ML20247J063
| ML20247J063 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/01/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247J054 | List: |
| References | |
| NUDOCS 8909200186 | |
| Download: ML20247J063 (10) | |
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7 g-NUCLEAR REGULATORY COMMISSION j,
WASHINGTON, D. C. 20S55 l
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING l
l AMENDMENT NO. :150 TO FACILITY OPERATING LICENSES NO. DPR-56 j
PHILADELPHIA ELECTRIC COMPANY PUBLIC 5ERVICE ELELIMIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY T LANTIC CITY ELECTRIC CUPPANY
= PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 3 DOCKET NO. 50-278
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1.0 INTRODUCTION
By letter dated July 7, 1988, Philadelphia Electric Company n !uested
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an amendment to Facility Operating License No. DPR-56 for the Peach Bottom Atomic Power Station, Unit No. 3.
This amendment would revise the Technical Specifications (TS) to:
(1) incorporate the operating' limits for all fuel types for Cycle 8 operation, (2) incorporate a change in slope of the flow biased Average Power Range Monitor (APRM) scram and rod blocksetpointstoprovideincreasedoperating(4)clarifyadefinition flexibility during power ascension, (3) correct.a typographical error,
- of; Average Planar Linear Heat Generation Rate (APLHGR), and (5) make various changes tn the Bases resulting from the core reload. TS changes were proposed for the operation of Peach Botton Atomic Power Station, Unit No. 3 for Cycle 8 (PB3C8) with a reload using General Electric (GE) manufactured fuel assemblies and GE analyses and methodologies. Enclosed were the requested TS changes and reports (Refs.'2 and 3) discussing the reload and analyses done to support and justify Cycle 8 operation and extended power-flow operating regions.
The reload for Cycle 8 is generally a normal reload with no unusual core features or characteristics. TS changes are primaril Average Planar Linear Heat Generation Rate (MAPLHGR) y related to Maximum and Linear Heat Generation Rate (LHGR) limits for the new fuel and Minimum Critical Power Ratio (MCPR) limits for all of the fuel using Cycle 8 core and transient parameters.- The new fuel is the GE extended burnup fuel which is extensively used in GE fueled core reloads.
The submittal also proposes extensions of the standard allowed operating regions-on the reactor temperature end power-flow map).
The Extended Lead Lire Limit Analysis (ELLLA), Increased Core Flow (ICF, and the Final Feedwater Temperature Reduction (FFWTR) proposed modes of extended operation are similar to those approved on a number of other BWRs in recent years. Except for changes to the flow-biased APRM scram and rod block setpoints for ELLLA and some additional MCPR limits for ICF, they require no other changes to Cycle 8 TS.
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. As noted above, the supporting analysis was performed by GE for PECO using NRC approved methods and codes.
In the subject Peach Bottom, Unit 3 submittal, as well as in the previous PEC0 submittal for Peach Bottom, Unit 2, Cycle 8, GE has proposed a new approach of only including the curves for the most limiting and least limiting MAPLHGRs as a function of planar average exposure values for each fuel type in the TS. During power operation, the process computer would check that the APLHGR for each type of fuel as a function of axial location and average planar exposure was within the limits based on the applicable APLHGR limit values which had been approved for the respective axial lattice of a given fuel type. When the process computer is not available and hand calculations are required, the most limiting lattice ArtHGR limits for each fuel type will apply to every lattice of that fuel type. The lattice dependent MAPLHGR values are provided in Reference 3.
This procedure for treating lattice-dependent MAPLHGRs for a given fuel type has been reviewed and approved by the staff and provided in Reference 4.
2.0 EVALUATION 2.1 Reload Description The PB3C8 reload will retain 284 P8X8R, 284 BP8X8R, and 4 Lead Test Assemblies (LTAs) from the previous cycle and add 192 new GE8X8EB fuel assemblies. All of the fuel assemblies are GE manufactured. The reload is based on a previous cycle core nominal average exposure of 15.9 GWd/ST and Cycle 8 end of cycle exposure of 18.9 GWJ/ST. The loadirg will be a conventional scatter pattern with low reactivity fuel on the core periphery.
2.2 Fuel Design The new fuel for Cycle 8 is the.GE extended burnup fuel GE8X8EB. The fuel designations are BD319A and BD321A. This fuel type has been approved in the Safety Evaluation Report for Amendment 10 to GESTAR II (Refs. 5 and 6). The specific descriptions of this fuel are presented for PB3C8 in Reference 3.
