ML20012D443

Proposed Tech Specs Re Idle Recirculation Loops

ML20012D443
Person / Time
Site:	Oyster Creek
Issue date:	03/19/1990
From:	GENERAL PUBLIC UTILITIES CORP.
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ML20012D440	List: ML20012D439 ML20012D441 ML20012D443
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NUDOCS 9003270327
Download: ML20012D443 (11)

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bb%# . OYSTER CREEK NUCLEAR GENERATING STATION PROVISIONAL. OPERATING LICENSE NO. DPR-16 i DOCKET NO. 50-219 TECHNICAL' SPECIFICATION CHANGE. REQUEST NO. 186 h- (

A' Applicant hereby. requests the Commission to change Appendix A to the above -'

captioned' license as below, and pursuant to 10 CFR 50.91, an analysis concerning the determination of no significant hazards considerations is also e ~ presented: '

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1. Section'to be Chanced [

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. Section 3.3 3

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2) Extent of Chance Revise Technical Specification 3.3.F.2 to include limitations on operation' 1 i with an idle recirculation loop which is isolated. A revision to Section 3.3 and 3.10Lbases is,also needed.

N 3. Chances Raouested 2 o

s The requested ~ changes are shown on attached Technical Specification pages-

[s 3.3-3 and 3.3-3a. .In addition, the Section 3.3 bases will be revised on page 3.3-8 as well as the Section 3.10 bases on_page 3.10-4. . Bases pages .[

.3.3-Ba, 3.10-5 and 3.10-6 are changed due to pagination. Bases page "

3.10-6a is added due to pagination.

4. Discussion Technical Specification :3.3.F.2 presently requires an idle recirculation

' loop to.be unisolated during power, operation. The purpose of this LTechnical Specification-Change-Request is to permit reactor operation with one recirculation' loop fully isolated 1.e., suction, discharge and discharge bypass valve closed. This change can be utilized to control-leakage.from pump seals by isolating the pump, when necessary.

l l The Reactor Recirculation System has been designed to perform the following i functions:

f pagy 4i4': A.- To provide forced circulation of reactor water through the core to

g; overcome the power density limitation of the fuel..

• vA C n .mNC . B. To provide a variable moderator (coolant) flow through the core to H , 50a Ob control reactor power without manipulation of the control rods.

, ' Olh So The Reactor Recirculation System has been sized to provide a total flow 'l gy capacity equal to the required flow at rated load plus a. design margin of 1 l

g 10 percent.

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$i" A recirculation loop-will be fully isolated by closing its suction, I j

Q discharge and discharge bypass valves. Once a loop is out of service and j L ,

oc fully isolated, the remaining four recirculation pumps will provide higher j I *1" flow by increasing their speed. Normal power operation can continue with 1

-four loops.

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j"-ni e . Current Technical Specification 3.3.F.2 permits reactor operation with one

? idle recirculation loop provided the loop is not isolated from the reactor vessel. iThis restriction minimizes the occurrence of a severe cold water addition transient during pump startup in an idle recirculation loop.

, Further, the discharge bypass and suction valves in an idle loop are

• maintained open in order to have the coolant inventory in the' loop available during a LOCA blowdown.

The latest Oyster Creek LOCA analyses (Reference 4 in Technical specification section 3.10) includes an analysis of the design basis LOCA Jb with one of the recirculation loops fully isolated from the reactor i vesse1 The results from this analysis conclude that 10CFR50.46 criteria for aak cladding temperature and maximum cladding oxidation are met, ,

provided that an appropriate MAPLHGR multiplier is applied to the initial reactor power conditions. These multipliers are given belows t MAPLHGR Fuel Tvoe Exposure Ranoe -Multiolier a \

-i h P8X8R E S 15.0 GWd/MTU 0.99 ,

P8X8R E > 15.0 GWd/MTU 0.98 GE8X8EB All Exposures 0.98 In addition to the GE fuel assemblies, there are 29 Exxon fuel assemblies which were loaded in Cycle 10 or earlier that are now located on the core periphery. The Exxon four loop analysis does not address an isolated recirculation loop. However, the impact of isolating a recirculation loop on the MAPLHGR limits is small as seen from the GE analysis. The core periphery is a Jow power region where fuel assemblies operate at 50% or less of the MAPLHGR limits. Therefore, allowing four loop operation with a fully isolated loop for Exxon fuel assemblies will not pose a concern for exceeding 10CFR50.46 limits.

