures are identified and demonstrate that adequate control is provided.
The analysis required to directly transfer fuel from the reactor core to the spent fuel pool Region 2 requires a demonstration that adequate procedures and administrative controls are in place to insure that the spent fuel meets the acceptance criteria defined in Technical Specification Figure 2-10 prior to the transfer.
The Fort Calhoun Station Unit No; 1 Operating Procedure OP-11, " Reactor Core Refueling Procedure" defines the steps required to provide a safe and organized method of refueling the reactor. This procedure covers the fuel movement sequence (Appendix A), and post core loading verification checks. The Low Power Physics Testing (SP-PRCPT-1) procedure is an additional check that ensures the core is loaded as planned and analyzed.
The tsansfer of fuel from the reactor vessel to the spent fuel pool is covered by the same procedure (0P-11) along with the fuel. movement sequence. In this
- case the fuel burnup determination,- performed using Special Procedure SP-BURNVP-1, must be completed prior to fuel movement into Region 2 of the spent- fuel pool. This procedure is used-to determine and verify the burnup of each fuel assembly.
The administrative procedures covering the movement of fuel are considered to adequately ensure that the fuel will meet the acceptance criteria for burnup prior to movement to Region 2 of the spent fuel pool. Furthermore, the potential risk of dropping or mislocating a fuel assembly with a two-step operation is minimized.
JUSTIFICATION The direct transfer of spent fuel from the reactor vessel to the spent fuel pool Region 2 allows for safe storage on the basis that all administrative procedures are followed. Furthermore, the potential risk of dropping or mislocating a fuel assembly with the additional handling required in a two-step operation is minimized.
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- c. NO SIGNIFICANT HAZARDS 1
< The proposed amendment to the Technical- Specifications will specifically allow L ~
. spent = fuel. transfer directly from the reactor core to the spent fuel pool Region 2. ,
- The Technica1' Specification changes are defined on page 2-38 of Section 2.8.
c Prior to the transfer of a fuel assembly during refueling operations, the .
L Special Procedure SP-BURNVP-1 must be complete to insure an independent review of the assembly burnup and'to verify the burnup is within the acceptable ranges
.identif ei d in' Technical Specification Figure 2-10.
L ' BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION i This proposed amendment does not involve a significant hazards consideration ,
because the operation of Fort Calhoun Station in accordance with this amendment '
would not:
- 1) Involve a significant increase in the probability or consequences of an accident.previously evaluated. This change allows the direct transfer. of fuel from the reactor core to Region 2 of the spent fuel pool provided all.
procedures have been completed and the assembly burnup meets the acceptance criteria with no changes in administrative specifications. The final storage location of any given fuel element will be the same using a one-step or a two-step transfer process. Therefore,-this change does not-a significantly increase the probability or consequences of a previously-evaluated accident.
Create the possibility of a new or different kind of accident from any 2) accident previously evah ated. It has been determined that a new or
- different kind of &ccident is not created because no new or different modes of operation ara proposed for the plant. The continued use of the existing Technical Specification administrative controls prevents the possibility of a new or different kind of accident.
. 3) Involve a significant reduction in a margin of safety. Administrative specifications involving the transfer of spent fuel to the storage pool conform to current plant conditions and, therefore, preserve the margin of safety. Changes in the fuel transfer procedures contain strict administrative procedures-and ensure that the burnup requirements are verified-as acceptable and, therefore, will not reduce the margin of safety.
l Based on the above considerations, OPPD does not believe that this amendment l ,
involves a significant hazards consideration.
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' DELETE REQUIREMENT FOR FUEL. 4 PERFORMANCE REPORT AT END :3 0F EACH CYCLE.
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! DISCUSSION OF CHANGE-LThis request is an administrative change to the' Fort Calhoun Station Technical Specifications previously requested by 0 PPD and approved in- Amondment No.- 77 to facility Operating License No. DPR-40.
The requested Technical Specification revision is administrative because of the previously approved request documented in Amendment No. 77 (Reference 1). The
. additional: reporting was performed as part of high fuel burnup demonstration project that was completed, hence the comprehensive fuel inspections and reporting- are no longer required.
The analysis discussions are contained in the reference documents identified
.herein. Copies of the appropriate pages are furnished for your convenience.
