ML20199L729
| ML20199L729 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 01/30/1998 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20199L711 | List: |
| References | |
| NUDOCS 9802100018 | |
| Download: ML20199L729 (7) | |
Text
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The 11cen c amend ent is contingent on the limitation c' =0ritering of the long tern load factor to : ure it doc not exceed the a:Sumed value of 0.77 and that a reevaluation of hend of licen:0 fluence with
( b ENDF/B V! cros: tection: and updated uncertaintics wil' be performed to a: Cure that the value of the RTm willnotexceedthescreeningcriterionj onitoring, w / x, x / -
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Fire Protection Proaram Omaha Public Power District shall implement and' maintain in effect all 3rovisions of the approved Fire Protection Program as described in the hdated Safety Analysis Report for the facility and s' approved in the SERs dated February -14 anc August 23, 1978. November 17,1980. April 8.
and August 12.1982 -July 3. and November 5.1985. July 1.1986. December
- 20. 1988. Novenber 14. 1990. March 17, 1993, and January 14, 1994, subject to the following provision:
Omaha Dublic Power District may.make changes to the approvad Fire Protection Program without prior approval of the Commissior only 4 f those changes would not adversely affect the ability to ac'lieve aad maintain safe shutdown in the event of a fire, G
[G.f Additional Conditions The Additional Conditions contained in Appendix B as revised thraugh
' Amendment
, are hereby incorporated into this license.
Omaha Public r)wer District shall operate the. facility in. accordance with the
,lditional Conditiors.
4.
1his amended license is effective as of the date of issuance and shall expire fit mic;.ight on August 9, 2013.-
FOR THE ATOMIC ENERGY COMMISSION Original signed by:
A. Giambusso-l A. Giambusso. Deputy ector j
for Reactor Projcets Directorate of Licensing
)
Enclosures:
- 1. -Appendix A - Technical i
Specifications 7
2-Apper. dix B -
Additional Conditions Date of 1ssuance: August 9. 1973 Amendment No. 155.158.160, W j
9802100018 980130 PDR ADOCK 05000285 P
PDR.
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U.S. Nuclear Regulatory Comission LIC-98-0009 ATTACHMENT B
DISCUSSION, JUSTIFICATION, AND NO SIGNIFICANT HAZARDS ANALYSIS DISCUSSION AND JUSTIFICATION:
The 0maha Public Power Dit'.rict (OPPD) proposes to delete Section 3.E, License Term from Fort Calhoun Station (FCS) Unit 1 Operating License No. DPR-40.
The long term load factor described in Section 3.E is used for calculation of the RT,n value to ensure that the screening criteria for reactor vessel integrity are not exceeded. The previous fluence analysis performed by Combustion Engineering (ABB/CE) used a 0.77 load factor in conjunction with the ENDF/B-IV cross section-library.
As shown in Attachment C, Westinghouse Electric Corporation (W) has completed an ana!ysis (Westinghouse calculation SE-REA-95-003, Fast Neutron Fluence Evaluations for the Fort Calhoun Unit 1 Reactor Pressure Vessel, dated November 1995) to update the ABu/CE calculation.
-In the updated analysis, the long tena load factor was increased from 0.77 to-0.85 to reflect improvement in FCS Unit 1 operating efficiency.
The updated analysis also used the ENDF/B-VI cross section library with updated uncertainties as required by Operating License No. DPR-40, Section 3.E.
The neutron fluence calculations are carried out using forward and adjoint fornN1ations in r,e geometry of the two dimensional Discrete Ordinates Transport (00T) code.
The anisotropic scattering is treated with a P expansion of the scattering cross 3
section and the angular discretization is modeled with a S, order of quadrature.
The actual core power distribution and neutron source distributions from 14' cycles of operation (13.6 Effective Fuli Power Years) were utilized, which included the spectral changes due to plutonium accumulation.
The BUGLE-93 cross section library which is based on the data set of the Evaluated Nuclear Data-File /B-VI (ENDF/B-VI) was used. - The Westinghouse 00T code was benchmarked to the ENDF/B-VI cross sections using the Poolside Critical Assembly (PCA) simulator experiment at the Oak Ridge National Laboratory (ORNL), surveillance capsule and cavity dosimetry _ measurements.
The results of these fluence evaluations demonstrate that the best estimate fast neutron exposure of the pressure vessel can be determined with a la uncertainty of 113% for & (E>1.0MeV),119% for & (E>0.1MeV) and 114% for dpa.
These uncertainties.are within the *20% guidelines. contained in (Draft Regulatory Guide, DG-1053, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence).
