ML20058G937
| ML20058G937 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/03/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058G926 | List: |
| References | |
| NUDOCS 9312100185 | |
| Download: ML20058G937 (10) | |
Text
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p%9 k UNITED STATES
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(' S NUCLEAR REGULATORY COMMISSION ,
i f WASHINGTON D.C. 20556-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.158 TO FACILITY OPERATING LICENSE NO. DPR-40
- OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION. UNIT NO. 1 DOCKET NO. 50-285
1.0 INTRODUCTION
By letter of July 17, 1986, as supplemented by letters dated April 30, May 15, and December 21, 1987, May 18, 1989, December 16, 1991, March 17, May 15, July 6, 1992, and June 23, August 12, September 17, October 15, and ,
October 27, 1993, Omaha Public Power District (0 PPD) submitted a request for !
changes to the Fort Calhoun Station, Unit No. I license. The requested change would change the expiration date for the Fort Calhoun Station, Unit 1 Operating License from June 7, 2008 to August 9, 2013. The original work request performed using TAC No. M61976, was closed December 8, 1988.
The staff issued an Environmental Assessment (EA) dated August 3,1987 (52 FR 28767), as required by 10 CFR 51.21 and 51.22, in which it concluded that the August 1972 Final Environmental Statement for Fort Calhoun Station ,
remains valid and pursuant to 10 CFR 51.31 an environmental impact statement need not be prepared for this action.
The April 30, May 15, and December 21, 1987, May 18, 1989, December 16, 1991, March 17, May 15, July 6, 1992, and June 23, August 12, September 17, October 15, and October 27, 1993, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
2.0 EVALUATION Section 103.c of the Atomic Energy Act of 1954, as amended, provides that a i license is to be issuad for a specified period not exceeding 40 years. The i Code of Federal Reguh . ions in 10 CFR 50.51 specifies that each license will be issued for a fixed period of time, to be specified in the license, not to exceed 40 years from date of issuance. Also, 10 CFR 50.56 and 10 CFR 50.57 .
allow the issuance of an operating license pursuant to 10 CFR 50.51 after the -
construction of the facility has been substantially completed, in conformity' with the construction permit and when other provisions specified in 10 CFR v 9312100185 931203 5 DR ADDCK 0500 '
. . l 50.57 are met. The currently licensed term for Fort Calhoun Station is 40 years, commencing with the issuance of the construction permit on June 7, ,
1968. Accounting for the time that was required for plant construction, this represents an effective operating license term of less than 35 years.
Consistent with Section 103.c of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.51, 50.56 and 50.57 of the Commission's regulations, the licensee, by its application of July 17, 1986, seeks extension of the ,
operating license term from the date of operating license issuance, namely '
40 years from August 9, 1973. This action would extend the period of operation to the full 40 years provided by the Atomic Energy Act and the Code of Federal Regulations.
The impact of the additional period of operation on the general population and environment in the vicinity of the Fort Calhoun Station is addressed in the NRC staff's Environmental Assessment dated July 29, 1987. '
The licensee's request for extension of the operating license is based on the fact that a 40 year service life was considered during the design and '
construction of the plant. Although this does not mean that some components will not wear out during the plant lifetime, design features were incorporated which maximize the inspectability of structures, systems and equipment. The reactor coolant system components and support systems are analyzed for the integration effects of radiation damage and cyclic loadings (with added -
margin) which could reasonably be expected to occur in a 40-year lifetime.
Surveillance and maintenance practices which were implemented in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for Inservice Inspection and Inservice Testing of Pumps and Valves and the facility Technical Specifications (TS) provide assurance that any unexpectei degradation in plant equipment will be identified and corrected.
- These TS are part of the plant's operating license and have been approved by the NRC, as are all subsequent changes to the TS. The specific provisions and requirements for ASME Code testing are set forth in 10 CFR 50.55a.
