BSEP-96-0123, Application for Amends to Licenses DPR-62 & DPR-71,revising TS to Allow Uprate of Facilities to 105% of Rated Thermal Power.Proprietary & Nonproprietary Versions of GE Power Uprate SAR for Bsep... Encl.Proprietary GE Rept Withheld
| ML20101N468 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/02/1996 |
| From: | Campbell W CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20013A238 | List: |
| References | |
| BSEP-96-0123, BSEP-96-123, NUDOCS 9604080397 | |
| Download: ML20101N468 (27) | |
Text
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CP&L Carolina Power & Light Company William R. Campbell PO Box 10429 Vice President Southport NC 28461-0429 Brunswick Nuclear Plant SERIAL: BSEP 96-0123 10 CFR 50.90 TSC 94TSB16 i
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS 105% THERMAL POWER UPRATE Gentlemen:
In accordance with the Code of Federal Regulations, Title 10, Parts 50.90 and 2.101, Carolina Power & Light Company hereby requests a revision to the Technical Specifications for the Brunswick Steam Electric Plant (BSEP), Unit Nos.1 and 2. These proposed amendments revise the BSEP Technical Specifications to allow uprate of the units to 105% of rated thermal power. The License Topical Report supporting this amendment request was submitted by .790|letter dated November 20,1995]], and is supplemented herein to support this license amendment request. In addition, setpoints and allowable values that are not directly impacted by power uprate are being revised in this submittal to reflect calculations reconstituted during the power uprate project.
Enclosure 1 provides a detailed description of the proposed changes and the basis for the changes.
Enclosure 2 detalis the basis for the Company's determination that the proposed changes do not involve a significant hazards consideration.
l Enclosure 3 provides an environmental evaluation which demonstrates that the proposrA l amendments meet the eligibility for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental assessment needs to be prepared in connection with the issuance of the amendment.
Enclosure 4 provides the marked-up Technical Specification pages for Unit 1.
Enclosure 5 provides the marked-up Technical Specification pages for Unit 2.
g Enclosure 6 provides Supplement 1 to NEDC-32466P, proprietary version inc!uding affidavit. In enrdance with 10 CFR 2.790, this enclosure siiouid be withheld from public disclosure.
9604080397 960402 PDR ADOCK 05000324 h e
(},'f f ll Tel 910 457 2496 Fox 910 457 2803 /
Document Control Desk BSEP 96-0123 / Page 2 Enclost re 7 provides Supplement 1 to NEDO-32466, non-proprietary version.
Typed pages for these changes will be provided at a later date.
Carolina Power & Light Company is providing, in accordance with 10 CFR 50.91(b), Mr. Dayne H. Brown of the State of North Carolina, Division of Radiation Protection, with a copy of the proposed license amendments.
CP&L plans to implement power uprate during the Fall 1996 Unit 1 refuel outage (B111R1) and Fall 1997 Unit 2 refuel outage (B213R1); therefore, CP&L requests that the proposed amendment, once approved by the NRC, be issued with an effective date upon completion of the B111R1 and B213R1 refuel outages.
Please refer any questions regarding this submittal to Mr. George Honma at (910) 457-2741.
Sincerely, h .
William R. Campbell VPUkah
Enclosures:
- 1. Basis for Change Request
- 2. 10CFR50.92 Evaluation
- 3. Environmental Considerations
- 4. Marked-up Technical Specification Pages - Unit 1
- 5. Marked-up Technical Specification Pages - Unit 2
- 6. NEDC-32466P Supplement 1(Proprietary version with affidavit)- WITHHOLD FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.790
- 7. NEDO-32466 Supplement 1(Non-Proprietary version)
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Document Control Desk BSEP 96-0123 / Page 3 William R. Campbell, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, and agents of Carolina Power & Light Company.
I 9 11Ab w Notary $eal) d My commission expires: b lJ,199/,
pc: Mr. D. H. Brown, Director, Division of Radiation Protection, NCDEHNR Mr. S. D. Ebneter, Regional Administrator, Region ll Mr. D. C. Trimble, Jr., NRR Project Manager - Brunswick Units 1 and 2 Mr. C. A. Patterson, Brunswick NRC Senior Resident inspector The Honorable H. Wells, Chairman - North Carolina Utilities Commission i
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General Electric Company AFFIDAVIT I, George B. Stramback, being duly sworn, depose and state r follows:
(1) I am Project Manager, Regulatory Services, General Electric Company ("GE") and have been deLgated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply ~for its withholding.
(2) The information sought to be withheld is contained in the proprietary report NEDC-32466P, Supplement 1, Power Uprate Safety Analysis Reportfor Brunswick Steam Electric Plant Units 1 and 2 (Supplement 1), Class 3 (GE Proprietary Information),
dated March 1996. This document, taken as a whole, constitutes a proprietary compilation ofinformation, some ofit also independently proprietary, prepared by '
the General Electric Company. The independently proprietary elements are delineated by bars marked in the margin adjacent to the specific material.
(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FCIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec. 1905, r.nd NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and 2.790(d)(1) for " trade secrets and commercial or fi _.ncial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclow e is here sought is all " confidential commercial ,
information", and some portions also qualify under the narrower definition of " trade l secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission 975F2d871 (DC Cir.1992), and Public Citizen Health Research Groun
- v. FDA,704F2dl280 (DC Cir.1983).
(4) Some examples of categories of information which fit into the definition of proprietary information are:
- a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention ofits use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies; Exempt from Public Disclosure GBS-96 ~,-afCPLpu2.d M da M ageI Per 10 CFR 2.790
i I
- b. Information which, if used by a competitor, would reduce his expenditure of !
resources or improve his compentive position in the design, manufacture, !
shipment, installation, assurance of quality, or licensing of a similar product;
- c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;
- d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric;
- e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
Both the compilation as a whole and the marked independently proprietary elements incorporated in that compilation are considered proprietary for the reasons set forth in paragraphs (4)a. and (4)b., above.
(5) The information sought to be withheld is being submitted to NRC in confidence.
The information (both the entire body of information in the form compiled in this document, and the marked individual proprietary compilations or elements) is of a sort customarily held in confidence by GE, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) followmg.
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.
(7) The procedure for approval of extemal release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent I
authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
GBS-96-3-afCPLpu2. doc Affidavit Page 2 Exempt From Public Disclosure Per 10 CFR 2.790
(8) The information identified by bars in the margin is classified as proprietary because it contains detailed results and conclusions from these evaluations, utilizing analytical models and methods, including computer codes, which GE has developed, obtained NRC approval of, and applied to perform evaluations of transient and accident conditions in the GE Boiling Water Reactor ("BWR"). The development and approval of these system, component, and thermal-hydraulic models and computer codes was achieved at a significant cost to GE, on the order of several million dollars.
