ML20054J041
| ML20054J041 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 06/17/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054J040 | List: |
| References | |
| NUDOCS 8206280007 | |
| Download: ML20054J041 (12) | |
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UNITED STATES 3
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NUCLEAR REGULATORY COMMISSION I
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W ASHINGTON, D. C. 20665 e
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SAFETY EVALUATION _BY THE OFFICE OF NUCLEAR REACTOR REG TO FACILITY OPERATING LICENSE NO. DPR 44 SUPPORTING AMENDMENT NO. 86 PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 2 DOCKET NO. 50-277_
l.0 Introduction The Philadelphia Electric Company (the licensee) has proposed changet to the Technical Specifications of the Peach Bottom Atomic Power Station, Unit No. 2 (Ref.1). The proposed changes relate to the replar.ement of 276 fuel assemblies constituting refueling of the rea'ctor core for 6th cycle operation at power levels up to 3293 Mwt (100% power).
2.0 Fuel Design Evaluation i:
The reload application (Ref.1) contains five fuel-design-related changes: (1) analysis of the safety considerations involved in the reactor refueling and the Cycle 6 operating limits, (2) continued operation with two previously irradiated fuel assemblies following reconstitution, (3) continued operation with develop-mental fuel channel boxes, (4) incorporation of new and revised Maximum Average Planar Linear Heat Generation Rate (MAPLFGR) limits for the Cycle 6 fuel including extended exposure MAPLHGR limits for standard and lead test assembifes and (5) additien of a generic MAPLHGR curve for General Electric P8X8R fuel.
2.1 Safety Analysis of Cycle 6 Ooerating Limits The licensee's analysis of the safety considerations involved in the reactor refueling and the Cycle 6 operating limits are set forth in the Peach Bottom Unit 2 Cycle 6 Reload Report (Ref. 2).
In all fuel-design-related areas except those identified in Section 2.0 above, the Reload Report reif es on a generic document, the General Electric Reload Fuel Application Report (Ref. 3).
I B206200007 820617 b
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1 The latter. report has been reviewed and a'pproved (Ref. 4) by'the NRC staff. We conclude that additional staff review of those portions of Reference 2 con-cerning the standard fuel design is unnecese.sry for Cycle 6 operation.
In addition to the 210 standard 8x8R and 552 prepressurized 8x8R fuel assemblies, two previously irradiated high-burnup lead test assemblies (LTAs) will be insertedsin'the Cycle 6 core. The operation of these lead test assemblies was previously approved through Cycle 5.
The safety analysis for operation of these assemblies during Cycle 6 is described in Appendix C of Reference 2.
Although Appendix C generally follows the fuel design criteria used for the standard fuel (Ref. 3), some of the analyses used to demonstrate that the LTAs meet these criteria appear to have been perfomed at exposures higher than nomally encountered. From the description of the LTA design analysis presented in Appendix C, we are unable to detemine (a) differences between that analysis and those contained in the General Electric Reload Fuel Application Report (Ref. 3) and (b) differences between that analysis and those identified in the Standard Revicw Plan (Ref. 5), lle note, however, that the design criteria used, and the level of detail presented in Appendix C, are typical of that pre-viously accepted for LTA programs. The analysis is acceptable because (a) the allowable power rating of these assemblies at high exposures is significantly lower than the rest of the core and (b) only two lead test
, bundles are involved.
2.2 Reconstituted Fuel Assemblies The licensee has proposed reconstitution and continued use of two previously irradiated standard 8x8R fuel assemblies. The purpose of the reconstitution is to obtain fission ' gas release data in conjunction with a General Electric,-
extended burnup test program. Six rods in each fuel assembly will be replaced
.with, fresh rods.
The twelve removed rods will be subjected to fission gas
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pressure measurements (puncture tests). An analysis of the safety considerations involved in the continued operation of the reconstituted fuel is described in Appendix B of Reference 2.