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' REFERENCES
- r , 1. U. S.cNRC e1'tter from.E. G. Tourigny to W..C. Jones-(OPPD),-dated 4/26/84; issuing Amendment No.677
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. NOTE:L Copies of selected pages from the above reference are attached.
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4/27/84' AWEt% GPB,' lCJ, J10G, K7M,. RIJ,;WGG, TW ~y1 D0ll, = PMS, FAT, ITF, JJF - (3) ,' D7M, TRR, -
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js UNITED STATES .
9,,p[T j fu
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,1 NUCLEAR REGULATORY COMMISSION'
' WASHINGTON, D. C. 20566
~ ,u ;o Y ,
,.6,,,. c APR 2 6<1984
~
y ^ Docket Noi 50-285i '[
m
-Mr. W. C. Jonesc .
l
Division Manager, Production-4*. Operations, i
Omaha Public Power District
- 1623 Harney. Street j w Omaha,
- Nebraska 68102 y7
Dear Mr. Jones:
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!The Commission has issued the enclosed' Amendment No. 77 to Facility.0perating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1. This amendment, l
!b ' iconsists of changes to the Technical. Specifications in response to your applica- 7 l
l tion < dated February.14,'1984, as supplemented by letters dated March 28,
- 4. ! April,4_and 16,.1984'.' .
The amendment authorbes-changes.to the Fort Calhoun Station, Unit No. 1.t
,m ,
Technical Specifications which are required to support the operation ~of f the unit at full rated power during Cycle 9. Specifically, the followina' ~
, 4 specifications are changed: minimum departure from nucleate boiling (DNB) '
- ratio, total unrodded
- planar l radial peak, unrodded integrated total radial l peak, lower bound'of the moderator. temperature coefficient and linear heat' ,
rate: measurement-calculation uncertainty factor. Because these specifica- :
tions, are changed, the followino figures;are changed: Thermal Margin / Low I
. LPressure Safety Limit Thermal Margin / Low Pressure Limiting Safety System LSettings; Limiting Condition for Operation for Excore Monitoring and. inear 3
' Heft Rate,. Limiting Condition for Operation for 'DNB Honitoring, and F .and 3 F' and Core Power Limitations.. A steam generator differential press re >
'rNetorprotectivesystemtripisaddedtothetechnicalspecifications.
~
~ Instrument. Operating Requirements and Minimum _ Frequencies for Checks, s Calibration, and Testing are added for this.new trip. i a
The' amendment also revises the Technical Specifications for the reactor coolant: system pressure-temperature limits for operation to 8.5 effective full- power years. 'The current ~ technical specifications permit operations to 7.0' effective full- power years, which will occur during Cycle 9 opera-tion. This necessitates a change in the following figure: Heatup-Reactor Not i
Critical, Cooldown-Reactor Not Critical, and Predicted Radiation Induced '
NDTT Shift. Last, the amendment deletes the spent fuel inspection re-
, 'quirements.
FC
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k TA copy'of'the Safety Evaluation 1s also enclosed. The' Notice of Issu.snee.w'ill f : belincluded in the Commission's next regular monthly Federal- Register notice.
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$p:n w, O G..fTourig.
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% <f, P oject Manager -
a s W . Operating.Rebetor' Branch #3 ,
J.. Division of Licensing na
' Enclosure':
.1. : Amendment' No. 77 to DPR 2. ' Safety Evaluation
. cc w/ enclosure:'
See next page.-
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- Omaha Public Power District
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Harry,H. Voigt, Esq.
LeBoeuf, . Lamb, Leiby 8 MacRae- l LJ >$' l 1333'New Hampshire Avenue, N.W.
Washington,' D. C. 20036
'Mr. Jack Jensen Chairman, Washington County -
Board of Supervisors-c ' Blair, Nebraska - 68023 U.S. Environmental Protect 1on Agency
-Region;VII ATTN: Regional Radiation '
Representative 3
,' 324 East 11th Street s
.. Kansas C1,ty, Missouri 64106'- - >!
Metropo.litan Planning Agency- '
C ATTN::.Dagnia Prieditis'
- 7000 West Center Road Omaha, Nebraska 68107- !
Mr. - Larry .Yand'e1.1 U.S.N.R.C. Resident Inspector P. 0.-Box >309 Fort Calhoun,: Nei m 6B023 a
Mr. Charles B. Br1m '.n Manager'- Washington Nuclear-Operations : ,
C-E Power _ Systems .
Combustion Engineering, Inc.