The remainder of the information identified in paragraph (2), above, is classified as proprietary because it constitutes a confidential compilation of information, including detailed results of analytical models, methods, and processes, including computer codes, and conclusions from these applications, which represent, as a whole, an integrated process or approach which GE has developed, obtained NRC approval of, and applied to perform evaluations of the safety-significant changes necessary to demonstrate the regulatory acceptability of a given increase in licensed power output for a GE BWR. The development and approval of this overall approach was achieved at a s.gnificant additional cost to GE, in excess of a million dollars, over and above the very large cost of developing the underlying individual proprietan/ analyses.
To effect a change to the licensing basis of a plant requires a thorough evaluation of the impact of the change on all postulated accident and transient events, and all other regulatory requirements and commitments included in the plant's Updated Final ,
Safety Analysis Report. The analytical process to perform and document these evaluations for a proposed power uprate was developed at a substantial investment in GE resources and expertise. The results from these evaluations identify those BWR systems and components, and those postulated events, which are impacted by the changes required to accommodate operation at increased power levels, and, just as importantly, those which are not so impacted, and the technical justification for not considering the latter in changing the licensing basis. The scope thus determined forms the basis for GE's offerings to support utilities in both performing analyses and providing licensing consulting services. Clearly, the scope and magnitude of effort of any attempt by a competitor to effect a similar licensing change can be narrowed considerably based upon these results. Having invested in the initial evaluations and developed the solution strategy and process described in the subject document, GE derives an important competitive advantage in selling and performing these services. However, the mere knowledge of the impact on each system and component reveals the process, and provides a guide to the solution strategy.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR technology base, and its commercial value extends beyond the original GBS-96-3-afCPLpu2. doc Exempt From Public Disclosure AfTidavit Page 3 Per 10 CFR 2.790
l . e t E :
!~ l l development cost. The value of the technology base goes beyond the extensive I L physical database and analytical methodology, and includes development of the !
l expertise to determine and apply the appropriate evaluation process. In addition, the l technology base includes the value derived from providing analyses done with NRC- ;
approved methods, including justifications for not including certain analyses in ,
applications to change the licensing basis.
GE's competitive advantage will be lost if its competitors are able to use the results !
of the GE experience to avoid fruitless avenues, or to normalize or verify their own j process, or to claim an equivalent understanding by demonstrating that they can j l arrive at the same or similar conclusions. In particular, the specific areas addressed i
! by any document and submittal to support a change in the safety or licensing bases l l
of the plant will clearly reveal those areas whcre detailed evaluations must be performed and specific analyses revised, and also, by omission, reveal those areas [
not so affected. !
l While some of the underlying analyses, and some of the gross structure of the :
, process, may at various times have been publicly revealed, enough of both the !
analyses and the detailed structural framework of the process have been held in confidence that this information, in this compiled form, continues to have great ,
competitive value to GE. The value of this information to GE would be lost if the information as a whole, in the context and level of detail provided in the subject document, were disclosed to the public. Making such information available to - r competitors without their having been required to undertake a similar expenditure of resources, including that required to determine the areas that are not affected by a i power uprate and are therefore blind alleys, would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing its analytical ;
processes.
)
P f
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I Exempt From Public Disclosure GBS-96-3-afCPLpu2. doc Per 10 M W Affidavit Page 4 l l
STATE OF CALIFORNIA )
) ss:
COUNTY OF SANTA CLARA )
George B. Stramback, being duly sworn, deposes and says:
That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.
Executed at San Jose, California, this4 6 day of M ,1996.
P .k. des
' " Go6rge B. stramback General Electric Company Subscribed and sworn before me thisd day of Y 0.cch ,1996.
3 . ... PAULA F. HUSSEY
= No? ub$c omio e M4 C wc # E docc $ wa Notary Public, State of Californid Exempt From Public Disclosure Per 10 CFR 2.790 GBS-96-3-afCPLpu2. doc Affidavit Page 5
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ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC Pt. ANT UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 ;
OPERATING LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS 105% THERMAL POWER UPRATE Backaround of Prooosed Chances This proposed amendment consists of a number of changes which will permit uprated power operation of the Brunswick Steam Electric Plant Units 1 and 2. In addition, setpoints and allowable values, that are not directly impacted by power uprate, are being revised in this submittal to reflect calculations reconstituted during the power uprate project.
Brunswick is a dual-unit General Electric (GE) BWR/4 with a Mark I containment. Other plants similar in design to Brunswick have received NRC approval to operate at uprated conditions. 1 The Brunswick units, like most BWR plants, have equipment and system capability to accommodate steam flow rates at least 5% above the original rating. Because of the significant ,
economic advantages of operating at a higher power level, Carolina Power & Light Company ['
(CP&L) is proposing a permanent amendment to the operating license for each Brunswick unit, which will enable them to be operated at power levels up to 105% (2558 MWt) of current rated power level. f The analyses and evaluations supporting the changes directly related to power uprate were completed using the guidelines in General Electric Topical Report NEDC-31897P-A, " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate", (Reference 2). Certain issues are evaluated generically and have been submitted to the NRC in General Electric Topical Report NEDC-31984P, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate", (Reference 3). Both of the above GE reports have been approved by the NRC and form the basis for Brunswick's thermal power uprate evaluation. A Brunswick plant specific analysis, NEDC-32466P, " Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2", (Reference 1) has been performed by GE. This report was submitted for -
NRC review in November of 1995. Enclosure 6 to this letter provides Supplement 1 to NEDC-32466P.
' The method for achieving higher reactor power is to extend the power / flow map by increasing core flow along existing flow control lines; however, there will not be an increase in the maximum recirculation flow limit over the pre-uprate value. Additional control rod withdrawal provides the increase in reactor power. Uprated operation will also require slightly higher reactor vessel dome pressure to provide adequate inlet pressure conditions at the turic v (accounting for the larger pressure drop through the steam lines at higher flow) and to provide sufficient pressure control and turbine flow capability.
Twenty-six setpoint calculations were reconstituted during the Power Uprate Project. These setpoint calculations were performed in accordance with CP&L Design Guide DG-Vill.0050, which was developed from techniques and recommendations put forth in ISA's Recommended Practice S67.04, Parts 1 and 2.
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r The instrument setpoint calculations reconstituted during the power uprate project include .
several conservatisms and conservative assumptions with respect to current operational !
setpoints, including:
o Instrument drift is based on a 30 month calibration frequency. Current plant setpoints are based on 22.5 months.
o Use of existing plant calibration tolerances rather than using the minimum achievable calibration tolerance.
o Power Uprate to 105% of current rated thermal power (2558 MWt).
