ML20126F512
| ML20126F512 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/03/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126F505 | List: |
| References | |
| TAC-56776, NUDOCS 8506170511 | |
| Download: ML20126F512 (8) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 93 TO FACILITY OPERATING LICENSE N0. DPR-46 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298
1.0 INTRODUCTION
By letters dated January 10, 1985 (Ref. 1) and February 28, 1985 (Ref. 8),
the Nebraska Public Power District (NPPD) proposed to revise the Technical Specifications for Cooper Nuclear Station (CNS). The proposed amendment would support operation of CNS during the upcoming fuel Cycle 10 (Reload 9) and expand the flexibility of plant limits to permit operation with barrier-type fuel and hafnium (General Electric Hybrid I) control rods. The proposed amendment would revise the following areas:
(1) rod block monitor (RBM) upscale trip setting, (2) maximum average planar linear heat generation rate (MAPLHGR) curves, (3) minimum critical power ratio (MCPR) curves, and (4) description of control rod materials. A Supplemental Reload Licensing Submittal, prepared by General Electric and dated November 1984 (Ref. 2),
was transmitted by Reference 1 to support the proposed amendment.
The Cooper Nuclear Station Cycle 10 core will consist of 548 fuel assemblies of which 432 are from previous cycles and 116 are new. The new fuel assemblies are of types P8DRB265L and P8DRB283 which are the same types of fuel added in the previous Cycle 9.
The core for Cycle 10 consists entirely of these two types of fuel except for 12 type 80RB283 fuel assemblies which were loaded originally in Cycle 5.
Type 80RB283 fuel assemblies are the same as the type P8DRB283 fuel assemblies except that they are not pressurized.
2.0 EVALUATION 2.1 Fuel Mechanical Design The 116 new General Electric (GE) fuel assemblies (88 of type P8DRB265L and 28 of type P8DRB283) to be loaded in Cycle 10 are identical to the fuel loaded in previous cycles. These fuel types are standard General Electric reload designs as described in Reference 3.
This reference has been approved (Ref. 4) for such use and we conclude that no further review of the fuel design is required. The licensee has also' proposed using barrier type fuel (BP80RB265L and BP8DRB283). This type of fuel has been approved (Ref. 5) and we conclude that no further review of this type of fuel design is required.
3 8506170511 850 298 PDR ADOCK PDR P
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. 2.2 Nuclear Design The nuclear design of the Cycle 10 reload core has been performed with methods and techniques that are described in Reference 3.
The results of the analyses are given in Reference 2.
Those results are within the range of those usually encountered for boiling water reactor (BWR) reloads and are acceptable.
In particular the shutdown margin is 0.011 in AK.
The Standby Liquid Control System is capable of making the system subcritical at cold Xenon-free conditions assuming 600 ppm of boron by an amount of 0.038 AK.
These results are acceptable. Since they have been obtained by previously approved methods, we conclude that the nuclear design of the Cycle 10 reload core is acceptable.
2.3 Thermal-Hydraulic Design The thermal-hydraulic (T-H) analysis of the Cycle 10 core was performed with methods and techniques described in Reference 3.
Analyses were done at a power level of 2482 thermal megawatts.
2.4 Minimum Critcial Power Ratio A safety limit value of 1.07 for the core-wide minimum critical power ratio (MCPR) is used for Cycle 10. This value is generic for BWR reloads and is acceptable for Cycle 10.
The operating limit MCPR is obtained by performing analyses of anticipated events in order to determinc the reduction in critical power (aCPR) resulting from them. Analysis methods, including treatment of uncertainties, are described in Reference 3.
The operating limit MCPR is established by adding the largest value of ACPR to the safety limit value.
2.5 Thermal-Hydraulic Stability The Cooper reload submittal relies on the GE cycle specific analysis procedure (GESTAR) to demonstrate that the reactor has sufficient margin to be free of thermal-hydraulic instabilities. The maximum deca Cooper submittal is.86.
Our evaluation (Ref. 6) y ratio calculated in the of the GE T-H stability methodology has shown that their method has an uncertainty of.2 in calculated decay ratio. Since the Cooper (.86) decay ratio is based on a best estimate calculation their true decay ratio could be as high as 1.06 (.86 +.2).
Since a decay ratio greater than 1.00 indicates an undamped oscillation the Cooper analysis does not show any margins from undamped oscillations.
Our evaluation (Ref. 6) of the GE T-H stability methodology also concludes j
that a core design consisting of approved GE fuel bundles in conjunction with General Electric Service Information Letter (SIL)-380 operating recommendations incorporated into the Technical Specifications is in compliance with GDC 10 and 12 requirements. Since the licensee could not show through analysis that T-H instabilities are prevented by design, he has committed (Ref. 11) to incorporate GE SIL-380 recommendations into plant operating procedures prior 1
t l to startup of Cycle 10 and to translate the procedures into plant Technical Specifications in a timely manner.