The licenses has stated that the mechanical d'esign changes in the new rods were minor and that the initial enrichments of the new rods were selected to assure that the power peaking in the reconstituted assemblies will be similar to the non-reconstituted assemblies. Therefore, the results of the fuel rod themal and mechanical design evaluations in Reference 3 remain applicable to the reconstituted assemblies. We agree with this conclusion and find the use of the reconstituted fuel assemblies for Cycle 6 to be acceptable.
2.3 Developmental Fuel Channel Boxes The licensee has requested approval for the continued operation of twelve develop-mental (i.e., experimental) fuel assembly channel boxes, which were initially installed during the first reload of Peach Bottom Unit 2.
These channel boxes utilize various wall thicknesses and heat treatments as part of a study of oxide growth and corrosion behavior. The analysis of the safety considerations involved in continuing the use of the developmental channel boxes is presented in Appendix E of Reference 2 and in Reference 6.
Neither report specifically identifies any burnup linitation on the analysis although the Cycle 6 Application (Ref.1)' states that a 40 GWd/MtU limit is observed. Based on previous NRC staff approval of this program and on the developmental nature of these channel boxes, we continue to find their use acceptable at Peach Bottom Unit 2.
2.4 MAPLHGR Limits The licensee has submitted new and revised MAPLHGR limits for all Cycle 6 fuel These types including extended exposure limits for standard and LTA fuel.
limits were generated by methods (Ref. 7) submitted as part of the application.
Although the methodology used is generally applicable for these limits, we believe that the effects of enhanced fission gas release in high burnup fuel (above 20 GWd/Mtu) were not adequately considered in the generic analysis.'
In response to this concern, the General Electric Company requested (Refs, 8-9) that credit for approved, but unapplied, emergency core cooling system (ECCS) evaluation model changes be used to avoid MAPLHGR penalities at higher burnup.
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O This proposal was found acceptable (Ref.10) provided that certain plant-specific conditions were met.
In a letter dated July 15,1981 (Ref,11), Philadelphia Electric Company found the General Electric proposal applicable to both Peach Bottom Units 2 and 3.
On the basis of this finding, we conclude that t'te MAPLHGR limits proposed for Peach Bottom Unit 2 Cycle 6 are acceptab1r..
It should be noted that these MAPLHGR limits have been provided for average planar exposures of up to 40 GWd/STU for all fuel types except the two LTAs, which have MAPLHGR limits specified for average planar exposures of up to 50 GWd/STU. We regard these proposed Technical Specifications as limiting on both power and burnup, and therefore the peak planar average exposure during Cycle 6 operation must be limited to 40 GWd/STU for all but the two LTAs by the proposed Technical Specifications. It should also be noted that a basis for MAPLHGR extensions beyond 40 GWd/STU average planar exposure has not yet been accepted by 'the NRC staff for other than LTA operation. A General Electric submittal that would justify such high-burnup application is expected in the near future. However, communication with the licensee has revealed that average planar exposures beyond 40 GWd/STU are not anticipated for 80RB284L fuel and, therefore, such a submittal is not required to support the Cycle'6 safety analysis. We thus find the MAPLHGR limits acceptable as submitted.
2.5 Generic MAPLHGR Limit The licensee has proposed the addition of a generic P8X8R l'APLHGR' curve to the plant Technical Specifications. This curve would be added for the purpose of re-ducing the need for future cycle-dependent revisions to the Technical Specific-ations.
It was constructed to bound the Reload 4 and Reload 5 P8x8R fuel at Peach Bottom Unit 2.
Because this curve was generated from a number of specif,ic fuel types, rather than from the set of all possible P8x8R fuel loadings, it is necessary that the licensee detemine that the generic P8x8R MAPLHGR curve is 4
bounding for future specific iMPLHGR limits supplied' by the fuel vendor (e.g.
General Electric). Should this be the case, we would accept use of the generic
!%PLHGR curve without additional modification to the plant Technical Specific-ations.
