ML20010J260
| ML20010J260 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/16/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20010J258 | List: |
| References | |
| NUDOCS 8109300053 | |
| Download: ML20010J260 (6) | |
Text
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- p se )E UNITED STATES co f
j NUCLEAR REGULATORY COMMISSION 1 0 j
wAsHiNGTom. o. c. 20sss
+,.....,o SAFETY EVALUATION BY THE OFFICE OF NUCuEnR REACTOR REGULATION SUPPORTING AMENDMENT NO.79 TO FACILITY GPERATING LICENSE NO. OPR-56
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PHILADELPHIA ELECTRIC C0fiPANY.
PUBLIC SERVICE ELECTRIC AND GAS COMPAi!Y DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY
_. ___ PEACH BOTTOM ATCMIC POWER STATION, UNIT NO. 3 OrCKET NO. 50-278 Introduction By letter dated September 30, 1980, Philadelphia Electric Company (PECo or the licensee) made application to modify the Technical Specifications (TSs) for Peach Bottom Atomic Power Station, Units 2 and 3, to permit an extension of the maximum average planar linear heat generation rate (MAPLHGR) f.om 30,000 to 40,000 megawatt days per short ton of uranium (mwd /T). By letter dated itay 20, 1981, we issued TSs extending the MAPLHGR for Unit 2 only.
This license amendment evaluates the requested change for Unit 3.
In addition, by letter dated March 30, 1981, as supplemented April 24, June 30 and July 15, 1981, the licensee made application to modify the TSs for Peach Bottom. Unit 3 to permit operation with the reload numoer 4 core (Cycle 5).
Evaluation Thermal-Hydraulic Design Peach Bottom Unit 3 Reload 4 consists of 216 new P8x8R fuel bundles which have drilled lower tie plates and finger springs to regulate bypass flow.
This makes a total of 764 bundles with drilled lower tie plates for Reload 4 (or Cycle 5). Reload 4 has a total of 489 P8x8R fuel bundles, 252 8x8R fuel bundles and 23 8x8 fuel bundles. Assumed cycle exposure is increased from 17,160 tWd/T (Reload 3) to 18,208 PWd/T (Reload 4). Also, for operational flexibility, Minimum Critical Power Ratio (MCPR) operating limits with dif-ferent options were providad in the proposed TSs. Our review consisted of the following:
(a) Fuel Cladding Integrity Safety Limit, (b) Operating Limit
[1CPRs (0LMCPRs), (c) Thermal-Hydraulic Stability, and (d) TS modifications.
The objective of the review is to confirm that the thermal-hydraulic design of the reload core has been accomplished using acceptable methods, that it provides an acceptable margin of safety from conditions which would lead to fuel damage during normal operation and anticipated operational transients, and is not susceptible to thermal-hydraulic instability.
8109300033 s10735-Fuel Cladding Integrity Safety Limit yDRADOCK 05000278 PDR As stated in Ref. 4, the minimum allowable critical power ratio for core-wide or localized transients is 1.07.
This limit has been imposed to assure
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~PB-3 that during transients 99.9% of the fuel rods will avoid boiling transition.
There has been no change in the safety limit MCPR for Peach Bottom Unit 3 from Cycle 4 to Cycle 5.
OLMCPRs Various transients could serve to reduce the actual MCPR below the intended safety limit MCPR (SLMCPR) during Cycle 5 operation. The most limiting of these operational transients have been analyzed by the licensee to deter-mine which event could potentially induce the largest reduction in the initial critical power ratio (ACPR).
The transients evaluated were the generator load rejection without bypass, the feedwater controller failure, loss of 100 F feedwater heating, the control rod withdrawal error and the fuel loading error.
Transients were analyzed on the basis of the initial conditions given in Section 6 of Ref. 2.
The initial MCPR assumed in the calculation of the aCPR for the generator load rejection without bypass at End : f Cycle (E0C)-2000 !4ld/T for the PTA/P8x3R fuel is 0.01 below the OLMCPR.
ihe licensee has provided justification for this assumption in Ref. 7.
This is acceptable to the NRC staff. The aCPR values given in Section 9 (Ref. 2) are plant-specific deterministic values calculated by using the ODYN transient code (Ref. 6).
The value of aCPRs 5
for the same fuel types (8x8 and 8x8R) for Cycle 5 is 0.26 ccmpared to 0.23 for Cycle 4, and for P8x8R and PTA fuel for Cycle 5 is 0.29 compared to 0.25 for Cycle 4.
This difference is due to the use of the (Ref. 3) 0DYN transient code compared to the REDY code used in Cycle 4.
We have evaluated the ODYN code and found it acceptable for transient analyses of the Cycle 5 core.
Fuel Loading Error aCPRs The licensee stated (Ref. 2) that the mislocated bundle loading error event analysis will no longer be reported for each cycle as per Ref. 5.