When a recirculation loop is fully isolated at power, the isolated portion between the suction and discharge valves will cool to near ambient temperature. Before the pump in the fully isolated recirculation loop can be restarted, the loop temperature must be warmed to within 50'F of the bulk coolant temperature in order to avold the injection of cold water into the reactor core to prevent.a transient and to avoid thermal stresses to the reactor vessel nozzles and CRD housings. This requirement cannot be satisfied with the current system: configuration. Therefore, a fully isolated loop will not be restarted once it is isolated unless the reactor is in the cold shutdown condition. The suction valve, discharge valve and discharge bypass valve in the fully isolated loop shall be in the closed position and the associated motor breakers shall be opened and defeated to prevent cold water injection into the vessel.

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. With a: loop fully isoluted, the temperature of the pump and the piping f' .

E between the isolation valves.will drop from the normal' 548'F to ambient. A i'; temperature gradient in the fully isolated loop will exist between the.

isolation valves and the reactor vessel nozzle. An-analysis based on a conservativt temperature profile along the piping indicates-that no adverse effect on the piping will occur as a result of this condition.

Recirculation loops A and E are connected to isolation condenser piping.

L When the isolation condenser is initiated, the performance of the isolation condenser will not be affected. The condensate return from the isolation.

condenser will flow to the downcomer region of the reactor and will be

. carried over to the reactor via either the remaining-running recirculation pumps or other open recirculation loop (s).

Recirculation loop A also has a reactor water sample line connected to it  !

on the auction side upstream of the isolation valve. If the auction valve is closed, a representative sample may not be obtained due to no flow circulation in the piping. Upon this condition, reactor water sample can I be alternately obtained from the cleanup system. l The reactor water cleanup system is connected to recirculation loop B. If loop B is fully isolated, the. cleanup system can function normally since the fully isolated loop will prevent any backflow and short circuiting of .l cleanup flow. I J

The shutdown cooling (SDC) system is connected to recirculation loop E.

-There is no effect on cooldown by SDC since normal practice is to close the-E loop discharge valve when the pump is idle to prevent short circuiting SDC flow in the loop.

5. Determination GPU Nuclear has determined that operation of the Oyster Creek Nuclear Genorating Station in accordance with the proposed technical specifications- ~) '

does'not' involve a significant hazard. . The proposed change would not:

i 1.EInvolve a significant increase in the probability or the consequences of I any accident previously evaluated. The Oyster Creek LOCA analysis concludes that there is no increase in peak cladding temperature with one recirculation loop isolated provided that a KAPLHGR multiplier of

'O.98 is used.

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2. Create the possibility of a new or different kind of accident from any .i accident previously evaluated. No physical changes are being made to e, the facility. Operation in the proposed manner precludes the occurrence of a severe cold water addition transient which is an analyzed transient of moderate frequency.

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3. Involve a significant reduction in a margin of safety. The LOCA analysis shows that changes in the calculated values of applicable parameters are insignificant. Appendix K limits are met.

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E. Reactor Coolant Ouality l i

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q, , 1. -The reactor coolant quality during power operation with steaming  ;

g, rates to the turbine-condenser of less.than 100,000 pounds per hour shall be limited-to conductivity 2 us/cm [S=mhos at 25'C(77'F)]  ;

chloride ion 0.1 ppm '

nc 2. When the conductivity and chloride concentration limits given in 3.3.E.1 are exceeded, an orderly shutdown shall be initiated immediately, and the reactor coolant temperature shall be reduced ik to--less than 212'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

n i; 3. 'The reactor coolant quality during' power operation with steaming.

