ML20210Q902
| ML20210Q902 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 08/26/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20210Q898 | List: |
| References | |
| NUDOCS 9709020201 | |
| Download: ML20210Q902 (6) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WADHINGTON, D.C. 3004He01
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t SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 97 TO FACILITY OPERATING LICENSE NO. NPF-47 I
ENTERGY OPERATIONS. INC.
RIVER BEND STATION. UNIT 1 DOCKET NO. 50-458 1.0 INTRODut FON By application dated January 20, 1997, Entergy Operations, Inc. (the licensee) requested changes to the Technical Specifications (Appendix A to Facility Operating License No. NPF-47) for the River Bend Station (RBS), Unit 1.
The proposed changes would revise the Technical Specifications (TS) to allow the use of flow control spectral shift strategies to increase cycle energy. The request would allow incorporation of the Maximum Extended Load Line Limit (MELLL) Analysis for RBS.
The licensee also provided additional information for the staff's review. The licensee's letter dated July 7, 1997, provided clarifying information and did not change the staff's initial no significant hazards consideration determination.
2.0 BACKGROUND
in their request dated January 20, 1997, the licensee also submitted a safety evaluation for the request, including NEDC-32611P, " Maximum Extended Load Line Limit Analyses for River Bend Station Reload 6 Cycle 7," dated November 1996.
The Maximum Extended Load Line Limit (MELLL) o)eration mode and the associated TS changes expand the operating domain along t1e 121% rod line to the power / flow point of 100% power and 75% core flow.
An enlarged power / flow map for RBS would permit improved power ascension capability by extending plant o)eration at rated power with less than rated core flow during the fuel cycle.
Tiie average power range monitoring (APRM) flow-biased serair and rod-block setpoints would also be increased to accommodate the MELLL region on power / flow map. Additionally, TS would be revised to accommodate single loop operation (SLO) in the HELLL reginn.
Use of HELLL has been previously approved for other BWR/6 designs and is acceptable for use at RBS.
By letter dated June 26, 1997, the staff requested additional information on MELLL transient analysis including the Rod Withdrawal Error (RWE) analysis and the revised Average Power Range Monitor (APRM) flow-biased scram and rod block
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setpoints. The licensee provided a response by letter dated July 7,1997.
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9709020201 970826 PDR ADOCK 05000458 P_
i 3.0 EVALUATION Anticipated Operational Occurrences for MELLL Operation The limiting anticipated operational occurrences (A00s) (pressurization and non-pressurization events)itical power ratio (0LMCPR) requirements are for the MELLL region were examined to ensure that operating limit minimum cr satisfied. Maintaining the OLMCPR provides sufficient margin so that the safety limit MCPR (SLMCPR) will-not be exceeded. Maintain'ng this limit
- assures that 99.9 percent of the fuel rods are expected to avoid boiling transition. The analysis show that the OLMCPR for operation in the MElil region remains bounded by the OLMCPR established for current rated conditions of 100% power and 107% flow. The following core-wide events were considered to bound OLMCPR requirements, and were evaluated for MELLL operation:
Generator Load Rejection with No Bypass (LRNBP)
Feedwater Controller Failure (FWCF) to Maximum Demand Fuel Loading Error (FLE)
Pressure Regulator Failure (PRFDS) Downscale The transient and accipent analysis methodologies used for RBS cycle 7 are described in GESTAR-II.
These events were examined for OLMCPR 1mpact when o>erating in the MELLL region. The analysis results for RBS showed that for tte pressurization transients at the MELLL condition of 100% power and 75%
flow, the LRNBP event yields the most limiting fuel thermal responses for both Gell and GE8x8EB fuel, and are bounded by the previously analyzed 100% power, 107% flow and 100% power,100% flow conditions. The power and flow dependent thermal limits (MCPR-used in the analysis.p, MCPR-f) for the current operating cycle (Cycle 7) wereThese li for future cycles and will be reported as required in the Core Operating Limits Report (COLR).-
Loss of Coolant Accident (LOCA)
An ECCS/LOCA analysis was conducted to determine the impact of MELLL operation on the RBS LOCA peak cladding temperature (PCT). The result was an estimated increase of-less than 5'F from the value for operation in the currently I
aI> proved power / flow regime. The evaluation was conducted using the current MS licensing basis methodology SAFE /REFLOOD, The licensee has stated that prior to implementation of MELLL in Cycle 8, adequate PCT margin will exist to absorb the additional PCT impact. The licensee's preliminary SAFER /GESTR analysis was conducted with inputs which bound both Cycle 7 and 8, and have resulted in a PCT of about 1300'F. This is acceptable to the staff.
