ML20205J767

Proposed Tech Specs Increasing Min Required RCS Flow Rate & Changing SRs for RCS Flow Rate

ML20205J767
Person / Time
Site:	Fort Calhoun
Issue date:	03/31/1999
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20205J756	List: LIC-99-0031, Forwards Application for Amend to License DPR-40,increasing Min Required RCS Flow Rate & Changing SRs for RCS Flow Rate ML20205J763 ML20205J767
References
NUDOCS 9904120186
Download: ML20205J767 (11)

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U.S. Nuclear Remilatory Commission I l

LIC-99-003 . )

Attachment A l Requested Changes of Technical Specifications set forth in Appendix A of the l Facility Operating License i No. DPR-40 i i

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05 85 P PDR 52 l

' 2:0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued)

-(5) DNBR Margin During Power Operation Above 15% of Rated Power (a) The following limits on DNB-related parameters shall be maintained: "4'k:- the ""*!e, '^" -

(i) Cold Leg Temperature (4)-as specified in the COLR (Core Inlet Temperature)

(ii) Pressurizer Pressure 2 2075 psia *

(iii) Reactor Coolant Flowrate 2 o?,000 206 000 gpnt* indicated (iv) Axial Shape Index,-% (2)-as specified in the COLR (b) With any of the above parameters exceeding the limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce power to less than 15% of rated power within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Basis ,

The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200 F.

Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System, or the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limit. The Excore Detector Monitoring System performs this function by continuously monitoring the axial shape index with the operable quadrant symmetric excore neutron flux detectors and verifying that the axial shape index is maintained within the allowable limits of the Limiting Condition for Operation for Excore Monitoring of LHR Figure provided in the COLR as adjusted by Specification 2.10.4(1)(c) for the allowed linear heat rate of the Allowable Peak Linear Heat Rate vs. Burnup Figure provided in the COLR, RC Pump configuration, and F,y' of the Fa\ F,y' and Core Power Limitations Figure provided in the COLR. In conjunction with the use of the excore monitoring system and in establishing the axial shape index limits, the following assumptions are made: (1) the CEA insertion limits of Specification 2.10.l(6) and long term insertion limits of Specification 2.10.l(7) are satisfied, (2) the flux peaking tugmentation factors are as shown in Figure 2-8, and (3) the total planar radial peaking factor does not exceed the limits of Specification 2.10.4(3).

• Limit not applicable during either a themial power ramp in excess of 5% of rated thermal power per minute or a thermal power step of greater than 10% of rated thermal power.

• • - ""r'ber 4 r 2c'"e! " t rd ce~erp^^& '^ r :-direted n^"' rete ^r 202,50^ gp-' ^"

~her v2!ue " @!r "~: ; re :ndir'ed 2!ue rd :nc!"A= r 2!!^":2nce r^- errement "nce-+-+y (e.g., 513 r, :"d!~"ed, 2!!^" - r^ r ect" Tc ef 515 F).

(i) un.n:- .ng u.,,:. cm c gre 1.,i .g 7e.,,m,.2.m g 77m, gggg !., ne ent n (3)una,:-.ng :,,, : . :,,; cm.a:1:m, cm npr.:mm i cm. rwn u~,:.m.:gg r:gmg m.m.gg :- .ng COLR, 2-57c Amendment No. 32,4,52,30,23,92,! no,' ",! A !, ! 5^

2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued)

In order for these objectives to be met, the reactor must be operated consistent with the operating limits specified for margin to DNB.

The parameter limits given in (5) and the Fa', F,y' and Core Power Limitations Figure provided in the COLR along with the parameter limits on quadrant tilt and control element assembly position (Power Dependent Insertion Limit Figu'e provided in the COLR) provide a high degree of assurance that the DNB overpower margin will be maintained during steady state operation.

The actions specified assure that the reactor is brought to a safe condition.

Se c+^ cee!ent pump d!ffe--t! prrure ~^ he :ng rytter my be used 'e neerure F-"

The Reactor Coolant System flow rate of 206,000 gallons per minute is the indicated value. It does not include instrumentation uncertainties.

The calorimetric methodology shall be used to measure the Reactor Coolant System flow rate.

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l 2-57e Amendment No. 32,57,1d!,157,449 (next page is 2-59)

3.0 SURVEILLANCE REQUIREMENTS

• 3.1 Instrumentation and Control (Continued)

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

The minimum calibration frequencies of once-per-day (heat balance adjustment only) for the power rage safety channels, and once each refueling shutdown for the process system channels, are considered adequate.

The minimum testing frequency for those instrument channels connected to the Reactor Proiective System and Engineered Safety Features is based on ABB/CE probabilistic risk analyses and the accumulation of specific operating history. The quarterly frequency for the channel functional tests for these systems is based on the analyses presented in the NRC approved topical report CEN-327-A, "RPS/ESFAS Extended Test Interval Evaluation," as supplemented, and OPPD's Engineering Analysis EA-FC-93-064, "RPS/ESF Functional Test Drift Analysis."

