ML20151E922
| ML20151E922 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/12/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20151E914 | List: |
| References | |
| NUDOCS 8804150436 | |
| Download: ML20151E922 (3) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.151 TO FACILITY GPERATING LICENSE NO. DPR-62 CAROLINA POWER & LIGHT COMPANY et al.
BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 DOCKET NO. 50-324
1.0 INTRODUCTION
By letter dated February 3,1988, as supplemented March 30, 1988, the Carolina Power & Light Company submitted a request for changes to the Brunswick Steam Electric Plant, Unit 2, (BSEP-2) Technical Specifications (TS) to incorporate upgraded Minimum Critical Power Ratio (MCPR) values applicable to the operation of BSEP-2, Cycle 8.
On March 30, 1988, the licensee provided clarification with respect to NRC staff concerns.
In addition, the submittal provided a changed MCPR value which had been inadvertently omitted in the February 3,1988 submittal. The March 30, 1988 submittal did not substantially change the action noticed, or alter the staff's initial determination published, in the F_ederal Reoister on March 9, 1988, 2.0 EVALUATION 2.1 MCPR Safety Limit The MCPR fuel cladding integrity safety limit of 1.07, currently used for BSEP-2 reload cores, was established in 1978. This safety limit was designed to provide a level of conservatism for establishing operating limit MCPR values, based on fuel design characteristics typical of those utilized at that time. The level of conservatism built into the safety limit provides adequate margin to ssure that more than 99.9% of the fuel rods in the core are expected to ay11d boiling transition.
The increase in conservatism has been recognized because of current fuel designs. An updated safety limit of 1.04, specified in Amendment 14 to NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel' (GESTAR II), has been reviewed and approved by the NRC for D-lattice plants when applied to second successive reload cores of P8x8R, BP8x8R, GE8x8E or GE8x8EB fuel types with high bundle R-factor ( 1.04).
BSEP-2 is such a 0-lattice plant, with Cycle 8 being the third successive reload core with high bundle R-factor ( 1.04) fuel designs.
Therefore, the staff finds that the proposed amendment for the changing the MCPR safety limit specified in the BSEF-2 TS from 1.07 to 1.04 is acceptable.
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- 2.2 Operating Limit MCPR Values Operating limit MCPR (0LMCPR) values are designed to limit the conse-quences of operational transients previously evaluated.
Since the upgraded safety limit MCPR was reviewed and generically approved by the i
staff, adjustment of the operating limit MCPR values was proposed by CP&L for a plant specific application. The licensee also provided a letter on March 30, 1988, to clarify the staff's concern on the deletion of the MCPR adder.
The staff has reviewed the February 3 and March 30, 1988 submittals and found that the Cycle 8 reload for BSEP-2 meets the criteria set for the application of the upgraded safaty limit MCPR and that the clarification for deletion of the MCPR adder is acceptable.
Therefore, the proposed adjustment of the OLMCPR values for BSEP-2 Cycle 8 reload is acceptable.
2.3 Technical Specifications The Technical Specification changes are for the most part related to the approved upgraded safety limit MCPR.
Details of the specification changes follow:
(1) Specifications 2.1.2, 3.1.4.3 and 3.3.4 and Bases 2.0, 2.2.1, 3/4.1.3, 3/4.2.3.
The amendment changes the MCPR safety limit, specified in the BSEP-2 TS, from 1.07 to 1.04.
This change is based on the generically approved amendment to GESTAR II.
Therefore, the staff finds the proposed change is acceptable.
(2) Bases 3/4.2.3 and Table 3.2.3.2-1 New operating limit MCPR values correspond to the generically approved upgraded MCPR safety limit of 1.04.
The new OLMCPR value is 0.05 smaller than that of the original value given in the BSEP-2 Cycle 8 reload analysis. This 0.05 difference is due to 0.03 gained from the upgraded MCPR safety limit and 0.02 MCPR adder deleted from this proposal. The BSEP-2 licensee stated in the March 30, 1988 letter that the operator will take all the necessary corrective actions to bring the reactor to a safe operating condition by reducing the reactor power in case of opera-tional occurrence', such as a main steam line isolation valve out-of-service or a fewater heater out-of-service.