These fuel descriptions are acceptable.
In operation the GEBX8EB fuel will be assigned a number of lattice regions and appropriate MAPLHGR limits, which have been determined by approved thermal-mechanical and loss-of-coolant accident (LOCA) analyses, will be applied to each of these regions. The process computer contains, and acts on, full details of the MAPLHGR information. The TS present the least and most limiting lattice MAPLHGRs as a function of average planar exposure.
When hand calculations of MAPLHGRs are required (process computer inoperative), the most limiting MAPLHGR values are used for all lattices of a given fuel type.
These TS are acceptable. A proprietary report, reviewed by the NRC staff and, available to the Peach Bottom, Unit 3 staff, provides complete details of the lattice definitions and MAPLHGR limits.
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..The' proposed LHGR limit for the GE8X8EB fuel is 14.4 KW/ft rather than the 13.4.KW/ft for the other GE fuel in PB3C8. This LPGR limit has been reviewed and accepted for this. fuel in the GE extended burnup fuel review (Ref. 5).
(See the referrals in Reference.5 to References 18 and 19.
1 These references are responses to questions and presentations relating to the GE8X8E8 fuel which provide information on the 14.4 KW/ft LHGR.) This LHGR limit is acceptable for the GE8X8EB fuel in PB3C8.
2.3 Nuclear Design The nuclear design for PB3C8 has been performed by GE with the approved methodology described in GESTAR-II (Ref. 6). The results of these analyses are given in the reload report (Ref. 2) in standard GESTAR-II format.- The results are within the range of those usually encountered for BWR reloads.
In particular the shutdown margin is 2.2% and 1.3% delta K at beginning-of-cycle (BOC),and at the exposure of minimum shutdown margin, respectively, thus fully meeting the required 0.38% delta K.
The Standby Liquid Control System also meets shutdown requirements with a shutdown margin of 4.2% delta K.
Because these and other PB3C8 nuclear design paremeters have been obtained with previously approved methods and fall within expected ranges, the nuclear design is acceptable.
2.4 -Thermal-Hydraulic Design The thermal-hydraulic design for PB3C8 has been performed by GE with the
. approved methodology described in GESTAR-II (Ref. 6) and the results are given'in the reload report (Ref. 2). The parameters used for the analysis are those approved in Reference 6 for the Peach Bottom BWR-4 class except for the parameters listed in Appendix C of Reference 2. ' The GEMINI system of methods (approved in Ref. 7) was used for the relevant transient analyses.
The Operating Limit MCPR l0LMCPR) values are determined by the Ifmiting transients, which are usually Rod Withdrawal Error (RWE), Feedwater ControllerFailure(FWCF)[LRWBP). Loss of Feedwater Heating (LFH), and Load Rejection Without Bypass The analyses of these events for PB3C8, using the ODYN Option A and B approaches for pressurization trcnsients, provide new Cycle 8 TS values of the OLMCPR as a function of average scram time, for operation in both standard and extended operating regions.
For PB3C8 PEC0 has elected, following standard practice, to have exposure dependent OLMCPRs. Two exposure regions from BOC to end-of-cycle (EOC) were analyzed:
(1) BOC to E0C - 2 GWd/ST and (2) EOC - 2 GWG/ST to EOC.
For standard operating conditions, the LRWBP event is controlling at both Option A and B limits except at BOC Option B where the RWE event is controlling. These OLMCPR results are reflected in TS changes. Approved methods (Ref. 6) were used to analyze there events (and others which could be limiting) and the analyses and results are acceptable and fall within expected ranges.
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. The Safety. Limit MCPR (SLMCPR) is set so that less than 0.1 percent of the
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fuel pins in the core are subject to boiling transition when some fuel in the core is at the SLMCPR. The SLMCPR is being changed to 1.04 (1.05 for single recirculation loop operation) for PB3C8. This change has been approved by the NRC for D-lattice cores operating with the second successive reload core of high bundle R-factor fuel types (Ref.10). The licensee states that all fuel types to be loaded for Cycle 8 operation (that is, SP/P8X8R, LTA, and GE8X8EB) are high bundle R-factor fuel types. Thus the change to the SLMCPR is acceptable.
