ML20082U925
| ML20082U925 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/12/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20082U919 | List: |
| References | |
| NUDOCS 9109230036 | |
| Download: ML20082U925 (8) | |
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WASHINGTON, D.C 20565 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOP REGULATION RELATED TO AMENDMENT NO.154 TO FACILITY OPERATING LICENSE NO. DPR-71 AND AMENDMENT N0.- 186 TO FACILITY OPERATING LICENSE NO. DPR-62 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324
_1.0_ INTRODUCTION By. letter dated August 17, 1987, as supplemented May 30 and June 29, 1990,_ Carolina Power & Light Company (CP&L or the'-licensee) requested
. amendments to Facility Operating Licenses OPR-71 and DPR-62 for the Brunswick Steam Electric Plant, Units 1 and 2 (BSEP1' and BSEP2). The proposed _ amendments would exter.d the expiration dates of these licenses from February 7, 2010,-to September 8, 2016, for BSEP1, and from February 6, 2010, to December 27, 2014, for BSEP2, effectively recapturing the construction time as operating time. The letters provided by the licensee
.on August 8 and August 29, 1991, provided clarification and did not alter the requested action or staff findings.
2.0 DISCUSSION
-Title 10 of the CodeLof Federal Regulations, Section 50,51 (10 t.FR 50.51),
specifies that each license will be issued for a fixed oriod of time not to exceed 40 years from the date of issuance. The cum ent term shown in L
the licenses for BSEP1:and BSEP2 is 40 years-commencing with the issuance of-the construction permit. Accounting for the time that was required for construction, the effective operating license terms were 33 years and 5-months.for BSEP1, and 35 years and 2 months for BSEP2. Consistent with 10 CFR 50,51: of the Commission's regulations, the licensee, by the
. August 17, 1987, application, requested extensions of-the operating license times for BSEP1 and BSEP2. This request would set the fixed periods of the L
licenses from-the dates of-issuance of the operating licenses rather than L
from the dates of the construction permits.
o
-3.0 EVALUATION
-The staff has evaluated the safety issues associated with issuance of the proposed license amendments. These proposed amendments would allow additional periods of operation of 6 years, 7 months for BSEP1 and 4 years, 10 months, for BSEP2._-The issues addressed consist of additional radiation exposure to the licensee's operating staff, potential increased impacts on the offsite population, and the general aging of plant structures 9109230036 910912 PDR ADOCK 05000324 P
PDR.
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and equipment.
The impact of additional radiation exposure to the facility operating staff and the impact on the general population in the vicinity of BSEP are addressed in the NRC staff's Environmental Assessment (56 FR 46016) dated September 9, 1991.
The licensee discussed the impact of license extension on major, difficult to replace components, equipment and plant structures in their May 30, 1990, letter. The items considered in this category are reactor vessel, mechanical equipment, and plant structures.
BSEP was designed for a 40-year operating life. The reactor vessels, which are generally regarded as the limiting item for the purpose of plant operating life, were designed for 40 years of normal operation.
Mechanical equipment is designed for 40 years of operation and is subjected to comprehensive surveillance and maintenance programs to assess aging-(including the Inservice Inspection (ISI) Program and Inservice Testing Program). Related electrical equipment is also designed for 40 years of operation or is subject to the Environmental Qualification (EQ)
Program.
3.1 Reactor Vestals and Internals The BSEP reactor vessels were designed and fabricated to meet the requirements of 10 CFR 50.55a and Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME Code) (1965 Edition, including Summer 1967 Addenda). The licensee has stated that the vessels will retain structural integrity throughout their design life with an adequate margin. The _ detailed designs and analyses were performed by Chicago Bridge and Iron Company and were approved by Gereral Electric Company.
The reactor vessel analyses-were performed in three parts:
thermal evaluation, stress calculations, and fatigue evaluations. All areas were qualified with adequate margin for the design life of 40 years.
The-BSEP reactor vessels and. associated surveillances comply with the.
requirements of 10 CFR Part 50, Appendix H and the ASME Code.Section XI.
As required by 10 CFR Part 50, Appendix H, a reactor vessel material surveillance program was established at the time the operating licenses l
were issued to monitor the effects of radiation-induced changes on the l
material properties of the reactor vessel.