This supports the deletion of the 0.02 MCM adder for BSEP-2 Cycle 8 operation.
Therefore, we find the proposed new OLMCPR values are acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
S This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the
. types, of any effluents that may be released off site; and that there should be no significant increase in individual or cumulative occupa-tional radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration, and there has been no public comment on such finding.
Accordingly, this amendment meets the eligibility) criteria for cate-gorical exclusion set forth in 10 CFR $51.22(c)(9. Pursuant to 10 CFR 051.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
4.0 CONCLUSION
The Comission ma@ a proposed determination that this amendment involves no significant hazards consideration which was published in the FEDERAL REGISTER (53 FR 7585) on March 9, 1988, and consulted with the State of North Carolina.
No public comments or requests for hearing were received and the State of North Carolina did not have any coments.
The staff 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, and (2) such activities will be conducted in compliance with the Comission's regula-tions, and the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Principal Contributor:
T. Huang Dated:
April 12, 1988 i
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UNITED STATES l'
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April 12, 1988 Docket No. 50-324 Mr. E. E. Utley Senior Executive Vice President Power Supply and Engineering & Construction Carolina Power & Light Company Post Office Box 1551 Raleigh, North Carolina 27602
Dear Mr. Utley:
SUBJECT:
ISSUANCE OF AMENDMENT N0.151 TO FACILITY OPERATING LICENSE NO. DPR BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2, REGARDING UPGRADED MCPR SAFETY LIMIT, CYCLE 8 (TAC NO. 67128)
The Nuclear Regulatory Connission has issued the enclosed Amendment No.151 to Facility Operating) License No. DPR-62 for the Brunswick Steam Electric Plant, Unit 2 (BSEP-2. The amendment consists of changes to the Technical Specifications in response to your submittal dated February 3,1988, as supplemented March 30, 1988.
The amendment changes the Technical Specifications to incorporate an upgraded Minimum Critical Power Ratio (MCPR) fuel cladding integrity safety limit and associated operating limit MCPR values applicable to the operation of BSEP-2, Cycle 8.
A copy of the related Safety Evaluation is also enclosed.
Notice of issuance will be included in the Conmission's Bi-Weekly Federal Register Notice.
Sincerely, M h.
4 Ernest D. Sylvester, Project Manager Project Directorate 11-1 Division of Reactor Project I/II
Enclosures:
1.
Amendment No.151 to License No. DPR-62 2.
Safety Evaluation cc w/ enclosures:
See next page 1
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Mr. E. E. Utley Brunswick Steam Electric Plant Carolina Power & Light Company Units 1 and 2 cc:
Mr. P. W. Howe Mr. C. R. Dietz Vice President Plant General Manager Brunswick Nuclear Project Brunswick Nuclear Project Box 10429 Box 10429 Southport, North Carolina 28461 Southport, North Carolina 28461 Mr. R. E. Jones, General Counsel Mr. H. A. Cole Carolina Power & Light Company Special Deputy Attorney General P. O. Box 1551 State of North Carolina Raleigh, North Carolina 27602 Post Office Box $29 Raleigh, North Carolina 27602 Mr. Mark S. Calvert Associate General Counsel Mr. Robert P. Gruber Carolina Power & Light Company Executive Director Post Office Box 1551 Public Staff - NCUC Raleigh, North Carolina 27602 Post Office Box 29520 Raleigh, North Carolina 27626-0520 Mr. Christopher Chappell, Chairman Board of Commissioners Post Office Box 249 Bolivia, North Carolina 28422 Mrs. Chrys Baggett State Clearinghouse Budget and Management 116 West Jones Street Raleigh, North Carolina 27603 Resident Inspector U. S. Nuclear Regulatory Comission Star Route 1 Post Office Box 208
.l Southport, North Carolina 28461 Regional Administrator, Region II U. S. Nuclear Regulatory Commission 101 Marietta Street, Suite 2900 Atlanta, Georgia 30303 Mr. Dayne H. Brown, Chief Radiation Protection Branch Division of Facility Services N. C. Department of Human Resources 701 Barbour Drive Raleigh, North Carolina 27603-2008
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, CAROLINA POWER & LIGHT COMPANY, et al.
DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMEND l.2NT TO FACILITY OPERATING LICENSE Amendment No.151 License No. DPR-62 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendnent by Carolina Power & Light Company (the licensee), dated February 3,1988, as supplemented March 30.
1988, complies with the standards and re Energy Act of 1954, as amended (the Act)quirements of the Atomic
, and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
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2.
Accordingly, the license is amended by changes to the Technical Specificaticns, as indicated in the attachment to this license airendment; and paragraph 2.C.(2) of Facility Operating License No. DPR-62 is hereby amended to read as follows:
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. (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.151, are hereby incorporated in the license.
Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
i 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION j
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Elinor G. Adensam, Director Project Directorate 11-1 Divisi]n of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: April 12, 1988 l
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ATTACHMENT TO LICENSE AMENDMENT NO.151 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET N0. 50-324 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marginal lines.
Rcmove Pages Insert Pages 2-1 2-1 B2-1 B2-1 B2-2 B2-2 B2-3 B2-3 B2-4 82-4 B2-5 B2-5 B2-6 82-6 B2-7 B2-7 B2-8 B2-8 B2-9 B2-10 B2-11 B2-12 I
B2-13 3/4 1-17 3/4 1-17 3/4 2-8 3/4 2-8 3/4 2-12 3/4 2-12 B3/4 1-2 B3/4 1-2 B3/4 2-3 B3/4 2-3
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER (Low Pressure or Low Flow)
THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the 2.1.1 reactor vessel steam does pressure less than 800 psia or core flow less than 10% of rated flow.
APPLICABILITY: COWDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of BATED THERMAL POWER and the reactor vessel steam dome pressure less than 800 psia or core flow less than 10% of rated l
flow, be in at least HOT SHUTDOWW within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
THERMAL POWER (High Pressure and High Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.04 l
with the reactor vessel staan does pressure greater than 800 psia and core flow greater than 10% of rated flow.
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APPLICABILITY: CONDITIONS 1 and 2.
ACTION:
With MCPR less than 1.04 and the reactor vessel steam done pressure greater l
than 800 psia and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
l REACIOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 pais.
APPLICABILITY: CONDITIONS 1, 2, 3, and 4.
ACTION:
With ':he reactor coolant system pressure, as sensured in the reactor vessel steam done, above 1325 psis, be in at least HOT SHUTDOWW with reactor coolant system pressure < 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
2-1 Amendment No. 4, 151 l
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2.1 SAFETY LIMITS l
i BASES 2.0 The fuel cladding, reactor pressure vessel, and primary system piping are the principal barriers to the release of radioactive materials to tbv environs. Safety linics are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity limit i,s set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit auch that the MINIMUM CRITICAL PCWER RATIO (MCPI) is no less than 1.04.
MCPR > 1.04 l
represen:s a ::nservative mar b relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom frorn perforations or cracking. Although some corrosion or use related cracking may
. occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.
Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is justi as measurable as that from use-related cracking, the thermally caused cladding perforations sighal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would prcduce onset of transition boiling, MCPR of 1.0.
These conditions represent a significant departure from the condition intended by design for planned operation.
2.1.1 THERMAL POWER (Low Pressure or Low Flow)
The use of the NRC approved CPR correlation is not valid for all critical power calculations at pressures below 800 psia or core flows less than 10% of rated flow.
Therefore, the fuel cladding integrity limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis.
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows wi withaflowof2Sx10}1alwaysbegreaterthan4.5 psi. Analyses show that lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 pai. )Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10 lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.
With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of LATED THERMAL POWER for reactor pressure below 800 psia is conservative.
BRUNSWICK - UNIT 2 8 2-1 Amendment No. O,151
SAFETY LIMITS BASES (Continued)
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2.1.2 THERMAL POWER (High Pressure and High Flow)
The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not riolated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the esgion where fuel damage could occur. Although it is recognised that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power, result in an uncertainty in the value of the critical power.
Therefore, the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.
The Safecy Limit MCpR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power.
