ML20028D931
| ML20028D931 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/10/1983 |
| From: | Balukjian H, Kopp L, Powers D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20028D916 | List: |
| References | |
| NUDOCS 8301200095 | |
| Download: ML20028D931 (28) | |
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UNITED STATES l
NUCLEAR REGULATORY COMMISSION
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I SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 61 TO FACILITY OPERATING LICENSE NO.-DPR-69 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO. 2 DOCKET NO. 50-318' a
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B301200095 830110 DR ADOCK 05000
Table of-Contents Page
--1.0 Intro' duction 1
2.0 Fuel System Design
,2 2.1 -Fuel Mechanical Design 2
2.2 Fuel Thermal Design 3
3.0 -Nuclear Design 6
3.1 Physics Characteristics 6
e 3.2 Power Distribution 7
'3.3 Safety Related Data.
8 4.0 ' Thermal-Hydraulic Design 9
4.1 DNBR Analysis 9
4.2 Effects of Fuel Rod Bowing on DNB Margin 10 5.0 Safety Analyses-12 5.1 Anticipated Operational Occurrences 12 5. 2-CEA Ejection Transient 12 5.3 Other Accidents and Transients 13 v
6.0 Radiological Consequences of the Extended Fuel Burn-up 14 7.0 Technical Specifications 15 7.1 Thermal Margin Safety Limits 15 7.2 Peripheral Axial Shape Index 15 7.3 Minimum DNBR and Power Uncertainty 15 7.4 Use of Excore Detectors for Linear Heat Rate Monitoring 16 7.5 Implementation of BASSS 17 7.' 6 Shutdown Margin 17 7.7 RTD Response Time 18 7.8 Fuel Enrichment 18 7.9 Pressure Transmitters 18 7.10 Plant Modifications 19 7.11 Post-Accident Monitoring Instrumentation 21 21 8.0 Conclusions 8.1 Environmental. Considerations 22 8.2 Safety Conclusions 22 9.0 Refereness 23 i
. 130 Introduction
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By appitcation dated October 15, 1982, the Baltimore Gas and Electric Company (BG&E) submitted an application.for license amendment to address Cycle 5 Operation of Calvert Cliffs Unit 2 at a rated-thermal power of 2700 MWt. Our evaluation of Cycle 5 includes the consideration of ex-tended burnup, equivalent to approximately 18 months of plant nperation, and also the use of new analytic methods in predicting fuel and thermal-hydraulic performance.
In addition to Technical Specification changes associated with Cycle-5 operation,"we have reviewed a special test ex-ception which will facilitate startup testing of the auxiliary feedwater systen. Changes to the requirements for post-accident monitoring have also been reviewed.
The October 15, 1982 application was supplemented by lett.er.
dated ttovember 17, 1982 and other documents as indicated Section 9.0.-
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i 2-1 2.0 Fuel System Design This section provides an evaluation of the fuel system design and performance analysis (Ref.1) submitted in' support of the proposed fifth:
cycle of operation at Calvert Cliffs Unit 2.
The licensee's analyses of the reload core characteristics have been conducted with respect to the.
Calvert Cliffs Unit 1 Cycle 6 safety analysis (Refs. 2 and 3). The use of_ the recent Unit 1 safety analysis as a reference cycle for the Cycle 5 operation of Unit 2 has been proposed because of the similarity in the g
basic system chara'cteristics between the two reload cores.
The fuel system design and performance analysis in the reference cycle was previ-ously reviewed and approved (Ref. 4) by the NRC. Where the reference cycle analyses envelop the proposed cycle conditions, or where the re-vised analyses rely on previously-ap,7 roved methods, we continue to find those analyses acceptable.
2.1 Fuel Mechanical Design The mechanical design of the 76 fresh Batch G fuel assemblies is identical y
to that of the standard Batch H fuel described in the reference cycle submittal.
It is also identical to the fresh assemblies previously inserted in Calvert Cliffs Unit 2 Cycle 4 with the exception of a 0.200 inch reduction in the overall length of the fuel rods. The length reduction will provide additional clearance for fuel rod length increases during the extended burnur lifetime of this fuel.
The length reduction is small and the change was previously found acceptable for the reference cycle. Therefore, we continue to accept the change.
There is recent evidence (Ref. 5) of insufficient shoulder gap clea.ance in the Combustion Engineering 16X16 fuel at ANO-2, and the NRC staff is formally investigating this issue for the 16X16 fuel in all C-E plants.
No evidence' has been found to indicate a significant problem with 14X14 fuel at Calvert Cliffs.
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- BG&E has measured shoulder gap clearances in ~17 assemblies during.the present
. outage and plans further observations during the next outage to' monitor shoulder gep? clearances. The NRC staff will keep; informed of:these activities and take-appropriate action if the-need arises.
The fuel vendor, Combustion Engineering, has also performed analytical predictions of the cladding collapse time' forLall Calvert Cliffs. Unit'2 Cycle 5 fuel.and has concluded that: the collapse resistance of 'all e
standard fuel rods is' sufficient to preclude collapse during their design li fetime. Although these predictions were made for Cycle 5 conditions, they utilized methods of analysis previously accepted in' the reference cycle analysis. We. conclude that the. creep collapse calculations have been perforned in an acceptable manner.