With the licensee's comitments relative to the GE SIL-380 recomendations (Ref.11), the staff concludes that themal hydraulic instability does not pose a safety concern for continued operation of Cooper.
l 2.6 Transient and Accident Analyses Transient and accident analysis methods are described in Reference 3.
These j
are the same methods that have been used in previous cycles for CNS and they are acceptable for Cycle 10.
2.7 Pressurization Events The one-dimensional transient code ODYN has been used to analyze these events.
The licensee has elected to use ODYN Option B in which measured rod scram times are used.
For this option the pressurization events are not limiting.
If Option A scram times are used the Load Rejection Without Bypass event is limiting. Use of the Option B mode is widespread in boiling water reactors and its use is acceptable for CNS.
2.8 Non-Pressurization Events The licensee has elected to use the generic analysis results for the rod withdrawal event. This has been approved by the staff for BWR reloads and is acceptable for CNS. The rod withdrawal error event is the limiting event' for Option B and establishes the operating limit MCPR for Cycle 10 of 1.22.
The analysis of this event has been performed by the approved methods of Reference 3 and is acceptable.
2.9 Loss-of-Coolant Accident The loss-of-coolant accident has been analyzed for Cycle 10 of CNS at a power of 2532 themal megawatts (2482 x 1.02). The analysis has been j
performed by the approved methods of Reference 7 and is acceptable.
2.10 Rod Drop Accident A cycle specific rod drop accident analysis has been performed for Cycle 10 of CNS for both the hot and cold shutdown cases since the parameters of the generic analysis were not bounding for these cases. The result is less than the NRC criterion of 280 calories per gram for the peak enthalpy in both analyses. Since this meets our criterion for this event it is acceptable.
2.11 Technical Specifications l
The changes to be made to the Technical Specifications are due to the following l
circumstances:
t l
l
. 1.
Page 61 (Table 3.2.C) - The equation for the Rod Block Monitor (RBM)
Upscale (Flow Bias) was changed.
2.
Page 62 (Notes for Table 3.2.C.) - The note for the RBM in Table 3.2.C was changed.
In addition, the staff requested the reference for the method of calculating the value of N, the RBM setpoint selected.
This was provided in Reference 8.
3.
Page 211b - Figures 3.11-1.5 and 3.11-1.6 - Maximum Average Planar Linear Heat Generation (MAPLHGR) Rate versus Exposure with LPCI Modification and Bypass Flow Holes Plugged. The titles to these figures were changed to include their application to the new befrier fuel.
4.
Page 212d - Figure 3.11-2c - Minimum Critical Power Ratio (MCPR) vs T(based on tested measured scram time as defined in Reference 9) for 8x8R Fuel (80C to E0C - 1000 mwd /ST). This figure is the result of new analyses.
5.
Page 212e - Fi mwd /ST to E0C)gure 3.11-2d - MCPR vs i for 8x8R Fuel (E0C - 1000 This figure is the result of new analyses.
6.
Page 212f - Figure 3.11-2e - MCPR vs T for P8x8R and BP8x8R Fuel (BOC to E0C - 1000 mwd /ST). This figure is the result of new analyses.
7.
Page 212g - Figure 3.11-2f - MCPR vs i for P8x8R and BP8x8R Fuel (E0C - 1000 mwd /ST to EOC). This figure is the result of new analyses.
8.
Page 217 - Section 5.2.B currently describes the control rod material as boron carbide power (B C) compacted to approximately 70 percent g
theoretical density.
In inticipation of the use of the GE Hybrid I Control Rod (HICR) which contains approximately 15 percent hafnium, the licensee wishes to also include this type in the description.
Each of these is discussed below.
2.12 RBM Upscale (Flow Bias) Equation Modification The trip level setting equation for the RBM Upscale (Flow Bias) was changed to 0.66W + (N-66) from (0.66W+40%). This allows the value of N the RBM setpoint (%), to be calculated in accordance with the latest NRC approved version of NEDE-24011-A. This change follows approved procedures and is, therefore, acceptable.
2.13 Note to RBM Upscale Equation Modification This change serves to explain the tenns used for the RBM Upscale equation and is acceptable.
e 1
- 2,14 MAPLHGR vs Exposure - Figures 3.11-1.5 and 3.11-1.6 The titles to Figures 3.11-1.5 and 3.11-1.6 were changed to include BP80RB265L and BP8DRB283 barrier fuel respectively which exhibits the same values as the existing fuel. The tem " barrier fuel" stems from the use of a 0.003-inch thick, high purity zirconium liner, i.e., barrier bonded to the inner surface of the Zircaloy-2 portion of the fuel rod cladding. The overall di,ensions of the fuel rods are the same as for the GE 8x8 prepressurized retrofit bundle. The use of the barrier fuel was approved in 8 'erence 5 and is, therefore, acceptable.