2.1 Reactor Vessel The design of the reactor vessel and its internals considered the effects of 40 years of operation at full power and a comprehensive vessel material surveillance program is maintained in accordance with 10 CFR Part 50, Appendix H that ensures the fracture toughness requirements of Appendix G are met. As stated in the Updated Safety Analysis Report (USAR), reactor vessel surveillance capsules are provided for post-irradiation testing of Charry V-notch and tensile specimens.
The Pressurized Thermal Shock (PTS) rule,10 CFR 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," adopted l on July 23, 1985, establishes screening criteria that define a limiting level of embrittlement beyond which operation cannot c,ontinue without further plant-specific evaluation. Thescreeningcriteriaareggvenintermsofreference temperature, 0RT, The screening criteria are 270 f for plates and axial welds and 300 F Eo.r circumferential welds. The RT 1 is defined as the sum of :
(a) the unirradiated reference temperature, (b) th"e, margin to be added to '
1 cover uncertainties in the initial properties, copper and nickel contents, fluence, and calculation procedures, aad (c) the increase in RT,,, caused by irradiation. The amount of increase in RT is based on the amount of 1
neutron irradiation and the amount of copp,er, and nickel in the material. The ,
greater the amounts of copper, nickel and neutron fluence, the greater the increase in RT,,, for the material and the lower its fracture resistance.
The PTS rule was amended on May 15, 1991. The amended rule changed the method '
i of calculating embrittlement to the method recommended in Regulatory Guide (RG) 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials",
and requires licensees to consider the effect of reactor vessel operating The temperature and surveillance licensee performed resultsin this assessment ona the lettercalculated RT,5,value.
dated July 1992, which contained the licensee's response to Generic Letter 92-01, Revision 1,
" Reactor Vessel Structural Integrity,10 CFR 50.54(f)".
In addition to the PTS screening criterion, licensees are required by 10 CFR Part 50, Appendix G to maintain the Charpy upper-shelf energy (USE) at or above 50 ft-lb unless lower values will provide margins of safety against fracture equivalent to those required by Appendix G of the ASME Code. The percentage decrease in the Charpy USE is dependent upon the amount of copper in the material and the amount of neutron irradiation. The method recommended by the staff for determining the effect of neutron irradiation on the Charpy USE is documented in RG 1.99, Revision 2. In a letter dated October 27, 1993, the licensee provided information to demonstrate that the Fort Calhoun reactor vessel with Charpy USE less than 50 ft-lb would provide margins of safety against fracture equivalent to those required by Appendix G of the ASME Code.
2.1.1 Pressurized Thermal Shock (PTS) Evaluation The Fort Calhoun reactor vessel was fabricateel by Combustion Engineering. Its reactor vessel beltline consists of six axia!!y oriented welds, one circumferentially oriented weld and six plates. The axially oriented welds '
were fabricated using a tandem arc process and the circumferentially oriented '
weld was fabricated using a single arc process. Combustion Engineering ,
maintained adequate records to determine the heat number of all plates and weld wire used in the fabrication of the beltline welds and plates. The amount of copper and nickel in each plate was determined from a chemical analysis of each plate by the plate fabricator. These data indicate that the RT,3* values for the plates will be less than 200,F on August 2013, and will not be limiting.
The PTS rule,10 CFR 50.61, defines the weight percent of copper and nickel as ;
the best-estimate value, which would normally be the mean of the measured values for a plate or forging or for the weld samples made with the weld wire heat numbers that matches the critical vessel weld. The amount of copper and nickel in each weld was determined from measurements of weld deposits fabricated using the same heat number or combination of heats of weld wire as used in the fabrication of the beltline welds. Using the methodology i values, the beltline weld with documented in 10value the greatest RT,,, CFR on50.61 to calculate August 2013, is RT,,,he t 3-410 axial weld, which was
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.f fabricated using weld wire heat number 27204. The RTm value projected for.
this weld on August 2013, is 264.3 0F. The staff agrees with the licensees method of determining the amounts of copper and nickel in each beltline weld, l except for the axial weld fabricated with heat number 13253 weld wire. The l licensee determined the amount of copper and nickel in this weld by averaging !