' NEDE-240ll-P-A-ll, " General Electric Standard Application for Reactor Fuel, GESTAR II," and NEDE~240ll-P-A-ll-US, "GESTAR II U.S. Supplement,"
November 1995.
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3 Rod Withdrawal Error The rod withdrawal error (RWE) event is a localized transient event not significantly affected by MELLL operation. The characteristics of the RWE event are such that the most important parameters affecting the transient I
response are the initial control rod pattern and the error rod position.-
These parameters are not affected by the low flow operation such as in the u
MELLL domain. The severity of the RWE event for a PWR/6 is mitigated-by-the 1
Rod Withdrawal Limiter (RWL) function which is not affected by the MELLL
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operation. -In addition, the APRM flow-biased rod block setpoint is revised to initiate the clamp at 75% core flow ct,ndition instead of at 100% core flow to bound the MELLL' domain.- Therefore, the previously-analyzed RWE remains valid for MELLL operation.-
Thermal:- Hydraulic stability l
The licensee has-stated that RBS has implemented the GE Servica Information l
Letter (dations, and that RBS is also in compliance with-the interin measures
$1L) 380, Revision 1 "BWR Core Thermal Hydraulic Stability,*
recommen of Bulletin 88-07, Supplement 1, " Power Oscillations in Boiling Water Reactors (BWRs)." In addition, core / channel decay ratio and the stability exclusion region are reanalyzed for new fuel designs-to provide assurance of stability performance.. RBS is scheduled to implement option I-A included in the BWROG topical report NED0-32339-A> " Reactor Stability Long-Term Solutions:
Enhanced Option-1-A." This is acceptable to the staff.
Vessel overpressure Protection The Main Steam Isolation Valva (MSIV) closure event was re-analyzed and demonstrated conformance to the ASME code in the MELLL region. The results of this analysis confimed that=the peak pressure of 1295 psig for MELLL operation-is-bounded by the' current cycle 7 result of 1311 psig and is within-ASME Section 111 limit of 1375 psig.
Containment Pressurization ~ Response / Dynamic Loading' Bounding short-term containment response analyses of the design basis LOCA event were performed-to demonstrate that operation in the MELLL domain will not result in exceeding' containment design limit. -The RBS final-safety-
. analysis report (FSAR) stecaline break analysis-is applicable to the MELLL
- conditions. A main steam line break,'although more limiting for peak temperature response was not re-analyzed for MELLL operation because the vessel dome pressure,is unchanged. Therefore, the analysis.results in SAR
.section 6.2-for steamline break remain valid in the MELLL region.
The recirculation line break in the MELLL region was analyzed to 'how that the s
drywell-wetwell pressure differential response remains below the 25 psid design value., The drywell airspace temperature-is within the current 1
recirculation line break analysis results in the-FSAR. The licensee submittal also stated:that MELLL operation will not significantly affect containment
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d namic loads.
sfort-ters LOCA analysis.The LOCA containment dynamic loads analysis is based upo The LOCA dynamic loads considered for MELLL i
o>eration include pool swell, condensation oscillation and chugging loads.
Tiese loads are bounded by previously-specified design values.
Control Rod Drop Accident I
Banked position withdrawal sequenct (BPWS) and rod patterns are used for RBS.
For plants using BPWS the control rod drop accident (CRDA)ith a high degree of has been j
statistically analyzed generically and it was found that w
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confidence the peak fuel enthalpy would not approach the 280 cal / gram 2
acce>tance criteria for this event.
In addition, the CRDA is a startup event
-whici would not be affected by MELLL operation.