CP&L has reviewed the setpoint calculations reconstituted under the Power Uprate Project and ;
has not identified any operability concerns for our current licensed power level, based on the )
conservatisms and bounding assumptions of the power uprate calculations. CP&L is
{
reconstituting the remainder of the TS functional unit setpoint calculations under the LSSS l
setpoint reconstitution and 24-month fuel cycle projects. CP&L intends to include TS changes associated with these calculations in the improved Technical Specification conversion submittal.
To date, CP&L has identified certain physical plant changes necessary to support power uprate at the Brunswick Plant. These changes include instrument setpoint changes, indicating meter ,
scale changes for the RWCU System flow and Main Steam Flow indicators, Leak Detection, Process Computer, ERFIS, and Feedwater System software changes, SRV setpoint changes, and changes to the plant simulator.
J Attachment A is a table listing the TS changes proposed by this amendment request. The numbers listed in the discussion of change section refer to the change number in the table.
Those changes directly related to uprating the thermal power of the Brunswick Plant are uniquely identified in the table. These changes include changes resulting from operating parameter changes (e.g. temperature, pressure, flow) and those necessary to maintain existing operational margins (e.g. Vessel Pressure High ATWS-RPT). A discussion of the changes in Attachment A follows.
Prooosed Chanae No.1
- 1. Rated Thermal Power is increased to 2558 MWt in the Unit 1 and 2 Operating Licenses and }
Section 1.1 (Definitions), Page 1-6, of the Technical Specifications. i I
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- v. --,
l Basis for Chanae No.1 This increase and redefinition of rated thermal power for Brunswick Units 1 and 2 follows the generic guidelines of NEDC-31897P-A (Reference 2). NEDC-31897P-A provides the j generic licensing criteria, methodology, and a defined scope of analytical evaluations and I equipment revists to be performed to demonstrate the ability to operate safely at the l
uprated power ;wel. Technical Specifications parameter values, which are expressed as a I percentage of rated reactor thermal power or steam flow, were not changed since the uprated values were used in the bounding analyses and evaluations required by Reference 2 unless otherwise specified in this submittal. NEDC-32466P, " Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2" (Reference 1) provides the results of the evaluations supporting the proposed uprated power operation consistent l with the methodology presented in Reference 2. The report concludes that an uprated l power rating of 2558 MWt can be achieved without a significant impact to equipment or I safety analyses. 1 Prooosed Chanae Nos. 2. 3. & 23 2 Table 2.2.1-1, Function 2.b, " Flow Biased Simulated Thermal Power - High" scram function Allowable Value is being changed from s (0.66W + 67%) with a maximum of s 115.5% of RATED THERMAL POWER to s (0.66W + 61%) with a maximum of s 115.3% of RATED THERMAL POWER. The Trip Setpoint is being changed from s (0.66W + 64%) with a >
maximum of s 113.5% of RATED THERMAL POWER to s (0.66W + 59.6%) with a maximum of s 113.6% of RATED THERMAL POWER. Figure 2.2.1-1,"APRM Flow Bias Scram Relationship to Normal Operating Conditions" will be rescaled.
3 Table 2.2.1-1 Function 2.c, " Fixed Neutron Flux -High" scram function Allowable Value is being changed from s 120% of RATED THERMAL POWER to s 118% of RATED THERMAL POWER. The Trip Setpoint is being change from s 120% of RATED THERMAL POWER to s 116.3% of RATED THERMAL POWER.
23 Table 3.3.4-2 Function 1.a "APRM Upscale (Flow Biased) Rod Block" Allowable Value is being changed from s (0.66W + 61%) with a maximum of s 110% o' RATED THERMAL ,
POWER to s (0.66W + 56%) with a maximum of s 111% of RATED THERMAL POWER. l The Trip Setpoint is being changed from s (0.66W + 58%) with a inaximum of s 108% of RATED THERMAL POWER to s (0.66W + 54.6%) with a maximum of s 109.3% of RATED THERMAL POWER.
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I Basis for Chanae Nos. 2. 3. & 23 The analytical limit for the Flow Biased Simulated Thermal Power - High and Fixed Neutron i Flux - High scram functions and APRM Upscale Rod Block do not change for power uprate, i since they are based on a percentage of rated thermal power. The TS Allowable Value and l Nominal Trip Setpoint have been recalculated due to the power / flow map boundary change. The power / flow map changes are an extension of the boundary along the 120%
rod line from the existing 100% thermal power (2436 MWt) to 105% thermal power (2558 ,
MWt). From this point the boundary is " clamped" so that any increase in flow will not result in an increase in thermal power (see Reference 1). This clamp occurs at approximately 75% flow currently and will change to approximatley 81% flow under power uprate. This change in reference flow requires a change in the zero flow intercept point. The revised TS Allowable Value and Trip Setpoint, calculated in accordance with CP&L's setpoint methodology, ensure adequate margin exists to the analytical limit.
Prooosed Chance No. 4
]
- 4. Table 2.2.1-1, Function 3;
- Reactor Vessel Steam Dome Pressure - High" allowable value is i changed from s 1045 psig to s 1070 psig. The Trip Setpoint is changed from s 1045 psig to s 1067.9 psig.
Basis for Chance No. 4 The reactor steam dome high pressure scram limit is increased as a result of the increase in the reactor steam dome normal operating pressure. Operating pressure for uprated power is increased to assure that satisfactory reactor pressure control is maintained. The operating pressure was chosen on the basis of steam line pressure drop characteristics and i the steam flow capability of the turbine. Satisfactory reactor pressure control requires an )
adequate flow margin between the uprated operating condition and the steam flow capability of the turbine control valves at their maximum stroke. An operating dome i pressure of 1030 psig, which is 25 psi higher than the current operating dome pressure, is !
expected. Therefore, the high pressure scram is increased the same amount to preserve i existing margins to reactor trips. !
The high pressure scram terminates a pressurization transient not terminated by direct scram or high neutron flux scram. The setting is maintained above the nominal reactor vessel operating pressure and below the specified analytical trip limit used in the safety analyses. The revised high pressure scram setpoint will preserve the hierarchy of pressure setpoints. This means that the high pressure scram setpoint will remain below the opening setpoint of the safety / relief valves (SRVs). The SRV nominal setpoints are also increased 25 psi, as discussed in proposed change 27. This hierarchy of setpoints provides assurance that there is a low probability of opening an SRV without scram intervention.