The mean and standard deviation of the control rod scram speed data that are used to compute the adjusted mean scram time (Tau) are being changed.
These new values of the mean and standard deviation results in a calculation of a small and more restrictive value of Tau. These changes lead to a conservative determination of Tau and OLMCPR and are therefore acceptable.
The Peach Bottom 3 TS have staff approved provisions for incore neutron detector monitoring of thermal-hydraulic stability according to the recor4nendation of GE SIL-380. Thus cycle specific stability calculations are not required, either for standard conditions or the extended temperature 2.6) power-flow co.1ditions proposed for Cycle 8 operation (see Section and 2.5 Transient and Accident Analyses The transient and accident analysis methodologies used for PB3C8 are described and NRC approval indicated in GESTAR II (Ref. 6). The GEMINI system of methods (Ref. 7) option was used for transient analyses. The limiting MCPR events for PB3C8 are indicated in Sections 2.4 and 2.6.
The core wide transient analysis methodologies are acceptable and fall within expected ranges.
The RWE was analyzed on a plant and cycle specific basis (as opposed to the statistical approach) and a rod block monitor setpoint of 107 was selected to provide an OLMCPR of 1.21 for all fuel types. The mislocated 1
assembly event is not analyzed for reload cores on the basis of (NRC approved, see Ref. S.2-59 of Ref. 6) studies indicating the small probability of an event exceeding MCPR limits. The disorientation event was analyzed with standard methods for the PB3C8 D lattice fuel, giving a j
nonlimiting MCPR of 1.18. The local transient event analyses are thus acceptable.
The limiting pressurization event for establishing overpressure protection margin, the main steam isolation valve closure event with flux scram, was analyzed with standard GESTAR II methods and gave results for peak steam dome and vessel pressures well under required limits. These are acceptable methodologies and results.
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5-l LOCA analysis, using approved methodologies (SAFE /REFLOOD/ CHASTE) and i
parameters, were performed to provide MAPLHGR values for the new reload fuel assemblies (GE8X8EB). These analyses and results are acceptable.
Since some parameters of the generic rod drop accident (RDA) were not l
I bounding for PB3C8, cycle specific RDA analyses were done for cold and hot j
shutdown conditions. These were done with standard, approved GE methods.
The results were well within the required 280 cal /gm limit. The analyses and results are acceptable.
2.6 Changes to APRM and RBit Setpoints The PB3C8 reload submittal proposes extensions to standard cperating regions in the. GESTAR II category of " Operating Flexibility Options." The selected options are ELLA, ICF, and FFWTR. These have become commonly selected and approved options for a number of reactors in recent years.
These options are described and discussed in appendices A and B of the reload report (Ref. 2) and in GE topical reports (Refs. 8 and 9). These appendices provide the results of transient analyses for setting MCPR limits for Cycle 8 and the topical reports provide generic analyses of transients and accic'ents, applicable for follow-on cycles as well.
The proposed ELI,LA changes the APRM flow-biased scram and rod block lines on the power-flow map, and permits operation up to the new APRM flow-biased rod block line (0.58W + 50%) up to the intersection with the 100 percent power line at a core flow of 877.
In the above, W is the recirculation drive flow in rercent of rated. Similarly, the new APRM flow-biased scram line is giver a (0.58W + 62%). These are standard changes for ELLA. For ICF, the proposed flow increase is to 105 percent core flow at 100 percent power. The increased flow would be allowed throughout the cycle and after normal EOC 8 (with or without FFWTR) with reactivity coast down to 70 percent power. After E0C 8, the ICF would be bounded by 110 percent of rated flow. The proposed FFWTR involves valving out last stage feedwater heaters (going to a feedwater temperature of about 328"F) and is proposed only for operation after normal EOC 8.
For the ELLLA extension, the topical report (Ref. 9) discusses a full range of transient and accident events relevant to the region extension, and presents results of calculations or previously approved conclusions.
