ML20062F192

Safety Eval Supporting Amend 43 to Facil Oper Lic DPR-40. Concludes Public Will Not Be Endangered

ML20062F192
Person / Time
Site:	Fort Calhoun
Issue date:	12/05/1978
From:	Office of Nuclear Reactor Regulation
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ML20062F178	List: ML20062F172 ML20062F186 ML20062F192 ML20062F200
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NUDOCS 7812150117
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/  % UNITEO STATES

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,. ; NUCLEAR REGULATORY COMMISSION W ASHINGTON, 0, C. 20555 s,<,, *>.. . L*n.*/

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i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 43 TO FACILITY OPERATING LICENSE NO. OPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1

DOCKET NO. 50-285 Introduction By applications dated Parch 22, 1978 and August 7,1978, as

'1 revised by letters dated October 31, 1978 and November 7 and 27, 1978,

Omaha Public Power District (OPPD or the licensee) requested a revision of the Technical Specifications to allow operation of Fort Calhoun following Cycle 5 fuel reload and to delete i Technical Specificaticns which permitted partial insertion of control element assemblies (CEAs).

Tvalua tion 1.0 Core Characteristics The reload for Cycle 5 operation will consist of the discharge j of 44 burnt fuel assemblies and the introduction of 44 fresh Batch G fuel assemblies. These assemblies are of the same i design as the Batch F fuel assemblies which were approved for use in Cycle 4 No fuel assemolies in the Fort Calhoun core will contain fixed burnable poison reds.

Design calculations for Cycle 5 were based on an assumed Cycle 4 average fuel exposure of between 7900 and 8200 7812150111

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Megawatt days per metric tonne of uranium metal (MWD /MTU).

j Actual Cycle 4 fuel exposure was 8135 MWD /MTU and hence is within the design window.

i Clad creep collapse was computed by Combustion Engineering (CE) for OPPD using the CEPAN code which has been approved by the NRC. The licensee has predicted that clad collapse I

will not occur for individual fuel assemblies which incur an accumulated assembly average exposure of 33,100 effective I

full power hours (EFPH). The licensse has also predicted j that the maximum exposure of any 'sel assembly in the core at the end of Cycle 5 will be 30,170 EFPH. Based on our l review of the licensee's evaluation, we find that,there

, exists ample margin to predicted clad collapse and use of i

fuel densification augumentation factors based on non-collapsed clad is justified.

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Ceparture from nucleate boiling ratio (DNBR) analyses were performed by CE fer the licensee using methods used and approved

for Cycle 4 with the exception of statistical methods of I

i combining uncertainties. This issue is discussed in detail later in this evaluation.

t Core lifetime for Cycle 5 is predicted by the licensee to be 10,000 MWD /MTU as compared to tne 8100 M'dD/MTU incurred in Cycle 4 The longer cycle is to be achieved by loading i

a full third of the core (44 of 133 fuel assemblies) with i fresh fuel. This loading is consistent with the Fort Calhoun long term fuel management. Cycle 4 was an intentionally

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short cycle which was achieved by loading only 36 fresh fuel assemblies. The use of more fresh fuel assemblies in Cycle 5 will result in higher predicted radial peaking factors at beginning of Cycle 5 (a feature of the design and not a necessary consequence) and more severe c~.ial power distributions at end of Cycle 5 (a natural consequence of the greater mismatch of the axial burnup distribution of new and previously loaded fuel assemblies). The larger peaks and more severe power cistributions have been properly incor-porated in the safety analyses. The execre linear heat rate (LHR) limiting condition for operation (LCO) axial shape index (ASI) tent, Technical Specification Figure 2-6, was recomputed to account for these phenomena.

Predicted core kinetics parameters exhibit small changes from the Cycle 4 values (Reference 1, Table 3-1). The variation of all the parameters is within the normal range of cycle to cycle variation and no change represents a safety concern. The variations have been properly accounted for in the safety analysis for Cycle 5.

Predicted control element assembly reactivity worths exhibit small changes frcm the Cycle 4 values (Reference 1, Table 3-2).

The variation of the reactivity worths is with1n the normal cycle to cycle variation. There is accle predicted reactivity worth for adequate control and shutdown.

