ML20116F014
| ML20116F014 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/31/1996 |
| From: | Hutchinson C ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19311C160 | List: |
| References | |
| GNRO-96-0087, GNRO-96-87, NUDOCS 9608060201 | |
| Download: ML20116F014 (12) | |
Text
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. 8" ENTERGY lnN!*"""' '"*-
Port Gcscn MS 39150 Tel 601437 2800 C. R. Hutchinson vce Preurjent 3 Cx] Ikd(Of CtfFM)
July 31,1996 U.S. Nuclear Regulatory Commission Mail Station P1-37 Washington, D.C. 20555 Attention: Document Control Desk
Subject:
Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF-29 Cycle 9 Reload Revision to Proposed Amendment to the Operating License (PCOL- 96/008, Revision 1)
GNRO-96/0087 Gentlemen:
Entergy Operations, Inc. is submitting by this letter a revision to the previously submitted proposed amendment to the Grand Gulf Nuclear Station (GGNS)
Operating License under GNRO-96/00053, dated May 9,1996. This revision supersedes the previous submittal in it's entirety.
The proposed amendment requested changes to those Technical Specifications (TS) required to support Grand Gulf Nuclear Station, Unit 1 Cycle 9 (Reload 8).
These changes included a change to the minimum critical power safety limit (SLMCPR) and changes to the references for the analytical methods used to determine the core operating limits. Cycle 9 will be the first cycle of operation with a mixed core of Siemens Power Corporation (SPC) 9x9-5 and General Electric (GE) GE11 reload fuel. The proposed amendment reflected the revised SLMCPRs for two-loop and single-loop operation This submittal revises the amendment submitted by letter dated May 9,1996, which supported MCPR Safety Limits of 1.10 and 1.11 for two-loop and single-loop operation respectively. General Electric has recently identified an error in their core simulator code, PANACEA, which affects the results of the GGNS safety limit analysis.
The PANACEA code uses the number of axial TIP measurements to calculate the axial elevations of the TIP/LPRM measurements. Entergy provided TIP data to
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4 July 31, 1996 GNRO-96/0087 Page 2 of 4 GE containing twenty-five axial nodes to facilitate modeling of the Cycle 9 non-GE fuel. However, in the safety limit analysis, PANACEA erroneously read this data as having twenty-four axial nodes and subsequently mapped the axial data incorrectly. The impact of this error on the GGNS Cycle 9 MCPR Safety Limit was determined to be a 0.02 increase in the two-loop MCPR safety limit and a 0.03 increase in the single-loop MCPR Safety Limit. As such, the revised GGNS Cycle 9 two-loop and single-loop MCPR Safety Limits are calculated to be 1.12 and 1.14. GE has indicated that this error does not affect any of their other fuel customers.
In accordance with the provisions of 10CFR50.4, the signed original of the requested revision to the previously submitted amendment is enclosed.
Attachment 2 provides the discussion and justification to support the proposed amendment revision. The original amendment request had been reviewed and accepted by the Plant Safety Review Committee and the Safety Review Committee. The conclusions of the Significant Hazards Considerations for this revision remain unchanged.
Based on the guidelines in 10CFR50.92, Entergy Operations has concluded that this revision to the amendment involves no additional significant hazards considerations as detailed in Attachment 2.
The revised General Electric report provided as Attachment 4 supports the proposed TS changes. General Electric considers the information contained in this report to be proprietary. In accordance with the requirements to 10CFR2.790(b), an affidavit is enclosed to support the withholding of the information contained in this report from public disclosure (Attachment 5).
Entergy Operations requests NRC approval and issuance of Technical Specifications changes by October 01,1996 to allow related work activities to be implemented.
Yours truly,
.Q. c.R. Mc4.~ ~
CRH/ACG/mte i
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- July 31, 1996 GNRO-96/0087 Page 3 of 4 i
4 attachments: 1. Affirmation per 10CFR50.30 2 2. GGNS PCOL-96/ 008, Revision 1 l 3. Mark-up of Affected Technical Specification Pages
- cc: Mr. J. E. Tedrow (w/a)
Mr. R. B. McGehee (w/a)
Mr. N. S. Reynolds (w/a) i Mr. H. L. Thomas (w/o) j Mr. J. W. Yelverton (w/a) 4 Mr. L. J. Callan (w/a) i Regional Administrator U.S. Nuclear Regulatory Commission i
Region IV 611 Ryan Plaza Drive, Suite 400
, Arlington, TX 76011 j Mr. J. N. Donohew, Project Manager (w/a) '
l Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13H3
- Washington, D.C. 20555 Dr. E. F. Thompson (w/a)
State Health Officer State Board of Health P.O. Box 1700 Jackson, Mississippi 39205 j
July 31, 1996 GNRO-96/0087 Page 4 of 4 bcc: Mr. D. G. Bost (w/a) l Mr. C. A. Bottemiller (w/a) !