The methodology used, as summarized above, is the same as the neutron fluence calculation section of WCAP-14040, Revisior,1, Methodology used to Develop Cold
' Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Curves (TAC #
M91749).
Application of the exposure methodology to the FCS reactor vessel indicates that at the conclusion of Cycle 14 the critical weld material (i.e.,
weld 3-410). had accumulated a maximum fast neutron fluence (E>1.0MeV) of 9.24E18 2
n/cm and ' had reached a corresponding RT,n value of 235.9'F based on the correlations provided in Regulatory Guide 1.99, Revision 2.
d
DISCUSSION AND JUSTIFICATION: (Continutd)
Based on thc use of low leakage fuel management as embodied in the design of FCS.
Operating Cycles 15 and 16, including actual power generation through Cycle 17 to date and projections' through the last operating-cycle in which Operating License DPR-40 expires, the critical weld material will have accrued a niaimum fast neutron fluence of 1.53E19 n/cm and will have reached a RT,33 value of I
2 approximately 269'F, using a long tena load factor of 0.85.
Westinghouse Calculation SE-REA-95-003 (attached) indicates that RT,35 is 265.8'F.
.The difference from the above noted value (269'F)is due to the recent change in limiting chemistry factor (revised during Generic Letter 92-01, Rev. 01, Supplement 1 evaluations) and recent actual plant-performance / generation.
4 In accordance witt,10 CFR 50.61, this assessment must be updated whenever thcre is a-significant change in projected values of RT,33 or upon request for a change j
in the expiration date of the facility. Thus, Section 3.E can be deleted from Operating-License No. OPR-40 based upon the analysis contained in Attachment C and the fact that Secticn 3.E is redundant to 10 CFR 50.61 requirements.
A
BASIS FOR N0 SIGNIFICANT HAZARDS CONSIDERAT30N8
- The proposed change does not involve a significant hazards consideration because -
operation of Fort Calhoun Station (FCS) Unit 1 in accordance with this. change would not:
(1)
Involve a significant increase in the probability or consequences of an accident previously evaluated.
The previously evaluated accidents affected by this change are limited to the pressurized thermal shock (PTS) events, Vessel embrittlement due to fast neutron associated damage to the limiting beltline region reactor vessel material, which for Fort Calhoun Station is the lower course axial welds, is a component in the PTS analysis.
The fast neutron, thermal-neutron ard dpa values of the FCS reactor vessel were recalculated using actual power history values for - Cycles 1 through 14 rather than conservative estimates, with the revised BUGLE-93 cross sections from the ENDF/B-VI cross section library to appropriately account for the iron atoms in the thermal shield and a methodology that the NRC has previously approved for neutron fluence calculations performed by Westinghouse.
The evaluation included data from the three surveillance capsules (W-225, W-265, and W-275) previously removed and analyzed.
The evaluation results indicate that the FCS reactor vessel is able to reach current licensed life without exceeding the 10 CFR 50.61 screening criteria for RTns of 270*F for limiting axial welds.
In accordance with 10 CFR 50.61, this assessment must be updated-whenever there is a significant change in projected values of RTns or upon request-for a change in.the expiration date of the facility.
Since these requirements are contained in -10 CFR 50.61, Section 3.E can be deleted from Operating License No. DPR-40 without resulting in a significant 4
increase in the probability or consequences of any accident previously evaluated.
- (2)
Create the possibility of a new or different kind of accident from any previously analyzed.
The proposed change does not physically alter the configuration of the plant and no new or different mode of operation is proposed.
Increasing the long term load factor from 0.77 to 0.85 m:ra accurately projects RTns by accounting for improvement in FCS operating cycle-efficiency.
Requirements for assessing and repgrting RTns are contained in 10 CFR 50.61
-and therefore, the proposed change does not crev.e the possibility of a new or different kind of accident from any previously analyzed.
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4 BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION: (Centinued)
(3)
Involve a significant reduction in a margin of safety.
The margin of-safety is defined by the draft regulatory guide DG-1053 for i
neutron fluence calculations which requires the methodology to be capable
- of providing best estimate fluence evaluations within 120 percent (la).
The_ analysisi shows that the applicable regulatory criteria are met and therefore, the proposed change does not involve a significant reduction in a_ margin of safety.
I Therefore. based on-the above, it is OPPD's position that this proposed amendment does not involve a significant hazards consideration as defined by 10 CFR 50.92 and the proposed change will not result in a condition which significantly alters the impact of the Station on the environment, i
Thus, the proposed change meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and pursuant to 10 CFR 51.22(b) i no environmental assessment need be prepared.
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U.S. Nuclear Regulatory Commission LIC-98-0009 ATTACHMENT C i
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