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2.1 Procosed Technical Specification Changes (1) Refueling Baron Concentration, which was previously specified as corresponding to at least a 5% shutdown margin, would be quantified as at least 1700 Parts Per 111111on (PPfi) boren. [TS Definitions, TS 4.4.2]

(2) The upper limit of the measured va?ue of the total T

unrodded planar radial peak, F yy, the LHR LCO, would be increased from 1.57 to 1.62. [TS 1.1, TS 2.10.4(2)]

(3) The upper limit of the measured value of the unrodded T

integrated total radial peak, F g, the CNBR LCO, would be increased from 1.50 to 1.57.

[TS 1.1 TS 2.10.4(2)]

(4) The Baron concentration in the Safety injection and Refueling Water (SIRW) tank would be reduced frca 1900 PPM to 1700 PPM.

[TS 2.2 Basis (2), TS 2.3(1), TS 2.3.(3), TS 2.8 Basis, TS 2.14(5)]

(5) The required shutdown margin for hot shutdown, hot standby, or power operation would be increased from 3.4* 4k/k to 3,75 ak/k.

[TS 2.10.2(1), TS 2.10.2 Basis]

(6) The required shutdown margin for cold shutdown would be increased from 1.05 ak/k to 2.0% ak/k. [75 2.10.2(2), TS 2.10.2 Basis]

(7) The measurerent uncertainty to be applied to LHR naasurements would be decreased from 10.0% to 5.8%. [TS 2.10.4(1)(2)]

(8) The LHR uncertainty factor due to fuel densification would be decreased from 0.7% to 0.2%. [TS2.10.4(1)4]

(9) A new requirement on excore men' torina would be imposed which requires that the CEA's be withdrawn to beyond their Long Term Insertion Limit (LTIL) when the LHR LCO is monitored by excore detectors. [TS2.10.4(c)]

(10) The Excore LHR LC0 ASI Tent Scaling Equation would be changed.

[TS2.10.4(c)]

(11) Figure 2-6, the Excore LHR LCO ASI Tent, would be modified.

[TS Figure 2-6]

(12) Figure 2-8, the Flux Peaking Augmentation Factors (i.e.,

Power Spiking due to fuel densification and the consequent formation of gaps in the fuel pellet colur.n), would be modified. [TS Figure 2-8]

(13) Figure 2-9, the F I R and FT xy vs Core Power Limit would be modified.

[TS Figure 2-9]

(14) Section 6.4, which incoses limits on CEA insertion, would be deleted. [TS6.4]

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2.2 Evaluation of Proposed TS Changes (1) & (4) The SIRW tank boron concentration of 1900 PPM was detemined in Cycle 2 usfr.g conservative methods. For Cycle 5, OPPD and CE performed low temperature shutdown calculations in more detail than had been done in Cycle 2. The result of this study showed that 1700 PPM boron is adequate to assure at least 54 shutdown margin.

(2) The water hole peaking penalty of a.6%,as discussed in Section 4, b

has been applied to the INCA POR pin / box power ratios. The result is that for a given set of measured incore activations, the I.NCA computedpowerpeak,Ffy, will be 4.6% higher using the Cycle 5 methodology than it would have been using in the Cycle 4 methodology.

WiththisincreaseitisdifficulttocontroltotheoldTSFfy of 1.57, and the TS limit has been raised to 1.62. This change in itself has no safety significance in that the LC0 value of F{y is an input, not an output, of the safety analysis, and its value requires no justification with respect to measurement uncertainties or comoutational conservatism credits.

(3) For the reasons given in (2) above, the Ff LCO has been increased from 1.50 to 1.57, and as explained above,the value of Ff requires no justification with res::ect to measurement uncertainties or computational conservatism credits.

(4) Treated above with (1)

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(5) The increase in required shutdown margin from 3.4% sk/k to 3.7% ak/k is to assure that no return to criticality occurs during i

the full length' CEA drop transient from 100% power as explained

fa Section 3.

(6) The increase in cold shutdown required shutdown margin from 1%

1 ak/k to 2% ak/k is me;e to assure that there is sufficient time for operator action during the limiting boron dilution incident as explained in Section 3.