the amounts of these elements in a fort Calhoun closure head weld, a Cook j surveillance weld and a Salem surveillance weld, which were fabricated using i heat number 13253 weld wire. The Fort Calhoun closure head weld and the Salem !
surveillance weld were fabricated using a single arc process. The Cook :I surveillance weld was fabricated using a tandem arc process. Since the l chemical analyses from single arc welds represent the composition of a single !
coil of weld wire and the chemical analysis from a tandem arc process represents the average composition of two coils of weld wire, a simple average [
of all values does not represent the best estimate of the chemical :
composition. The staff believes that the best estimate of the chemical s composition for this case is the weighted average, where single arc welds are !
given a weight of one and tandem arc welds are given a weight of two. A !
weighted average increase the amount of copper in the weld with heat number !'
13253 weld wire to 0.23% and has no significant effect upon the amount of nickel. The increase in the amount of copper will increase the amount of embrittlement. However, the amount of embrittlement will be less than the t amount for the axial weld fabricated using weld wire heat number 27204. Since I this value is less than the screening criterion in the PTS rule, PTS is not a I ccncern for the Fort Calhoun reactor vessel through August 2013. ;
2.1.1.1 Charov Upper-Shelf Enerav (USE) Evaluation l
t The unirradiated Charpy USE of each plate was determined from samples removed i from each plate. One plate was tested with both transversely and !
longitudinally oriented Charpy specimens and five plates were tuted with only ,
longitudinally oriented specimens. To convert the longitudinal data to !
tr:nsverse data, the licensee reduced the USE values by 65%, in accordance j with the correction factor identified in Section 1.2 of Branch Technical ;
Position - MIEB 5-2, " Fracture Toughness Requirements", which is documented in !
.itandard Review Plan 5.3.2 in NUREG-0800, Revision 1, July 1981. Using the !
methodology documented in RG 1.99, Revision 2, all plates are projected to ;
have USE greater than 50 ft-lb through August 2013. !
The unirradiated Charpy USE for the welds was determined from tests of weld f deposits fabricated using the same heat or combination of heats of weld wire ,
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used to fabricate the beltline welds, except for the axial welds with heat .
number 51989 and the axial welds with heat numbers 13253 and 12008. The r licensee's-data search could.not find any unirradiated Charpy USE data for I welds with heat number 51989 or with heat numbers 13253 and 12008. For these ;
weld wire heats the licensee determined generic unirradiated values of Charpy i USE. !
The welds fabricated with heat number 51989 weld wire were welded using Linde 124 flux. The licensee's data search identified 19 unirradiated Charpy USE l values from 16 Certified Material Test Reports and 3 PWR surveillance welds !
- that were fabricated by Combustion Engineering using Linde 124 flux. l l
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The mean value of this data,101.3 ft-lb, was used by the licensee as the unirradiated Charpy USE for the welds with heat number 51989.
The welds fabricated with heat numbers 13253 and 12008 weld wire were welded using Linde 1092 flux. The licensee's data search identified 13 PWR surveillance welds that were fabricated by Combustion Engineering using Linde 1092 flux. The mean value of this data,112 ft-lb, was used by the licensee as the unirradiated Charpy USE for the welds with heat numbers 13253 and 12008.
Using th methodology documented in RG 1.99, Revisien 2 and the previously discuued amounts of copper and unirradiated Charp/ USE, the licensee's evaluation indicates that all welds are projected to have Charpy USE greater than 50 ft-lb through August 2013.
The staff agrees with the licensee that generic values of unirradiated Charpy USE may be used when heat specific data is unavailable. However, the unirradiated value to be used should be the mean minus two standard deviation value to account for uncertainty in the data and the generic value should be calculated using data from all flux types that were used by Combustion Engineering to fabricate the beltline welds because the licensee has not demonstrated that the different flux types used by Combustion Engineering to fabricate the beltline welds result in different unirradiated Charpy USE.
