ML20076J161
| ML20076J161 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 08/26/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20076J159 | List: |
| References | |
| NUDOCS 8309060456 | |
| Download: ML20076J161 (4) | |
Text
.
pco ous%[o UNITED STATES
!,, s. c *' g NUCLEAR REGULATORY COMMISSION g
E j WASHINGTON, D. C. 20553 k.~.v /
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 74 TO FACILITY OPERATING LICENSE NO. DPR-59 POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NU. CLEAR POWER PLANT DOCKET NO. 50-333 1.0 Introduction In its application for amendment dated May 25, 1983, the Power Authority of the State of New York (the licensee) proposed changes to the Technical Specifications for the James A. FitzPatrick Nuclear Power Plant (the facility) as a result of new fuel loaded during the current refueling outage (Reload 5) to support operation in Cycle 6.
The reload involves removing depleted fuel assemblies in about one-third of the nuclear reactor core and replacing them with new fuel at the same type previously loaded in the core.
The proposed changes pertain to revisions in the Minimum Critical Power Ratio (MCPR) Operating Limits to accommodate the new fuel; add additional Rod Block Monitor Trip Level Settings to facilitate control rod withdrawals; and, delete referenced to the fuel types and supporting analyses for the fuel removed from the core.
In support of the of the reload application, the licensee has also enclosed the GE BWR supplemental licensing submittal for the facility in Reference 1.
Our evaluation of the licensee's submittal and proposed Technical Specification change is provided below.
2.0 Evaluation 2.1 Fuel Mechanical Design The Cycle 6 core consists of 12 GE 8x8R bundles and 548 GE P8x8R bundles.
Two hundred of the GE P8x8R bundles contain fresh fuel.
The pbx 8R fuel is of the current GE standard design as described in Reference 3 which has been approved by the staff in Reference 5.
Both 8x8R and P8x8R fuel types have 62 fuel rods and two water rods.
2.2 MAPLHGR Limit The maximum average planar linear heat generation rate (MAPLHGR) for the fresh fuel labeled Type P8DRB299 is identical to the previously approved MAPLHGR limit for the existing Cycle 5 fuel of the same type.
We thus find the MAPLHGR limit acceptable for Cycle 6.
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2.3 Nuclear Design The nuclear design analysis was performed with the methods and procedures described in Reference 3 which has been approved by the staff in Refe ence 5 for reload applications.
The nuclear parameters for the reload core are within the range of those normally obtained and are acceptable.
2.4 Thermal-Hydraulic Design y
The objective of the review is to confirm the thermal-hydraulic iesign of j
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 operation and anticipated operational transients, and is not susceptible to thermal-hydraulic instability.
The review includes the following areas: (1) safety limit minimum critical power ratio (MCPR), (2) operating limit MCPR, and (3) thermal-hydraulic s tabil ity.
The licensee has submitted the analysis report for Cycle 6 operation at rated core flow conditions (Ref. 2).
Discussion of the review concerning the thermal-hydraulic design for Cycle 6 operation follows.
Safety Limit MCPR 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 3, the approved safety limit MCPR is 1.07.
The safety limit MCPR of 1.07 is used for FitzPatrick Cycle 6 operation.
Operating Limit MCPR The most limiting events have been analyzed by the licensee to determine which event could potentially induce the largest reduction in the initial critical power ratio (aCPR).
The ACPR values given in Section 9 of Reference 2 are plant-specific values calculated by using approved methods including ODYN methods.
The calculated ACPRs are adjusted to reflect either Option A or Option B ACPR by employing the conversion methods described in Reference 4, which was approved by the staff in Reference 6.
The MCPR values are determined by adding the adjusted aCPRs to the safety limit MCPR.
Section l
11 of Reference 2 presents both the cycle MCPR values of the pressurization and non-pressurization transients.
The maximum cycle MCPR values (0ptions A and B) in Section 11 are specified as the operating limit MCPRs and incorporated into the Technical Specifications.
We found that the approved method was used to determine the operating limit MCPRs to avoid violation of the safety limit MCPR in the event of any anticipated transients.
Therefore, we conclude that these limits are acceptable.
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Thermal-Hydraulic Stability b
The resul ts of thermal-hydraulic analyses (Ref. 2) show that the maximum k
core stability decay ratio is 0.93, as compared to 0.87 for Cycle 5 core, which has been previously approved.
Since the calculated maximum core stability decay ratio is less than some of the operating plants (for g
example, Peach Bottom Units 2 and 3 have decay ratio of 0.98) and since
$k operation in the natural circulation mode is prohibited by Technical Specification 3.5.J, there will be additional margin to the stability y
limit.
Therefore, we conclude that the thermal-hydraulic stability results are acceptable for Cycle 6 operation.
N t-2.5 T.9nsient and Accident Analyses h
A cycle specific analysis of the rod drop accident was performed for b
Cycle 6 because the accident reactivity shape function was not bounded h
by the generic curve for both cold and hot standby conditions.
In both n
cases the calculated peak fuel enthalpy was less than our acceptance F
criterion of 280. calories per gram.
We conclude that the analysis of the
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rod drop accident is acceptable.
D ine analysis of the rod withdrawal error at power was extended to include C
larger values of the rod block monitor trip setting with accompanying larger changes in the critical power ratio.
This permits greater freedom p
of rod motion at times when other events are limiting with respect to the operating value of MCPR.
This is an acceptable procedure.
5 2.6 Technical Soecifications
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mE With the exceptions noted below, the proposed Technical Specification E2 changes are related to the discharge of the last Sx8 fuel from the core or are typographical error corrections or clarifications.
The only fuel y
remaining in the core is of the 8x8R type.
These changes are acceptable.
g The two additional equations for the Rod Block Monitor setpoints are added in Specification 3.1.B and in the Notes for Table 3.2-3.
These are acceptable.
Figure 3.1-2 has been altered to reflect the MCPR operating limit as a function of T for the new cycle and MAPLHGR limits for the new fuel (Type P8DRB299) are added in Figure 3.5-10.
These changes are acceptable.
3.0 Concl usi on 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 this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
S. L. Wu S. Sun W. Brooks
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j Dated: August 26, 1983
4-4.0 References 1.
J. P. Bayne (PASNY) to D. Vassallo (NRC), May 25, 1983.
2.
" Supplemental Reload Licensing Submittal for James A. FitzPatrick Nuclear Power Plant, Reload 5," GE Report Y 1003J01 A56, March,1983.
3.
" General Electric Standard Application for Reactor Fuel," GE Report NEDE-240ll-P-A-4, January,1982.
4.
" Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," GE Report NEDE-24154-P, October,1978.
5.
D. Eisenhut (NRC) to R. Gridley (GE), May 12, 1978, Safety Evaluation for General Electric Standard Application for Reactor Fuel.
6.
R. Tedesco (NRC) to G. Sherwcod (GE)', February 14, 1981, Acceptance for Referencing GE Report NEDE-24154-P, in reload submittals.
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