(7) The flRC accepts as adequate a 10.0% uncertainty in the linear heat rate measurement, and the use of 5.8% uncertainty represents a 4.2% deficit. As explained in Section 4, OPPD has

• i demonstrated a credit of 5% to offset the 4.2% deficit, and on i

this basis we find the use of the assumed 5.8% uncertainty acceptable.

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i (8) The linear heat rate uncertainty factor due to axial fuel densification and thermal expansion is reduced from 0.7% to 0.2%.

Since this reload removed the remaining less stable fuel, all fuel used in the Cycle 5 core is more stable with respect to

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fuel densification. In view of this and based on our evalu-

] ation of the licensee's analysis, we find that a reduction in j this uncertainty factor to 0.2% is justified and acceptable.

(9) The excore LHR LCO ASI tent, TS Figure 2-6, is generated assuming whatever CEA insertion limit is specified for monitoring on excores in the TS. The more restrictive the CEA insertion limit, the less restrictive the Excore LHR LCO ASI Tent will be. With the TS change the CEA insertion limit during periods of mr,'.itoring on excores

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is now the more restrictive long term insertion limit (LTIL),

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rather than the power dependent insertion limit (PDIL) which was the limit prior to the introduction of this TS change.

i Therefore, the less restrictive excore LHR LCO ASI trip-tent l 1s acceptable. The Cycle 5 insertion limits in combination with i the corresponding LHR LCO ASI trip tent provides protection which is substantially equivalent to that provided by the Cycle 4 in-sertion limit and corresponding tent.

(10-11) The loss of coolant accident (LOCA) analysis for Cycle 4 was performed assuming an initial peak linear heat rate of 14.7 kw/ft. The plant must be operated (limiting condition for operation or LCO) such that this value is not exceeded.

Cycle 4 operation was bah on linear heat rate vs axial shape 4

index tents generated assuming a 15.5 kw/ft LCO and subsequent j scaling to 14.7 kw/ft. The p ; posed TS are a consistent package that includes the linear heat rate vs axial shape index tents and i a scaling equation based on an LCO of the more conservative 14.7 kw/ft, and also includes the water hole peaking penalty and are there-j fore acceptable.

(12) The existing TS Figure 2-8 is being updated to be consistent with the augmentation factors discussed on page C-13 of Reference 1.

The calculational model used was approved for the Cycle 4 design, is applicable for Cycle 5, and thus needs no further review for application to Cycle 5. Because of the higher density fuel used in the Cycle 5 core, the aug:rentation factors ray be reduced from -!

the Cycle 4 values without loss of safety margin and are acceptable. ,

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9-(13) The explanation for the change in the two curves in TS Figure 2-9 is the same as the explanation given in (2) and (3) above.

(14) TS 6.4 is deleted because no guide tube wear has been observed at Fort Calhoun as reported by OPPD (References 5 and 6) and reviewed by the NRC (Reference 7). TS 6.4 was originally -

imposed at Fort Calhoun because of CEA guide tube wear experienced at other plants. Based on inspections, no guide tube wear problems exist at Fort Calhoun and, therefore, this TS is no longer necessary.

3.0 Transients and Accidents The transients and accidents considered in the safety analysis are shown in Table 1 and include all occurrences which are considered for a reload.

For all transients and accidents not reanalyzed, the reference analysis parameter values bound the Cycle 5 parameter values. Hence, for these transients and accidents during Cycle 5, the reference analysis is conservative, applicable and acceptable.

For all trarsients and accidents reanaly:ed with Cycle 5 parameter values, the me hods used were accepted for the analysis of previous cycles are applicable for Cycle 5 and hence are acceptable for use in Cycle 5.

In all cases,1f the result of a new analysis for Cycle 5 was found more limiting tnan the previous analysis, the appropriate TS limits were altered so that there is no loss of safety margin, i

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• TABLE 1 FT. CALHOUN, CYCLE 5 INCIDENTS CONSIDERED IN TRANSIENT AND ACCIDENT ANALYSIS Analysis Status Anticipated Operational Occurrences for which the RPS Assures no Violation of SAFDLs: .

