ML20216C157
| ML20216C157 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 05/08/1998 |
| From: | Donohew J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216C161 | List: |
| References | |
| NUDOCS 9805190119 | |
| Download: ML20216C157 (8) | |
Text
~
Acacg b
3-4 UNITED STATES g
,g NUCLEAR REGULATORY COMMISSION
't WASHINGTON, D.C. 20656.0001
\\...../
ENTERGY OPERATIONS. INC.
SYSTEM ENERGY RESOURCES. INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION ENTERGY MISSISSIPPl. INC.
DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.136 License No. NPF-29 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee) dated September 18,1997, and supplemented by letter dated February 24,1998, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in corJormity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the Commission's regulations; O.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9805190119 980508 PDR ADOCK 05000416 P
PDR C
l 1
l
\\
. 2.
Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby amended to read as followt:
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. '.136, l
are hereby incorporated into this license. Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSIGN I
1 M.
h
/ Jack N. Donohew, Senior Project Manager N' Project Directorate IV-1 i
Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 8, 1998 I
ATTACHMENT TO LICENSE AMENDMENT NO.136 FACILITY OPERATING LICENSE NO NPF-29 DOCKET NO. 50-416 Replace the following pages of the Appendix A Technical Specifications with the attached pages.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
REMOVE INSERT l
2.0-1 2.0-1 5.0-20 5.0-20 B2.0-3 B2.0-3 83.2-8 B3.2-8 l
l l
i i
l 1
1
SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%
rated cor.ow:
THERMAL POWER shall be s 25% RTP.
- 2.1.1.2 With the reactor steam dome pressure 2 785 psig and core flow 210% rated core flow:
MCPR shall be 21.11 for two recirculation loop operation or 21.12 for single recirculation loop operation.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s 1325 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed:
2.2.1 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance with 10 CFR 50.72.
2.2.2 Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert allinsertable control rods.
)
2.2.3 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the plant manager cnd the corporate executive responsible for overall plant nuclear safety.
l
- MCPR values in T.S. 2.1.1.2 are applicable only for cycle 10 operation.
GRAND GULF 2.0-1 Amendment No. 120,101.136 I
h l
i Rsporting Rzquir&ments 5.6 5.6 Reporting Requirements t
5.6.5 Core Onerating Umits Renort (COLR)(continued) 10.
XN-NF-85-74(P)(A), "RODEX2A (BWR): Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA.
l 11.
XN-CC-33(P)(A),"HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," Exxon Nuclear Company, Inc.,
i Richland, WA.
12.
XN-NF-825(P)(A),"BWR/6 Generic Rod Withdrawal Error Analysis, MCPR, for Plant Operation Within the Extended Operating Domain,"
Exxon Nuclear Company, Inc., Richland, WA.
13.
XN-NF-81-51(P)(A),"LOCA-Seismic Structural Response of an Exxon Nuclear Company BWR Jet Pump Fuel Assembly," Exxon Nuclear Company, Inc., Richland, WA.
14.
XN-NF-84-97(P)(A),"LOCA-Seismic Structural Response of an ENC 9x9 BWR Jet Pump Fuel Assembly," Advanced Nuclear Fuels Corporation, Richland, WA.
15.
XN-NF-86-37(P)," Generic LOCA Break Spectrum Analysis for BWR/6 Plants," Exxon Nuclear Company, Inc., Richland, WA.
16.
XN-NF-82-07(P)(A)," Exxon Nuclear Company ECCS Cladding Swel!ing and Rupture Model," Exxon Nuclear Company, Inc., Richland, WA.
17.
XN-NF-80-19(A), Volumes 2,2A,2B, & 2C," Exxon Nuclear Methodology for Bolling Water Reactors EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA.
18.
XN-NF-79-59(P)(A)," Methodology for Calculation for Pressure Drop in BWR Fuel Assemblies," Exxon Nuclear Company, Inc., Richland, WA.
- 19.
NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR-il) with exception to the misplaced fuel bundle analyses as discussed in GNRO-96/00087 and the generic MCPR Safety Limit analysis as discussed in the generic MCPR Safety Limit anlaysis as discussed in GNRO-96/00100, letters from C. R. Hutchinson to USNRC.
- 20.
J11-02863SLMCPR, Revision 1,"GGNS Cycle 9 Safety Limit MCPR Analysis."
(continued)
- ltems 19 and 20 of TS 5.6.5.b are applicable only for Cycle 10 operation.
GRAND GULF 5.0-20 Amendment No. 120,131,136
Reactor Core SLs B 2.1.1 BASES APPLICABLE U.l.1 Fuel Claddina Intearity (continued)
SAFETY ANALYSES ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER > 50% RTP. Thus a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig is conservative. Because of the design thermal hydraulic compatibility of the reload fuel designs with the cycle I fuel, this justification and the associated low pressure and low flow limits remain applicable for future cycles of cores containing these fuel designs.
2.1.1.2 MCPR The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an A00 from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,
MCPR = 1.00) and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One spncific uncertainty included in the SL is the uncertainty inherent in the critical power correlation.
References 6 and 7 describe the methodology used in determining the MCPR SL.
The calculated MCPR safety limit is reported to the customary three significant digits (i.e., X.XX); the MCPR operating limit is developed based on the calculated MCPR safety limit to ensure that at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.
The fuel vendor's critical power correlations are based on a significant body of practical test data, providing a high degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual critical power being estimated. As long as the core pressure and flow are within the range of validity of the correlations, the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. These conservatisms and the (continued)
GRAND GULF B 2.0-3 LDC 98033
Reactcr Core SLs i
B 2.1.1 BASE 5 APPLICABLE 2.1.1.2 MCPR (continued)
I SAFETY ANALYSES inherent accuracy of tha fuel vendor's correlation provide a reasonable degree of assurance that 99.9% of the rods in the core would not be susceptible to transition boiling during sustained operation at the MCPR SL.
If boiling transition were to occur, there is reason to believe that the integrity i
i l
(continued)
GRAND GULF B 2.0-3a LDC 98033
MCPR B 3.2.2 BASES (continued)
SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is a 25% RTP and than every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER reaches 2 25% RTP 1ls acceptable given the large inherent margin to operating, limits at low power levels.
REFERENCES 1.
NUREG-0562, " Fuel Failures As A Consequence of Nucleate Boiling or Dry Out," June 1979.
2.
NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR-II).
3.
UFSAR, Chapter 15, Appendix 15B.
4.
UFSAR, Chapter 15, Appendix 15C.
5.
UFSAR, Chapter 15, Appendix 150.
6.
NEDE-30130-P-A, Steady State Nuclear Methods.
I 7.
NED0-24154, Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors.
8.
Deleted 9.
GNRI-xx/xxx, Amendment,_ to the Operating License.
l GRAND GULF B 3.2-8 LDC 98033