2.6 Conclusions We'have reviewed those sections of the reload report for Peach Bottom Unit 2, Cycle 6, dealing with changes to the fuel system design and its analysis. We find those portions of the application acceptable.
3.0 Nuclear Design Evaluation The reload report follows the procedures described in Reference 3.
Reference 3 has been approved for use in the nuclear design and analysis of reloads for boiling water reactors (Reference 4).
Its use is acceptable for Peach Bottom.
Separate cycle-specific analyses were done for the rotated bundle event, the rod withdrawal error and the control rod drop accident. The latter analysis was necessary because the scram curve for Cycle 6 is non-conservative with respect to the generic curve. The results of the cycle-specific analyses meet the relevant criteria and are acceptable.
We have reviewed the nuclear aspects of the fuel assembly rod replacement and extended exposure LTAs. Six fuel rods are to be removed from each of two assemblies and replaced with fresh rods having enrichments that are
~ designed to compensate for the depletion of the removed rods. The licensee concludes that this replacement will have a negligible effect on the nuclear characteristics of the assemalies and of the core. We concur with this con-clusion.
Two LTAs, which are part of a program to assess the effect of extended
~ burnup on boiling water reactor fuel, will renain in Peach B6ttom in Cycle 6.
The effect of the presence of these assemblies on the nuclear char-acteristics of the core has been analyzed. The reactivity (K m)'of the bundles 5
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decreases monotonically with burnup. The ' local peaking factors tend to increase but the bundle powers decrease (due to the reactivity decrease) so that no limits are approached during the extended burnup.
Doppler and void reactivity coefficients remain essentially constant.
In summary, the presence of the two LTAs will have a negligible effect on the core nuclear chadacteristics.. ~~
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We conclude that the proposed Cycle 6 of the Peach Bottom Unit 2 reactor is acceptable with respect to its core physics aspects.
4.0 Thermal-Hydraulic Evaluation Peach Bottom Unit 2, Reload 5 fuel assemblies are identical in mechanical design to P8x8R assemblies previously licensed and operated in Peach Bottom-2, Reload 4' (Ref. 12). The new fuel assemblies differ from the existing fuel assemblies in the core only in having higher U-235 enrichment. This change was accounted for in the submitted reload analysis.
This review includes the following areas:
(1) safety limit Minimum Critical Power Ratio (MCPR), (2) operating limit MCPR, (3) thermal-hydraulic stability, and (4) change to Technical Specifications 3.5.K and 4.5.K.
The objective of this review is to confim that the thermal-hydraulic design of the reload core has been accomplished using acceptable methods, and provides acceptable margin of safety from conditions which could lead to fuel damage during nomal operation and anticipated operational transients, and is not susceptible to themal-hydraulic instability. The themal-hydr'aulic models and reload methodology used are described in Reference 3.
i 4.1 Safety Limit itCPR The safety limit !!CPR has been established to assure that at.least 99.9 percent i
of the fuel rods in the core do not experience boiling transition during t.he worst anticipated operational occurrence. As stated in Reference 3, the safety i
limit itCPR is 1.07. There has been no change in the safety linit itCPR for Peach
. Bottom-2 from Reload 4 to Reload 5.
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4.2.
Operating Limit MCPR Various transients could reduce the MCPR below the intended safety limit'MCPR during Cycle 6 operation. The anticipated operational transients have been analyzed by the licensee to detemine which event could potentially induce the largest reduction in the initial critical power ratio (ACPR).
The ACPR values given in Section 11 (Ref. 2) are plant specific values which include results for the transients calculated by using the ODYif methods '(Refs.13 and 14).
The maximum value of ACPR resulting from the limiting transient, the generator load rejection without bypass transient; is 0.36 for Reload 5 as compared to 0.23 for Reload 4 (Refs. 2 and 15). The large difference of a CPR for this transient is due to the use of the ODYft methods compared to the REDY methods used in Reload 4.
The calculated ACPRs were adjusted to reflect either Option A or Option B ACPR by employing the conversion method described in Reference 16. The MCPRs are then detemined by adding the ACPR to the safety limit.