Control Element Assembly Withdrawal Reanalyzed Boron Dilution Reanalyzed .

Startup of an Inactive Reactor Coolant Pump Not Reanalyzed Excess Load Not Reanalyzed Loss of Load Not Reanalyzed Loss of Feedwater Flow Not Reanalyzed Excess Heat Removal due to Feedwater Malfunction Not Reanalyzed Reactor Coolant System Depressurization Not Reanalyzed Loss of Coolant F1cw l Not Reanalyzed Loss of AC Power . Not Reanalyzed Anticipated Operational Occurrences which are Dependent on Initial Overpcwer Margin for Protection Against Violation of SAFDLs:

Loss of Coolant Flcw I Not Reanalyzed Loss of AC Power Not Reanalyzed full Length CEA Drep Reanalyzed Part length CEA Drop Not Reanalyzed Part length CEA Malpesitioning Not Reanalyzed Transients Resulting frca Malfunction of One Not Reanalyzed Steam Generator Postulated Accidents:

CEA Ejection Reanalyzed Steam Line Rupture Reanalyzed Steam Generator Tube Rupture Not Reanalyzed Seized Rotor Reanalyzed I Requires Low Flow Trip.

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- -11 3.1 CEA Withdrawal Analysis This analysis would affect the Temperature Margin / Low Pressure 4

(TM/LP).setpoint. The result of the analysis demonstrated that the Cycle 4 TM/LP setpoints are adequate, and hence we find that I no TS change is required.

3.2 Boron Dilution Incident 3

The TS requires that the operator should have at least 15 minutes

]' to respond to a Boron Dilution Incident. The previous TS 2.10.2(2) required a shutdown margin for cold shutdown of 1% t.k/k. The-Baron Oilution Incident Analysis indicated that with 1% ' shutdown margin the operator would have 17 minutes to respond to the Baron J

Dilution Incident. OPPD felt that this was too close to the 15 minute limit, and raised the TS margin to 2". to give the operator more time to respond to a Boron Oflution Incident. Since this change provides additional time over and above the 15 minute response time <

set out in the TS, to deal with the Baron Dilution Incident, we find

/ r, it acceptable.

3.3 Full Length CEA Oroo Incident .

This incident was reanalyzed, and it was found that the present

TS limits provide adequate ratection against this incident. Hence 1 f we find that no TS change is required. f 3.4 Steam Line Ruoture Incident '

l In the analysis of the steam line rupture, it was found that if

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the present TS were to remain in effect, there would be a return  ;

i to criticality in the event of a steam line rupture. To prevent I s

this return to criticality,the required shutdown margin during

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,,ower operation in TS 2.10.2(1) was increased from 3.4% ak/k to i 3.7% ak/k. Based on our review, we find tnat requiring a shutdown '

4 l margin of 3.7% ak/k will preclude a return to criticality in the i event of the limiting steam line rupture incident and is, therefore. I aCCeota ble.

12 3.5 Siezed Rotor Analysis The Slezed Rotor Incident was reanalyzed. Based on that reanalysis, we find that the present TS limits provide adequate protection against this incident.

3.6 CEA Ejection Incident The CEA Ejection Incident was reanalyzed. Based on that reanalysis, we find that the present TS limits provide adequate protection against this incident.

3.7 Emergency Core Cooling Systen (ECCS) Analysis The ECCS parameters for Cycle 4 were found to bound the corresponding parameters for Cycle 5, and, hence, we find that the ECCS analysis for Cycle 4 is also applicable to Cycle 5.

3.8 Conclusions We have reviewed the licensee's transient and accident analyses and, based on that review, we have concluded that they are acceptable and adequately support Fort Calhoun Cycle 5 cpc ation with the TS changes discussel above.

4.0 Uncertainty in Nuclear Power Peaking Factors This section describes the current nuclear peaking factor uncertainties used in the licensee's safety calculations. It is concluded that, although certain values used in the safety calculations are considered ncnconservative, there exists of' setting conser-vatisms to justify Cycle 5 operation.

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i 4.1 Documentation of Uncertainties Reference 8, which is still under review by the NRC staff, documents

the assumed uncertainties in Ff and F of 5.1% and 5.8%, respectively, and Reference 9 documents the 4.6% water hole power peaking bias.