7910 Woodmont Avenue ,
r Bethesda, Maryland- 20814 Regional Administrator Nuclear Regulatory Commission, Region IV Office of Executive Oirector for Operations -
-611 Ryan Plaza Drive Suite 1000 '
Arlington, Texas 76011 r
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D E The ASGTPTF can be tested with the reactor in operation or shut down.
The prcposed Technical Specification changes relating to the ASGTPTF modification include provision for monthly testing of the SG pressure instrument channels (i.e., a bistable trip test), a channel check to -
be performed each shift, and calibration of the SG pressure sensors during each refueling outage. Similar Technical Specification sur-veillance reouirements currently existing for the SG low pressure RPS instrumentation will remain unchanged.
Based on our review of the licensee's submittal, we conclude that the electrical, instrumentation and cowTrel aspects of the proposed ASGTPTT modifications to the Reactor Protection System and the associated pro-posed Technical Specifications comply with the applicable criteria of Section 7.2 " Reactor Trip System" of the Standard Review Plan (NUREG-0800), ,
and therefore, are acceptable. -
f 8.0 DELETION OF SPENT FUEL INSPECTION REQUIREMENTS i '
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OPPD has requested that Section 5.9.3.h of the Fort Calhoun Technical Specifi-
cations be eliminated. This specification requires a report on fuel performance following each refueling outage. A comprehensive fuci assembly demonstratior, project has recently been completed by OPPD which i L resulted 'in the conclusion that no rod perforations nor anomalies existed in
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i a lead denonstntion assembly after six cycles (approximately 56 GWD/MTU) of /
, irradiation.
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/ Based on this and on the discusatan presented in Reference 16, we conclude '
- l. that the elimination of this requirement involves no significant hazards
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s considerations in accordance with the criteria contained in 10CFR50.92.
- Reference 16 points out that we currently do not require reporting for routine surveillance of standard fuel designs. However, it refers to the N .
f' Licensee Event Report System Rule (10CFRSO.73) which became effective this
(
) year for guidance as to when reports should be issued. The supporting \-
document for this rule states that reportable situations include " fuel ^
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cladding failures in the reactor or in the storage pool that exceed expected values, that are unique or widespread, or that resulted from unexpected
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9.0 CONCLUSION
S . PHYSICS, THERMAL-HYDRAULICS, TRANSIENTS AND FUELS The staff has reviewed the infomation presented in the Fort Calhoun Cycle 9 reload report and in OPPD responses to our requests for additional infomation.
We find the proposed reload and the associated modified Technical Specifications ,
acceptable.
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, t wr - 1 'g' h requested Technical Specification revisions should be provided as .
- previously requested and., approved.
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e- i NO SIGNIFICANT' HAZARDS ;
i r The proposed Technical Specification change deletes the requirement to provide ;
a fuel performance report after each cycle of operation. i
! The Technical Specification document change is defined on page 5-15 of Section -
[ 5.9. ;
The proposed change for deleting Technical Specification requirement 5.9.3.h i' has been previously approved in the Safety Evaluation Report for Amendment No.
L 77 to the facility operating license. Thus, this change is considered administrative and does not affect the safe operation of the plant. The :
requirement to provide the fuel performance report is duplicate to the Licensee Event Report System Rule (10CFR50.73) to report unique, widespread or :
a unexpected fuel cladding failures in the reactor or spent fuel pool, j I -
BASIS FOR NO SIGNIFICANT HA7ARDS DETERMINATION ;
This proposed amendment does not involve a significant hazards consideration i because the operation of Fort Calhoun Station in accordance with this amendment !
would not: t
- 1) Involve an increase in the probability or consequence of an accident previously evaluated. This change implements a previously approved -
revision to the Technical Specifications. Therefore, this change does not ;
increase the probability or consequences of a previously evaluated :
accident. This change is considered administrative. "j
- 2) Create the possibility of a new or different kind of accident from any :
accident previously evaluated. it has been determined that a new or different kind of accident is-not created because no new or different modes !
of operation-are proposed for the plant. The continued use of the current Technical Specification administrative controls prevents the possibility of anew or different kind of accident. !
7 3) Involve a significant reduction in a margin of safety. Administrative specifications involving the deletion of the fuel performance report will not reduce the margin of safety.
1 L Based on the above considerations, OPPD does not believe that this amendment involves a significant hazards consideration.
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