The staff used the certified material test data reported by the licensee and the unirradiated Charpy USE data reported by all PWR and BWR licensees with surveillance welds fabricated by Combustion Engineering to determine a generic unirradiated Charpy USE for Arcos B-5, Linde 0091, Linde 124 and Linde 1092 flux welds fabricated by Combustion Engineering. The mean minus two standard deviation unirradiated Charpy USE value is 75 ft-lb. Using this value of unirradiated USE and the methodology in RG 1.99, Revision 2, the axial welds with heat number 51989 are projected to have USE greater than 50 ft-lb through August 2013, but the axial welds with heat numbers 13253 and 12008 are projected to have USE of 49 ft-lb.
The licensee has provided an analysis to demonstrate that the axial welds in the Fort Calhoun reactor vessel beltline could meet the margins of safety against fracture equivalent to those required by Appendix G of the ASME Code if the Charpy USE falls below 50 ft-lb. To meet these safety margins, the licensee must compare the material's fracture resistance (J ) to the driving force for fracture (Jg ). The material fracture resistance used in the licensee's analysis was Utermined using the Charpy weld methodology in NUREG/CR-5729. This is an acceptable method of determining the material fracture resistance for the Fort Calhoun reactor vessel axial beltline welds.
The licensee determined the driving force for fracture for their size vessel from the Combustion Engineering Owners Group Report No. CEN-604, Revision 01,
" Final Evaluation of Low Upper Shelf Energy for Combustion Engineering Nuclear Steam Supply Systems Reactor Pressure Vessels," September 1993. This report was 1993.
submitted to the NRC for information only in a letter dated September 23, The difference between the material fracture resistance and the driving force for fracture that is documented in the licensee's report demonstrates
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that the axial welds in Fort Calhoun will satisfy the safety margins against fracture equivalent to those required by Appendix G of the ASME Code on August 9, 2013. This conclusion is supported by generic analyses of reactor ,
vessels similar to Fort Calhoun that are documented in NUREG/CR-6023. !
NUREG/CR-6023 indicates that reactor vessels with Charpy USE less than 1 49 ft-lb could meet the margins of safety against fracture equivalent to those !
required by Appendix G of the ASME Code. j
.i 2.1.1.2 Irradiation Temperature and Surveillance Material Test Results !
As discussed previously, the decrease in Charpy USE and the increase in RTp ,, :
caused by irradiation was predicted using the methodologies recommended in :
RG 1.99, Revision 2. This guide indicates that: (a) the procedures for l estjmating embrittlement are val,id for a nominal irradiation temperature of i 550 F, (b) irradiation below 525 F should be considered to produce greatec (
emb-ittlement, and (c) the gtandard deviations for the increase in RTp7 , +
caused by irradiation is 28 F for welds and 170 F for plates, j The licensee reported that for approximately 31% of fuel cycle 2 the Fort i Calhoun reactor vessel operated with an average inlet temperature of 522.7"F. ;
This represents approximately 1% of the vessel's total operating time. The i methodology in RG 1.99, Revision 2 for estimating embrittlement should be 3 applicable to Fort Calhoun because the Fort Calhoun reactor vessel operated l 99% of its time with an average inlet temperature greater than 525"F. This .
conclusion is substantiated by the test results from the Fort Calhoun l surveillance program. Two surveillance capsules have been withdrawn from the !
1 Fort Calhoun reactor vessel. The test results from these capsules indicate ;
that the measured increase in RT is within one standard deviation of the j RG1.99, Revision 2predictedval7,ue and the measured percentage decrease in ;
- Charpy USE is less than the RG 1.99, Revision 2 predicted decrease in Charpy. j l USE. Since the capsules were irradiated at the lower irradiation temperature d
and the test results are in good agreement with the RG 1.99, Revision 2 :
I predicted values, the RG 1.99, Revision 2 methodologies are applicable to Fort
- Calhoun. j 2.1.1.3 Conclusions l
a) The Fort Calhoun surveillance data indicates that the methodology in i RG 1.99, Revision 2 for predicting the effect of neutron irradiation on i reactor vessel materials is applicable to the Fort Calhoun reactor vessel. ,
i b) The materials data provided by the licensee indicates that RTp7 value for {
the Fort Calhoun reactor vessel materials will be less than the scr,eening t criteria in ths PTS rule, 10 CFR Part 50, on August 9, 2013, and PTS is not a j concern for the Fort Calhoun reactor vessel. i c) The materials data provided by the licensee indicates that the Charpy USE ?