"m rates to the turbine-condenser of greater than or equal to 100,000 3, pounds per hour shall be limited tos O conductivity 10 uS/cm [S=mhos at 25'C(77'F))

chloride ion 0.5 ppm

4. When the maximum conductivity or chloride concentration limits given in 3.3.E.3-are exceeded, an orderly shutdown shall be initiated immediately,.and the reactor coolant temperature shall be reduced to less than 212'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />..

5. During power operation with steaming rates on the turbine-condenser of greater than or equal to 100,000 pounds per hour, the time limit above 1.0 us/cm at 25'C (77'F) and 0.2 ppm chloride shall not exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for any single incident.

/ 6 .' When the. time limits for 3.3.E.5 are exceeded, an orderly shutdown l shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

F. Recirculation Looo Ooerability L

! 1. During POWER OPERATION, all five recirculation loops shall be OPERATING except as specified in Specification 3.3.F.2.

2. POWER OPERATION with one idle recirculation loop or one fully isolated loop per F.2.c is permitted. When the idle loop is isolated the following conditions shall'be mett l

H a. The average planar linear heat generation rate (APLHGR) of all 3 fuel rods in any fuel assembly, as a function of average planar exposure, at any axial location shall not exceed 98% of the b, '

limits given.in the specifications for APLHGR in Section 3.10.A. The action to bring the' core to 98% of the APLHGR .

i limits shall be completed prior to isolating the recirculation y loop.

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b. The associated recirculation pump motor generator set circuit breaker shall be opened and defeated from operation.

OYSTER CREEK 3.3-3 Amendment No: 42, 93, 135 f" TSCR186 1

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{ 4f n! , c. The suction valve, discharge valve and discharge bypass valve

, p in the isolated-loop shall be in the closed position and 4

associated motor breakers shall be opened and defeated from- i operation.

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$ d. The fully isolated loop as in 3.3.F.2.C above shall not be :s returned to service unless the reactor is in the COLD SHUTDOWN condition.

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3. If. Specifications 3.3.F.1 and 3.3.F.2 are not met, an orderly  ;

i shutdown shall be initiated immediately until all operable. control

, ,- rods are fully inserted and the reactor is in either the REFUEL r MODE or SHUTDOWN CONDITION within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. i

4. With reactor coolant temperature greater then 212'F and irradiated fuel in the reactor vessel, at least one recirculation loop.

discharge valve and its associated suction valve shall be.in the ~ 'i full open position.

5. If Specification 3.3.F.4 is not met, immediately open one i recirculation loop discharge valve and its associated suction valve.

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6. With reactor coolant temperature less than 212'F and irradiated t i- ' fuel in the reactor vessel,-at least one recirculation loop discharge valve and its associated suction valve shall be in the i full open position unless the reactor vessel is flooded to a level-above 185 inches TAF or unless the steam separator and. dryer are removed.

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OYSTER CREEK 3.3-3a Amendment No. 135

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, , 'pH, chloride, and other_ chemical parameters are made~to determine-the cause of the unusual conductivity and instigate proper corrective action. These can be done before. limiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceeded. Several techniques are available to correct off-standard reactor water l quality conditions including removal of impurities from reactor water -t by the cleanup system, reducing-input of impurities causing  ?

off-standard conditions by reducing power and reducing the .}

reactor coolant. temperature.to less than:212'F. The major benefit of reducing.the reactor coolant temperature'to less than 212'F is to reduce the temperature dependent corrosion i rates and thereby provide time for the cleanup system to re-estabilish proper water quality.

Specifications 3.3.F.1 and 3.3.F.2 require a minimum of four OPERATING recirculation loops during reactor POWER OPERATION.