1-The licensee has-also informed us thct all standard fuel assemblies used
~ in control element-assembly (CEA) positions during Cycle 5 operation will have stainless steel sleeves installed in the guide tubes. This modifi-w cation, which is used to prevent guide tube wear,'was previously accepted I
in the reference cycle Safety Evaluation.
The Cycle 5 core'will also contain _one previously-irradiated prototype CEA.
This prototype will. undergo a third cycle of irradiation during Cycle 5.
We find these' proposed Cycle 5 con-ditions acceptable.
2.2.
Fuel Thermal Design The performance of the fuel in the Calvert Cliffs, Unit 2, Cycle-5 core has been analyzed using a revised version (Ref. 6) of the Combustion l
Engineering fuel performance code (Ref. 7), called FATES-3.
The code is I
used in a number of areas in the safety analysis, including fuel rod initial conditions for the analysis of the LOCA and other transients and 4
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accidents, the power-to-centerline melt lir3it, minimum and maximum core-average gap conductance,1 fuel stored energy for containment analy' sis, maximum end-of-life rod pressure, and fuel mechanical design limits.
The Calvert Cliffs Unit 2 Cycle 5 reference cycle (i.e., Unit 1 Cycle-6) safety analysis was the first time'the FATES-3 code was used in a licensing application. At the time our evaluation of the Unit 1 Cycle 6 safety analysis was issued, we had not yet completed our review of the
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FATES-3 code.
However, the review had identified several staff concerns, notably code conservatism and fission gas release.
Because the reload schedule for Unit 1 Cycle 6 did not permit resolution of these ~ issues, we obtained cyclo specific conditions from the licensee (Ref. 8) and reproduced a number of the fuel performance calculations described in that reload report with a staff audit code (Ref. 9).
Those calculations, which include LOCA initial conditions, power-to-centerline melt, maximum fuel average temperatures, and end-of-life rod pressure, were expected u
to be most limiting in the reload safety analysis and most affected by those issues identified by the staff.
In two areas, LOCA initial conditicns and end-of-life rod pressure, the analyses presented by the licensee were less limiting than those pro-duced by the staff.
In response to the first area of concern, the licensee reported results (Ref.10) of a supplemental calculation which showed the calculated peak cladding temperature, peak local, and core-average cladding oxidation levels remained below the 10 CFR 50.46 acceptance limits.
On tnis basis, we found the Unit 1 Cycle 6 Technical Specification limit of 15.5 kW/ft on peak linear heat generation rate acceptable without further review of FATES-3. The LOCA analysis per-formed for Unit 2 Cycle 5 used input data, including the reload fuel e
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I l 1 performance parameters, which apply to both Unit 2 Cycle 5 and Unit 1 LCycle 6.
Hence, our approval of the reference cycle ' conditions continues to apply to Unit 2 Cycle 5 for operation at a peak linear. heat generation rate of 15.5 kW/ft and at a power level of 102% (2754 MWt).
With regard to the end-of-life rod pressure limit, we concluded that this limit would be met for assembly-average burnups below approximately 38,000 MWD /MTV. This value was based on our own. audit calculation as well as one produced by the previous version of the FATES code (Ref. 7) using a fission gas release burnup enhancement factor supplied by the staff (Ref.11).
Based on information contained in the Unit 2 Cycle 5' reload report, the end-of-cycle burnups are predicted to be below this limit for all assemblies in the Cycle 5 core.
With' respect to the reference cycle safety analysis, the licensee has identified one exception in the Unit 2 Cycle 5 fuel thermal design e
analysis. This is the power-to-centerline melt limit.
For the Unit 2 Cycle 5 analysis, this limit was determined by taking some credit for the decrease in power peaking which takes place at high burnup. The calculation assumes that cladding temperatures remain near those at normal operating conditions. This is assured by simultaneous appli+
cation of the DNBR limit.
This new calculation results in an increase in the power-to-centerline melt limit from 21 to 22 kW/ft.
Based on our previous audit calculation of this limit, we find the proposed modification acceptable.
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'3.0 fluclear Design
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3.1 Physics' Characteristics
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The Cycle 5 burnup is expected to be between 13,100 MWD /T-and 13,700
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MWD /T, depending on the final-Cycle 4 termination point. The Cycle 5 -
physics characteristics were calculated for Cycle 4 tenninations be-tween 16,300 and 17,300 MWD /T and the loading pattern presented is applicable.to any Cycle 4 termination point within this _ band. Based upon conversations with representatives of BG&E, the Cycle;4 discharge was determined to be 16,533 MWD /T and, therefore, the physics charac-teristics are valid.
Cycle 5 will contain 48 fuel assemblies (Batch G) of 4.0 w/o enrichment.
This higher' enrichment has. been accounted for in the determination of Cycle 5 physics parameters and in the safety analyses. Also, approval has been given for storage of up to 4.1 w/o enriched fuel in Calvert--
Cliffs Units 1 and 2 (Ref.12. ) Therefore, operation with the higher enrichment Batch G assemblies is acceptable.
