ML20055C883

Amend 140 to License DPR-16,revising Tech Spec 3.3.F.2 to Include Limitations on Operation W/Idle Recirculation Loop Isolated

ML20055C883
Person / Time
Site:	Oyster Creek
Issue date:	06/18/1990
From:	Stolz J Office of Nuclear Reactor Regulation
To:	GPU Nuclear Corp, Jersey Central Power & Light Co
Shared Package
ML20055C884	List: ML20055C883 ML20055C885 ML20055C886
References
DPR-16-A-140 NUDOCS 9006250331
Download: ML20055C883 (11)

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UNITED STATES

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WASHINGTON, D. C. 20555 1

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GPU HUCLEAR CORPORATION AND i

JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION j

AMENDMENT ~TO PROVISIONAL OPERATING LICENSE Amendment No.140 License No. DPR-16

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The Nuclear Regulatory Comission (the Comission) has found that:

1 A.

The application for arandment by GPU Nuclear Corporation, et al.,,

(thelicensee),datedMarch 19, 1990 complies with the. standards and

. requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Counission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will. operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

Thereisreasonable.asturance(1)thattheactivitiesauthorized by this-amendment can be conducted without endangering the health and safety of the public, and (ii) that such~ activities will be conducted in compliance'with the Comission's regulations; D.

The issuance sf _ this amendment will not be inimical to the common defense and security or to the health and' safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

9006?50331 900618

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Accordingly, the license is amended by changes to the Technical y

F Specifications as indicated in the attachment to this license amendment, l-and paragraph 2.C.(2) of Provisional Operating License No. DPR-16 is hereby L

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amended to read.as follows:

(2) Tectnical Specifications I

The Technical Specifications contained in Appendices A and B, as revised through Amendment No.140, are hereby incorporated in the 4

license..GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.

M 3.

This license amendment is effective as of the'date of issuance, to be p

implemented within 30 days of issuance.

'n FOR THE NUCLEAR REGULATORY COMMISSION E

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F-F. Stolz, Director fr ject Directorate I-4 ~

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p' Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation L

Attachment:

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Changes to the Technical 4

Specifications

'Date of Issuance: June 18, 1990, w

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ATTACHMENT TO LICENSE AMENDMENT NO.140 i

' PROVISIONAL OPERATING LICENSE NO.-DPR-16 1

DOCKET NO. 50-219 1

Peplace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

1 Remove Insert Page 3.3-3 Page 3.3-3 Page 3.3-3a Page 3.3-3a Page 3.3 Page 3.3-8 1

Page 3.3-8a Page 3.3-8a Page 3.10-4 Page 3.10-4 Page 3.10 Page 3.10-5 Page 3.10-6 Page 3.10-6 Page 3.10-6a i

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Reactor coolant Duality i

1.

The reactor coolant quality during. power operation with steaming rates to the turbine-condenser-of less than 100,000 pounds per hour shall be limited tos l

L conductivity. 2 us/cm [s=mhos at 25'C(77'F)]

O chloride ion. 0.1 ppm l

2.

When the conductivity and chloride concenseation ILaitr given in L"

3.3.E.1 are exceeded, an orderly shutdown sh?ll be.ir.itiated 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 />.

3.

The.xeactor coolant quality during power operation with steaming rates to the turbine-condenser of greater than or equal to 100,000 pounds per hour shall be limited to conductivity 10 us/cm (8=mhos at 25'C(77'F)]

o chloride ion 0.5 ppm 4.

When the maximum conductivity or chloride concentration limite 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 l-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 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 Loon Ooerability 1.

During POWER OPERATION, all five recirculation loopa shall be OPERATING except as specified in specification 3.3.F.2.

2.

POWER OPERATION with one idle recirculation loop or ona fully l

isolated loop per F.2.c is permitted. When the idle looy is isolated the following-conditions shall be. nets

a. The average planar linear heat generation rate'(APLHOR) of all l

fuel rode in any fuel assembly, as a function of average planar l,

exposure, at any axial location shall not exceed 98% of the limite given in the specifications for APLHOR in Section 3.10.A..The action to bring the core to 98% of the APLHGR L,

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

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b. The associated recirculation pump motor generator set circuit L

breaker shall be opened and defeated from operation.

1 OYSTER CREEK 3.3-3 Amendment No: 36 %, g, 140

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c._The suction valve, discharge valve and discharge bypass valve l

in the isolated loop shall be in the closed position and i!

4 associated motor breakers shall be opened =and defeated from operation.

d. The fully isolated loop as in 3.3.F.2.C above shall not be returned to service unless the reactor is in the cold BNUTDOWN condition.

3.

If specifications 3.3.F.1 and 3.3.F.2 are not met, an orderly shutdown shall be initiated immediately until all operable control rods ago fully inserted and the reactor is in either the REFUEL 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 />.

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

-l recirculation loop discharge valve and its associated suction i

valve.

i 6.

With reactor coolant temperature less than 212'F and irradiated 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 pooltion unless the reactor vessel'is flooded to a levelt above'185 inches TAF or unless the steam separator'and dryer are i

removed.

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OYSTER CREEK 3.3-3a Amendment No. )Jer,140

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pH, chloride, and other chemical parameters are made to determine the cause of.the unusual conductivity and instigate i

proper corrective action. _These can be done before limiting conditions, with respect to variables affecting the boundaries

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of the reactor coolant, are exceeCed. Several techniques are available to correct off-standard reactor water quality conditions including removal of impurities from reactor water 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 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 j

OPERATING recirculation loops during reactor POWER OPERATIOC.