Core parameters have not been established for POWER OPERATION l , with less than four OPERATING loops. Therefore, specification  ;

13.'3.F.3 requires reactor POWER OPERATION to be terminated and

< the reactor placed in the REFUEL HODE or SHUTDOWN CONDITION ,

within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

F During.four loop POWER OPERATION.the idle loop, when it is not .i isolated,-is required to.have its discharge valve closed and- Li its discharge bypass and suction valves-open. This provides  !

and limits reactor coolant backflow through an idle loop and thus minimizes the occurrence of a severe cold' water addition transient during startup of an idle loop. .In addition, with the discharge bypass and suction-valves in an idle loop open the coolant inventory in the loop-is available during LOCA blowdown. 3 The requirements of Specification 3.3.F.2 for partial loop operation in which the idle loop is isolated, preclude the (

inadvertent startup of a recirculation pump with a-cold leg thus avoiding any reactivity addition transient or reactor vessel nozzle thermal stress concerns.

Specifications 3.3.F.4 and 3.3.F.6 assure that an adequate flow path exists from the annular space, between the pressure vessel ,

wall'and the core shroud, to the core region. This provides sufficient hydraulic communication betwoon these areas, thus-assuring that reactor water instrument readings are indicative of the level in the core region. For the bounding loss of feedwater transient (2) , a single fully open recirculation loop transfers coolant from the annulus to the core region at approximately)five circulation ( .

times the bolloff rate with no forced With the reactor vessel flooded to a level above 185 inches TAF or when the steam separator and dryer are removed, the core region and all recirculation loops can therefore be isolated. When the steam separator and dryer are removed, safety limit 2.1.D ensures water level is maintained above the core shroud.

OYSTER CREEK 3.3-8 Amendment No. 42, 93, 135 TSCR186 x____-

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Ref erences:- (1)- FDSAR, Volume I, Section IV-2' ;l (2) Letter to NRC. dated May 19, 1979, " Transient of May 2,'1979"

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(3). General Electric Co.' Letter G-EN-9-55, " Revised Natural

Circulation Flow Calculation",' dated May 29,-1979 ..

(4)' Licensing Application. Amendment 16, Design o -Requirements Section (5) (Deleted)

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. (6) FDSAR,. Volume I,'Section IV-2.3.3 and Volume'II, Appendix H i (7)_ FDSAR, Volume I,' Table IV-2-1 (8) Licensing Application Amendment 34, Question 14 (9)J Licensing Application Amendment 28, Item III-B-2 (10) Licensing Application Amendment-32,' Question 15 (11) (Deleted) ,

(12) (Deleted)

E (13) Licensing Application Amendment 16, Page 1 (14).GPUN.TDR 725'Rev. Os Testing and Evaluation of

. Irradiated Reactor Vessel Materials Surveillance-Program Specimens i

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OYSTER CREEK 3.3-8a Amendment No. 135 i

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.The' maximum average-planar LHGR limits of_ fuel types V and VB are shown in-Figuro 3.10-1 for five: loop operation and in Figure 3.10 for four loop operation, and are the result of LOCA analyses performed

'by Exxon Nuclear Company utilizing an evaluation model developed by Exxon Nuclear Company in compliance with Appendix K to 10 CFR 50.(1).

Operation is permitted with the four-loop limits of Figure 3.10 provided the fifth-loop has its_ discharge valve closed and its-bypass

.and suction valves open. Four loop operation is permitted with the idle loop isolated (suction, discharge and discharge bypass valves closed),with Exxon fuel assemblies since the Exxon assemblies are located only on the core periphery and operate at significantly lower MAPLHGR values than the rest of the core. The MAPLHGR multiplier in Figure 3.10-3 is further reduced for an isolated idle loop consistent-with the multiplier for GE fuel. Additional requirements for isolated

-idle loop operation are given in Specification 3.3.F.2. In addition, the maximum average' planar LHGR. limits shown in Figures 3.10-1 and 3.10-2 for Type V and VB' fuel were analyzed with 100% of the spray cooling coefficients specified in Appendix K to 10 CFR Part 50 for 7 x 7~ fuel. These spray heat transfer coefficients;were justified in the ENO Spray Cooling Heat Transfer Test Program (2).