The Cycic 5 moderator temperature coefficient (MTC) is calculated to be ap/ F for beginning of cycle (B0C) and -2.1 x 10-4 ap/ F
-0.1 x 10-4 U
0 for end of cycle (E0C).
The Doppler coefficient for Cycle 5 is.a best estimate value expected to be accurate to within 15%.
These values are bounded by the values used in the safety analyses for the reference cycle (Calvert Cliffs Unit 1 Cycle 6). We, therefore, find the values of the MTCs and Doppler coefficients to be acceptable.
Other Cycle 5 physics parameters are also comparable with the reference cycle value,s and are acceptable.
The values of these parameters as used in the safety analyses are appropriately modified for calculational uncertainties so as to conservatively bound the predicted values.
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I The zero power steam line-break' accident occurring at EOC is the most limiting and.provides the basis-for-establishing the Technical Speci-fication required shutdown margin which for Cycle 5 is 5.2% ap.' At the enij of Cycle 5, the calculated hot zero power (HZP) ret.ctivity worth of all CEAs inserted assuming the highest worth CEA is stuck out of the core is 7.6% an..The CEA bite, which accounts for the possibility of the CEAs being slightly inserted rather than fully withdrawn, reduces the worth by an additional 1.8% ap, resulting in a calculated scram worth of 5.8% ap. Assuming a 10% calculational uncertainty, the net available calculated HZP scram worth at E0C is 5.2% ap. Since this is equal to the Technical Specification shutdown margin and includes a 10% calculational uncertainty, it is acceptable.
3.2 Power Distribution Radial power distributions are given for the all rods out (ARO). con-
' dition at beginning, middle, and end of cycle that are characteristic of '
t the high burnup end of the previous cycle (Cycle 4) shutdown window.
The high burnup end of the Cycle 4 shutdown window tends to increase the power peaking in the axial central region of the core for Cycle 5.
The single rod power peaking values include an allowance for the increased peaking that is characteristic of fuel rods adjoining the water holes in the fuel assembly lattice.
The safety and setpoint analyses conservatively include uncertainties and other allowances so that the power peaking values actually used are higher than those expected to occur at any time during Cycle 5.
Incore detector measurements are used to compute the core peaking factors using the INCA code (Ref. 14).
The coeffi.cients required to perform this data re' duction are obtained using the methodology described in the topical report.
For Cycle 5 operation, the power distribution measurement uncertainties used will be 6 percent for the total integrated t/
radial peaking factor (F ) and 6.2 percent for the total power peaking O
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8-factor (F ).
These are the approved. values from Reference 14 and are, 9
therefore, acceptable.
3.3 Safety Related Data Other safety related data such as limiting parameters of dropped CEA reactivity worth and the augmentation factors (used to account for the power density spik,es due to axial gaps caused by fuel densification) for Cycle 5 are identical to the values used in the reference cycle and are, therefore, acceptable.
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4.0 Ther.aal-Hydraulic Design 4.1 DNBR Analysis The thermal margin methodology for Calvert Cliffs Unit 2 Cycle 5 has been modified and is in agreement with the approved methodology used for Calvert Cliffs Unit'l Cycle 6, a duplicate plant which references the same FSAR.
This methodology uses the CETOP code (Refs.15 and 16) which is a variant of the TORC code (Ref.17). The CETOP code was approved
. (Ref. 4) for use on the Calvert Cliffs Units and is optimized for simplified modeling and computer cost reduction.
It is used in con-junction with the CE-1 critical heat flux correlation (Ref.18). With the application of statistical combination of uncertainties (Ref.19) a new design limit in the CE-1 minimum DNBR of 1.23 has been approved (Ref. 4) for Calvert Cliffs Unit 1 Cycle 6 and is also applicable to Calvert Cliffs Unit 2 Cycle 5.
Since the uncertainties of system parameters such 'as engineering heat flux factor, hot channel factor, and rod pitch and cladding diameter factors are the same for both Calvert Cliffs Unit I q
and 2, the DNBR limit of 1.23 is also applicable to Calvert Cliffs Unit 2.
Table 4.1 lists the pertinent thermal-hydraulic parameters used for both safety analysis and for generating reactor protection system nc+eoints.
Due to the use of statistical combination of uncertainties ($cv) operating state parameters, which affect the limiting safety system settings and the limiting conditions for operation, the nominal inlet temperature and nominal primary system pressure were used. The SCU methodology for LSSS and LC0 has been previously approved for Calvert Cliffs Unit 1.
Since the same methodology is used (Ref. 20), the staff concludes that the SCU for Calvert Cliffs Unit 2 is also acceptable.
However, the staff is currently performing a gener.ic review regarding the instrument measurement uncertainties which are acceptable in the SCU analysis for LSSS and LCO.
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Table 4-1 compares thermal - hydraulic parameters at full. power fo.r Cycle 5 with the reference cycle which is Unit 1 Cycle 6.
It is noted that all the values shown are identical except for the slightly greater heat flux (0.7%) for Cycle 5 which is due to the smaller (0.7%) total heat transfer area for Cycle 5 accounted for by the axial densification factor. Note also that the heat transfer area for Unit 2 Cycle 5 is also smaller than for Unit i Cycle 6 due to a larger number of shims (928 for Unit 2 and 672 for Unit 1.) This causes'a small increase in the Unit 2 Cycle 5 core average heat flux and average linear heat rate compared to Unit 1.