Mr. B. M. Burmeister (w/a) l Mr. R. W. Byrd (w/o)
Mr. L. F. Dale (w/a)
Mr. L. F. Daughtery (w/a)
Mr. J. G. Dewease (w/a)
Mr. M. A. Dietrich (w/o)
Mr. C. W. Elisaesser (w/o)
Mr. C. B. Franklin (w/o) '
Mr. A. C. Goel (w/a)
Mr. J. .!. Hagan (w/o)
Mr. C. C. Hayes, Jr. (w/a)
Mr. J. J. Fisicaro (w3) (w/a)
I Mr. J. B. Lee (w/a)
Mr. M. D. McDowell (w/a)
Mr. M. J. Meisner (w/o)
Mr. D. C. Mims (ANO) (wia)
Mr. R. L. Patterson (w/a)
Mr. F. B. Rives (w/a)
Mr. J. E. Venable (w/a)
Mr. K. L. Walker (w/a)
Mr. M. D. Withrow (w/a)
File (LCTS/RPTS)(w/a)
File (Hard Copy)(w/a)
File (NS&RA)(w/a)
File (Central)(w/a)(57 )
Mr. D. A. Shelton (w/a)
Mail Stop V-920 lilinois Power Company Clinton Power Station P. O. Box 678 Clinton, IL 61727 Mr. B. S. Ferrell (w/a)
Mail Stop E 210 Cleveland Electric illuminating Company Perry Power Station P. O. Box 97 Perry, Ohio 44081
Attachment 1 to GNRO - 96/0087 BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMIF l LICENSE NO. NPF-29 i DOCKET NO. 50-416 IN THE MATTER OF MISSISSIPPI POWER & LIGHT COMPANY and SYSTEM ENERGY RESOURCES, INC.
and SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION and 4 ENTERGY OPERATIONS, INC.
AFFIRMATION 1, J. J. Hagan, being duly sworn, state that I am Acting Vice President, Operations GGNS of Entergy Operations, Inc.; that on behalf of Entergy Operations, Inc., System Energy Resources, Inc., and South Mississippi Electric Power Association I am authorized by Entergy Operations, Inc. to sign and file with the Nuclear Regulatory Commission, this application for amendment of the Operating License of the Grand Gulf Nuclear Station; that I signed this application as Acting Vice President, Operations GGNS of Entergy Operations, Inc.; and that the statements made and the matters set forth therein are true and correct to the best of my knowledge, information and belief.
J . Ha'gan STATE OF MISSISSIPPI COUNTY OF CLAIBORNE SUBSCRIBED AND SWORN TO before me, a Notary Public, in and for the County and State above named, this Sl" day of ALY .1996.
(SEAL) dlk%Lb5 Plss:::WISTATEWe2 NOTAW FUDuc Notah Public
!!Y COM.4!S!!O'l C:PCI] JC. 2's. 2003 My commission expires: temTo sm0Ta Srm .
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Attachment 2 to GNRO-96/0087 i
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GGNS PCOL-96/008 1
(REVISION 1) i i
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l . Grand GulfNuclear Station Attachment 2 to GNRO - 96/0087 Cycle 9, PCOL-96/008, Revision 1 Page 1 of 6 A. AFFECTED TECHNICAL SPECIFICATIONS
- 1. The following Technical Specifications are affected by the proposed change:
l 2.1.1 Reactor Core Safety Limits l 5.6.5 Core Operating Limits Report '
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- 2. The following Technical Specification Bases are affected by the proposed change.
Since Technical Specification Bases are controlled under 10CFR50.59 Program, l the markup of the Bases Sections are provided for information only:
1 B 2.1.1.1 Fuel Cladding Integrity B 2.1.1.2 MCPR B 2.0 References l
B 3.2.2 Minimum Critical Power Ratio (MCPR)
B 3.2 References B. DESCRIPTION OF CHANGES
- 1. Technical Specification 2.1.1.2: Change the Safety Limit MCPR for Two Loop Operation and Single Loop Operation to 1.12 and 1.14, respectively. l
- 2. Technical Specifications 5.6.5: Add the following:
- 19. NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR-II) with exception to the misplaced fuel bundle analysis as discussed in GNRO 96-00053, letter from C. R.
liutchinson to USNRC dated May 09,1996.