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i Prooosed Chanae Nos. 5. 6. & 9 5 . Table 2.2.1-1 Function 4, Table 3.3.2-2 Function 1.a.1 and Function 5.a. Table 3.3.3-2, Function 4.c; " Reactor Vessel Water Level- Low, Level 1" Allowable Value is being i
changed from a +162.5 inches to a +153 inches. The Trip Setpoint is being changed from ;
! a +162.5 inches to a +153.2 inches. +
6 Table 3.3.2-2 Function 1.a.2, Table 3.3.3-2 Trip Function 1.a, Function 2.b, Function 4.b;
- Reactor Vessel Water Level- Low, Level 3" Allowable Value is being changed from a +2.5
, inches to a +13 inches. The Trip Setpoint is being changed from a +2.5 inches to a +14.1 J' inches.
a
, 9. Table 3.3.2-2 Function 2.c, Function 3.e, Table 3.3.3-2 Function 3.a, Table 3.3.6.1-2 i Function 1, Table 3.3.7-2 Function 1; " Reactor Vessel Water Level- Low, Level 2"
)
, Allowable Value is being changed from a +112 inches to a +103 inches. The Trip Setpoint
- is being changed from a +112 inches to a +104.1 inches.
1 Basis for Chanae Nos. 5. 6. & 9 i
For power uprate the setpoints for the reactor vessel water level instruments were
, calculated based on the new normal operating pressures and temperatures. CP&L's setpoint methodology was applied in order to determine the revised TS A!!owable Value and l
. Nominal Trip setpoint. The analytical limits for the ECCS functions are contained in the BNP 5 specific SAFER /GESTR-LOCA analysis (Reference 5), which is the design basis analysis I i for Brunswick's ECCS system's response to the DBA LOCA. In this analysis the analytical l limit was established for Low Level 2 (LL2) at +92 inches, and Low Level 3 (LL3) at +2.5 4
l inches (all values are referenced to instrument zero). The power uprate OPL-3 (Reference :
- 6) described the analytical limits for the reactor scram parameters. These parameters are l used in the limiting transient analysis to confirm that at power uprate conditions, no new accidents or an increase in consequences result. The OPL-3 established the analytical limit for Low Level 1 (LL1) at + 150 inches. Applying CP&L's setpoint methodology and using the analytical limit from SAFER /GESTR-LOCA and OPL-3, revised TS Allowable Values i and Trip Setpoints were calculated and are reflected in this proposed change.
1 l Prooosed Chanae Nos. 7. 8.10 - 22. & 31 7 Table 3.3.2-2 Trip Function 1.c.3, " Main Steam Line Flow - High" Allowable Value is being changed from s 140% of rated flow to s 138% of rated flow. The Trip Setpoint is being changed from s 140% of rated flow to s 137% of rated flow.
i 1
8 Table 3.3.2-2 Trip Function 1.e; " Condenser Vacuum - Low" Allowable Value is being changed from a 7 inches Hg vacuum to 2 7.5 inches Hg vacuum. The Trip Setpoint is being changed from 2 7 inches Hg vacuum to 2 7.6 inches Hg vacuum.
1 i 10 Table 3.3.2-2 Trip Function 4.a.1; "HPCI Steam Line Flow- High" Allowable Value is being i changed from s 300% of rated flow to s 275% of rated flow. The Trip Setpoint is being changed from s 300% of rated flow to s 272% of rated flow.
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11 Table 3.3.2-2 Trip Function 4.a.3; "HPCI Steam Supply Pressure - Low" Allowable Value is being changed from = 100 psig to a 104 psig. The Trip Setpoint is being changed from a 100 psig to a 106.6 psig.
12 Table 3.3.2-2 Trip Function 4.a.6; "HPCI Exhaust Diaphragm Pressure - High" Allowable Value is being changed from s 10 psig to s 9.0 psig. The Trip Setpoint is being changed from s 10 psig to s 8.5 psig.
13 Table 3.3.2-2 Trip Function 4.b.1; *RCIC Steam Line Flow - High" Allowable Value is being changed from s 300% of rated flow to s 275% of rated flow. The Trip Setpoint is being changed from s 300% of rated flow to s 272% of rated flow.
14 Table 3.3.2-2 Trip Function 4.b.3; "RCIC Steam Supply Pressure - Low" Allowable Value is being changed from 2 50 psig to a 53 psig. The Trip Setpoint is being changed from x 50 psig to a 55.6 psig. t 15 Table 3.3.2-2 Trip Function 4.b.6; *RCIC Exhaust Diaphragm Pressure - High" Allowable Value is being changed from s 10 psig to s 6.0 psig. The Trip Setpoint is being changed from s 10 psig to s 5.5 psig. .
i 16 Table 3.3.2-2 Function 5.b;" Reactor Steam Dome Pressure - High" Allowable Value is being changed from s 140 psig to s 137 psig. The Trip Setpoint is being changed from !
s 140 psig to s 130.8 psig.
17 Table 3.3.3-2 Function 1.b; " Reactor Steam Dome Pressure - Low" Allowable Value is being changed from 410 i 15 psig to a 404 psig. The Trip Setpoint is being changed from j 410 i 15 psig to 2 406.7 psig. i 18 Table 3.3.3-2 Function 2.d.1; " Reactor Steam Dome Pressure - Low / RHR Pump Start and LPCI Valve Actuation" Allowable Value is being changed from 410115 psig to a 404 psig.
The Trip Setpoint is being changed from 410 i 15 psig to a 406.7 psig.
19 Table 3.3.3-2 Function 2.d.2;" Reactor Steam Dome Pressure Low I Recirculation Pump Discharge Valve Actuation" Allowable Value is being changed from 310 i 15 psig to a 304 psig. The Trip Setpoint is being changed from 310115 psig to a 306.7 psig.
20 Table 3.3.3-2 Function 4.d;" ADS Timer" Allowable Value is being changed from s 120 seconds to s 108 seconds. The Trip Setpoint is being changed from s 120 seconds to s 83 seconds.
21 Table 3.3.3-2 Trip Function 4.e; " Core Spray Discharge Pressure - High" Allowable Value is being changed from a 100 psig to a 102 psig. The Trip Setpoint is being changed from a 100 psig to 2112.1 psig.
22 Table 3.3.3-2 Function 4.f; "RHR (LPCI MODE) Pump Discharge Pressure - High" Allowable Value is being changed from 2100 psig to a 102 psig. The Trip Setpoint is being changed from a 100 psig to a 111.1 psig.
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31 Table 3.3.2-2 Trip Function 1.c.4, " Main Steam Line Flow - High" (Unit 2 only) Allowable Value is being changed from s 40% of rated flow to s 32% of rated flow. The Trip Setpoint is being changed from s 40% of rated flow to s 30% of rated flow.