6 The design reactor coolant flow rate is 381,600 gpm (143.8 x 10 lb/hr) whereas the actual measured flow rate from the reference c,scle (Unit 1, 6
cycle 6) is 403,700 gpm (152 X 10 lbm/hr) (Ref. 21) indicating ample margin.
9 The effect of a CEA guide tube wear prob ~iem and sleeve repair were investigated (Ref. 21) and it was concluded that there is no adverse affect on DNBR margin (Ref. 4).
4.2 Effects of Fuel Rod Bowing on DNB Margin The fuel rod bowing effects on DNB margin for Calvert Cliffs. Unit 2
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Cycle 5 were evaluated in accordance with the guidelines of Reference '4.
The following table of DNBR penalty versus bundle average burnup was obtained f' rom Reference 21.
i Bundle Average Burnup GWD/T DNBR Penalty %
24.0 0.00
, 25.8 0.60 30.5 12.20 34.8 3.60 35.2 3.75 i
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. A total of 141 fuel assemblies will exceed the NRC specified DNB penalty threshold burnup of 24 GWD/T (Ref. 23) during Cycle'5, the maxiInum assembly burnup reaching'35.2 GWD/T by-the end of the cycle.
For those assemblies which wiil experience a burnup of between 24 and 35.2 GWD/T at any time during Cycle. 5, the minimum best estimate margin available relative to more limiting peaking values present in other assemblies is greater than 11%.
The DNB rod bow penalty for this burnup range, as detennined from Reference 23, varies from 0 to 3.8%.
Since the magnitude of the margin available is considerably greater than the corresponding DNB rod bow penalty, no power penalty for fuel rod bowing is required in Cycle 5.
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TABLE 4-1 Calvert Cliffs Unit 2 Thermal-Hydraulic Parameters at Full Power Reference +
General Characteristics Unit Unit 1, Cycle 6**
Cycle 5**
TotalHeatOutput(coreonly)
MW 2,700 2,700 6
10 Btu /hr 9,215 9,21 5 Fraction of Heat Generated in Fuel Rod
.975
.975 Primary System Pressure (nominal) gsia 2,250
'2,250 Inlet Temperature F
548 548 Total Reactor Coolant Flow gpm 381,600 381,600 (steady state) 10 lb/hr 143.8 143.8 6
Coolant Flow Through Core 10 lb/h'r 138.5 138.5 '
Hydraulic Diameter (nominal channel) ft 0.044 0.044 6
2 Average Mass Velocity 10 lb/hr-ft 2.59 2.59 Pressure Drop Across Core (steady state psi 11.1 11.1 flow irreversible op over entire fuel assembly)
Total Pressure Drop Across Ve.ssel (based psi 34.7 34.7 on steady state flow & nominal dimensions) 2 Core Average Heat Flux (accounts for above Btu /hr-ft 184,266***
185,532****
fraction of heat generated in fuel rod &
axial densification factor) 2 Total Heat Transfer Area (accounts for ft 48,748***
48,415****
axial densification factor) 2 Film Coefficient at Average Conditions Btu /gr-ft_o F 5,930 5,930 Average Film Temperature Difference F
31 31 Average Linear Heat Rate of Undensified kw/ft 6.16***
6.20****
Fuel Rod (accounts for above fraction of heat generated in fuel rod)
Average Core Enthalpy Rise Btg/lb 66.5 66.5 Average Core Temperature Rise F
49.0++
49.0++
Maximum Clad Surface Temperature F
657 657 4
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Table 4-1 (Cont'd)
Reference +
Calculational Factors Unit 1, Cycle 6**
Cycle 5**
Engineering Heat Flux Factor 1.03*
1.03*
Engineering Factor on Hot Channel Heat Input 1.02*
1.02*
Rod Pitch & Clad Diameter Factor 1.065*
1.065*
Fuel Densification Factor (axial) 1.01 1.01 Total Planer Radial Peaking Factors Unrodded Region For DNB Margin Analyses (Fr) 1.70 1.70 For kW/ft Limit Analyses (Fxy) 1.70 1.70 Peak Allowable Linear Heat Generation Rate (kW/ft) 15.5 15.5 Limiting Transient Minimum DNBR CEA Drop
>1.23
>1.23 Loss of Flow sl.23 51.2 3 Minimum Allowable DNBR 1.23 1.23
- These factors have been combined statistically with other uncertainty factors at 95/95 confidence / probability level (Ref.19) to define a new design limit on CE-1 minimum DNBR when iterating on power as discussed in Reference 19 and approved.by the NRC in Reference 4.
- Due to the statistical combination of uncertainties described in References 19, 24, and 25, the nominal inlet temperature and nominal primary system pressure were used to calculate some of these parameters.
- Based on Unit 1, Cycle 6 specific value of 672 shims.
- Based on Unit 2, Cycle 5 specific value of 928 shims.
+ Reference cycle (Unit 1, Cycle 6) analysis is contained in Reference 2.
++ Reference 21.