for all Fort Calhoun reactor ve::sel materials, except for the axial welds !
fabricated with heat numbus 13253 and 12003 weld wire will be greater than i the 50 ft-lb criteria in 10 CFR Part 50, Appendix G, on August 9, 2013. The i staff has performed a conservative analysis of generic unirradiated Charpy j i
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I data and determined that the Charpy USE for the axial welds fabricated with heat numbers 13253 and 12008 weld wire could be 49 ft-lb on August 9, 2013. l d) Based on the difference between the material fracture resistance and the i driving force for fracture that is documented in the licensee's report and the j generic analyses of reactor vessels similar to Fort Calhoun that are ;
documented in NUREG/CR-6023, the staff believes that the Fort Calhoun reactor vessel will be able to satisfy the requirements of 10 CFR Part 50, Appendix G i on August 9, 2013. l e) Based on the above conclusions, we believe the Fort Calhoun reactor vessel '
can be safely operated until August 9, 2013, and its operating license expiration date may be extended to August 9, 2013. l B
2.1.2 Fast Neutron Fluence ;
The limiting (critical) element in the Ft. Calhoun pressure vessel is axial l weld 3-410 which has relatively high copper and nickel contents. This weld is subject to high rate of embrittlement and loss of fracture toughness. The licensee has implemented aggressive flux reduction measures to reduce the :
peripheral assembly power and thus reduce the fast neutron leakage. These steps consisted of: (1) use of full length hafnium rods in the assemblies '
facing the critical weld (2) use of natural uranium fuel rods and (3) use of integral absorber rods, comprised of ZrB (zirconium diboride) coated fuel ;
pellets, which is used to maintain lower, radial flux peaking and thus maintain 1 thermal margin. l The flux estimation was performed using the DOT 4 code which was benchmarked to ,
calculations related to the pool critical assembly (PCA). The present flux . l estimates are based on a P3 scattering and a Se quadrature approximations and ,
ENDF/B-IV based cross sections. The source estimates are based on a pin-by< l pin power distribution in a DOT 4 p(r,6) solution and axial peaking from a i d(r,z) solution. The Pu buildup in the peripheral assemblies was taken into account in the estimation of the neutron source. Dimensions were derived from '
as-built drawings. The licensee is planning a second phase in the fluence computations, subject to the outcome of staff research effort currently under l way at Brookhaven National Laboratory. In this second phase, the computational methodology will include an updated set of cross sections based ,
on ENDF/B-VI and updated uncertainties evaluation.
The methodology and the approximations used in the fluence estimate are l acceptable. However, there are two issues which need to be brought out: i (1) the licensee used a load factor of 0.77 in estimating the end of license i flue,ce. This value was based on Ft. Calhoun historical performance, however, it is slightly lower than the 0.80 value recommended in 10 CFR 50.61; and (2) Ft. Calhoun utilizes a thermal shield, and when the second phase is i implemented (ENDF/B-VI based cross sections and appropriate uncertainties) it '
is very likely that the estimated value will increase, making the estimated RT'7s value higher and possibly higher than the 10 CFR 50.61 screening criteria. In a meeting with the staff on April 7, 1993, the licensee understood this likelihood and opted to proceed with the present methodology.
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Therefore, this approval is subject to the limitations of monitoring of the long term load factor to assure it does not exceed the assumed value of 0.77, and the expectation that the licensee will perform a reevaluation of the end of license fluence with ENDF/B-VI cross sections and updated uncertainties to assure that the value of the RTn , will not exceed the screening criterion.