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

3.3 F.3 requires reactor POWER OPERATION to be terminated and

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the reactor placed in the REFUEL 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 />.

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During four loop POWER OPERATION the idle loop,'when it is not isolated, is required to have its discharge valve closed and its discharge bypass and suction. valves open. This provides and limits reactor coolant backflow through an idle loop and thus minimises 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.

The requiremente 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 th'to avoiding any reactivity addition transient or reactor vessel nossle thermal stress concerne.

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 i

sufficient hydraulic communication between these areas, thus assuring that reactor water instrument readings are-indicative ofthelevelinthegoreregion.- For the bounding loss of feedwater transient ( I, a single fully open recirculation loop transfers coolant from the annulus to the core region at approximatelgfivetimestheboiloffratewithnoforced circulation (

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. ft, p5,1&f,140

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j References - (1) - FDSAR, Volume I, Section:!V-2

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Letter to NRC dated May 19, 1979, " Transient of May 2, 1979"

-(3) General Electric Co. Letter G-EN-9-55, " Revised Natural circulation F4w Calculation", dated May 29, 1979 (4)' Licensing Applisstion Amendment 16, Design.

l Requirements Sect.9n (5)

(Deleted)

(6)

FDSAR, Volume I, Sec% ion IV-2.3.3 and Volume II, Appendix H (7) FDSAR, Volume 7, Table IV-2-1 (8) Licensing Application Amendment 34, Question 14

.(9) Licensing Application Amendment'28, Item III-3-2 (10) Licensing Application Amendment 32, Question 15 (11) (Deleted)

(12) (Deleted)

(13) Licensing Application Amendment 16, Page 1 (14) GPUN TDR 725 Rev. Os Testing and Evaluation of Irradiated Reactor Vessel Materiale Surveillance Program Specimens l

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OYSTER CREEK 3.3-8a Amendment No. g, 140

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The maximum average planar LHOR limita of fuel-types V and VB are shown in Figure 3.10-1 for five loop operation and in Figure 3.10-2 for four loop operation, and are the result of toCA analysts performed by Exxon Nuclear Company utilising 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 3.10-2 provided the fifth loop has its discharge valve closed and its bypass and suction valves open. Four loop operation is permitted with tne idle loop isolated (suction, discharge and discharge bypass valves I

clocad) with Exxon fuel assemblies since the Exxon assemblies are i

located only on the core periphery and operate at significantly lower-MAPLHOR values than the rest of the core. The MAPLHOR multiplier in Figure 3.10-3 is further reduced for an isolated idle loop consistent 1

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 LHOR limits shown in Figures 3.10-1 and j

3.10-2 for Type V and VB fuel were analysed with 100% of the spray cooling coefficients specified in Appendix R to 10 CFR Part 50 for 7 x 7 fuel. These spray heat transfer coefficients were justified in the ENC Spray Cooling Heat Transfer Test Program (2).

The maximum average planar LHCR 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 MAPLHOR multiplier 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 q

will not exceed the 2200'F limit.

i The ef fect of axial power profile peak location for fuel types V and VB is evaluated for the worst break size by performing a series of fue1~ heat-up calculations. A set of multipliers is devised to reduce the allowable bottom skewed axial power peake 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 1

deterioration in heat transfer during the extended downward flow period during blowdown. The MAPLHOR 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 in1the 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.: J5,})1(n140

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Ccic11ations 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 axial shrinkage due to densification, the analyses of peak clad temperatures do not consider any change in LHOR 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 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 KN-174.

Pellet-clad thermal conductance for Type V and VB fuel was calculated using the CAPEX 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 analysia 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. 'che ' 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 flow condition for the steady-state core operation results in rapid voiding of the core during the blowdown period of the LOCA. The rapid 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 1024 power provides a conservative basis for evaluation of the peak cladding temperature and the maximum average planar linear heat. generation rate (MAPLHGR) for the reactor.

OYSTER CREEK 3.10-5 Amendment No.:

K, JFf, ps,140

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Tho APRM rosponoo 10 uoCd to predict whon tho rod bicek cecuroLin tho-

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analysis of the-red withdrawal error transient.' The transient rod position at the rod. black 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 the 3

worst case APRM status condition (APRM Status 1) is determined in tne rod withdrawal error transient analysis.

The steady state MCPR values for APRM status conditions 1, 2, and 3 will be evaluated each cycle.

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 withdrawal' 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 eystem while in a, condition which prevents the possibility of this transient occurring.

The steady-state MCPR limit was selected to provide margin to accommodate transiente 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 eladding integrity MCPR limit designated in specification 2.1.

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

The purpose of the Kg factor.is to define-operating limits at other than rated flow conditions. At.less that 100% flow the required MCPR is the product of-the operating limit McPR and the K factor.

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

The K factor curves shown in Figure 3.10-6 were developed g

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.

Por the manual flow control mode, the K factors g

were calculated such that at the maximum flow state (as limited by the pump accop 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 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.g OYSTER CREEK 3.10-6 Amendment No. t JT, JYI, g,140

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REFERENCES

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-(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 4 EWR Fuel 60 and 63 Active Rods, Interim Report," October 1975.

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

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

. SAFER /COREC00L/GESTR-LOCA LOSS-OF-COOLANT ACCIDENT ANALYSIS,"

August 1987.

4 OYSTER CREEK 3.10-6a Amendment No.:

FI, 121f, 140 l

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