Basis for Chance Nos. 7. 8.10 - 22. & 31 Change numbers 7,10,13, & 31 are as a direct result of power uprate. The Analytical Limits for these functions do not change, since they are based on a percentage of flow; however, applying CP&L's setpoint methodology and utilizing the uprated design flow, revised TS Allowable Values and Trip Setpoints were calculated.
The remaining proposed changes are not as a direct result of power uprate. These changes are included in this submittal to eliminate the need for additional administrative control.
Setpoint calculations, utilizing CP&L's setpoint methodology were performed to determine the new TS Allowable Value and Nominal Trip Setpoint for each function. The Allowable Values for these functions have changed in a more conservative direction, providing revised margin between the Analytical Limit and the new Allowable Value.
Change numbers 17,18, & 19 have been revised to eliminate the range currently shown in the Technical Specifications. These functions are permissives for the low pressure ECCS systems, which ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure. This design basis information is accounted for in the setpoint calculation for these functions; therefore, the TS Allowable Values and Trip Setpoints only reflect the minimum required pressure.
Procosed Chance No. 24 24 Table 3.3.6.1-2 Trip Function 2; " Reactor Vessel Pressure - High" Allowable Value is being changed from s 1120 psig to s 1143 psig. The Trip Setpoint is being changed from s 1120 psig to s 1137.8 psig.
Basis for Chanae No. 24 The ATWS-RPT for high pressure trip function initiates a trip of the recirculation pumps in the event of an ATWS transient. The recirculation pump trips reduce reactivity by reducing core flow during the initial part of the ATWS event.
The current analytical limit for the ATWS-RPT high pressure trip is 1120 psig. This value was increased 50 psi in the power uprate ATWS evaluation to account for the increase in vessel operating pressure, SRV setpoints, and maintain operational margin. The current allowable value in the Technical Specifications is s 1120 psig. Raising the ATWS-RPT high pressure setpoint prevents unnecessary recirculation pump trips following pressurization transients with a reactor scram (e.g. turbine trip or load rejection with bypass), while maintaining the requirements of 10 CFR 50.62. These recirculation pump trips lead to thermal stratification within the reactor vessel. This problem has been noted by the NRC in Information Notice 93-62 (Reference 4). Continued recirculation pump operation following a scram negates this phenomenon by allowing for better mixing of the reactor coolant and reduces thermal stratification in the vessel. Supplement 1 to the License Topical Report (Reference 1) addresses the revised analysis.
E1-7
Prooosed Chanae No. 25 l
25 Table 3.3.7-2 Trip Function 2; " Reactor Vessel Water Level- High" Allowable Value is being changed from s 208 inches to s 207 inches. The Trip Setpoint is being changed from s 208 -!
inches to s 206.8 inches. !
Basis for Chance No. 25 !
The Analytical Limit for this function does not change as a result of power uprate. The Allowable Value has now been calculated in accordance with CP&L's setpoint methodology.
The Allowable Value is high enough to preclude closing the injection valve of the RCIC .
system during normal operation, yet low enough to shutdown the RCIC system prior to the water level overflowing into the Main Steam Lines.
Proposed Chanae No. 26 26 Figure 3.4.1.1-1, " Thermal Power Limitations"; Rescale the figure by 1/1.05 in order to account for the increase in thermal power . ;
Basis for Chance No. 26 The change in reactor thermal power will necessitate the change in scale for this figure.
With power uprate there is no increase in core flow; therefore, the figure is shifted downward by the ratio of current rated thermal power to uprate thermal power (1/1.05).
This shift will maintain existing operational flexibility with only one recirculation loop in operation. !
Prooosed_C.ht. nae No. 27 27 Each of the SRV lift setpoints in Limiting Condition for Operation (LCO) 3.4.2 will be i
increased by 25 psi and the drift allowance will be increased from i 1% to
- 3%. The !
minimum number of safety valves required is reduced to 10. The Action Statements and Bases have been changed to reflect the reduced requirement of valves. ;
Basis for Chanae No. 27 The SRVs are designed to prevent over pressurization of the reactor pressure vessel during abnormal operational transients. The SRV lift setpoints are increased to accommodate the increase in operating pressure that accompanies power uprate. The increase in SRV setpoints ensures that adequate margins are maintained so that the increase in dome pressure during normal operation does not result in an increase in the number of unnecersary SRV actuations. The setpoint increase also maintains the hierarchy of pressure setpoints described in these proposed changes. The transient analyses performed for power uprate include the
- 3 percent tolerance to the nominal setpoints. The results of these analyses show the reactor vessel pressure increases by approximately 4.3%, but remains below the 1375 psig ASME Code limit (Reference 1).
The analysis performed for the over pressure event assumed 2 SRVs out of service. For the ATWS event, one SRV was assumed out of service. The minimum number of SRVs required by the LCO is being changed to reflect the minimum number of valves assumed in E1-6
the ATWS event. This transient is the most limiting and bounds the over pressure event.
The action statements require the plant be in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Cold Shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of less than the required number of SRVs are operable.
The adequacy of BWR SRVs to operate at uprated temperatures and pressures has been evaluated generically in Section 4.6 of Reference 3. The reactor operating pressure and temperature increases of less than 40 psi and 5'F, respectively, used in that evaluation bound the uprated BNP operating conditions.
The impact of power uprate on the BNP containment dynamic loads due to SRV discharge has also been evaluated. As discussed in Section 4.1.2 of Reference 1, the vent thrust loads with power uprate were calculated to be less than the loads used in the containment analysis. The effect of power uprate on SRV air-clearing of the SRV discharge line, the pool pressure boundary, and torus submerged structure drag loads is also discussed in Section 4.1.2 of Reference 1. That discussion concludes that the smallincrease in setpoint pressure is within the margin in the SRV loads defined in the Mark l Containment Long-Term Program. Therefore, this proposed change does not impact the BNP load definitions used in the containment analysis.
Prooosed Chance No. 28 28 The maximum Reactor Steam Dome Pressure in Limiting Condition for Operation (LCO) and Action Statement for Specification 3.4.6.2 and Surveillance Requirement (SR) 4.4.6.2 and associated Bases will be increased 25 psi (from less than 1020 psig to less than 1045 psig). The Bases for 3/4.6.2 will be revised to reflect the new maximum operating pressure i of 1045 psig. '
Basis for Chance No. 28 As discussed in Section 3.2 of Reference 1, the maximum reactor dome pressure is an initial condition of the vessel over pressure protection analysis, which assumes a fast !
isolation of the four main steam lines by the main steam isolation valves. The reactor scram !
signal generated directly by the valve closure is assumed defeated for this analysis.
Instead, the scram signalis Generated by high neutron flux. The over pressure analysis for I power uprate assumed an initial dome pressure of 1060 psia (1045 psig), which represents I an increase of 25 psi. This initial pressure was chosen because it is 15 psi above the rated dome pressure of 1045 psia, consistent with current Technical Specifications.