At the present stage of review, we have concluded that the 5.1% and

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j 5.8% are nonconsarvative. In addition, the 4.6% should have an

! uncertainty associated with it which the licensee has not factored i

into its analysis. However, whatever uncertainty is inherent in j the 4.6% could logically be applied to the 5.1% and 5.8%, and our present position is to accept the 4.6% fully and assign any uncertainties in the water hole peaking to the uncertainties in FT7and F , which are j the factors affected by the water hole peaking.

4.2 Uncertainties in Fh and Fd Used in the Safety Analysis In Reference 9, which documents the water hole peaking bias of 4.6%,

certain computational conservatisms are cited which could act as credits to mitigate the effects of the additional 4.6% peaking.

Based on these credits, we agreed with the CE licensees at a December 16, 1977 meeting that an additional penalty of 2.8% rather than 4.6% could be justified. However, with the newly adopted use

! of thermal margin / low pressure methodology and statistical methods, I this reduction in penalty can no longer be justified generically for i i l all CE licensees, and credits must be considered on a-case by case j casis.

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-i4-The 4,6% bias has been applied to the computed peak pin powers which are then used in the remainder of the safety analysis. This method accounts for the 4.6% bias completely, and it need not be incorpor-1 ated in the uncertainty treatment. OPPD used uncertainties of 5.1% and 5.8% for Ff and F respectively.

i l 4.3 Status of NRC Staff Review of CE Uncertainties i

We have sutmitted an extensive list of questions concerning the justification for the 5.1% and 5.8%.00) These questions were answered in part in Reference 11. However, a number of responses were not-complete, and many of the questions of Reference 10 have not yet been addressed. From the available data, we conclude that uncertainties of 8% for FTr and 10% for F can be justified, and in Reference 2, i

Question 2 OPPD has demonstrated sufficient credit to offset the difference i between the 8% and 10% and the 5.1% and 5.8% used in the safety i

l analysis, i.

4.4 Conservatisms in the Safety Analysis to Offset Uncertainties As stated above, in lieu of justifying the 5.1% and 5.8%, OPPD has cited a number of known conservatisms in the safety analysis methodology for which they do not take credit. These are enumerated j in the following sections.

! 4.4.1 DNBR Conservatism i

j ihere are three areas in which conservatisms are known to exist in I

• he DNBR safety analysis. In the first two of these, the degree i

l of conservatism is known, since comparisons with a more exact model

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s have been performed. In the latter, the degree of conservatism has'  ;

not been evaluated; however, it is clear that it is conservative.

These three areas of conservatism are:

Conservatism in DNBR Limits due to Thermal-Hydraulic Model I i The Thermal-Hydraulic analysis for Cycle 5 was done with the f

l COSMO-INTHERMIC computer codes and the W3 DNBR correlation. In Reference 9 the licensee states, and we agree, that COSMO-INTHERMIC'-

W3 predicts results which are conservative by at least 3% when com-pared with the momaccurate TORC /CE I Thermal-Hydraulic model. This  !

3% credit is applicable to both the DNBR limiting safety system t

settings (LSSS) and DNBR LCO.

i Conservatism in DNB Limits Due to Statistical Combinations In the ONB limit analysis, the assumed uncertainties in various i

measured parameters are not combined in a single equation, but l f

are factored into functional relationships as biases at various i points in the analysis (I2) . This biasing of functional relation-ships throughout the analysis is equivalent to adding the absolute [

t i power uncertainties equivalent to the uncertainties in the various j

measured parameters and applying the total power uncertainty to  ;

the best estimate calculation. The specific uncertainties along

with their equivalent ower uncertainties are given below.  ;

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j ASI 0.L3 ASIU > 2.2%

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, Pressur. 22 PSI > o,g; I Temperature 2F 1.0.9%

4T i j Flow 2. 5.0%

Power 5% (LSSS) 2.3.5%

2% (LCO) 3.l.4%

In the OPPD analysis, the equivalent sum of these uncertainties is 12.4% for LSSS,10.3% for LCO. These measurement uncertainties are statistically independent, and, hence, the proper method for com-bining them is Root Sum Square (RSS). The RSS combination yields 6.6% for LSSS, 5.8% for LCO, giving a net conservatism in the analysis of 5.8% for LSSS, 4.5% for LCO.