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Anticipated Transient Without scram'(ATWS) l The basis for ATWS requirements is 10 CFR 50.62.
For BWRs, the rule includes requirements for an ATWS Recirculation Pump Trip (RPT), Alternate Rod Insertion (ARI) system and an equivalent Standby Liquid Control System (SLCS) j injection rate. A plant-specific ATWS analysis was performed to support RBS operation in the MELLL domain.
Results of the re-evaluation of a number of ATWS limiting transients demonstrated continued conformance to FSAR ATWS acceptance criteria. This is acceptable to the staff.
l Single Loop operation Single Loop Operation (SLO) analyses were reviewed to ensure their ap)1icability in the MELLL region.
The maximum power / flow state point acilevable with one recirculation pump operation is-83% power /54% flow. The proposed changes revise TS 3.4.1 to reflect the new power limit of 83% rated thermal power and SLO limits will be reviewed for adequacy for each future operating cycle, heactor Vessel Internals Integrity An evaluation was perfomed to identify potential increases in reactor internal pressure differences following a recirculation pump runout along the MELLL rod line. This evaluation shows that the pressure differences for MELLL operation resulting from recirculation pump runout are bounded by.the current design basis results for the upset condition. The licensee has also stated that reactor internal vibrations during HELLL operation will be bounded by previously established acceptance criteria. Therefore, it has been concluded that RBS can operate in the MELLL region without any detrimental effects on the reactor internals due to flow induced vibration or reactor internal pressure differences.
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. APRM Setpoint Changes The flow-biased APRM thermal power scram line and the APRM flow-biased rod l
block line are not credited in any RBS safety licensing analysis.
For the current licensed power / flow map, the flow-biased APRM thermal power scram line was defined as 0.66W+ 0.51, and the flow biased APRM rod block line was set at 0.66W40.457 where W is the recirculation drive flow in percent of rated.
These values are reflected in TS Table 3.3.1.1-1 which is being revised. With the current power / flow map expansion to include the MELLL domain, the upper boundary of the licensed operating domain is now extended to approximately the 121% rod line. To accomodate this expanded operating domain above the rated rod line, the setpoints for the flow-biased scram line and the flow-biased rod block line are now redefined as follows: nominal 0.66W+0.67 and 0.66W+0.61 (in note (b) of Table 3.3.1-1).
The licensee has stated that the establish-ment of the revised setpoints is consistent with the previously approved GE setpoint methodology.
These proposed TS changes are acceptable.
4.0
SUMMARY
E01 has performed the analyses for implementation of MELLL using current cycle limits to determine the impact of operation in the MELLL region.
The analyses have shown that the SLMCPR will not be exceeded for limiting A00s, and have also show that the impact on LOCA PCT is minimal for Cycle 7 and 8.
The licensee has stated that the im)act on PCT will be addressed prior to implementation to ensure that tle PCT limits in 10 CFR 50.46 are addressed.
The report shows that RBS can be operated within the MELLL region while continuing to support required safety margins.
The licensee also arovided changes to the technical specifications to reflect the changes for ME.LL implementation (see EVALUATI0li section " Single loop 0)eration" and "APRM Setpoint Changes" above).
These changes are approved as t1ey are consistent with the safety analysis.
Page changes from the Bases were also provided consistent with these changes and are acceptable.
5.0 STATE CONSULTATION
in accordance with the Comission's regulations, the Louisiana State Official was notified of the proposed issuance of the amendment. The State official had no coments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no d*H-e ipd+ w. e J:l4W Jess gui-mai*
O o public comment on such finding (62 FR 8799). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c) environme(9).
Pursuant to 10 CFR 51.22 ntal assessment need be prepar(b) no environmental impact statement or ed in connection with the issuance of the amendment.
7.0 CONCLUSION
The Commission has concluded based on the considerations discussed above, public w(1) there is reasonable assurance that the health and safety of the that:
ill not be endangered by operation in the proposed manner, activities will be conducted.in compliance with the Commission's reg (2) such
- ulations, and-(3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
G. Golub Date: August 26, 1997 I
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