Prooosed Chance No. 29 & 30 29 in Surveillance Requirement 4.5.1.b, the reactor pressure and HPCI turbine steam supply pressure are increased 25 psi (from 21000 psig to 21025 psig and from 1000 +20, -80 psig to 1025 +20, -80 psig respectively).
30 In Surveillance Requirement 4.7.4.b, the RCIC turbine steam supply pressure is increased 1 25 psi (from 1000 +20, -80 psig to 1025 +20, -80 psig).
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Basigfor Chance Nos. 29 & 30 The allowable HPCI and RCIC surveillance test pressure is increased to correspond with the increase in normal reactor operating pressure and LCO/SR on maximum reactor pressure that accompanies power uprate. The requested changes will allow the quarterly demonstration of HPCI and RCIC capability to be performed at normal reactor operating pressures, which meets the intent of the current Technical Specifications.
References
- 1. NEDC-32466P " Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2," September 1995, including Supplement 1, dated March 1996.
- 2. NEDC-31897P-A " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," June 1991.
- 3. NEDC-31984P, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," Volumes I, ll; Supplements 1, 2, July 1991.
- 4. NRC Information Notice 93-62, " Thermal Stratification of Water in BWR Reactor Vessels"
- 5. NEDC-31624P " Brunswick Steam Electric Plant Units 1 and 2 SAFER /GESTR-LOCA Loss of Coolant Accident Analysis," Revision 2, July 1990.
- 6. OPL-3, " Plant Operating Licensing Analysis for Brunswick Steam Electric Plant, Unit 1" ,
E1-10
l ATTACHMENTA t
M
~
^
PROPOSED TECHIGCAI1SPECFICATION CHANGES?
a
- > s s s
p;g u e ,
(Noj NedinEnlSpecslication - AllowableV$11uel (Trip'Selpoint '
s cm Comments ' '
s s -
Cument - Proposed '
Currard Proposed s 1* License C.1- RATED THERMAL 2436 Mwt 2558 Mwt WA WA See NEDC-32466P .>
& Definitions POWER ;
i 2* Table 2.2.1-1.2b Flow biased Simulated s (0.66W+67%) s (0.66W+61%) s (0.66W+64%) s (0.66W+59.6%) Setpoet Calculaten ;
max sus.5% max sus.3% max sn3.5% max sn3,6% oC51 mot
& Figure 2.2.1-1 Thermal Power- High 3* Table 2.2.1-1.2c Fixed Neutron Flux - $120% Rated s118% Rated s120% Rated s116.3% Rated Setpoint Calculation ,
Thennal Power Thermal Power Thermal Power Thermal Power oC51-ooot High 4* Tab a 2.2.1-1.3 Reactor Vessel Steam s 1045 psig s 1070 psig s 1045 psig s 1067.9 psig Setpoint Calculaten ,
0B21-0073 Dome Pressure - High t
5* Table 2.2.1-1.4, Table 3.3.2-2.1.a.1, 2 -162.5 inches = +153 inches a +162.5 inches a +153.2 inches Setpoint Calculation 0s21-0069 s Table 3.3.2-2.5.a Table 3.3.3-2.4.c Reactor Vessel Water Level- Low, Level 1 6* Table 3.3.2-2.1.a.2, Table 3.3.3-2.1.a. > +2.5 inches = +13 inches = +2.5 inches a +14.1 inches Setpomt Calculation 0B21-0071 Table 3.3.3-2.2.b, Table 3.3.3-2.4.b Reactor Vessel Water Level - Low, Level 3 :
t 7* Table 3.3.2-2.1.c.3 Main Steam Line s140% of rated flow s138% of rated flow s140% or rated now s137% or rated now Setpomt Calculation 0s21-0068 Flow -High 8 Table 3.3.2-2.1.e Condenser a 7" Hg vacuum 2 7.5" Hg vacuum a 7"lig vacuum a 7.6" Hg vacuum Setpomt Calculation 0s21-0075 t Vacuum - Low i
- Indicates changes that are a direct result of power uprate analyses E1-11 ,
i
. .- - - - - . . . - , , . - . _ - , . - - . . . . . . , . . . . - - - . - , - - . - - . . . h
ATTACHMENT A t PROPOSED TECHNICAL SPECIFICATION CHANGES 1 "
p
- No.: (Technical Specification : Allowable Valuel tTrip Setpoint :-
. . ..- . Comments 3 iCurrent 6 Proposed? : Current - tProposed 9* Table 3.3.2-2.2.c, Table 3.3.2-2.3.e, a +112 inches a +103 inches a +112 inches a +104.1 inches Setpoint Calculation Table 3.3.3-2.3 a. Table 3.3.6.1-2.1, 0B21-0070 Table 3.3.7-2.1 Reactor Vessel Water Level - Low, Level 2
10* Table 3.3.2-2.4.a.1 HPCI Steam Line s300% of rated h s 275% of rated s300% of rated h s 272% of rated h Setpoint Calculation h OE41-0036 Flow - High 11 Table 3.3.2-2.4.a.3 HPCI Steam 2100 psig a 104 psig = 100 pss a 106.6 psig Setpoint Calculation OE41-0035 Supply Pressure -
Low 12 Table 3.3.2-2.4.a.6 HPCI Turbine s 10 psig s 9.0 psig s 10 psig s 8.5 psig Setpoint Calculation OE41-0037 Exhaust Diaphragm Pressure - High 13* Table 3.3.2-2.4.b.1 RCIC Steam Line s300% of rated flow s 275% cf rated s300% of rated h s 272% of rated h Setpoint Calculation fl w OE51-0026 Flow - High 14 Table 3.3.2-2.4.b.3 RCIC Steam a 50 psig a 53 psig a 50 psig a 55.6 psig Setpoint Calculatxm OE51-0025 Supply Pressure .