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5.0 Safety Analyses _
5.1 Anticipated Operational Occurrences The CEA group withdrawal and full length CEA drop anticipated operational occurrences (A00s) were re-evaluated for Cycle 5.
The key transient input parameters were conservative with respect to the reference cycle (Unit 1, Cycle 6)valuesand,therefore,noreanalysiswasnecessary.
The results and conclusions in the' reference cycle analysis are still valid for Unit 2. Cycle 5.
F 5.2 CEA Ejection Transient The CEA ejection transient was reanalyzed for Cycle 5 because of the lower available scram worth at trip and changes in the post ejected 4
three-dimensional pin power peak and ejected CEA worth as compared to the reference cycle.
The analytical method used in the reanalysis is the NRC approved Combustion Engineering CEA ejection method described in CENPD-190-A(Ref.26).
The_most limiting key safety parameters in Cycle 5 were used to bound the most adverse conditions.
These included the least negative Doppler coefficient, the most positive moderator temperature coefficient, and an E0C delayed neutron fraction to produce the highest power rise during the event.
The analyses show that both the zero power and full power cases result in peak fuel enthalpies less the the NRC limiting criterion of 280 cal /gm.
We conclude that the initial assumptions and analytical models used ensure that primary system integrity will be maintained in the event of a CEA ejectjon.
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. '5.3 Other Accidents and Transients The three design basis even,ts:
loss of feedwater, feedwater line break, and steam line break events were analyzed with and without loss of AC power (LOAC) on turbine trip to ensure that relevant acceptance criteria continue to be met in light of the new AFAS logic.
C A spectrum of feedwater and steam line break sizes were analyzed. The 2
2 for worst break size was 1.275 ft for the feedline break, 6.305 ft steam line break inside containment and 0.33 ft2 for steam line break 2
outside containment with LOAC on turbine trip. The 0.275 ft feed line break downstream of the check valve with LOAC on turbine trip resulted in the highest peak pressure but did not exceed the pressure limit of 2750 psia. The 6.305 ft2 steam line break inside containment resulted in the maximum post trip return to power and minimum DNBR. However, the DNBR still was above the design limit of 1.23.
The 0.33 ft2 steam line break outside containment resulted in the maximum number'of predicted fuel pin failures where 2f. of fuel pins experienced DNB, however, the resulting site boundary dose was well within the 10 CFR 100 guidelines.
The result of the loss of feedwater flow event with and without LOAC following reactor trip showed that the peak pressure did not exceed the upper limit of 2750 psia and that an adequate heat sink was maintained during the event.
The licensee performed an ECCS analysis using NRC accepted methods that were approved for Unit 1 Cycle 6.
They used input data applicable to both Unit 1 Cycle 6 and Unit 2 Cycle 5.
The ECCS analysis resulted in a peak clad temperature of 20380F, versus the acceptance criteria limit of 0
2200 F, The peak local clad _ oxidation was 8.5% compared to the acceptance cirteria limit of 17%, and the peak core wide clad oxidation was less than 0.51% versus the acceptance criteria limits of 1.0%. Thus we conclude that the ECCS analysis of Unit 2 Cycle 5 is acceptable.
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6.0 Radiological Consequences of the Extended Fuel Burn-up
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On June 24, 1982, the NRC issued License Amendment No. 71 for C,al' vert Cliffs Unit 1.
This amendment authorized Cycle 6 operation of Calvert Cliffs Unit 1 and included our evaluation of the radiological conseque'nces of design basis events for an extended fuel burn-up corresponding to approx-imately 18 months of plant operation.
By letter dated October 15, 1982, BG&E provided a Safety Analysis based on Calvert Cliffs Unit 1 Cycle 6 as the reference cycle. This cycle was chosen as the reference because, in addition to the great similarity in important parameters between the two units, the burnup considered is the same: 33,700 mwd /MTU batch average at discharge.
As indicated above, the staff previously reviewed in detail the effects of burnup on design basis accidents for the reference cycle. The conclusion of this review wasthatthedesignoftheplantandthecyclereloadconditidnswere adequate to mitigate the consequences of design basis accidents up to a v
batch average discharge burnup of 34,000 mwd /MTU.
The licensco's submittal for Unit 2 Cycle 5 was reviewed to determine if factors for the reload would negate a similar conclusion for.the requested action. No such factors were identified.
Based on the Intensive review of the reference cycle and the similarity of the two units and reload factors, it is concluded that operation of Unit 2 Cycle 5 is acceptable based upon the expected batch average discharge burnup.
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7.0 Technical Specifications
' Draft Technical Specifications changes were presented in the October 15, 1982 application and.in the November 17, 1982 supplement. These changes are addessed herein.
l7.1 Thermal Margin Safety Limits TS Figure 2;1-1, " Reactor Core Thermal Margin Safety. Limit", has.been changed to reflect higher. radial peaking factors and implementation of the margin recovery programs.* The corresponding Basis-.in B2.1.1 has been
. changed to reflect an increase in the maximum steady state peak linear heat generation rate-(centerline fuel melt) from 21 to 22.0 kw/ft. We' find these changes acceptable.