The probability of the occurrance of boiling transition is determined using an approved critical power correlation. Details of the fuel cladding integrity safety limit calculation are given in Reference 1.
Uncertainties used in the determination of the fuel cladding integrity safety limit and the bases of these encertainties are presented in Reference 1.
The power distribution is based on a typical 764 assembly core in which the rod p.sttern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution in Rrunswick Unit 2 during any fuel cycle could not be as severe as the distribution used in the analysis. The pressure safety limits are arbitrarily selected to be the lowest transient overpressures allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section 831.1.
Reference
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"Ceneral Electric Standard Application for Reactor Fuel," KEDE-240ll-P-A (latest approved revision).
l BRUNSWICK - UNIT 2 B 2-2 Amendment No,151
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SAFETY LIMITS BASES (Continued) 2.1.3 REACTOR COOLANT SYSTEM PRESSURE The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. However, the pressure safety limit is set high enough sech that no foreseeable circumstances can cause the system pressure to rise to this limit. The pressure safety limit is also selected to be the lowest transient everpressure allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III and USAS Piping Code, Section B 31.1.
2.1.4 REACTOR VESSEL WATER LEVEL With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat.
If the water level should drop below the top of the active fuel during this period, the ability to remove decay beat is reduced.
This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide an adequate margin for effective action.
BRUNSWICX - UNIT 2 B 2-3 Amendment No.151
2.2 LIMITINC SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINJT The Reactor Protection System Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints havo been selected to ensure that the reactor core and reactor ecolant systes are prevented from exceeding their safety limits.
Inter ediate Range Wa-iter, Neutron Flur - Righ The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5-decade.10 -esuze instrument. The trip setpoint of 120 divisions is active in each of the 10 ranges. Thus, as the IRM is ranged up to accournodate the in:vease in power level, the trip setpoint is also ranged up.
Brage 10 allows the IRM instruments to remain on scale at higher power levels to provide for additional overlap and also permits calibration at these higher powers.
The most significant source of reactivity change during the power increase is due to control rod withdrawal.
In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed in Section 7.5 of the FSAR. The most severe case involves an initial condition in which the reactor is just suberitical and the IRMs are not yee on scale.
Additional conservatism was taken in this analysis by assuming the IRN channel closest to the rod being withdrawn is bypassed. The results of this analysis show that the reactor is shut down and peak power is limited to 1% of RATED THERMAL POWER, thus maintaining MCPR above 1.04.
Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
2.
Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides an adequate thermal margin between the setpoint and the Safety Limits.
This margin acconinodates the anticipated maneuvers associated with power plant startup. Effects of increasing oressure at zero or low void content are minor; cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained by the RSCS and RWH. Of all BRUNSWICK - UNIT 2 B 2-4 Amendment No.151
2.2 LIMITING SAFETY SYSTEM SETTINCS BASES (Continued) 2.
Average Power Range Monitor (Continued) the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally, the heat flux is in near-equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THEMAL POWER per a:inute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.
The 15% APRM trip remains active until the mode switch is placed in the Run position.
The APRM flow-biased trip system is calibrated using heat balance data taken during steady state conditions.
Fission chambers provide the basic input to the system and, therefore, the monitors respond directly and quickly J
to changes due to transient operation; i.e.,
the thermal power of the fuel will be less than that indicated by the neutron Yluz due to the time constants of the heat transfer. Analyses demonstrate that with only the 120% trip sitting, none of. the abnormal operational tr' nsients analyzed violates the a
fuel safety limit and there is substantial margin' from fuel damage.
Therefore, the use of the flow-referenced trip setpoint, with the 120% fixed setpoint as backup, provides adequate margins of safety.
The APRM trip setpoint was selected to provide an adequate margin for the Safety Limits and yec allows an operating margin that reduces the possiblility of unnecessary shutdowns. The flow-referenced trip setpoint must be adjusted by the specified formula in order to maintain these margins.
3.
Reactor Vessel Steam Dome Pressure-High High Pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids, thus adding reactivity.
The trip will quickly reduce the neutron flux, counteracting the pressure increase by decreasing heat generation.