Low 15 Table 3.3.2-2.4.b.6 RCIC Turbine s 10 psig s 6.0 psig s to psig s 5.5 psig Setpoint Calculation OE51-0027 Exhaust Diaphragm Pressure - High 16 Table 3.3.2-2.5.b Reactor Steam s 140 psig s 137 psig s 140 psig s 130.8 psig Setpoint Calculation 0832 6 9 Dome Pressure -
High
- Indicates changes that are a direct result of power uprate analyses E1-12
ATTACHMENT A
- PROPOSED TECHNICAL SPECIFICATION CHANGES 3 No[ (TechnicalSpecificationj ?Allidabl$ Valt.ies 2 Trip S$thnt} . .
m . Comments; 1 Current: ; Propos$d? E Current ; i Proposed 17 Table 3.3.3-2.1.b Reactor Steam 410 i 15 psig 2404 psig 410 i 15 psig 240s.7 psig setpoint calcuration OB21-0074 Dome Pressure -
Low 18 Table 3.3.3-2.2.d 1 Reactor Steam 410
- 15 psig 2404 psig 410
- 15 psig 240s.7 psig setpoint calculation OB21-0074 Dome Pressure -
Low / RHR Pump Start & LPCI Valve Actuation 19 Table 3.3.3-2.2.d 2 Reactor Steam 310
- 15 psig 2304 psig 310 i 15 psig 2306 7 psig setpoint caiculation OB21m74 Dome Pressure -
Low / Recirculation Pump Discharge Valve Actuation 20 Table 3.3.3-2.4.d ADS Timer s 120 seconds s 108 seconds s 120 seconds s 83 seconds setpoint Calculation 0821-007s 21 Table 3.3.3-2.4.e Core Spray Pump 2100 psig a 102 psig = 100 psig = 112.1 psig setpoint calculation OE21-0013 Discharge Pressure
- High 22 Table 3.3.3-2.4.f RHR (LPCI Mode) 2100 pse = 102 psig > 100 psig a 111.1 psig setpoint calculation OE11-0022 Pump Discharge Pressure - High 23* Table 3 3.4-2.1.a APRM Upscale s (0 esw+s1%) s (0.ssw+56%) s (0.ssW+58%) s (0.66w+54.6%) setpoint calculation max suG% d max sn1% d max s108% M max sin 3% d Oc51-0001 (Flow Biased) Rod RATED THERMAL RATED THERMAL RATED THERMAL RATED THERMAL Block POWER POWER POWER POWER
- Indicates changes that are a direct result of power uprate analyses E1-13
LPROPOSED TECHNICAL SPECIFICATION CHANGES 3 No.: STechnical Spedin,ativ6 y Allowabie Valuei ETrip Setpoint3 '
~ Commentst s Current 1 JProposed 1 Current: 1 Proposed 24* Table 3.3.6.1-2.2 Reactor Vessel s 1120 psig s 1143 psig s 1120 psig s 1137.8 psig setpoint Calcutation 0B21-0077 Pressure - High 25* Table 3.3.7-2.2 Reactor Vessel Water s +208 inches s +207 inches s +208 inches s +206.8 inches setpoint Calculation OB21-0072 Level- High 26' Figure 3.4.1.1-1 Thermal Power N/A N/A N/A N/A Rescale Figure by (1/1.05)
Limitations 27* 3.4.2 Safety Relief Valve lift 4 @ 1105 psig 1% 4 @ 1130 psig 13% N/A N/A NEDC-32466P 4 @ 1115 psig i 1% 4 @ 1140 psig i3%
& Bases setting 3 @ 1125 psig i1% 3 @ 1150 psig i3%
28* 3.4.6.2 Reactor Steam < 1020 psug < 1045 psig N/A N/A NEDC-32466P 4.4.6.2 Dome Pressure
& B 3/4.6.2 29* 4.5.1.b HPCI Operability Test Rx Press e1000 psig Rx Press 21025 N/A N/A Revised to include the new Turb. Press 1000 psig psig plant operating pressure
+20 -80 psig Turb. Press 1025 psig +20,-80 psig 30* 4.7.4.b RCIC Operability Test Turb. Press 1000 psig Turb. Press 1025 N/A N/A Revised to include the new
+20. -80 psig psig +20,-80 psig plant operating pressure 31* Table 3.3.2-2.1.c.4 Main Steam Line s40% of rated now s32% of rated h s40% of rated h s30% of rated now setpoint Cateuration 0821-0068 (Unit 2 only) Flow -High
- Indicates changes that are a direct result of power uprate analyses E1-14
l ENCLOSURE 2 r BRUNSWICK STEAM ELECTRIC PLANT UNIT NOS.1 AND 2 i NRC DOCKET NOS. 50-325 AND 50-324 !
OPERATING LICENSE NOS. DPR-71 AND DPR-62 !
REQUEST FOR LICENSE AMENDMENTS ,
105% THERMAL POWER UPRATE 10 CFR 50.92 EVALUATION '
The Commission has provided standards in 10 CFR 50.92 for determining whether a significant I hazards consideration exists. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or ,
consequences of an accident previously evaluated, (2) create the possibility of a new or !
different kind of accident from any accident previously evaluated, or (3) involve a significant ,
reduction in a margin of safety. Carolina Power & Light Company has reviewed these proposed j license amendment requests and believes that their adoption would not involve a significant 'l hazards consideration. The basis for this determination follows. !
i
- 1. May the proposed activity involve a significant increase in the probability or ;
consequences of an accident evaluated previously in the Safety Analysis Report? j The increase in power level, steam flow, feedwater flow and associated instrument setpoint changes will not significantly increase the probability or consequences of an l accident previously evaluated. l The probability (frequency of occurrence) of Design Basis Accidents occurring is not ;
affected by the increase in power level, as plant equipment will remain in compliance !
with the applicable regulatory criteria (ASME Codes, IEEE Standards, NEMA Standards, j Regulatory Guide criteria, etc.) The physical plant changes necessary to support power i uprate include instrument setpoint changes, indicating meter scale changes for the RWCU System flow and Main Steam Flow indicators, Leak Detection, Process . !
Computer, ERFIS, and Feedwater System software changes, and SRV setpoint ,
changes. The setpoints were calculated in accordance with the CP&L Setpoint Methodology. Utilizing this methodology ensures scram setpoints (instrument settings that initiate automatic plant shutdowns) will be established such that there is no significant increase in scram frequency due to uprate. No new challenges to safety related equipment will result from power uprate. l t
The changes in consequences of hypothetical accidents which would occur from 102%
of the uprated power (2609 MWt), compared to those previously evaluated from a 102%
of the original power (2485 MWt), are not significant, because the accident evaluations at uprated power will not result in exceeding the NRC approved acceptance limits. The spectrum of hypothetical accidents and transients has been investigated, and those accidents / transients currently evaluated in the UFSAR were shown to meet the plant's current regulatory criteria at uprated conditions (105%). In the area of core design, for E2-1
1 l
l 1
l l
example, the fuel operating limits will still be met at the uprated power level, and fuel '
reload analyses show plant transients will still meet the criteria accepted by the NRC as specified in NEDO-24011 "GESTAR 11." Challenges to fuel or ECCS performance have been evaluated and shown to meet the criteria of 10CFR50 Appendix K. Challenges to the containment have been evaluated and still meet 10CFR50 Appendix A Criterion 38, Long Term Cooling, and Criterion 50, Containment. Bounding events involving radiological releases have been evaluated and were shown to be well within the criteria of 10CFR100. ;
1
- 2. May the proposed activity create the possibility of a new or different kind of accident from any accident previously evaluated in the Safety Analysis Report?