2.2 Reactor Coolant System As discussed above, the useful life of Fort Calhoun Station was intended to be 40 years. The Fort Calhoun reactor vessel was designed for transients considered to envelop design conditions over a 40-year operating period. The licensee has stated that Fort Calhoun monitors a number of these cycles and has determined that in the most limiting case, Loss of Turbine Load, only 40%
of the design cycles have been actually experienced during the first 50 %
(approximately 20 years) of plant operating life. Extrapolating this data indicates that Fort Calhoun can operate for its full 40-year design life without exceeding the design number of vessel cycles.
2.3 Electrical Eouipment Aging analysis has been performed for all safety-related electrical equipment in accordance with 10 CFR 50.49, " Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants," identifying qualified lifetimes for this equipment. These lifetimes have been incorporated into plant equipment maintenance and replacement practices to ensure that all safety-related electrical equipment remains qualified and available to perform its safety-related function regardless of the overall age of the plant.
The staff's Safety Evaluation for environmental qualification (EQ) of safety-related electrical equipment was issued in a letter dated February 15, 1985.
In the February 15, 1985, letter, the staff concluded that the Fort Calhoun EQ Program is acceptable and that compliance with 10 CFR 50.49 has been demonstrated.
3.0
SUMMARY
OF FINDINGS The staff published its original Safety Evaluation for Fort Calhoun Station on August 9, 1972. While changes have been made to the plant design since the original plant construction was completed, such as spent fuel pool modifications, major changes for fire protection in response to Appendix R, many TMI Task Action Plan modifications, and various other less major design changes, each of these changes where it involved safety-related components has been reviewed and approved by the staff with the details being documented in the staff's related Safety Evaluation. Further, as required by 10 CFR 50.71(e), these changes and their effect on accident analyses, if any, are routinely updated in the USAR. Based on the ongoing review process, the staff has not identified any concerns associated with approval of the proposed amendment to extend the expiration date of the license that are not already addressed by licensee commitments, operating procedures, and license requirements.
The NRC staff concluded in the Environmental Assessment that the annual radiological effects during the additional years of operation that would be '
authorized by the proposed license amendments are not more than were previously estimated in the Final Environmental Statement, and are acceptable.
The staff concludes from its considerations of the design, operation, testing, and monitoring of the mechanical equipment, structures, and the reactor vessel 1 that an extension of the operating license for Fort Calhoun Station to a 40-year service life is consistent with the USAR, Safety Evaluation Reports (SERs), and submittals made by the licensee, and that there is reasonable i
assurance that the unit will be able to continue to operate safely for the additional period authorized by this amendment. The plant is operated in compliance with the Commission's regulations, and issues associated with plant '
degradation and population changes have been adequately addressed.
In summary, the staff finds that the extension of the operating license for L Fort Calhoun Station to allow 40-year service life is consistent with the Final Environmental Statement and the Safety Evaluation Report for Fort Calhoun Station and that the Commission's previous findings are not changed.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
A Notice of Issuance of Environmental Assessment and Finding of No Significant Impact relating to the proposed extension of the Facility Operating License termination dates for the Fort Calhoun Station was published in the Federal Reaister on August 3, 1987 (52 FR 28767).
6.0 CONCLUSION
f The Commission has concluded, 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, (2) such activities will be conduct" in compliance with the Commission's regulations, '
and (3) the issuance of tha anendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
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- 1. Letter from W. G. Gates, Omaha Public Power District to USNRC " Updated 1 Assessment for Reaching 10 CFR 50.61 Pressurized Thermal Shock (PTS) Screening Criterion for Ft. Calhoun Station (FCS)" August 12, 1993. ,
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- 2. Letter from W.G. Gates, Omaha Public Power District to USNRC " Supplemental Information to support Extension of the Ft. Calhoun Station (FCS) Operating '
License Expiration Date" August 12, 1993.
Principal Contributor: Steven Bloom '
Barry Elliot Lambros Lois Date: December 3, 1993
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