In addition, Appendix B of the PB3C8 reload report (Ref. 2) presents additional calculations of limiting MCPR transients specifically for PB3C8. The transient analyses demonstrate that for reactors such as Peach Bottom 3 which do not have Recirculation Pump Trip for transient scram response assistance, the licensing basis results (e.g., 100 percent flow, 100 percent power for pressurization transients) bound the ELLLA region results (e.g., 87 percent flow, 100 percent power). These conclusions apply to all relevant MCPR events such as pressurization, rod withdrawal, and slow flow runout events. Changes to MCPR TS are not required because of adoption of the ELLLA option. Other relevant areas such as overpressure
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. f protection, LOCA, and containment analyses have also been examined by the licensee, and the analyses indicate that results are within allowable design limits. Thermal-hydraulic stability will be provided for by appropriate surveillance. The analyses have been done with approved methodologies and the results are similar to previously approved ELLLA i
extensions. Thus operation within the ELLLA region is acceptable for PB3C8.
For the ICF and FFWTR extensions, similar to the ELLLA presentation, the topical report (Ref. 8) discusses a full range of relevant transient and accident events and other potential problem areas, and Appendix A of the reload report (Ref. 2) provides analyses of limiting MCPR events for PB3C8. Unlike the situation for ELLLA, the analysis of MCPR events leads, in some situations, to more restrictive MCPR limits, which are cycle dependent. Appendix A presents the results of calculations, using standard methodology, for the most limiting events at the most limiting combination of ICF and/or FFWTR conditions for PB3C8. These are presented for Options A and B and for both exposure ranges considered for standard operating conditions (FFWTR is allowed only for E0C). The results are reflected in the TS which are changed to provide a new MCPR limit for EOC-2000 GWD/ST to EOC operation with ICF.
FFWTR operation, within the bounds to be used, is not limiting with or without ICF, compared to standard operating conditions. The RWE results for the standard operation region are not affected with the RBM clipped at 107 in the ICF region.
The MCPR analyses for ICF and FFWTR extensions use standard methods, follow previously approved trends, and are acceptable.
The licensee has also evaluated other events and affected system components related to these extensions. These include overpressure protection, fuel loading error, rod drop accident, and LOCA events, none of which are significantly affected by the extensions. As in the case for ELLLA, the thermal-hydraulic stability will be appropriately monitored using GE SIL-380 surveillance, and will thus present no new problem. The licensee has evaluated the effects of ICF induced increased pressure differentials and vibration response on reactor internals, fuel channels, and fuel bundles, and has determined that design limits will not be j
exceeded. The containment LOCA response was analyzed and the results show l
no significant impact of ICF and FFWTP,. The feedwater nozzle and sparger fatigue usage factors were examined for the effects of extreme programs of FFWTR and EOC power coast down. The analysis leads to the conclusion that there is no significant impact beyond a slightly increased nozzle refurbishment schedule (based on monitored seal leakage). The review of l
these various licensee evaluations leads to the conclusion that suitable analyses were performed and the results are compatible with other reviews and are acceptable for Peach Botton 3.
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l 1 2.7,Techncial Specifications 2.7.1
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i F e TS changes for PB3C8 associated with the reload and operating i
tiexibility options are primarily to provide for:
1 (a) - The new ELLLA APRM flow-biased scram and rod block setpoints. The I
changes are to TS 2.1.A.1, TS 2.1.B. Figure 1.1-1, Table 3.1-1, and Table 3.2.C and are acceptable.
(b) The new MCPR limits for Cycle 8 and ICF operation. The changes are to TS 3/4.5.K Tables 3.5.K.2 and 3.5.K.3, and Figures 3.5.K.1 and
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3.5.K.2 and are acceptable.
(c) The 14.4 KW/ft LHGR limit for the new GE8X8EB fuel. The changes are to TS 2.1.A TS 2.1.B, TS 3.5.J. Table 3.1.1 and Table 3,2.C and are acceptable.
(d) MAPLHGR limits for the new fuel. The changes are to TS 3.5.I, Figure 3.5.1.f and Figure 3.5.1.G and are acceptable.
2.7.2 One typographical error is being corrected by the amendtrict as follows:
(a) TS 3.5.K is incorrectly identified on page 133b as TS 3.5.K.1.
I This change is a correction to a TS identification and is acceptable, i
2.7.3 Page IV of the list of figures in the Table of Contents is being revised to reflect the MCPR and MAPLHGR figures changed by this amendment.
2.7.4 l
In the Definitions Section of the TS, a paragraph is being added to define
" Average Planar Linear Heat Generation Rate." This is a desirable addition because the term is not defined in the current TS.
2.7.5 The bases for TS 1.1 on pages 13 and 15 are being modified to include a i
reference for single-loop operation for Peach Bottom Units 2 and 3.