- 3. Bases 2.1.1.1: Change "ANFB" to " Fuel Vendor's Critical Power". Change 585 psig to 785 psig, >0.25 to >0.3, and Ref. 2 to Ref. 6.
- 4. Bases 2.1.1.2: Change "ANFB" to " Fuel Vendors", " correlation is" to
" Correlations are". Change Reference 2 to Reference 6. Delete "ANFB" and change " correlation" to " correlations". Delete sentence "Still further ... . .. . boiling i transition". Reword, as marked, line 16 of second paragraph of Section 2.1.1.2. !
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l 5. Bases 2.0 (References): Add References to GESTAR-II as item 6.
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. Grand GulfNuclear Station Attachment 2 to GNRO - 96/0087 Cycle 9, PCOL-96/008, Revision 1 Page 2 of 6
- 6. Bases 3.2.2: Delete last part ofline 8 in the first paragraph "and the . . .. . ...(Ref.
7)", and change Ref. 8 to Ref. 7 in the third line of the second paragraph.
- 7. Bases 3.2
References:
Replace References 2 with "NEDE-24011-P-A, General l Electric Standard Application for Reactor Fuel (GESTAR-II)", Reference 6 with l "NEDE-30130-P-A, Steady State Nuclear Methods", and Reference 7 with 1 "NEDO-24154, Qualification of the One-Dimensional Core Transient Model for
{
Boiling Water Reactof'.
I C. BACKGROUND The core for the current cycle (Cycle 8) at Grand Gulf Nuclear Station (GGNS) is composed entirely of Siemens Power Corporation (SPC) 9x9-5 fuel bundles. Entergy has l recently contracted with General Electric Nuclear Energy (GE) to provide the next three i batches of reload fuel. The next reload batch will be composed entirely of GE's Gell .
9x9 design. As such, the MCPR safety limit for the mixed-vendor core and the core operating limits must be developed with General Electric's analytical methods.
Representatives from Entergy Operations, Inc., and General Electric (GE) Company met with the NRC staff on December 12,1995 to discuss the mixed-core licensing approach for Cycle 9. As presented at this meeting, a cycle-specific evaluation has been performed.
The Cycle 8 MCPR safety limit was calculated with SPC's NRC-approved safety limit methodology reported in ANF-524 (P)(A), AdvancedNuclear Fuels Corporation Critical PowerMethodologyfor Boiling Water Reactors. With this methadalogy, SPC calculated a MCPR safety limit that is tailored to the GGNS plant and Cycle 8 core design. This analysis produced a MCPR safety limit of 1.06 and considered NRC-approved uncertainties for the BWR/6 feedwater flow and SPC's POWERPLEX core monitoring system.
GE's approach to the MCPR safety limit is significantly different than SPC's. As described in the NRC-approved NEDE-24011-P-A, General Electric Standard Applicationfor Reactor Fuel (GESTAR-II), a bounding MCPR safety limit that is generic to each GE fuel type is developed.
l The generic MCPR safety limit for Gell fuel is reported as 1.07 in NEDE-31152-P, GeneralElectric Fuel Bundle Designs. However, since approximately two-thirds of the
! GGNS Cycle 9 core will be composed of SPC's fuel, the MCPR safety limit analysis must consider the effects of this mixed-vendor core. As such, a cycle-specific analysis was performed considering Cycle 9 core design. This approach is consistent with that presented in our meeting with the NRC in December 1995. Any additional changes to the Cycle 9 core design are expected to be minor and will be evaluated to ensure the MCPR safety limit is unaffected.
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Grand Gulf Nuclear Station Attachment 2 to GNRO - 96/0087 Cycle 9, PCOL-96/008, Revision 1 {
Page 3 of 6 GE GGNS Cycle-9 SLMCPR Report Jil-02863SLMCPR, Revision 1 (Attachment 4) documents the GGNS Cycle 9 MCPR safety limit analysis. This evaluation was performed with the plant uncertainty values reported in General Electric BWR Thermal Analysis Basis (GETAB). Additional uncertainties associated with the CPR prediction of the SPC fuel were included. This evaluation contains considerable conservatism, especially in the uncertainty values associated with power distribution monitoring. In order to evaluate the sensitivity of the MCPR safety limit to this parameter, an analysis was performed with an uncenainty value characteristic of the GGNS Cycle 9 core monitoring system in place of GE's original P1 core monitoring system. This evaluation concluded that significantly lower MCPR safety limit values are applicable to GGNS Cycle 9.