The change in reactor thermal power will not create the possibility of a new or different kind of accident from any accident previously evaluated.
Equipment that could be affected by power uprate has been evaluated. No new operating mode, safety related equipment lineup, accident scenario, or equipment failure mode was identified. The full spectrum of accident considerations defined in the BNP UFSAR has been evaluated and no new or different kind of accident has been identified. Uprate uses developed technology and applies it within the capabilities of existing plant equipment in accordance with existing regulatory criteria including NRC approved codes, standards, and methods. General Electric has designed BWRs of higher power levels than the uprated power of any of the currently uprated BWR/4 fleet and has not identified new power dependent accidents.
The changes to the Technical Specifications required to implement power uprate make little change to the plant's configuration. These changes fali!nto three major categories.
The first includes those changes resulting from power uprate parameter changes.
These parameter changes, such as the increase in vessel pressure, temperature and piping system flows are minor in nature. The evaluations have shown the plant is still within its design capabilities when operating under these conditions. The changes required as a result of power uprate will not affect the design function (s) of currently installed equipment; therefore, there is no possibility of a new or different kind of failure mode. The second set of changes is a result of applying setpoint methodology to calculate TS Allowable Values and Nominal Trip Setpoints for instruments that are directly affected by the parameter changes due to power uprate. By using CP&L's methodology, the TS values were calculated to ensure adequate margin exists between the analytical limit and the TS Allowable Value. The third change include setpoints that were reconstituted by the power uprate project. Again, CP&L methodology was applied and the results show the setpoints have moved to a more conservative value. This will reduce the likelihood of spurious scrams and unnecessary challenges to safety systems while ensuring initiation / actuation equipment continues to function consistent with existing accident analyses.
E2-2
['
l
- 3. Does the proposed activity involve a significant reduction in a margin of safety defined in the basis of any Operating License Technical Specification?
l Power Uprate will not involve a significant reduction in a margin of safety. The bounding events which had been analyzed in the UFSAR were roevaluated to demonstrate that j power uprate can be implemented without exceeding any analyzed limit. Because the j applicable safety analysis criteria and limits are satisfied for power uprate, the margin of l safety associated with the safety limits and other limits identified in the Technical '
Specifications will be maintained. I As discussed in Section 5 of GE Nuclear Energy's License Topical Report NEDO- !
31984P " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," l the safety margins prescribed by the Code of Federal Regulations (CFR) have been l maintained by meeting the appropriate regulatory criteria. Similarly, the margins i provided by the application of the ASME design criteria have been maintained. The i Brunswick unique analysis NEDC-32466P " Power Uprate Safety Analysis Report for i Brunswick Steam Electric Plant Units 1 and 2" discusses the effects of power uprate on l safety margins for (1) fuel thermal limits, (2) design basis accidents and the challenges ,
for fuel, containment and radiological releases, (3) transient analysis, (4) non-LOCA - ,
radiological releases, and (5) environmental consequences. These evaluations '
conclude that applicable safety analysis criteria and limits are satisfied, and thus, the margins of safety will be maintained.
l The changes to the Technical Specification instrumentation will not involve a reduction j in the margin of safety. The calculations performed for power uprate have established !
an analyticallimit and calculated the TS Allowable Value and Nominal Trip Setpoint ,
using formal setpoint methodology. This ensures the instrumentation functional j requirements are met. ;
References
- 1. NEDC-32466P, Class ill (Proprietary), September 1995, " Power Uprate Safety Analysis !
Report for Brunswick Steam Electric Plant Units 1 and 2", and Supplement i dated l March,1996
- 2. NEDC-31897P-A, Class lli (Proprietary), May 1992, " Generic Guidelines for General !
Electric Boiling Water Reactor Power Uprate" Licensing Topical Report
- 3. NEDC-31984P, Class 111 (Proprietary), July 1991, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate" Licensing Topical Report E2-3 i
f
I I
ENCLOSURE 3 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS,1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS 105% THERMAL POWER UPRATE ENVIRONMENTAL CONSIDERATIONS 10 CFR 51.22(c)(9) provides criterion for and identification of licensing and regulatory actions ;
eligible for categorical exclusion from performing an environmental assessment. A proposed !
amendment to an operating license for a facility requires no environmental assessment if '
operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (3) result in an increase in individual or cumulative occupational radiation exposure. Carolina Power & Light Company has reviewed this request and believes that the proposed amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) Pursuant to 10 CFR 51.22(b), no environmentalimpact statement of environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis for this determination follows.
- 1. These amendments do not involve a significant hazards consideration, as shown in - !
Enclosure 2.
- 2. The proposed license amendments do not result in a significant change in the types or a I significant increase in the amounts of any effluent that may be released offsite. License Topical Report NEDC-32466P provides the power uprate safety analysis report for Brunswick, as well as an assessment of the radiological effects of power uprate operation during both normal and postulated accident conditions. Sections 8.1 and 8.2 discuss the potential effect of power uprate on the liquid and gaseous radwaste systems. Sections 8.3,8.4 and 8.5 discuss the potential effect of power uprate on radiation sources within the plant and radiation levels during normal and post-accident conditions. Section 9.2 presents the results of the calculated whole body and thyroid doses at the exclusion area boundary and the low population zone that might result from the postulated design basis i radiological accidents. All offsite doses remain below established regulatory limits for power uprate operation.
A non-radiological environmental assessment was submitted with the License Topical Report in November of 1995. This report describes the effects of power uprate on the 4 environment. Power uprate does not change the methods of generating electricity or of '
handling any influents from the environment or effluents to the environment; therefore, no new or different environmental impacts are expected. The detailed evaluation presented in the non-radiological assessment concluded that non-radiological parameters affected i by power uprate will remain within the bounding conditions cited in the state National Pollutant Discharge Elimination System (NPDES) permit, Permit No. NC0007064. )
Therefore, no significant environmental impact will result from operation of Brunswick ;
under power uprate conditions.
E3-1
The proposed license amendments do not introduce any new equipment nor does it require any existing equipment or systems to perform a different type of function than they are presently designed to perform. The proposed license amendments do not alter the function of existing equipment and will ensure that the consequences of any previously evaluated accident do not increase. Therefore, CP&L has concluded that there will not be a significant increase in the types or amounts of any effluent that may be released c'%e and, as such, does not involve irreversible environmental consequences beyond those already associated with normal operation.
- 3. These amendments do not result in an increase in individual or cumulative occupational radiation exposure.
l l
E3-2