Conservatism in Axial Flux Shaces In the multiplicity of the axial flux shapes used in the safety analysis, as computed by t he QUIX code, more severe shapes are calculated than are expected to occur during actual operation (I2) . This results in conservatism in the safety analysis. We have not allowed credit for this conservatism because it has not been quantified.

4.4.2 Conservatism in LHR LSSS The TS Axial Shape Index (ASI) trip tent is constructed to lie within the safety analysis LSSS tent. The construction has been done so that the TS ASI trip tent contains at least 5% conservatism compared with the safety analysis tent. A TS ASI trip tent identical to the safety analysis ASI LSSStent would be acceptable to us. Therefore, we find it acceptable to use the 5% margin to offset nonconservatisms in the LHR LSSS.

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4.4.3 Conservatism in LHR LCO

! The safety analysis was performed assuming that the Cycle 5 fuel can withstand s1LHR of 14.7 kw/ft, whereas because of the high density fuel used in Cycle 5, the fuel can withstand anLHR of ,

j 16.0 kw/ft. This provides at least 5% credit in the LHR LCO.

NRC Staff Evaluation of Credits

! The total uncertainties and credits are as follows:

DNBR DNBR LHR LHR LSSS LCO LSSS LC0 i Uncertainty Assumed by OPPD in Safety Analysis 5.1% 5.1% 5.8% 5.8%

Total Available Credits 8.8% 7.5% 5.0% 5.0%

Total of Above Two 13.9% 12.6% 10.8% 10.8% i i

Uncertainty Approved by NRC 8.0% 8.0% 10.0% 10.0%

Since in all cases the uncertainty approved by the NRC is less i

1 than the sum of the uncertainty assumed in the safety analysis plus available credits, we find the uncertainties used in the OPPD safety

analysis acceptable. We also find that taking such credit does not '

l j significantly reduce safety margins. i 5.0 Environmental Consideration We have determined that the amendment does not authorize a change i

in effluent types or total amounts nor an increase in power level 1 and will not result in any significant environmental impact. Having l -

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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 5 51.5(d)(4), that an t

environmental impact statement, or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.

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, 6.0 Conclusion We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the l

, amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public l will not be endangered by operation in the proposed manner, and (3)

I 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.

Dated: December 5,1978 1

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7.0 References

1. Omaha Public Power District, Fort Calhoun Station Unit No.1, Cycle 5 Core Reload Application, August 1978

2. Letter from T. E. Short, CPPD to R. W. Reid, NRC, October 31, 1978

, addressing round 1 of formal NRC questions. ,

3. Letter from T. E. Short, OPPD, to R. W. Reid, NRC, November 7, 1978 addressing round 2 of formal NRC questions.

4 Letter from T. E. Short, OPPD, to R. W. Reid, NRC, addressing questions posed during miscellaneous phone conversations.

5. Letter from T. E. Short, OPPD, to George E. Lear, NRC, February 10, i 1978.

6. Letter from LeBoeut, Lamb, Lieby & MacRae, Attorneys for OPPD to Edison G. Case, NRC, March 22, 1978.

7. Letter from R. W. Reid, NRC, to T. E. Short, OPPD, August 7,1978.

l 8. CENPD-145, INCA: Method of Analyzing In-Core Detector Data in Power Reactor, Ober, Terney, and Marks, Combustion Engineering, April 1975.  :

9. CEN-85(B)-P, Solution to Increased Water Hole Peaking on Operating Reactors, Combustion Engineering, February 16, 1978.

10. Request for Additional Information, letter from R. W. Reid, NRC, j to A. E. Lundvall, Jr., 8G&E, April 4,1978.

11. Response to Request for Additional Information on CEN-85(B)-P, .

letter from A. E. Lundvall, Jr., BG&E, to R. W. Reid, NRC, April 28, 1978. i I

12. CENPD-199, CE Setpoint Methodology, Combustion Engineering, April 1976. ,

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