The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the BRUNSWICX - UNIT 2 B 2-5 Amendment No. 151 l
2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) 3.
Reactor Vessel Steam Dome Pressure-High (Continued) pressure measurement compared to the highest pressure that occurs in the system during a transient. This setpoint is effective at low power / flow conditions when the turbine stop valve closure is bypassed. For a turbine trip under these conditions, the transient analysis indicates a considerable margin to the thermal hydraulic limit.
4 Resete-Vessel Water Level-Lov. Level di The reactor water level trip point was chosen far enough below the normal operating level to avoid spurious scrams but high enough above the fuel to assure that there is adequate water to account for evaporation losses and displacement of cooling following the most severe transients. This setting was also used to develop the thermal-hydraulic limits of power versus flow.
5.
Main Steam Line Isolation Valve-Closure The low pressure isolation of the main steam line trip was provided to give protection against espid depressurization and resulting cooldown of the reactor vessel. Advantage was taken of the shutdova feature in the run mode which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low pressures does not occur. Thus, the combination of the low pressure isolation and isolation valve closure reactor trip with the mode switch in the Run position assures the availability of neutron flux protection over the entire range of the Safety Limits.
In addition, the isolation valve closure trip with the mode switch in the Run position anticipates the pressure and flus transients which occur during normal or inadvertent isolation valve closure.
6.
Main Steam Line Radiation - High The Main Steam Line Radiation detectors are provided to detect a gross failure of the fuel cladding. When the high radiation is de:ected, a scram is initiated to reduce the continued failure of fuel cladding. At the same time, the Main Steam Line Isolation Valves are closed to limit the release of fission products. The trip setting is high enough above background radiation levels to prevent spurious scrams, yet low enough to promptly detect gross failures in tne fuel cladding.
BRUNSWICK - UNIT 2 B 2-6 Amendment No.151 l
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LIMITINC SAFETY SYSTEM SETTINC BASES (Continued) 7.
Dryvell Pressure-High High pressure in the dryvell could indicate a break in the nuclear process systaas. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant.
The trip setting was selected as lov as possible without causing spurious trips.
8.
Scram Discharre Volume Water Level-Hieh The scram discharge tank receives the water displaced by the motion of the control rod drive pistons during a reactor scram.
Should this tank fill up to a point where there is insufficient volume to accept the displaced water, control rod movement would be hindered. The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods when they are tripped.
9.
Turbine Stop Valve-closure The turbine stop valve closure trip antigi' pates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves. With a trip setting of 10% of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed.
10.
Turbine control Valve Fast closure. Control Oil Pressure - Low Lov turbine control valve hydraulic pressure will initiate the Select Rod Insert function and the preselected group of control rods will be fully inserted. Select Rod Insert is an operational aid designed to insert a predetermined group of control rods immediately following either a generator load rejection, loss of turbine control-valve hydraulic pressure, or by manual operator action using a switch on the R-T-C board. The assignment of control rods to the Select Rod Insert function is based on the start-up and fuel warranty service associated with each control rod pattern, on RCS considerations, and on a dynamic function of both time and core patterns.
Approximately ten percent of the control rods in the reactor will be assigned to the Select Rod Insert function by the operator.
This selection vill be accomplished by moving the rod scram test switch for thosu rods from the Normal position to the Select Rad Insert position.
BRUNSWICK - UNIT 2 B 2-7 Amendment No.151
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LIMITING SAFETY SYSTEM SETTINGS BASES (Continued) 10.
Turbine control Valve Fast Closure, Control Oil Pressure - Low (Continued)
Any rod selected for Select Rod Insert shall also have other rods in its notch group selected to ensure that the RSCS criteria of plus-minus one notch position equality is met when the rod pattern is greater than 50% ROD DD4SITY and THERMAL POWER < 20% of RATED THERMAL POWER.
It is possible that a rod pattern within these limits may occur af ter the Select Rod Insert function operates.
In order to reduce the number of reactor scrams, a 200 millisecond time delay, referenced from the low curbine control valve hydraulic pressure and Select Rod Insert signals, was incorporated to determine turbine bypass valve status via limit switches prior to initiating a reactor scram.