I By letter dated May 9,1996, Entergy submitted MCPR Safety Limits of 1.10 and 1 1.11 for two-loop and single-loop operation respectively. However, General Electric has recently identified an error in their core simulator code, PANACEA, which j affects the results of the GGNS Safety Limit analysis. The GGNS Cycle 9 MCPR l Safety Limits predicted by the corrected code are 1.12 and 1.14 for two-loop and j single-loop operation respectively.
As discussed at the December meeting, GGNS will maintain its current licensing basis for l the misplaced bundle events. As such, the mis-oriented and mis-located bundle events will l be analyzed as accidents subject to an acceptance criterion of a small fraction (10%) of 10CFR100 as reported in NUREG-0800, Section 15.4.7.
D. PROPOSED TS CHANGES The proposed changes to the Technical Specifications are to change the MCPR safety I
limit values for two-loop and single-loop operation to those values calculated by GE's methodology for GGNS Cycle 9. These marked-up Technical Specifications are included ;
as Attachment 3. !
The COLR methodology references will also be updated to include GE's GESTAR-II ;
report. The GESTAR reference is added to the list of documents that have been reviewed and approved by the NRC without a revision number to maintain consistency with the l other COLR methodology referenv md to allow reference to the upcoming revision to l GESTAR which will include this cyca, _; :ific analytical approach for the MCPR safety j limit. The SPC reports currently listed in the Technical Specifications will be unaffected since SPC fuel will remain in the Cycle 9 core.
l E. JUSTIFICATION The MCPR Safety Limit is developed to assure compliance with General Design Criteria 10 of 10CFR50 Appendix A. The Bases to Technical Specification 2.1.1 states that "The MCPR SL ensures sufficient censervatism in the operating MCPR limit that, in the event of an Anticipated Operatior.al Occurrences (AOO) from the limiting condition of i
Grand Gulf Nuclear Station Attachment 2 to GNRO - 96/0087 Cycle 9, PCOL-96/008, Revision 1 Page 4 of 6 i
operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition". The new MCPR SL was developed with considerable conservatism in the methodology.
GE GGNS Cycle-9 SLMCPR Report Jil-02863SLMCPR, Revision 1 (Attachment 4) documents the GGNS Cycle 9 MCPR safety limit analysis. This evaluation was performed with the plant uncertainty values reported in GETAB. Additional uncertainties associated with the CPR prediction of the SPC fuel were included. This evaluation contains considerable conservatism, especially in the uncertainty values associated with power i
distribution monitoring. In order to evaluate the sensitivity of the MCPR safety limit to this parameter, an analysis was performed with an uncertainty value characteristic of the GGNS Cycle 9 core monitoring system (3D MONICORE) in place of GE's original .Pl l core monitoring system. This evaluation concluded that significantly lower MCPR safety limit values are applicable to GGNS Cycle 9. As discussed at the GE/NRC meeting on March 17,1996, GE will revise GESTAR to incorporate this cycle-specific analytical approach for the MCPR safety limit.
Recent plant-specific analyses have identified that the gel 1 generic MCPR safety limit may not be conservative for some plants. This GGNS analysis, however, explicitly calculates the MCPR safety limit for the GGNS Cycle 9 core design and does not credit the generic analyses. Therefore, this analysis is not affected by the current issues concerning the applicability of the generic MCPR safety limit values.
F. CONCLUSION For two-loop operation, a Safety Limit MCPR of 1.12 was demonstrated to be adequate to ensure that 99.9 percent of the rods in the core avoid a boiling transition during the most limiting AOO. For single--loop operation, the limit is increased by 0.02 to 1.14. The MCPR fuel cladding integrity safety limits for GGNS Cycle 9 two-loop and single-loop operation were determined by applying the generic GE Safety Limit MCPR methodology to the GGNS Cycle 9 core design. The SPC fuel was explicitly considered and found to be bounded by the limiting Gell bundles. This approach has been presented to the NRC Staff and contains considerable conservatism in the applied uncertainties. The resulting values, therefore, represent bounding measures of the GGNS Cycle 9 Safety Limit MCPRs.
G. SIGNIFICANT IIAZARDS CONSIDERATION i
Entergy Operations, Inc. proposes to change the current Grand Gulf Nuclear Station Technical Specifications. The specific change is to modify the Minimum Critical Power Ratio (MCPR) safety limits reported in Technical Specification 2.1.1.2, the list of references in Technical Specification 5.6.5, and associated Bases changes. The proposed change is necessary in order to switch reload fuel vendors.