The bw temperature setpoint power operated relief valve (PORV) CHANNEL FUNCTIONAL TEST verifies operability of the actuation circuitry using the installed test switches. PORV actuation could depressurize the reactor coolant system and is not required.

Calculation of the Reactor Coolant System (RCS) total flow rate by performance of a j precision calorimetric heat balance once every 18 months verifies that the actur.1 RCS flow rate is greater than or equal to the minimum required RCS flow rate (Table 3-3, Item 15, Reactor Coolant Flow).

The Frequency of 18 months reflects the importance of verifying flow after a refueling outage when the core has been altered, Steam Generator tubes plugged, or other activities, which may have caused an alteration of flow resistoce.

'Ihis requirement is modified by a footnote that requires the surveillance to be performed ethin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after 2 95% reactor thermal power (RTP) following power escalation from a refueling outage. The footnote is necessary to allow measurement of the flow rate at normal operating conditions at power in MODE 1.

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U.S. Nuclear Regulatory Commission LIC-99-0031 1

Attachment B Discussion, Justification and No Significant Hazards Consideration

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m DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT IIAZARDS CONSIDERATION DISCUSSION AND JUSTIFICATION:

1. Reactor Coolant System Flow Rate for Power Operation Above 15% of Rated Power The proposed amendment revises Section 2.10.4 of Technical Specifications to increase the minimum Reactor Coolant System (RCS) flow rate to a new indicated value of 206.000 gallons per minute (gpm), and (for clarification) deletes the footn.ote relevant to the RCS flow rate. This eliminates the need for a cross reference between the actual flow rate and the indicated flow rate, and still maintains conservatism. The increase in credited RCS flow rate results in higher operating thermal margin and additional operational flexibility.

The credited RCS flow rate has increased because in 1998, the Fort Calhoun Station (FCS)

Unit Number 1 stafTperformed modification MR-FC-97-005 (Steam Generate- Orifice Plate Removal) to remove the orifice plates from the steam generators. The modi', - on design basis included an increase in estimated RCS flow rate from 39.34 x 10' pou.2 per hour (lbm/hr) to 41.31 x 10' lbm/hr, which corresponds to 207,500 gpm and 217,904 gpm, respectively.

The RCS flow rate measured using calorimetric methodology has been 208,800 gpm average for Cycle 18 through January 1999. The proposed RCS flow rate of 206,000 gpm (indicated) ,

represents a conservative value that would allow for potential pump degradation and l measurement variability.

Currently, Technical Specifications Section 2.10.4 specifies a Reactor Coolant Flow rate of 2197,000 gpm with a footnote indicating that the value is an actual limit and corresponds to an indicated value of 202,500 gpm. The analysis value used for RCS flow rate conservatively accounts for instrumentation uncertainties.

i The proposed amendment also changes the wording format regarding the cold leg i temperature and the axial shape index in Section 2.10.4. The Basis ofSection 2.10.4 is being l revised to clarify that (1) the RCS flow rate of 206,000 gpm is the indicated and does not I include instrumentation uncertainties, and (2) the calorimetric methodology shall be used to measure the RCS flow rate.

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I DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT IIAZARDS CONSIDERATION

2. Surveillance Frequency for Reactor Coolant System Flow Rate Fort Calhoun Station is currently using the Reactor Coolant Pump AP methodology to satisfy item 15 of Table 3-3 (Minimum Frequencies for Checks, Calibrations and Testing of Miscellaneous Instrumentation and Controls). FCS also has the ability to perform calculation of the RCS Flow Rate using the calorimetric, heat-balance methodology, which is independent from the pump curves. If a reactor coolant pump's impeller is replaced resulting in a probable change to the pump curve, the calorimetric methodology will allow the determination of the RCS flow rate without the pump curves. In addition, the NUREG 1432 (Improved Technical Specifications for Combustion Engineering Plants) surveillance section 3.4.1.4 requires that the Surveillance Frequency be 18 months, with the precision heat balance (calorimetric) methodology used. Therefore, OPPD proposes to change the minimum frequency specified in item 15 of Technical Specifications Table 3-3 regardirig RCS flow rate calculation from "M" (Monthly) to "R" (Refueling). Also OPPD proposes to add a note that states the surveillance is required to be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after 2 95% reactor thermal power (RTP). This is consistent with requirements contained in the NUREG 1432. The Bases of NUREG 1432 state that the footnote is necessary to allow measurement of the flow rate at normal operating conditions at power in MODE 1. The surveillance cannot be performed in MODE 2 or below, and will yield more accurate results if performed above 95% RTP.

The Bases of Section 2.10.4 and Section 3.1 are revised to support these changes.