The maximum average planar LHGR limits of fuel types P8x8R and GE8x8EB are shown in Figure 3.10-4 and Figure 3.10-5, for both 5-loop and 4-loop operation when the idle loop is not isolated, and are based on calculations employing the models described in Reference 4. .Four loop operation is permitted with the idle loop isolated (suction, discharge and' discharge bypass valves closed) provided that a MAPLHGR multiplier _

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of 0.98 as shown in Reference 4, is applied to figures 3.10-4 and 3.10-5. Additional requirements for isolated idle loop operation are given in Specification 3.3.F.2. Power operation with LHGR's at or below those shown in Figures 3.10-4 and 3.10-5 assures.that the peak cladding temperature following a postulated loss-of-coolant accident will not exceed the 2200*F limit.

The effect of axial power profile peak location for fuel types V a.nd

-VB is evaluated for the worst break size by performing a series of fuel heat-up calculations. A set of multipliers is devised to reduce the allowable bottom skewed axial power-peaks relative to-center or above center peaked profiles. The major factors which lead to the lower MAPLHGR' limits with bottom skewed axial power profiles are the change in canister quench time at the axial peak location and a deterioration in heat transfer during the extended downward flow period during blowdown. The MAPLHGR multiplier in Figure 3.10-3'shall only be applied to MAPLHGR determined by the evaluation model described in reference 1.

The possible effects of fuel pellet densification are:

1) creep collapse of the cladding due to axial gap formation;

2) increase in the LHGR because of pellet column shortening;

3) power spikes due to axial gap formation; and

4) changes in stored energy due to increased radial gap size.

OYSTER CREEK 3.10-4 Amendment No.: 75, 129 TSCR186

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[ 's calculations show that clad collapse is' conservatively predicted not' to. occur during the. exposure lifetime of.the fuel, Therefore, clad' collapse is not considered in the analyses.

since axial thermal expansion of-the-fuel pellets is greater than j axial shrinkage due to densification, the analyses of peak clad 1 temperatures do not consider any change in LHGR due to pellet column shortening.- Although the formation of axial gaps might produce a

' local power spike at one location on any one rod in a fuel assembly.

the increase in local density.would be on the order of only 2% at the axial midplane. .Since small local variations in power distribution have a small effect on peak clad temperature, power spikes were not I

considered in the analysis of loss-of-coolant accidents (1).

Changes in gap size affect the peak clad temperatures by their effect on pellet clad thermal conductance and fuel pellet stored energy.

Treatment of this effect combined with the effects of pellet cracking, relocation and subsequent gap closure are discussed in XN-174.

Pellet-clad thermal conductance for Type V and VB fuel was calculated using the GAPEX model (XN-174).

The specification for local LHCR assures that the linear heat generation rate in any rod is less than the limiting linear heat generation rate even if fuel pellet densification is postulated. The power spike penalty for Type V and VB fuel is based on analyses presented in Facility Change Request No.6 and FDSAR Amendment No.76, respectively. The analysis assumes a linearly increasing variation in axial gaps between-core bottom and top, and assures with 95%

confidence that no more than one fuel rod exceeds the design linear ,

heat generation rate due to power spiking.

The power spike penalty for GE fuel is. described in Reference 3.

The loss of coolant accident (LOCA) analyses are performed using an

. initial core flow that is 70% of the rated:value. The rationale for use of this value of flow is based on the possibility of achieving full power (100% rate power) at a reduced flow condition. The

. magnitude of the reduced flow is limited by the flow relationship for' .

overpower scram. The low flow condition for the'LOCA analysis ensures a conservative analysis because this initial condition is associated with a higher initial quality in the core relative to higher flow-lower quality conditions at full power. The high quality-low p flow condition for the steady-state core operation results in rapid L voiding of the core during the blowdown period of the LOCA. The rapid

l. degradation of the coolant conditions due to voiding results in a decrease in the time to boiling transition and thus degradation of heat transfer with consequent higher peak cladding temperatures.