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Grand GulfNuclear Station Attachment 2 to GNRO - 96/0087 Cycle 9, PCOL-96/008. Revision 1 Page 5 of 6 The Commission has provided standards for determining whether no significant hazards considerations exists as stated in 10 CFR 50.92 (c). A proposed amendment to an operating license involves no significant hazards if operation of the facility in accordance 1 l with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
l Entergy Operations, Inc. has evaluated the no significant hazards consideration in its request for this license amendment and determined that no significant hazards considerations result from this change. In accordance with 10 CFR 50.91(a), Entergy Operations, Inc. is providing the analysis of the proposed amendment against the three standards in 10 CFR 50.92(c). A description of the no significant hazards consideration determination follows:
l I. The proposed change does not significantly increase the probability or consequences of an accident previously evaluated.
i l The Minimum Critical Power Ratio (MCPR) safety limit is defined in the Bases to l Technical Specification 2.1.1 as that limit which " ensures that during normal operation and during Anticipated Operational Occurrences (AOOs), at least 99.9%
of the fuel rods in the core do not experience transition boiling." The MCPR safety limit is re-evaluated for each reload and, for GGNS Cycle 9, the analyses have concluded that a two-loop MCPR safety limit of 1.12 based on the l l application of the generic GE MCPR methodology is necessary to ensure that this l acceptance criterion is satisfied. For single-loop operation, a MCPR safety limit of 1.14 based on the generic GE MCPR methodology was determined to be l necessary. Core MCPR operating limits are developed to support the Technical Specification 3.2 requirements and ensure these safety limits are maintained in the event of the worst-case transient. Since the MCPR safety limit will be maintained at all times, operation under the proposed changes will ensure at least 99.9% of the fuel rods in the core do not experience transition boiling. Therefore, The Minimum l
Critical Power Ratio (MCPR) safety limit change does no'. affect the probability or consequences of an accident.
l The implementation of GE's GESTAR-Il approved methodology has no effect on l the probability or consequences of any accidents presiously evaluated. One I
exception to GESTAR is that the mis-oriented and mis-located bundle events will continue to be analyzed as accidents subject to the acceptance criteria in the current licensing basis. The design of the gel 1 fuel bundles is such that the bundles are not likely to be mis-oriented or mis-located and the normal administrative controls will be in effect for assuring proper orientation and location. Therefore, the probability of a fuel loading error is not increased. This analysis ensures that postulated dose releases will not exceed a small fraction (10 l
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l Grand GulfNuclear Station Attachment 2 to GNRO - 96/0087 Cycle 9, PCOL-96/008, Revision 1 Page 6 of 6 l percent) of 10CFR100 limits. Therefore, the consequences of accidents previously evaluated are unchanged.
II. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The GE11 fuel to be used in Cycle 9 is of a design compatible with fuel present in i the core and used in the previous cycle. Therefore, the gel 1 fuel will not create .
the possibility of a new or different kind of accident. The proposed changes do not involve any new modes of operation, any changes to setpoints, or any plant ,
l modifications. They introduce revised MCPR safety limits that have been proved l to be acceptable for Cycle 9 operation. Compliance with the applicable criterion t
for incipient boiling transition continues to be ensured. The proposed MCPR safety limits do not result in the creation of any new precursors to an accident. I Therefore, the proposed changes do not create the possibility of a new or different type of accident from any accident previously evaluated. l l
III. The proposed change does not involve a significant reduction in a margin of :
safety.
The MCPR safety limits have been evaluated to ensure that during normal I operation and during AOOs, at least 99.9% of the fuel rods in the core do not experience transition boiling. Therefore, the implementation of the proposed
, changes in the MCPR safety limit ensure there is no reduction in the margin of ,
l safety.
I As with the current SPC methodology, GGNS will implement only the NRC-approved revisions to GE's GESTAR methodology. This GE methodology is similar to those SPC reports currently listed in TS 5.6.5 and it will be applied in a similar, conservative fashion. One exception to GESTAR is that the mis-oriented and mis-located bundle events will continue to be analyzed as accidents subject to l
the acceptance criteria in the current licensing basis. This analysis ensures that postulated dose releases will not exceed a small fraction (10 percent) of 10CFR100 limits. On this basis, the b 'nlementation of this GE methodology does not involve a significant reduction in a margin of safety.
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