7.2 Peripheral Axial Shape-Index TS Figure 2.2-1 has been changed to reflect the increase in maximum steady state peak linear heat generation rate (centerline fuel melt) as indicate'd b
above. We find this cha'nge acceptable.
7.3 Minimum DNBR and Power Uncertainty The minimum DNBR reflected in TS Figure 3.2-4_and as stated in Bases B.2.1.1,.B.2,2.1, and B.3/4.2.5 has been changed from 1.195 to 1.23 as a result of application of SCU.
In addition, the TM/LP trip description
- 0n July 17, 1980 the licensee met with the NRC staff to discuss the margin recovery program. This program involves the use of SCU and other analytic techniques which were necessary to "recov.er".the decrease in operating _ margins tilat would have resulted from the generally less advantageous parameters (such as peaking factors) associated with the extended burn-up fuel cycle.
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I r in Bases B.2.2.1 has been revised to reflec-t the SCU methodology., The NRC review of the SCU methodology is contained in Section 4.1 herein.
Another. change associated with SCU involves the 2% power uncertainty which is no longer required. This change is reflected in Bases B.2.1.1 and B.2.2.1.
We find these changes acceptable.
C 7.4 Use of Excore' Detectors for Linear Heat Rate Monitoring A modification to the use of excore detectors used to monitor linear heat rate (LHR) has been made. This technique will avoid unnecessary power level changes resulting from temporary on-line computer outages.
The computer is required to interpret the in-core detectors which had previously been the sole means of monitoring the peak LHR to justify operation at full power. The following changes.are associated with this use of excore detectors:
(1) Figure 3.2-2 is changed. This change also reflects the margin recovery programs and the increase in radial peaking.'
(2) TS 4.2.1.3 is revised to provide credit for the calculated value ofFx}whenmonitoringtheLHRLimitingConditionforOperation with the excore detectors.
(3) TS 3/4 2.2 has now been divided into 3/4 2.2.1 and 3/4 2.2.2 to reflect the option of using either incore or excore detectors to monitor L.HR at full power.
Figure 3.2-3 has been revised and re-numbered 3.2-3a and Figure 3.2-3b has been added. The above changes reflect the use of excore detectors to monitor LHR for short periods of time at full power and also an increase in radial peaking.
(4) Figure 3.2-3c has been added to reflect an increase in Cycle 5 radial peakingandtoseparatetheF[andFxy values.
We find these changes acceptable.
. -7.5 Implementation of BASSS The Better Axial Shape Selection System (BASSS) uses the fixed rhodium incore detector system' to monitor the Departure from Nuclear Boiling-Limiting Condition for Operation (DNB-LCO) rather than the excore de-tectors. Control Element Assembly (CEA) position and core average axial shape index are monitored to provide an alarm on power when the DNB-LCO is exceeded. Specifically, the BASSS determines allowable power level from knowledge of CEA position, total integrated radial peaking factor, and core average axial shape index using the on-line computer code PSINCA. The BASSS provides an alarm on power if the measured power level exceeds this allowable level. The BASSS method- "
ology was previously approved for use at Calvert Cliffs.
The following TS changes involve impelmentation of BASSS:
(1) Figure 3.1-2, the power dependent insertion limit (PDIL) is
' modified to indicate the BASSS operating region and reflect increased available SCRAM worth.
(2) TS 3.2.4, "DNB Parameters" and Table 3.2-1 are modified to require the axial shape index, core power, to be maintained via BASSS.
Table 3.2-1 also reflects a decrease in Pressurizer Pressure to increase operational flexibility.
(3) TS 3.2.3 is modified to incorporate BASSS and reflect an increase in radial peaking.
l We find the above changes acceptable.
7.6 Shutdown Margin The shutdown margin given in TS 3.1.1.1 is increased from 4.3 to 5.2%
AK/K. This change is also reflected in the Power, Dependent Insertion Limit (PDIL) in Figure 3.1-2.
The change in the shutdown margin results from the end of cycle (E'0C), hot zero power (HZP), steam line break analysis (see Section 3.1). Bases B 3/4 1.1.1 and 3/4 1.1.2 have been changed to be consistent with TS 3.1.1.1.
We find these changes acceptable.
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7.7 RTD Response Time
-The resistance temperature detector (RTD) response time has been 'in-creased from 1 8.0 to 1 12.0 seconds.
These response times are reflected in the following entries in Table 3.3-2, " Reactor Protective Instru-mentation Response Times":
(1) Power Level-High (2). Axial Flux Offset (3) Thermal Margin / Low Pressure
/
We find these changes acceptable as reflected in Cycle 5 analysis.
7.8 Fuel Enrichment The maximum enrichment for. reload fuel, specified in TS 5.3.1, " Fuel Assemblies", has been increased from 3.7 to 4.1 weight percent U 35-2 We find this change acceptable.
7.9 Pressure Transmitters The licensee has previously indicated that certain pressure transmitters located inside containment would be replaced by environmentally qualified transmitters, manufactured by Barton, in order to satisfy NRC concerns on environmental qualifications of electrical equipment.
