ML20211K991

Safety Evaluation Supporting Amend 106 to License DPR-46

ML20211K991
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Site:	Cooper
Issue date:	12/09/1986
From:	Office of Nuclear Reactor Regulation
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UNITED STATES g

8' NUCLEAR REGULATORY COMMISSION o

h WASHINGTON, D. C. 20555 kn....,/

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.106 TO FACILITY OPERATING LICENSE NO. DPR-46 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1. 0 INTRODUCTION By letter dated September 17, 1986 (Ref. 1), Nebraska Public Power District (the licensee) made application to modify the Cooper Nuclear Station (CNS) technical specifications to accommodate the Cycle 11 reload.

The letter contained an attachment describing proposed technical specification changes (revised pages 210, 212f, 214b and 217).

The licensee also provided a supplemental reload licensing submittal prepared by the General Electric Company dated May 1986 (Ref. 2) which describes the results of the engineering and reload licensing analyses.

The description of the Cycle 11 core is given below.

2.0 EVALUATION 2.1 Description of the Cycle 11 Core The Cooper Nuclear Station Cycle 11 core will consist of 548 fuel assemblies of which 396 are from previous cycles and 152 are new.

The new fuel assemblies are of type BP8DRB283 (barrier fuel) which has been approved for use at CNS (Amendment 93) but has not previously been used at CNS.

The core for Cycle 11 consists of the 152 new barrier fuel assemblies; 176 type P80RB283 assemblies (28 from Cycle 10, 60 from Cycle 9, 56 from Cycle 8, and 32 from Cycle 7); and 220 type P8DRB265L assemblies (88 from Cycle 10, 56 from Cycle 9, 56 from Cycle 8, and 20 from Cycle 7).

2.2 Fuel Mechanical Design The 152 new General Electric (GE) fuel assemblies to be loaded in Cycle 11 are of type BP8DRB283 barrier fuel which exhibits the same design specification parameters as the existing fuel.

The term " 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 portion of the fuel rod cladding.

The overall dimensions of the fuel rods are the same as for the GE 8x8 prepressurized retrofit bundle.

The use of the barrier fuel was approved in Reference 3 and is therefore acceptable for CNS.

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2 2.3 Nuclear Design The nuclear design of the Cycle 11 core has been performed with standard General Electric reload methods and techniques which are described in GESTAR II (Reference 4).

The results of the analyses are given in Reference 2 in standard GESTAR II format.

The shutdown margin of the new core meets the technical specification requirement that the core be at least.38% Ak subcritical in the most reactive condition when the highest worth control rod is fully withdrawn and all other rods are fully inserted.

For CNS Cycle 11, GE calculated that the cjbe,iththestrongestrodoutisequalto.985atbeginningofthe k

w which for this cycle is the core burnup providing minimum shutdown margin (1.5% Ak).

The standby liquid control system is capable of bringing the reactor from full power to a cold shutdown condition assuming none of the withdrawn control rods is inserted.

The 600 ppm boron concentration will bring the reactor subcritical to k condition (Ref. 2).

SincetheseresultIfba=ve.964at20 C xenon free been obtained by previously approved methods and fall within the expected range, we conclude that the nuclear design of the Cycle 11 reload core is acceptable.

2.4 Thermal-Hydraulic Design The thermal-hydraulic design for CNS has been performed with the methodology described in GESTAR II (Reference 4) and the results are given in Reference 2.

The parameters used for the analysis are those approved in Reference 4.

The objective of the review is to confirm that the thermal-hydraulic design of the core has been accomplished using acceptable methods, and provides an acceptable margin of safety from conditions which could lead to fuel damage during normal and anticipated operational transients, and is not susceptible to uncontrolled power oscillations due to thermal-hydraulic instability.

The review includes the following areas:

(1) safety limit minimum critical power ratio (MCPR), and (2) operating limit MCPR.

These are described below.

2.5 Thermal-Hydraulic Stability The CNS Cycle 11 reload was not analyzed for thermal-hydraulic instabilities.

In Amendment 94 the CNS technical specifications were amended to implement improved stability monitoring requirements consistent with GE Service Information Letter 380.

In accordance with the staff Safety Evaluation (Ref. 5) of the General Electric Topical Report NEDE-24011 Revision 6, Amendment 8, the facility is exempted from the requirement to perform a cycle specific stability analysis.

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3 2.6 Transient and Accident Analyses Transient and accident analysis methods are the approved General Electric methods described in Reference 4.

These are the same methods that have been used in previous cycles for CNS and they are acceptable for Cycle 11.

A safety limit MCPR has been imposed to assure that 99.9 percent of the fuel rods in the core are not expected to experience boiling transition during normal and anticipated operational transients. As stated in Reference 5, the approved safety limit MCPR is 1.07.

The safety limit MCPR of 1.07 is used for Cycle 11 operation.

The most limiting events have been analyzed by the licensee to determine which event could potentially induce the largest reduction (ACPR) in the initial critical power ratio.

The ACPR values given in Section 10 Reference 2 are plant specific values calculated by the approved methods including ODYN methods.

The calculated ACPRs are adjusted to reflect the calculational uncertainties by employing the conversion methods described in Reference 7.

The operating limit MCPR (0LMCPR) values are determined by adding the adjusted ACPRs to the safety limit MCPR.

Section 12 of Reference 2 presents the Cycle 11 MCPR values of both the pressurization and non pressurization transients.