Section 11 of Reference 2 presents the ACPR for both non-pressurization and pressurization events. The maximum MCPRs calculated by using the ACPR values in Section 11 are specified as the operating limit MCPRs and are incorporated in the Technical Specifications. We have reviewed the operating limit MCPR results discussed above and found these results acceptable.
4.3 Themal-Hydraulic Stability The results of the themal-hydraulic analysis (Ref. 2) show that the maximum Because themal hydraulic stability decay ratio is 0.85 for Reloads 5 and 4.
operation in the natural circulation mode will be prohibited by Technical Specification 2.1. A there will be added nargin to the core stability an.d this is acceptable to the NRC 's'taff.
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4.4. Changes to Technical Specifications 3.5.K and 4'.5.K The operating limit MCPR Technical Specification has been modified to include an Option B fomat where the operating limit itCPRs vary with the measured scram time (t). The specification is based on measurements to the 20 percent inserted position. Figures 3.5.K.1, 3.5.K.2 and 3.5.K.3 of the proposed Technical Specifications show operating limit MCPR vs 'c for 8x8 LTA, P8x8R and P8DRB285 fuels, respectively.
He find that the approved ODYN methods (Refs.13 and 14) were used and that the results of analyses are consistent with the proposed operating limit MCPR to avoid violation of the safety limit MCPR for the design transients. We conclude that this core reload will not adversely affect the capability for safe operation during Cycle 6 and that the proposed changes to Technical Specifications 3.5.K and 4.5.K. discussed above are acceptable.
5.0 Cycle 6 Transient Analyses Generic information relative to the reload analyses of boiling water reactor fuel is presented in General Electric Licensing Topical Report NEDE-24011-P-A,
" Generic Reload Fuel Applications," July 1979 (Ref. 3). This report is supplemented by plant-specific information contained in References 2 and 7.
I Together these documents provide the bases for the licensee's safety analysis for Reload 5 and the proposed Technical Specification changes associated with the reload (Cycle 6).
The licensee stated (Ref. 2) that all transients that are the basis of the Peach Bottom-2 Final Safety Analysis Report were reviewed for Cycle 6 and that those transients that were critical with respect to safety margins and sensitive to the core related parameter changes were reanalyzed.
The ODYN code is used to define input parameters for Critical Power Ratio (CPR) calculations during rapid pressurization transients. The ODYN code also is used to calculate the pressure transients more accurately than the REDY code and provides more detailed outputs. In Reference 2, the licensee has provided graphical results from the analysis of pressurization transients.
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The transients included are:
(1) Main Steam Isolation Valve Closure (2) Feedwater Controller Failure (3) Generator Load Rejection Without Bypass The licensee has performed the required analyses. As a consequence of using more than one code (ODYN and REDY) for the transients and the two options available with ODYN (Option A with straight penalty for uncertain-ties and Option B with statistical convolution of uncertainties and rod scram times), the limiting transient for Peach Bottom-2 is dependent upon periodic on-site measurements of average scram time. Depending upon the measured average scram time, the MCPR operating limit will change as shown on TS Figures 3.5.K.1, 2 and 3.
MCPRs are adjusted using Option B when all scram specifications in section 4.5.K.2.a. of the plant Technical Specifications are met. For operating i
limit MCPR values, s'ee TS Table 3.5.K.2.
In the event that the scram time specification is not met, a linear interpolation between the Option A i
MCPR and the Option B MCPR will be performed as given in Ref.16. Operating limit MCPRs adjusted using Option A are given in TS Table 3.5.K.3.
We have reviewed the General Electric generic scram time specification l
procedure using ODYN (Options A and B) and have found it to be acceptable (Ref.17 and 18). The licensee has duplicated these procedures, and we i
conclude that this is acceptable. We have also reviewed the licensee's h
proposed changes to Technical Specifications related to NCPR (pages 133b, l
133c, 133d, 133e, 142, 142a and 142b) and conclude that these changes are acceptable.