As a result of operational problems with these transmitters, experienced with their use at Calvert Cliffs Unit 1 the licensee has decided not to replace the existing t'ransmitters with the units manufactured by Barton. The following TS changes are requested, to reflect the safety analyses which assumed installation of the Barton transmitters:
(1) TS Table 2.2-1, " Reactor Protective Instrumentation Setpoint Limits" has been changed.
The Steam Generator Pressure-Low Trip Setpoint has been changed from 1 570 to 1 685 psia to reflect the uncertainty associated with Barton pressure transmitters during the Main Steam Line Break Event.
In addition, the note (2) in Table 2.2-1, relating to the Steam Generator-Low Trip Bypass has been we changed from bypass below 685 to bypass.below 785 psia to reflect the change in the Trip Setpoint.
This same change was incorporated into Table 3.3-l', " Reactor Protective Instrumentation" and Table 3.3-3, " Engineered Safety Feature Actuation System Instrumentation".
(2)
TS Bases B.2.2.1 is revised to describe the change in the Steam Generator Pressure-Low Trip Setpoint, addressed above.
In addition, the low pressure for the TM/LP Trip Setpoint has been changed from 1750 to 1875 phia to reflect the uncertainty associated with the Barton pressure transmitters during a LOCA.
(3) TS Table 3.3-4, " Engineered Safety Feature Actuation System Instru-mentation Trip Values" has been changed. The-SIAS Pressurizer Pressure-Low Trip Setpoint has been changed from > 1578 to > 1725 psia to reflect the uncertainty associated with the Barton trans-mitters during a LOCA.
In addition, Table 3.3-3, the SIAS Pressure-Low Bypass, has been changed from < 1700 to < 1800 psia to reflect the change in the actuation setpoint.
o (4) TS Table 3.3-4, the SGIS Setpoint, has been changed from > 570 to > 685 psia to reflect the uncertainty associated with Barton pressure transmitters during the Main Steam Line Break Event.
The instrument; uncertainties associated with the existing Fisher-Porter pressure transmitters are bounded by the analysis. We find the above changes to the TS to be acceptable and consistent with the Cycle 5 analyses.
7.10 Plant Modifications During the current Unit 2 outage, a number of plant modifications have been ccmpleted, some of which require changes to the TS.
Modifications to the auxiliary feedwater system have been completed which requires that a special test exception to the TS be established.
t Accordingly, a special test exception is proposed for TS 3.7.1.2.
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"AuxiliaryNeedwaterSystem". This exception would allow the auxiliary
- feedwater system (AFW) to be inoperable for a' single period, following startup from the Cycle 5 refueling, of up to 30 days at power levels up to.557. power, except with regard to the capability to manually in-itiate flow to either steam generator with the steam driven AFW pumps.
This proposed TS change would allow aux 11 hey feedwater system to be operated in its original design configuration as presently described in-the Calvert Cliffs FSAR.
Two steam driven trains, capable of being manually initiated from the control room, have more than adequate capacity to provide a minimum flowrate of 700 gpm at a Total Dynamic Head of 2490 ft, to the entrance of the Steam Generators.
A capacity of 450 gpm is sufficient to ensure that adequate AFW flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less 'than 300 F, at which 0
time, the shutdown cooling system may be placed on the line.
Accordingly, we find this change acceptable.
In addition to the above test exception, the following setpoints, or required values, must be established for AFW startup testing:
(1) TS Table 3.3-5, " Engineered Safety Features Response Times",
states that the manual initiation of the AFW actuation channels need not be response time tested.
In addition, a response time for the motor driven AFW pump, for starting ~on low steam generator level, is iridorporated in Table 3.3-5.
(2) TS Table 3.3-3, " Engineered Safety Feature Actuation System Instru-mentation" and Table 3.3-4, " Engineered Safety Actuation System Instrumentation Trip Values" and TS Table 4.3-2, " Engineered Safety Feature Actuation System Instrumentation Surveillance" are revised.
l These revisions result from modifications to the AFW actuation system, are consistent with the assumptions contained in the safety analysis and are therefore acceptable.
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Although these instruments would not be physically removed from their -installed locations, they would no longer be the subject of TS requirements. These instruments include:
(1)
Power Range Nuclear Flux Monitor - This instrument was judged to be unnecessary for post-trip reactor monitoring.
The funca tion of post-accident flux monitoring is adequately performed by the Wide Range Logarithmic Neutron Flux Monitor, which covers the power range flux indication, and is described in TS Tables 3.3-10 and 4.3-10.
(2) Reactor Coolant Total Flow - Reactor coolant system total flow is considered a non-essential channel of instrumentation in the post-trip condition because of the availability of reactor coolant. system (RCS) temperature indication, RCS e
subcooled margin and Reactor Coolant pumps status.
In the absence of adequate core flow, RCS temperature and subcooled margin are indirect indicators of core flow conditions and provide adequate disp?ay to ensure appropriate actions are initiated by operations personnel to recover from any abnormal conditions existing in the post-trip condition.
Based upon conversations with representatives of BG&E, 'we understand that neither the Total Reactor Coolant Flow nor the Power Range Nuclear Flux Monitor is referenced in the emergency operating procedures. Accordingly, it is appropriate to delete these monitors from TS Table 3.3-10 and 4.3-10.