We have accepted this for current Peach Bottom reloads (Ref. Sa). Tne licensee has done the rotated bundle loading error event analysis based on the new analysis procedure described in Ref. 6.
Analysis shows that the rotated bundle results in a MCPR greater than the safety limit of 1.07, and we find this analysis acceptable.
Rod Withdrawal Error (RWE) ACPRs RWE ACPRs given in Section 10 of Ref. 2 were calculated using previously approved methods (Ref. 4). The ACPR at the Rod Block Monitor (RBM) setpoint of 107% is 0.20 for Cycle 5 compared to 0.13 for Cycle 4.
This difference is due to less P8x8R fuel loaded in Cycle 5 compared to Cycle 4 and a dif-ferent loading pattern (_Ref. 2) in Cycle 5.
RWE aCPR is not limiting in Cycle 5.
Therefore, the RWE aCPR analysis is acceptable to the NRC staff.
Establishing OLMCPRs The aCPRs calculated above were adjusted to reflect either " Option A" or
" Option B" aCPRs by employing the conversion method described in Ref.10.
These adjustments are based on conservative factors.
The MCPR for the event is determined by adding the ACPR to the scfety limit.
Section 11 (Ref. 2) f presents both the MCPRs for the non-pressurization events as well as the adjusted fiCPRs (0ption A and Option B) for the pressurization events.
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.a) MCPRs were adjusted for Option B for all plants choosing to operate under Option B which meet all scram specifications given in Ref. 9.
b) - MCPRs were determined by a linear interpolation between the Option A MCPR and the Option B MCPR for all plants choosing to operate under Option B which do not meet the scram time specification. This interpolation is based on the tested measured scram time and is described in Ref. 9.
c) MCPRs were adjusted for Option A according to Ref. 9.
This option is to be used if the surveillance requirement of the TSs to scram time test contral rods is not perfonned.
We have reviewed all the OLMCPR results discussed above. These results are consistent with the previous Cycle 4 analysis and are more conservative for Cycle 5 than Cycle 4; therefore we find these results acceptable.
Thennal-Hydraulic Stability The results of the thermal-hydraulic stability analysis (Ref. 2) show that the channel hydrodynamic and reactor core stability decay ratios at the natural circulation-1057. rod line intersection are below the stability limit.
Decay ratio for Cycle 5 was 0.87 as compared to 0.90 for Cycle 4.
Because the operation in the natural circulation mode will be prohibited by the TSs, there will be added margin to the stability limit and we conclude this is acceptable.
Evaluation of TS Changes The licensee has submitted proposed changes to the Peach Bottom Unit 3 TSs (Ref. 3).. These changes:
(1) identify the operating limits for all fuel types for Cycle 5 operation, (2) incorporate f%PLHGR limits for the Reload 4 fuel and extended exposure MAPLHGR limits for the Re!aad 2 and Reload 3 fuel, (3) add a generic MAPLHGR curve for P8x8R fuel to reduce need for future cycle-dependent revisions.
OU1CPR TSs For All Fuel Types Based on our thennal-hydraulic design evaluation in this Safety Evaluation, changes in the TSs are found to be acceptable except tnat we have modified
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TS Figure 3.5.K.1, Page 142 (.Ref. 3) MCPR operating limit vs. r for 8x8 and 8x8R fuel. This modification makes the MCPR operating limit more conservative.
This change was agreed to by t'n licensee.
MAPLHGR Limit TS Curves For all fuel types the licensee proposed to extend the burnup time from 30,000 to 40,000 !Wd/T.
The licensee has stated (Ref.13) that they corply with General Electric letter (.Ref.12) for the MAPLHGR limits.
Therefore, as stated in Ref.11, we find the proposed extended exposure MAPLHGR limits for the Reloan 2 and Reload 3 fuel acceptable. The licensee's proposed MAPLHGR limits fer the Reload 4 fuel and the generic i%PLHGR curve for the P8x8R fuel to reduce the need for future cycle dependent revisions have been done witn
v a
S PB-3.,
currently approved methods and are in compliance with Ref.12; therefore, we conclude these _ revised. curves are acceptable.
Change in Control Rod Scram Time TS This proposed TS change calls for a 3.5 second average scram insertion time, rather'than 5.0 second average scram insertion time for the 90% rod insertion from the fully withdrawn position. 'This change is in conformance with the
-Cycle 5 unique transient analysis input presented in Reference 2.
Our review of the reduction in average scram insertion times indicates that there is no effect:on the course of previously analyzed transients.
Based on both the current use of these scram times in the General Electric Standard TSs and.the lack of change on transient results, we conclude that this change is-acceptable.
Evaluation of Fast Scram Control Rod Drive (FSCRD) Program During Cycle 5 In order to assist General Electric Corporation in developing a control rod drive system for the BWR/6 design, PECo has oeen using a single FSCRD in Unit 3 starting in Qycle 2.