If the turbine bypass valves opened in < 200 milliseconds, the reactor scram was bypassed.
It was found that during certain reload cycles the MCPR penalties involved with this time delay were more penalizing than the number of scrans savedi therefore, CP&L requested and received NRC approval to set this time at "0" in Amendment No. 14.
With the timer set at "0", Select Rod Insert and RPS trip will be initiated simultaneously.
The control valve closure time is approximately twice as long as that for the stop valves which means that resulting transients, while similar, are less severe than for stop valve closure.
No fuel damage occurs, and reactor system pressure does not exceed the safety relief valve setpoint. This is an anticipatory scram and results in reactor shutdown before any significant increase in pressure or neutron flux occurs.
This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by turbine first-stage pressure.
BRUNSWICX - UNIT 2 B 2-3 Amendment No. 151
REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITINC CONDITION FOR OPERATION 3.1.4.3 Both Rod Block Monitor (RBM) channels shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER.
ACTION:
a.
With one RBM channel inoperable, POWER OPERATION may continue provided that either:
1.
The inoperable RBM channel is restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or 2.
The redundant RBM is demonstrated OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the inoperable REM is restored to OPERABLE status within 7 days, or 3.
THERMAL POWER is limited such that MCPR will remain above 1.04, l
assuming a single error that results in complete withdrawal of any single control rod that is capable of withdrawal.
Otherwise, trip at least one rod block monitor channel; b.
With both RBM channels inoperable, trip at least one rod block monitor channel within one hour.
SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be d wonstrated OPERABLE by performance of s CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies and during the OPERATIONAL CONDITIONS specified in Table 4.3.4-1.
BRUNSWICK - UNIT 2 3/4 1-17 Amendment No.18, 7M,151
i POWER DISTRISUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of cure flow, shall be equal to or greater than the MCPR limit times the Xg shown in Figure 3.2.3-1 with the following MCPR limit adjustments:
Beginning-of-cycle (50C) to end-of-cycle (EOC) minus 2000 MWD /t with a.
ODYN OPTION A analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:
1.
NCPR for P8 x 8R fuel = 1.29 2.
MCPR for BPS x 8R fuel = 1.29 3.
MCPR for CE8 fuel = 1.29 b.
EOC minus 2000 MWD /c to EOC with ODYN OPTION A analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below 1.
MCPR for P8 x 81 fuel = 1.30 2.
MCPR for BP8 x 81 fuel = 1.30 3.
MCPR for CE8 fuel = 1.30 BOC to EOC minus 2000 MWD /t with ODYN OPTION 8 analyses in effect and c.
the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:
1.
MCPR for P8 x 8R fuel = 1.22 2.
MCPR for BP8 x 8R fuel = 1.22 3.
MCPR for CE8 fuel = 1.22 d.
EOC minus 2000 MWD /t to EOC with ODYN OPTION B analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:
1.
MCPR for P8 x 8R fuel = 1.26 2.
MCPR for BP8 x 8R fuel = 1.26 3.
MCPR for CE8 fuel = 1.26 APPLICABILITY:
OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER BRUNSWICK - UNIT 2 3/4 2-8 Amendment No. 707, 773, 151
Q E
TABLE 3.2.3.2-1 E
n TRANSIENT OPERATINC LINIT NCPR VALUES M
e S
FUEL TYPE TRANSIENT U
P8x8R BP3x8R CE8 u
i M)NPRESSURIZATION TRANSIENTS BOC + ECC 1.22 1.27 1.22 PRESSURIZATION TRANSIENTS
~
I R
A g
A B
A g
BOC + EOC - 2000 1.29 1.22 1.29 1.22 1.29 1.22
.L EOC - 2000 + EOC 1~.30 1.26 1.30 1.26 1.30 1.26 t
-l 3
m l
3 O
w
~
~
REACTIVITY CONTROL SYSTEM BASES CONTROL RODS (Cont'inued) potential effects of the rod ejection accident are limited. The ACTION statements permit variations from the basic requirements but at the same time
. impose more restrictive criteria for continued operation. A limitation on
)
inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum.