6.0 Incorporation of the 67B Control Rod Scram Time
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This Technical Specification change proposes that the 678 Control Rod Drive (CRD) scram times be incorporated into the Technical Specifications.
The 67B CPD scram times, in replacing the 67A scram times, require a 3.5 second average scram insertion time, rather than 5.0 second ayerage scram insertion time for the 90% inserted, from the fully withdrawn position.
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This change -is proposed to be in conformance with the reload-unique transient analysis input utilized in Reference 2.
The change is acceptable because the results of the transient analysis incorporating the change meet acceptable criteria for operating limit MCPRs as discussed above.
7.0 Environmental Considerations We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
8.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendnent does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: June 17,1982 The following NRC personnel have contributed to this' Safety Evaluation:
Morton Fairtile, George Thomas, Summer Sun, John Voglewede and Walter Brooks.
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.v' REFERENCES 1.
E. J. Bradley (PECO) letters to H. R. Denton 'and 'J. F. Stolz (NRC) dated February 19, 1982 and June 3, 1982.
2.
" Supplemental. Reload Licensing Submittal for Peach Bottom Atomic Power Station Unit 2. Reload No. 5," General Electric Company Report Y1003J01 A34, December 1981.
3.
" General Electric Boiling Water Reactor Generic Reload' Fuel Applications,"
General Electric Company Report NEDE-240ll-P-A-2 (Proprietary) and NEDO-240ll-A-2 (Non-Proprietary), July 1981.
4.
D. G. Eisenhut (NRC) letter to R. Gridley (GE) dated liay 12, 1978.
5.
U.S. Nuclear Regulatory Commission Standard Review Plan Section 4.2 (Rev. 2), " Fuel System Design," U.S. Nuclear Regulatory Commission Report NUREG-0800 (fomerly NUREG-75/087), July 1981.
"De' elopmental Channels-Supplemental Infomation for Reload 1 6.
v Licensing Submittal for Peach Bottom Atomic Power Station Unit 2."
General Electric Company Report NEDO-21172, Rev.1, Supplement 2, March 1976.
7.
" Loss-of-Coolant Accident for Peach Bottom Atomic Power Station Unit 2," General Electric Company Report NED0-24081, December 1977 (withAddenda1-7).
8.
R. E. Engel (GE) letter to T. A. Ippolito (NRC) dated May '6,1981.
f 9.
R. E. Engel (GE) letter to T. A. Ippolito (NRC) dated May 28, 1981.
10.
L. S. Rubenstein (NRC) memorandum for T. ft. Novak (NRC) on " Extension of General Electric Emergency Core Cooling Systems Perfomance Limits" dated June 25, 1981.
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- s.
a 11.
S. L. Daltroff (PECo) letter to J. F. Stolz (NRC) dated July 15, 1981.
- 12. Letter from R. Reid (NRC) to E. Bauee (PEco) dated June 13, 1980.
- 13. Letter from J. Quirk (GE) to P. Check (NRi?
ODYN Improvements, September 25, 1981.
- 14. Letter from J. Quirk (GE) to T. Speis (NRC), ODYN Improvements, October 13, 1981, 15.
Supplemental Reload Licensing Submittal for Peach Bottom Atomic Power Station Unit No. 2 Reload No. 4. NED0-24237-A, dated February 1980.
- 16. Letter from R. Buchholz (GE) to P. Check (NRC), "0DYN Adjustment Methods for Detemination of Operating Limits," dated January 19, 1981.
- 17. " Safety Evaluation for the General Electric Topical Report Qualification on the One-Dimensional Core Transient Model for Boiling Water Reactors,"
NED0-24154 and NEDE-24154P Volumes I, II, and III, June 1980.
18.
" Supplemental Safety Evaluation for the General Electric Topical Report l
Qualification of the Qne-Dimensionale. Core Model for Boiling Water Reactors,"
NED0-24154 and NEDE-24154P Volumes I, II, and III, January 1981.
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