8.0 Conclusions The environmental considerations presented in Section 8.1 are based upon the consideration of Cycle 5 operation of Ca1 vert Cliffs Unit 2 and, specifically, the effects of the extended fuel cycle as
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presented in Section 6.0.
The safety conclusions presented in Section 8.2 are based upon th NRC evaluation of accidents and transients. presented in Sec, tion 5.0 and the proposed TS changes presented in Section 7.0.
8.1 Environmental Considerations We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any sign'ificant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative delcaration and environ-mental impact appraisal-need not be prepared in connection with the e
issuance of this amendment.
8.2 Safety Conclusions We have conclude, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, ahd (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: January 10, 1983 Principal Contributors:
L. Kopp H. Balukjian D. Powers J. Vogelwede A. Gill J. Mitchell D. Jaffe
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l 9.0 References 1.-
Letter from A. E. Lundvall, Jr. to R. A. Clark, Calvert Cliffs.
Nuclear Power Plant Unit 2, Docket No. 50-318 Amendment to Operating License DPR-69 Fifth Cycle License Application dated October 15, 1982.
2.
A. E. Lundvall, Jr. (BG&E) letter to R. A. Clark (NRC) on " Sixth Cycle License Applic6 tion," Docket 50-317, dated February 17, 1982.
3.
A. E. Lundvall, Jr. (BG&E) letter to R. A. Clark (NRC) on " Supplement 1 to Sixth Cycle License Application," Docket 50-317, dated April 29, 1982.
4.
D. H. Jaffe (NRC) letter to A. E. Lundvall, Jr. (BG&E) on " Unit 1
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Cycle 6 License Approval. (Amendment No. 71 to DPR-53 and SER),"
dated June 24, 1982.
5.
Licensee Event Report No. 82-030/0lT-0, Arkansas Nuclear One -
Unit 2, Docket 50-368, October 6,1982.
6.
" Improvement to Fuel Evaluation' Model," Combustion Engineering Topical Report CEN-161(B)-P, July 1981.
7.
" Fuel Evaluation Model," Con.bustion Engineering Topical Report CENPD-139-P-A, July 1974.
8.
" Partial Response to'NRC Questions on CEN-161(B)-P, Improvements to Fuel Evaluation Mcdel," Combustion Engineering Topical Report CEN-193(B)-P, Supplement 2-P, March 21, 1982.
9.
C. E. Beyer et al.. "GAPCON-THERMAL-2: A Computer Program for Calculating the Thermal Behavior of an Oxide Fuel Rod," Battelle Pacific Northwest Laboratories Report BNWL-1898, November 1975.
~
10.
A. E. Lundvall (BG&E) letter to R. A. Clark (NRC) dated May 19, 1982.
11.
R. O. Meyer et al, " Fission Gas Release from Fuel at High Burnup,"
U.S. Nuclear Regulatory Commission Report NUREG-0418, March 1978.
12.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment Nos. 47 und 30 to Facility Operating License Nos. DPR-53 and DPR-69 Relating to Modification of the Spent Fuel Pool Calvert Cliffs Nuclear Power Plant Unit Nos.1 and 2.
- 13. (Deleted)
- 14.. CENPD-153-P, Revision 1, '.' Evaluation of Uncertainty in the Nuclear O
Power Peaking Measured by the Self-Powered Fixed In-Core Detector System" dated May 1980.
i i
15.
CENPD-206-P, " TORC Code, Verification and Simplified Modeling Methods" dated January 1977.
16.
Letter from P.' W. Kruse to W. J. Lippold, " Responses to First Round Questions on the SCU Program:
CETOP-D Code Structure and Modeling Methods, (CEN-124(B)-P, Part 2" dated tay 1981 and letter, P. W. Kruse to W. J. Lippold (above document), BGE-9676-576, May 1, 1981.
17.
CENPD 161-P, " TORC Code, A Computer Code for Determining the Thermal Margin of a P.eactor Core" dated July 1975.
18.
CENPD-162-P-A (Proprietary) and CENPD-162-A (nonproprietary),
" Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids Part 1 Uniform Axial Power Distribution" dated April 1975.
19.
CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 2" dated January 1980.
20.
Letter from A. E. Lundvall, Jr. (BG&E) to R. A. Clark (NRC) dated December 17, 1982.
21.
Letter from A. E. Lundvall, Jr. fPME) to R. A. Clark (NRC),
" Response to NRC Questions on Cytic 5 License Anplication" dated December 6, 1982.
22.
CEN-83(B)-P, "Calvert Cliffs Unit 1 Reactor Operation with Modified CEA Guide Tubes" dated February 8, 1978 and letter from A. E.
Lundvall, Jr. to V. Stello, Jr., " Reactor Operation with Modified CEA Guide Tubes" dated February 17, 1978.
23.
Letter from D. F. Ross and D. G. Eisenhut (NRC) to D. B. Vassallo and K. R. Goller (NRC), ' Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing in Thermal Margin Calculation for Light Water Reactors" dated February 16, 1977.
24.
CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 1" dated December 1979.
25.
CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 3" dated March 1980.
26.
CENPD-190-A, "CEA Ejection Analysis," January 1976.
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