The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigs:ed en a ti ely basis.
l l
Damage within the control rod drive mechanism could be a generic problem; therefore, with a control rod innovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.
Control rods that are inoperable for other reasons are permitted to be i
taken out of service, provided that those in the non-fully-inserted position are consistent with the SHUTDOWN MARGIN requirements.
Thendaberofcontrolrodspermittedtobeknoperablecouldbemorethan the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem, and the reactor must be shut down for investigation and resolution of the problem.
The control rod system is analysed to bring the reactor suberitical at a rate fast enough to prevent the MCPR from becoming less than 1.04 during the l
Limiting power transient analyzed in Section 14.3 of the FSAR. This analysis shows that the negative reactivity races resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MPCR remains greater than 1.04.
The occurrence of l
scram times longer than those specified should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.
Control rods with inopcrable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion BRUNSWICK - UNIT 2 B 3/4 1-2 Amendment No. 33, 151
POWER DISTRIBUTION LIMITS BASES 3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a TOTAL PEAKINC FACTOR of 2.39 for P8s8R and 3P8x8R fuel and 2.44 for CE8 fuel. The scram setting and rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than 1.0 in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the combination of ThiER.h POWER and pear fiax indicates a TOTAL PEAKING FACTOR greater than 2.39 for P8x8R and 8P8x8R fuel and 2.48 for CES fuel. This adjustaant may be accomplished by increasing the APRM gain and thus reducing the slo 9e and intercept point of the flow referenced APRM high fluz scram curve by the reciprocal of the APRM gain change. The method used to determine the design TPF shall be consistent with the method used to determine the WTPF.
3/4.2.3 MINIMUM CRITICAL POWER RATIO Tha required operating limit MCPts at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safe operationaltransients.gLimitMCPRof1.04,andananalysisofabnormal l
For any abnormal operating transient analysis ovaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any cias during the transient, assuming an instrument trip setting as given in Specification 2.2.1.
To assure that the fuel cladding integrity Safety Limit is not exceeded during I
cny anticipated abnormal operational transient, the most limiting transients have been analysed.co determine which result la the largest reduction la CRITICAL POWER RATIO (CPR). The type of transients evaluated were less of flow, increase in pressure and power, positive reactivity insertion, and coolant camperature decrease.
Unless otherwise stated in cycle specific reload analyses, the limiting transient which determines the required steady state MCPR limit is the turbine trip with failure of the turbine bypass. This transient yields the largest 4 MCPR. Prior to the analysis of abnorest operational transients an initial fuel bundle MCPR was determined. This parameter is based on the bundle flow calculated by a CR multichannel steady describedinSection4.4ofNED0-20360gateflowdistributionmodelas and on core parameters shown in Referencs 3, response to Items 2 and 9.
BRUNSWICK - UNIT 2 8 3/4 2-3 Amendment flo. 707. 173, 151
p UNITED STATES f'
h NUCLEAR REGULATORY COMMISSION h
j WASHINGTON, D. C. 20555 e
\\...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.151 TO FACILITY GPERATING LICENSE NO. OPR-62 CAROLINA POWER & LIGHT COMPANY et al.
BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 DOCKET N0. 50-324
1.0 INTRODUCTION
By letter dated February 3, 1988, as supplemented March 30, 1988, the Carolina Power & Light Company submitted a request for changes to the Brunswick Steam Electric Plant, Unit 2, (BSEP-2) Technical Specifications (TS) to incorporate upgraded Minimum Critical Power Ratio (MCPR) values applicable to the operation of BSEP-2, Cycle 8.
On March 30, 1988, the licensee provided clarification with respect to NRC staff concerns.
In addition, the submittal provided a changed MCPR value which had been inadvertently omitted in the February 3, 1988 submittal.
The March 30, 1988 submittal did not substantially change the action noticed, or alter the staff's initial determination published, in the Federal Register on March 9, 1988.
2.0 EVALUATION 2.1 MCPR Safety Limit The MCPR fuel cladding integrity safety limit of 1.07, currently used for BSEP-2 reload cores, was established in 1978.