BSEP established a surveillance capsule withdrawal schedule which uses three capsules. By letterm dated October 26, 1988, the licensee proposed revisions to the surveillance capsule withdrawal schedule in accordance with 10 CFR Part 50, Appendix.H, paragraph II.B.3.
The revised schedule was intended to be more compatible with boiling water reactor irradiation conditions and the BSEP materials.
The NRC staff. subsequently approved the withdrawal-schedule revisions by issuance of Amendment Nos.-140 and 172 to the Facility Operating Licenses, i
dated February 15, 1990. The-reactor material surveillance program requires the surveillance capsules to be withdrawn and examined to determine radiation-induced changes in material properties of the reactor vessels.
l-The data from the surveillance capsules will be used as descrfbed in 10 CFR Part 50, Appendix G, Sections IV and V and, based on the observed changes, the pressure / temperature limit curves along with surveillance
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- capsule withdrawal schedules will be revised and submitted to the NRC staff for review and approval in accordance with Section III of 10 CFR Part 50, Appendix H.
Therefore, the BSEP reactor vessel surveillance program complies with the requirement of 10 CFR Part 50, Appendix H, and provides adeauate assurance that the reactor vessel will be monitored for radiation-induced changes in material properties, throughout the licensed life of the plant including the proposed extension periods.
The licensee stated that, based on plant operating history and a projected 75 percent capacity for the remainder of the proposed 40 calendar year operating life of the plant, BSEP 1 is expected to achieve 76.9 effective full power years (EFPY) of operation. BSEP 2 is expected to achieve 25.8 EFPY of operation.. Based on the licensee's operating projections and the resulting vessel irradiation, the most limiting transition temperature shifts are' estimated to be 52* F (ID) and 45 F-(1/4T) for BSEP 1 and 72*
F (ID) and 63* F (1/4T) for BSEP 2.
Thus, the licensee expects that the transition temperature shifts at the end of 40 calendar years of plant life will be 1ess than 100' F, and the u'se of three surveillance capsules, as required by ASTM ".185-82, remains adequate.
The-ISI Program for the BSEP reactor vessels is based on the requirements of the ASME Code,Section XI, and applicable supplementary requirements.
The objective of the ISI Program is to observe if any age-related degradation occurs before it can significantly encroach on design margins.
Industry experience has shown that the quality of inspections and the extent of vessel regions inspected are constantly increasing. The ISI Program
-provides an ongoing confirmation of structural integrity of the reactor vessels during their entire operating life, including the proposed
' extension periods.
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Based on the above, it is concluded that the BSEP reactor vessels are fully-qualified for 40 years of plant operation.
The design of the BSEP reactor internals is in accordance with the applicable portions of-Section-III of the ASME Code, 1965 Edition through l-and including the Sumer 1967 Addenda..The design evaluation included a fatigue. assessment for the-40-year plant operational period-in accordance -
with the ASME Code. The major components within the reactor vessel were L
subjected to extensive-testing on a prototype plant coupled with a dynamic system analysis of BSEP to properly describe any resulting flow induced phenomena incurred from normal plant operation and from anticipated operational transients. Possible contributory sources of vibration were postulated from pump operation, flow-induced vibration caused by cross and/or parallel flow, and turbulent flow.- All flow-induced vibratory.
stresses were well within the fatigue allowable stresses established by.
the ASME Code. Based on the above,'it is concluded that the BSEP reactor internals are fully qualified for 40 years of plant operation.
3.2 Mechanical Equipment At the' time of licensing, the NRC concluded that the design of pressure retaining mechanical fluid systems within the boundaries of Atomic Energy Comission (AEC) Safety Classification A, B, and C were designed and
constructed in accordance with design criteria consistert with the coues listed in Regulatory Guide 1.26 and in conformante with section 50.55a of 10 CFR Part 50. The NRC furt her concluded that compliance "4+h the above design criteria provided reasonable assurance that th s.
nests' on ding quality level was adequate to safely withstand the plant de conditions and the combina' ion of loadings which the syster experience over the service lifetime without loss of stru.
....cegrity.
In addition, the NRC determined that the components satisPao the rcquirements of AEC General Design Criteria 1, 14, and 30.