The maximum MCPR value in Section 12 is the operating limit MCPR for the cycle and must be bounded by the technical specification (TS).

The value of operating limit MCPR is determined by the limiting transients, Rod Withdrawal Error (RWE), Feedwater Controller Failure (FWCF) and Load Rejection Without Bypass (LRWBP).

The analysis of these events for CNS, via the ODYN Option B approach, provide new Cycle 11 technical specification values of OLMCPR as a function of average scram time, T.

The proposed Cycle 11 TS change for OLMCPR is given in Figure 3.11-2e for BOC to E0C -1000 mwd /ST.

That is linear as OLMCPR =.14 T + 1.22, however, the Cycle 10 TS is OLMCPR =

1.22 when T 5.38 l

.225 (T

.38) + 1.22 when T h.38 therefore, the proposed Cycle 11 TS based on the analysis for the new fuels bounds the Cycle 10 TS values, and is acceptable.

The limiting overpressurization event, the main steam isolation valve (MSIV) closure with flux scram, analyzed with standard GESTAR II methods gave results for peak dome and vessel pressure well under l

required limits.

These are acceptable methodologies and results.

LOCA analyses, using approved methodologies and parameters (Reference l

4), were performed to provide MAPLHGR values for the new reload fuel l

assemblies (BP80RB283).

These analyses and results are acceptable.

A cycle specific rod drop accident analysis has been performed for Cycle 11 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.

4 2.7 Technical Specifications The changes to be made to the technical specifications are as follows:

A.

Page 210 (Limiting Condition for Operation - Average Planar Linear Heat Generation Rate) - The specification would be revised to include BP8x8R fuel among the types of fuel for which the APLHGR reduction factor applies.

B.

Page 212f - Figure 3.11-2e - Minimum Critical Power Ratio (MCPR) vs Tau (based on tested measured scram time as defined in Reference 8) for P8x8R and BP8x8R Fuel (BOC to E0C - 1000 mwd /ST).

This figure is the result of new analyses.

C.

Page 214b - Reference No. 9 (Reference 8) to be added to the bases of Section 3.11 which was inadvertently deleted by Amendment No.

94.

D.

Page 217 - (Major Design Features - Reactor) - The description would be revised to indicate that BP8x8R fuel is used.

Each of these is discussed below.

2.7.1 APHLGR Reduction Factor During single-loop operation, an APLHGR reduction factor is applied for each fuel design installed in accordance with Amendment 94.

At the time Amendment 94 was issued BP8x8R fuel was not installed and the reduction factor was thus not included.

The appropriate factor for BP8x8R fuel is 0.77, the same as for 8x8R and P8x8R fuel and would be added by this amendment. This is acceptable.

2.7.2 MCPR Specification Format - Figure 3.11-2e P8x8R and BP8x8R Fuel BOC to E0C - 1000 mwd /ST vs I:

The cycle Minimum Critical Power Ratio (MCPR) as a function of the parameter t is presented in curve form.

This curve is changed to reflect the new transient analyses as previously discussed.

The change is acceptable.

2.7.3 Reference No. 9 That Defines T (Tau) Used in Figure 3.11-2 This reference was previously in technical specifications but was inadvertently deleted by Amendment No. 94.

Therefore, the addition to page 214b is acceptable.

2.7.4 Major Design Features - Reactor Section 5.2.A of the Technical Specifications specifies the number and type of fuel assemblies in the core and would be revised to indicate that barrier fuel BP8x8R is included.

This is acceptable.

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3.0 ENVIRONMENTAL CONSIDERATION

S This 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.

Tne 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 the amendment involves no significant hazards consideration and there has been no public comment on such finding.

Accordingly, the 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 the amendment.

4.0 CONCLUSION

As a result of our review, which is described above we conclude that the proposed reload and Technical Specification changes are acceptable.

This conclusion is based on the following:

1.

Previously approve'd analysis methods and techniques are employed.

2.

The results of the transients and accidents which are affected by the reload are in conformance with the applicable regulation in our standard review criteria and therefore, are acceptable for Cycle 11.

3.

The revisions to the Technical Specifications have been found to be acceptable.

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 Com-mission'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 Contributor:

T. Haung Dated: December 9, 1986 i

6 REFERENCES 1.

Letter, L. C. Kuncl, Nebraska Public Power District, to D. R. Muller (NRC) dated September 17, 1986.

2.

" Supplemental Reload Licensing Submittal for Cooper Nuclear Power Station Unit 1, Reload 10," General Electric Company Report 23A4781 Class 1, May 1986.

3.

Letter, C. O. Thomas (NRC) to J. S. Charnley (GE) dated April 13, 1983.

4.

GESTAR II,' General Electric Standard Application for Reactor Fuel, NEDE24011f-A-7, dated August 1985.

5.

Letter, C. O. Thomas (NRC) to H. C. Pfefferlen (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment 8 Thermal Hydraulic Stability Amendment to GESTAR II," dated April 24, 1985.

6.

" Loss-of-Coolant Accident Analysis Report for Cooper Nuclear Power Station," NEDO-24095, August 1977 (as amended).

7.

" Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," GE Report NEDE-24154-P, October 1978.

8.

Letter (with attachment), R. H. 8uckholz (GE) to P. S. Check (NRC),

" Response to NRC Request for Information on ODYN Computer Model,"

September 5, 1980.

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