This safety limit was designed to provide a level of conservatism for establishing operating limit MCPR values, based on fuel design characteristics typical of those utilized at that time. The level of conservatism built into the safety limit provides adequate margin to assure that more thm 99.9% of the fuel rods in the core are expected to avoid boiling transition.
The increase in conservatism has been recognized because of current fuel designs. An updated safety limit of 1.04, specified in Amendment 14 to hEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (GESTAR II), has been reviewed and approved by the NRC for D-lattice plants when applied to second successive reload. cores of P8x8R, BP8x8R, GE8x8E or GE8x8EB fuel types with high bundle R-factor ( 1.04).
BSEP-2 is such a 0-lattice plant, with Cycle 8 being the third successive reload core with high bundle R-factor ( 1.04) fuel designs. Therefore, the staff finds that the proposed amendment for the changing the MCPR safety limit specified in the BSEP-2 TS from 1.07 to 1.04 is acceptable.
W m t n a;~~=
ag.
. 2.2 Operating Limit MCPR Values Operating limit MCPR (0LMCPR) values are designed to limit the conse-quences of operational transients previously evaluated. Since the upgraded safety limit MCPR was reviewed and generically approved by the staff, adjustment of the operating limit MCPR values was proposed by CP&L for a plant specific application.
The licensee also provided a letter on March 30, 1988, to clarify the staff's concern on the deletion of the MCPR adder.
The staff has reviewed the February 3 and March 30, 1988 submittals and found that the Cycle 8 reload for BSEP-2 meets the criteria set for the application of the upgraded safety limit MCPR and that the clarification for deletion of the MCPR adder is acceptable.
Therefore, the proposed adjustment of the OLMCPR values for BSEP-2 Cycle 8 reload is acceptable.
2.3 Technical Specifications The Technical Specification changes are for the most part related to the approved upgraded safety limit MCPR.
Details of the specification changes follow:
(1) Specifications 2.1.2, 3.1.4.3 and 3.3.4 and Bases 2.0, 2,2.1, 3/4.1.3, 3/4.2.3.
The amendment changes the MCPR safety limit, specified in the BSEP-2 TS, from 1.07 to 1.04.
This change is based on the generically approved amendment to GESTAR II. Therefore, the staff finds the proposed change is acceptable.
(2) Bases 3/4.2.3 and Table 3.2.3.2-1 New operating limit MCPR values correspond to the generically approved upgraded MCPR safety limit of 1.04.
The new Oi.MCPR value is 0.05 smaller than that of the original value given in the BSEP-2 Cycle 8 reload analysis.
This 0.05 difference is due to 0.03 gained from the upgraded MCPR safety limit and 0.02 MCPR adder deleted from this proposal. The BSEP-2 licensee stated in the March 30, 1988 letter that the operator will take all the necessary corrective actions to bring the reactor to a safe operating condition by reducing the reactor power in case of opera-tional occurrences, such as a main steam line isolation valve out-of-service or a feedwater heater out-of-service. This supports the deletion of the 0.02 MCPR adder for BSEP-2 Cycle 8 operation.
Therefore, we find the proposed new OLMCPR values are acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
S This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the i
types, of any effluents that may be released off site; and that there should be no significant increase in individual or cumulative occupa-tional radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration, and there has been no public comment on such finding.
Accordingly, this amendnent meets the eligibility) criteria for cate-gorical exclusion set forth in 10 CFR 651.22(c)(9.
Pursuant to 10 CFR 951.22(b), no environmental impact statemer t or environmental assessment need be prepared in connection with the issuance of the amendment.
4.0 CONCLUSION
The Comission made a proposed determination that this amendment involves no significant hazards consideration which was published in the FEDERAL REGISTER (53 FR 7585) on March 9, 1988, and consulted with the State of North Carolina.
No public comments or requests for hearing were received and the State of North Carolina did not have any coments.
The staff 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, and (2) such activities will be conducted in compliance with the Comission's regula-tions, and the issuance of the amendment will not be inimical to the ccmon defense and security or to the health and safety of the public.
Principal Contributor:
T. Huang Dated:
April 12, 1988