As suppoit for the proposed license amendment, CPAL has considered the
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pote.itial effect of the operating license extension on mechanical equipment and ccncludes that there will be no impact. Mechanical equipment for the BSEP was swecific.d to have a design life for 40 years of operation or is subject t se"teillance, testing, and maintenance requirements to detect de"o,stion and ensure corrective action.
For example, the nuclear t^tsn. 2;pply system mechanical equipment was designed and procured for a 40-year design life. All safety-related piping
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systems, incluaing piping supports were analyzed for 140-year design life, nsing dCC-approved methods and computer codes, ind conformed to the ASME B31.1.0 Power Piping Code, 1967 Eoition.
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lt was, and contin"as to be, understood that some items of equipment and subcomponents ure not expected to last 40 years.
Surveillance, maintenance, and testing of mechanical equipment are performed to verify operability of the equ'pment or detect potential degradation and ensure inadditIon,uipmentisreplacedorsomeotherappropriateaction that, when required eq subcomponents such as nonmetallics (e.g., gaskets, is taken, 0-rings) are inspected and periodically replaced, as necessary, as part of routine maintenance in order to ensure that the design life of the r
!r-equipment will be achieved.
These surveillance activities provide the necessary assurance that mechanical equipment will be maintained throughout the operating life of the plant, including the proposed license extensinn period.
3.3 Electrical Equipment Safety-related electrical equipment installed in the BSEP was designed for a full 40-year operating life. Exceptions include those cases where the equipment has some consumable quantity (e.g., neutron monitoring 0
detectors and bctteries,). Equipmant maintenance (wha"r required or anticipated for both preventive and corrective purposes) nas been considered within applicable plant maintenance procedures.
For those cases where less than a 40-year design service life applies, maintenance activities include equipment and component replacement. Additionally, required maintenance surycillance testing prac~cices have been implemented to maintain plant operating conditions within the plant Technical Specification (TS) limits.
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The existing design considerations and ongoing maintenance practices provide assurance that BSEP safety-related electrical equipment will a
remain operable through a full 40-year plant operating life (i.e.,
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throu h September 8, 2016, for DSEP1, and through December 27, 2014, for
!!SEP2.
The EQ program is one area in particular in which a programmatic activity is performed to maintain plant operating conditions within plant TS limits. The BSEP has in place an EQ program for safety-related electrical equipmen* and cables located in th' potentially harsh environments of the primary and secondary containments, to comply with the rcquirements of 10 CFR 50.49.
This program was found acceptable by the NRC, based upon a March 5, 1985, Safety Evaluation, as well as NRC audit inspections of the EQ Program during August 1985 and August 1989.
The BSEP EQ program includes the consideration of a " qualified life" for each item of electrical equipment / cable within its scope, in many cases, equipment and cable qualified lives have been deten 'ned to be greater than 40 years from the date of initial plant operation.
For those remaining cases where the qualified life has been determined to be less than 40 years, an EQ-related replacement and refurbishment process has been established, based upon EQ program documentation, and is being implemented as part of ongoing plant maintenance activities. Additional assurance of BSEP EQ equipment integrity was provided when a significant amount of equipment was replaced to assure compliance with 10 CPR 50.49.
In addi?. ion, EQ-related cable replacements in upper drywell elevations have been or will be performed for each unit based upon higher ambient temperature experienced to date (relative to operating conditions at lower drywell elevations).
Programmatic activities will continue to be performed irrespective of the proposed plant operating expiration date.
3.9 Structures l
All Seismic Category I structures for BSEP, including the containment, the concrete and structural steel internal structures, and the foundations were reviewed and found acceptable by the NRC at the time of licensing.
The structures were designed for dead loads, live loads, missiles, large breakloss-of-coolantauc':lents(LOCA),smallbreakLOCA,seismicevents.
l hurricane loads, and torrado loads in accordance with the applicable codes. The pre-stressed, post-tensioned concrete girders which support the fuel pool, steam separator and dryer pool, and reactor well were designed in accordance with the required ASME Code,Section III, and AmericanCohcreteInstituteStandard(ACI)-318.
The reinforced concrete containment is generally known not to be susceptible to significant degradation with time.
Nevertheless, the licensee has measures in place to ensure that any deterioration is detected and repaired. Throughout the service life of the plant, the containment structure is subject to the inspection and testing program of Appendix J. The Appendix J 1eak rate testing program is well documented
and provides reasontble assurance that the containment structure remains capable of performing its design function throughout the service lift of the facility, including the proposed extension periods.