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DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT IIAZARDS CONSIDERATION BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION:

The proposed changes to the Fort Calhoun Station Unit No.1 Technical Specifications do not involve significant hazards consideration because operation of FCS in accordance with the change would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

Nmbustion Engineering (ABB/CE) in Thermal-Hydraulic Report CR-94-19-CSE95-1131, ision 0 performed a comprehensive evaluation of the effects the removal of the orifice tes would have on steam generator tube degradation. It was concluded that the removal

'he orifice plates would increase the primary flow rate by approximately 5%.

The removal of the orifice plates was estimated to increase the probability of tubes requiring repair over the lifetime of the plant. However, the presence of the orifice plates had prevented inspection of approximately 22 % of the steam generator tubes for circumferential cracks on the hot leg side. Therefore, it was concluded that the removal of the orifice plates did not increase the probability of steam generator tube failure, given that the tubes previously covered by the plates are now inspected each outage in accordance with the Electrical Power Research Institute Pressurized Water Reactor (EPRI PWR) steam generator examination guidelines. Fort Calhoun Station is using the eddy current inspection technology to ensure that tubes showing evidence of a crack exceeding the present plugging criteria will be repaired or removed from service. Industry experience has shown that even in cases of severely degraded tubes, the resulting primary to secondary leak rates are insignificant compared to those j analyzed in the design basis steam generator tube rupture event.

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Calculation of the Reactor Coolant Flow Rate using the heat balance methodology once every refueling outage is consistent with requirements contained in the NUREG 1432, improved Technical Specifications for Combustion Engineering Plants' surveillance requirement {

3.4.1.4.

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l DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT HAZARDS CONSIDERATION (2) Create the possibility of a new or different kind of accident from any accident previously evaluated.

The original orifice plates were installed on each steam generator hot leg tube sheet in the primary inlet plenum as a field modification prior to the initial fuel load in the year 1973. The orifice plates were designed to increase the hydraulic resistance of the primary coolant flow rate in the associated tubes, theren reducing the primary coolant temperature inside the tubes.

Reduction of the primary coolant temperature and flow rate would decrease the heat flux, thus improving the steam quality and reducing the potential for dry-out and surface deposits on the outer surface of the tubes. liowever, due to inaccessibility, these originally installed orifice plates had prevented tube inspection in the hot lag tube sheet area, even with the latest state-of-the-art eddy current probe technology. The orifice plates also prevented normal repair techniques such as steam generator tube plugging and sleeving.

The original orifice plates were removed during the 1996 refueling outage. Ilowever, there were concerns related to Westinghcuse fuel failures as a result of flow-induced vibration. To address those concerns, new " removable" orifice plates were installed to maintain the RCS flow rate at the previous level. Since then, the remaining batches of the Westinghouse fuel l considered most susceptible to flow-induced vibration were replaced during the 1998 refueling outage, thus minimizing the concerns and allowing the permanent removal of the

" removable" orifice plates.

The removal of the " removable" orifice plates returned the steam generators to their original design configuration. RCS flow rate has increased by virtue ofdecreased hydraulic resistance through the steam generators. No other systems or components other than the steam generators have been affected. The resulting change in operational parameters (decreased reactor coolant Tno, temperature and increased flow rate) has been evaluated for the Updated Safety Analysis Report Chapter 14. Potential adverse consequences of the modifications were (1) increase in reactor vessel component vibration, (2) increase in hydraulic loading, and (3) increase in steam generator tube degradation for row 1-18 tubes. The potential adverse consequences were evaluated and found to be acceptable.

Calculation of the Reactor Coolant Flow Rate using the heat balance methodology once every refueling outage is consistent with requirements contained in the NUREG 1432, Improved l Technical Specifications for Combustion Engineering Plants' surveillance requirement 3.4.1.4.

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DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT IIAZARDS CONSIDERATION (3) Involve a significant reduction in a margin of safety.

The removal of the orifice plates has resulted in approximately a 5% increase in the reactor coolant flow rate. This has increased the margin for minimum reactor coolant system flow rate specified in Technical Specifications Section 2.10.4, Power Distribution Limits, item (5),

DNBR Margin During Power Operation Above 15% of Rated Power. Steam Generator tube inspections performed in accordance with Technical Specifications Section 3.17, Steam Generator Tubes, have not been adversely affected.

The increased flow rate has been analyzed for the thermal hydraulic effects on the reactor core and was found acceptable.

Calculation of the Peactor Coolant Flow Ratc using the heat balance methodology once every refueling outage is consistent with requirements contained in the NUREG 1432 [ Improved Technical Specifications for Combustion Engineering Plants] surveillance requirement 3.4.1.4.

Based on the above considerations, OPPD concludes that th ' proposed amendment to FCS Technical Specifications do not involve significant hazards considerations as defined by 10 CFR 50.92 and that the proposed amendment will not result in a condition which significantly alters the impact of the station on the environment. Thus the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22 (c) (9) and pursuant to 10 CFR 51.22 (b) no .

environmental assessment need be prepared.

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