This j
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single-loop operation, l
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2.7.6 The Bases for TS 3.5.1 are being modified to incorporate a new paragraph regarding APLHGR operating limits for multiple-lattice fuel bundles.
This change.is acceptable.
2.7.8 The Bases for the core thermal-hydraulic and physics analyses reference the GE topical report " General Electric Standard Application for Reactor
. Fuel," NEDE-24011-P-A in_a number of sections. This topical report is subiect to periodic revisforiby GE. Changes to the topical report have to be approved by the NRC. Wherever this is referenced in the Bases the words "as amended" are appended to refer to the most recent!y approved version. Bases pages affected are 15, 17, and 140c. There are also changes to the Bases in Sections 2.1 and 2.2 to eliminate information
'that is redundant to the topical report. These changes will eliminate the need to change the date on the references if the referenced topical report is revised. This approach has been adopted by most licensees and is endorsed by the staff.
2.7.9 A Scntence is being added in two places to the Bases for TS 2.1 on pages
- 17. and 18 that states that abnormal operational transients were analyzed at or above the maximum power level required by Regulatory Guide 1.49 to determine operating limit MCPRs. Reference to a specific power level (that is, 3440 MWt) is being deleted. The change reflects the analyses performed for.the core reload.
2.8' Conclusions We have reviewed the report submitted for the Cycle 8 operation of Peach Bottom 3 with extended operating regions. Based on this review we conclurie that appropriate material was submitted and that the fuel design, nuclear design, thermal-hydraulic design and transient and accident analyses are acceptable. The Technical Specification changes submitted for this reload suitability reflect the necessary modifications for operation in this cycle.
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9 2.9 References 1.
Letter from E. J. Bradley (PECO) to T. M. Murley (NRC), dated July 7, 1988.
2.
GE Report 23A5889 Rev. O, dated January 1988, " Supplemental Reload Licensing Submittal for Peach Bottom Atomic Power Station, Unit 3, Reload 7, Cycle 8."
3.
NEDE-24082-P-1, Supplement 1, Revision 1, January 1988, " Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station, Unit 3" (and Errata and Addenda Sheet No. 7).
4.
Letter from J. S. Charnley (GE) to M. W. Nodges (NRC), dated March 4, 1987, " Recommended MAPLHGR Technical Specifications for Multiple Lattice Fuel Designs."
5.
Letter (and attachment) from C. Thomas (NRC) to J. Charnley (GE),
dated May 28, 1985, " Acceptance for Referencing of Licensing Topical Report NEDE-24011-P-A-6, Amendment 10."
6.
GESTAR-II, hEbE-24011, Revision 8, " General Electric Standard Application for Reactor Fuel."
7.
Letter (and attachment) from G. Lainas (NRC) to J. Charnley (GE),
dated March 22, 1986, " Acceptance for Referencing of Licensing Topical Report NEDE-20411-P-A, 'GE Generic Licensing Reload Report,'
Supplement to Amendment 11."
8.
" Safety Review of Peach Bottom Atomic Power Station Unit No. 3 at Core Flow Conditions Above Rated Flow Throughout Cycle 6," NEDC-30519, March 1984.
9.
" General Electric Boiling Water Reactor Extended Load Line Limit Analysis % Pee h Bottom Unit 2, Cycle 7, and Peach Bottom Unit 3, Cycle 7," nDr-31298, May 1986.
- 10. Letter from Ashok C. Thadani (NRC) to J. S. Charnley (GE), " Acceptance for Referencing of Amendment 14 to General Electric Licensing Topical Report NEDE-24011-P-A, ' General Electric Standard Application for Reactor Frel,'" dated Deceaber 27, 1987.
In addition, typographical corrections were made to the licensee's incoming technical specification pages and bars were added to show the changes.
3.0 ENVIRONMENTAL C0i:SIDERATIONS This amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the l
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t types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed 4
J 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 categorical exclusion set forthin10CFR51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement nor environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Register (54 FR 27234) on June 28, 1989 and consulted with the State of Pennsylvania. No public comments were received and the State of Pennsylvania did not have any comments.
The staff has cencluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the p(2) such activities will be conducted in compliance with theublic will not be Commission's regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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. Principal Contributor:
D. Fieno Dated:
September 1, 1989 1
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