The plant's concrete and structural steel internal structures, including walls, compartments and floors, its other Seismic Category I structures (slabs, walls, beams and columns), and its foundations were 'ound adequate to meet General Design Ctiteria 2 and 4 These structures are generally known not to be susceptible to significant age-related degradation.
Nevertheless, surveillance and maintenance requirements set forth in the TS provide assurance of structural integrity and ensure that any degradation will 'ue detected and repaired.
3.5. Siting The NRC staff has concluded in its associated Environmental Assessment that the annual radiological effects during the proposed additional years of o>eration are not significantly greater than were previously estimated in tie Final Environmental Statement. These radiological effects are within acceptable limits.
The Exclusion Area is owned by CP&L and the Noith Carolina Eastern Municipal Power Agency (NCEMPA) and centro 11ed by CP&L.
No one lives within this area. As discussed in the Updhted Final Safety Analysis Report, all activities occurring within the exclusion area are either directly or indirectly related to plant operations.
CP&L owns and operates a rail line within the exclusion area. An agreement with Pfizer Corporation permits them to operate on an extension of the rail line to the Pfizer Plant.
This extension runs outside but parallel to a portion of the exclusion area with a 100-foot wide easement extending into the exclusion area. The casement allows Pfizer the right to operate the railroad, as well as mainitain an access road and underground pipeline for water and effluents. Normal operation and maintenance of this portion of the track does not extend into the Exclusion Area.
No change to these practices and conditions are anticipated through the requested extension periods for the operating licenses.
Pro,iecttd changes in population within the Low Population Zone (LPZ),
nearest population center distances and 10-mile radius Emergency Planning Zone (EPZ) have been found not to be significant for the period of the license extensions. Accordingly, the Commission's conclusions cegarding 10 CFR Part 100 siting criteria for BSEP are that the Exclusion Area, the LPZ, and population center distances meet the guidelines of 10 CFR Part 100 and are not changed by the proposed license extensions.
The staff concludes from its evaluation of the design, operation, testing and monitoring of the mechanical equipment, structures, reactor vessels, and electrical equipment / components, and siting that an extension of the operating licenses for BSEP 1 and BSEP 2, to a 40-year service life is consistent with the FSAR and the Safety Evaluation Report (SER), es supplemented.
A*
There is reasonable assurance that these units will continue to operate safely for the additional periods authorized by these amendments. Thc plants are opert.i.ed in coopliance with the Commission's regulations, and issues associated with plant degradation have been adequately addressed herein and in the previously issued evaluations relating to this matter.
4.0,5T ATE C0tlSULTAT10t1 In accordance with the Coneission's regulations, the State of I: orth Carolina official was notified of the proposed issuence of the amendment.
The State official had no comments.
5.0,Ellygp!MMM,CpK51DERATW4 Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessroent and finding of no significant impact have been prepared and published in the federal P3 ~1 ster on September 9, 1991 (56 FR 46016), t.ccordingly, based upon the enviriirin, ental assessment, we have determitied th6t the issuance of this amendment will not have a significant effect on the quality of the human environment.
6.0 C0tiCLUS10_t1 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 encangered by operation in the proposed manner, (p) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the ernendment will not be inimical to the cammon defense and security or to the health and safety of the public.
Principal Contributor:
B. tiozafari Date:
September 12, 1991
AttE!!Dt4Et1T 110. l'A TO FACILITY OPERATitlG LICENSE 140. OPR liRtitiSWICK, UtllT 1 AMENOMElli NO. 186 10 FAClllTY OPEPATlfiG LICENSE tiO. DPR BRtitiSWlCK, littlT P DocketFils f4RC +'9R Local PDR PD11-1 Reading T. Murley(1404)
S. Varga G. Lainas E. Adensam
/
P. Andercon ti. Le OGC D. Hagan (MilBB 3302)
E. Jordan (titlBh 3701)
G. Ih11 (4) (r?-137)
Wanda Jones (i>-130A)
C. Grimes (1103)
ACRS (10)
GPA/PA OC/LFith Brunswick file L. Reyes, Rll cc:
Brunswick Service List
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