Thus, analysis of the LOCA using 70% flow and 102% power provides a conservative basis for evaluation of the peak cladding temperature and i the maximum average planar linear heat generation rate (MAPLHGR) for the reactor.

l OYSTER CREEK 3.10-5 Amendment No.: 75, 111, 129 TSCR186

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1 The APRM r r ponse lo u:ed to pr; dict wh;n thy rod bicek cccurs in thi 4l ., analysis of the rod withdrawal error transient.. The~ transient-rod R ' position at the rod block and-corresponding MCPR can be determined.

The MCPR-has been evaluated for different APRM-responses which would result'from changes.in.the APRM status as'a. consequence of bypassed-APRM channel'and/or, failed / bypassed LPRM. inputs. The steady state MCPR required to protect the minimum transient CPR of 1.07 for tha worst. case APRM status condition (APRM Status 1) is determined.in the rod withdrawal error transient analysis. The steady state MCPR values for APRM status conditions 1, 2, and 3 will be evaluated each cycle.

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The time interval of eight (8) hours to adjust the steady. state of MCPR to account for a degradation in the APRM status is justified on the basis of instituting a control rod' block which precludes the possibility of experiencing a rod with.trawal error transient since rod withdrawal is physically prevented. This time interval is adequate to allow the operator to either increase the MCPR to the appropriate value or to upgrade the status of the APRM system while in a condition which prevents the possibility of this transient occurring, i

-The steady-state MCPR limit was selected to provide margin to

' accommodate transients and uncertainties in monitoring the core operating state, manufacturing, and in the critical power correlation ,

itself(3). .This limit was derived by addition of the CPR for the  !

most limiting abnormal operational transient' caused by,a single operator error or equipment malfunction to the fuel cladding integrity MCPR limit designated in Specification 2.1.

Transients analyzed each fuel cycle will be evaluated with respect to g' the steady-state MCPR' limit specified in this specification.

The purpose of the Kg. factor is to define operating limits at other ,

i than. rated flow conditions. At less that 100% flow the required MCPR y is the product of the operating l utt E PR and the Kg factor.

l - Specifically, the Kg factor provides the required thermal margin to protect against a flow increase transient.

The K g factor curves shown in Figure 3.10-6 were developed-generically using the flow control line corresponding to rated thermal power at rated core flow and are applicable to all BWR/2, BWR/3 and BWR/4 reactors. For the manual flow control mode, the Kg factors were calculated such that at the maximum flow state (as limited by the pump scoop tube set point) and the corresponding core power (along the

! rated flow control line), the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPR's were calculated at different points along the rated flow control line corresponding to different L'

b core flows. The' ratio of the MCPR calculated at a given point of core flow,-divided by the operating limit MCPR determines the value of K.

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OYSTER CREEK 3.10-6 Amendment No.: 75, 111, 129 TSCR186 n

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. REFERENCES ,

(1) XN-75-55-(A)', XN-75-55, supplement 1-(A), XN-75-55. supplement' 2-(A), Revision 2, " Exxon Nuclear Company WREM-Based NJP-BWR ECCS .;

Evaluation Model and Application-to the Oyster Creek plant," April  !

1977.

-(2) XN-75-36 (NP)-(A), XN-75-36 (NP) Supplement 1-(A), " Spray Cooling.

Heat Transfer phase Test Results, ENC - 8 x 8 BWR Fuel 60 and 63 Active Rods, Interim Report," October 1975. e

(' 3) NEDE-24195; General Electric Reload Fuel Application for Oyster Creek.

(4) NEDE-31462P; " OYSTER CREEK NUCLEAR GENERATING STATION ,

SAFER /CORECOOL/GESTR-LOCA LOSS-OF-COOLANT ACCIDENT ANALYSIS,"

August 1987.

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OYSTER CREEK 3.10-6a Amendment No.: 75, 129 TSCR186-

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