We have previously evaluated the use of this drive and found it acceptable.
In order to accumulate long term exposure of l
the drive in.an operating reactor environment, the licensee proposed to extend its use through Cycle 5.
Previous operating experience has been favorable.
The original drive, used in Qycle 2, has been removed, disassembled and in-j '
spected. The inspection provided support for continued use of a FSCR0 in i
Unit 3 through Gycle 5.
The current FSCRD was installed dur g Cycle 3.
These drives have no effect on the parameters used in the safety analyses.
We con-clude, based on the abov", that continued use of a FSCRD in Unit 3 during Cycle 5 is acceptable.
Environmental Considerations i
We have determined that-the amendment does not authorize a change in effluent types'or total amounts nor an ir;rease in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment j
involves an action which is insionificant from the standpoint of
~~ environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an 1
environmental impact statement, or negative declaration and environ-menta-1 impact _ appraisal need not be prepared in connection with the issuance of this amendment.
There is, however, an environmental consi(eration related to the amendment.
j 10 CFR 51.20g(2)(iii) states,'in part, "The average level of irradiation of the irradiated fuel from the reactor dJes not exceed 33,000 megawatt i
days per metric ton and...". The TS curves specify burnup in megawatt days per short ton. A short ton is 2,000 pounds and a metric ton is 2,205 pounds, thus a metric ton is 1.1 times greater than a short ton.
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In a previous Safety Evaluation performed for the Browns Ferry Nuclear Plant, n
Units Nes. I and 2, dated October 6,1980, we extended the irradiation to 40,000 megawatt days per short ton. This is the same request made by PECo j
for Peach Bottom Unit No. 3 in this amendment. We found that the Browns Ferry fuel when irradiated to 40,000 megawatt days per short ton did not exceed an average level of burnup of 33,000 megawatt days per metric ton.
The Peach Bottom 3 fuel is bounded by the evaluation done for the Browns Ferry fuel. We conclude, based on the bounding Browns Ferry analysis, that the proposed burnups to 40,000 MdD/Short Ton do not exceed the 10 CFR Part 51.20 limits of 33,000 MWD / Metric Ton.
'To ensure that the environmental con.iderations in 10 CFR 51 are evaluated, if MAPLHGR limits are extended in the future, we have, with the licensee's concurrence, added a note to the TS Figures related to MAPLHGR limits, stating the requirement of 10 CFR 51.20.
Conclusion We have concluded, based on the considerations discussed above, that:
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(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered 4
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 procosed 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 coracn defense and security or to the health and safety of the public.
Dated: September 15, 1981 f
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PB-3.
References
- 1. ' Letter, E. J. Bradley (PECo) to H. R. Denton (NRC), dated March 30,19G1.
2.
Supplemental Reload Licensing Submittal fer ?each Bottom Atomic Power Station Unit 3, Reload No. 4. Y1003J01A20, February 1981.
3.
" Prop'ssed Technical Specification Changes", alonc with technical descrip-tions and bases for such changes, March 30, 1981, enclosure of Ref.1.
3a Letter, S. L. Daltroff (PECo) to J. F. Stolz (NRC), dated June 30, 1981.
4.
" General Electric Boiling Water Reactpr Generic Reload Fuel Application",
July 1979, (NEDE-240ll-P-A-1).
5.
Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), "Cnange in General Electric Methods for Analysis of Mislocated Bundle Accident", November 14, 1980.
Sa. Letter, W. V. Johnston ";o T. A. Ippolito, " Change in Generr.1 Electric Analysis of Mislocated Bundle Accident", dated April 14, 1981.
6.
Safety Evaluation Report (letter), D. G. Eisenhut (NRC) to R. E. Engel (GE), MFN-200-78, dated May 8, 1973.
7.
" General Electric Boiling Water Reactor Generic Reload Fuel Application",
January 1979 (NEDE-240ll-P-A, Amendment 2).
S.
Letter, R. P. Denise (NRC) to G. G. Sherwood (GE), January 23, 1980.
9.
Letter (with attachment), R. H. Buchholz (GE) to P. S. Check (NRC), "Res onse to NRC Request for Information on ODYN Computer Model", September 5,1950.
10.
Letter, R. H. Buchholz (GE) to P. S. Check (NRC), "0DYN Adjustment Method for Determination of Operating Limits", January 19, 1981.
- 11. Letter, L. S. Rubenstein (NRC) to T. M. Novak (NRC), " Extension of General Electric Emergency Core Cooling System Performance Limits", dated June 25, 1981.
- 12. Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), " Additional Information Regarding Extension of Emergency Core Cooling Systen Performance Limits",
dated Pay 28, 1981.
13.
Letter,PECo to NRC dated June 30, 1981.
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