2.15 MCPR Specification Format - Figure 3.11-2c 8x8R Fuel (BOC to EOC - 1000 mwd /5T)
The cycle Minimum Critical Power Ratio (MCPR) as a function of tne parameter t is presented in curve form. This is from the analyses perfomed for non-pressurization events and pressurization events (for both the Option B scram time and the Option A [ Technical Specification] scram time) in order to establish end points on the curves. The curve of MCPR as a function of t is consistent with the results of the safety analysis (Ref. 2) and is acceptable.
The format of the Technical Specification is similar to that of other plants using Option B and is acceptable.
2.16 MCPR Specification Fomat - Figure 3.11-2d 8x8R Fuel (EOC-1000 mwd /5T to EOC)
This is acceptable for the same reasons as given for 2.15.
2.17 MCPR Specification Format - Figure 3.11-2e P8x8R and BP8x8R Fuel (BOC to EOC-1000 mwd /ST)
This is acceptable for the same reasons as given for 2.15.
2.18 MCPR Format - Figure 3.11-2f P8x8R and BP8x8R Fuel (EOC-1000 mwd /ST to E0C) 1 This is acceptable for the same reasons as given for 2.15.
2.19 Hybrid I Control Rods (HICR)
The description of the HICR control rods was submitted to NRC by General Electric in Topical Report NEDE-22290-A. The Safety Evaluation of the Type I HICR was reviewed and approved by the staff (Ref. 10). The probability of occurrence or the consequences of an accident would not be increased above those analyzed in the Final Safety Analysis Report (FSAR) because the weight and envelope of the HICR are identical to those of currently used assemblies, and the nuclear and mechanical properties of the HICR do not differ from currently used assemblies in a significant way. The staff has made similar assessments for Monticello, Dresden Unit 2. Hatch Unit 1, and Brunswick 1 and i
. 2.
Therefore, the use of the HICR control rods is acceptable and the addition of the HICR description in the Technical Specifications is acceptable.
3.0
SUMMARY
As a result of our review, which is described above, we conclude that the reload and Technical Specification changes proposed by the licensee letters dated January 10 and February 28, 1985, are acceptable. This conclusion is based on the following:
a.
Previously approved analysis methods and techniques are employed.
b.
The consequences of the transients and accidents which are affected by the reload are acceptable for Cycle 10.
)
c.
The revisions to the Technical Specifications have been found to be r
acceptable, i
d.
The licensee has committed (Ref. 11) to incorporate the GE SIL-380 recomendations relative to thermal-hydraulic stability (Section 4.2) into the plant operating procedures prior to startup of Cycle 10 and to translate the procedures into the plant Technical Specifications 1
in a timely manner.
4.0 ENVIRONMENTAL CONSIDERATION
S i
This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. -
The 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 Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
S We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and 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 Contributors:
H. Balukjian, G. Schwenk, W. Brooks Dated:
June 3,1985
l REFERENCES 1.
Letter, L. C. Kuncl, Nebraska Public Power District, to D. B. Vassallo I
(NRC) dated January 10, 1985.
i 2.
" Supplemental Reload Licensing Submittal for Cooper Nuclear Power Station Unit 1, Reload 9," General Electric Company Report 23A1813 Class 1 November 1984 j
3.
GESTAR II, General Electric Standard Application for Reactor Fuel, NEDE-24011-A, latest approved version.
4.
Approval letter, D. G. Eisenhut (NRC) to R. Gridley (GE) dated May 12, 1978, and Supplements thereto, fonning Appendix C to Reference 3.
5.
Letter, C. O. Thomas (NRC) to J. S. Charnley (GE) dated April 13, 1983.
6.
Letter, C. O. Thomas (NRC) to H. C. Pfefferlen (GE) dated April 24, 1985 - Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment 8 " Thermal Hydraulic Stability Amendment to GESTAR II."
7.
" Loss-of-Coolant Accident Analysis Report for Cooper Nuclear Power station," NED0-24095, August 1977 (as amended).
8.
Letter, J. M. Pilant, Nebraska Public Power District, to D. B. Vassallo (NRC), dated February 28, 1985.
9.
Letter, R. H. Buchholz (GE) to P. S. Check (NRC), dated September 5, 1980,
- 10. Letter, C. O. Thomas (NRC) to J. F. Kapproth (GE), " Acceptance of Referencing of Licensing Topical Report NEDE-22290, Safety Evaluation of the General Electric Hybrid I Control Assembly." dated August 22, 1983.
- 11. Letter, J. M. Pilant, Nebraska Public Power District, to D. B. Vassallo (NRC), dated April 4, 1985.
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