ML20149L851
| ML20149L851 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/14/1996 |
| From: | Bajwa S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20149L854 | List: |
| References | |
| NUDOCS 9611190180 | |
| Download: ML20149L851 (7) | |
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UNITED STATES p
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NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 2006H001 0*****l POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.238 License No. DPR-59 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Power Authority of the State of New York (the licensee) dated May 30, 1996, as supplemented October 17, 1996, and November 8, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the 1
Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; i
B.
The facility will operate in conformity 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 Comission's regulations; D.
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 Comission's regulations and all applicable requirements have been satisfied.
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. DPR-59 is hereby amended to read as follows:
9611190180 961114 PDR ADOCK 05000333 P
i (2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 238, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION e $macAf S.
ingh Bajwa, Acting Director Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical
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Specifications Date of Issuance: November 14, 1996
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ATTACHMENT TO LICENSE AMENDMENT NO. 238 FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO.' 50-333 Revise Appendix A as follows:
Remove Paaes Insert Paaes 7
7 12 12 13 13 14 14 l
JAFNPP 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Anobcatulttv:
Anoticabihty:
The Safety Lierwts estatsehed to preserve the fuel cloddag The Limiting Safety System Settings apply to trip settings of the integrity apply to those variables whch monitor the fuel thermal instruments and devices which are provided to prevent the fuel behavior.
cladding integrity Safety Limits from being exceeded.
Objective:
Objective:
The objective of the Safety Linuts is to establish limits below The objective of the Limiting Safety System Settmas is to define which the integnty of the fuel cledding is preserved.
the level of the process variables at which automate protectwo action is initiated to prevent the fuel cladding integrity Safety
,1 Limits from being exceedad.
i Snecifications:
Sneciftetsans:
A.
Trio Settmas A.
Reactor Pressure >785 one and Core Flow > 10% of Rated The limiting safety system trip settings sheE be es specified below:
The existence of a minunum cntical power ratio (MCPR) less then 1.09 shen constitute violation of the fuel 1.
Neutron Flux Trio Settmas cladding integrity safety limit, hereafter called the Safety Lwmt. An MCPR Safety Linut of 1.10 shall apply during e.
IRM - The IRM flux scram setting shall be set single-loop operation.
et s120/125 of fun scale.
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I I Note: TS 1.1.A lo app 5 cable for Cycle 13 only.
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Amendment No. ? ', ? ?, ^, '?, **, * ? ?, 157, 238 7
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I 1.1 BASES A. P-tor Pra*=_-e >785 a=ia and Core Flow > 10% of Rated 1.1 FUEL CLADDING INTEGRITY Onset of transition boiling results in a decrease in heat transfer The fuel claddmg integnty limit is set such that no calculated from the clad and, therefore, elevated clad temperature and the fuel damage would occur as a result of an abnormal possibility of clad failure. However, the existence of critical operational transsent. Because fuel damage is not directly power, or boiling transition, is not a directly observable observable, a step-back approach is used to establish a Safety parameter in an operating reactor. Therefore, the margin to bmit mmemum critical power ratio (MCPR). This Safety Umit boiling transition is calculated from plant operating parameters represents a conservative margin relative to the conditions such as core power, core flow, feedwater temperature, and core required to maintain fuel cladding integrity. The fuel cladding power distribution. The margin for each fuel assembly is
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is one of the physical barr:ers which separate radioactive characterized by the critical power ratio (CPR) which is the ratio i
materials from the environs. The integrity of this cladding of the bundle power which would produce onset of transition
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barrier is related to its relative freedom from perforations or boiling divided by the actual bundle power. The minimum value cracking. Although some corrosion or use related cracking of this ratio for any bundle in the core is the minimum critical l
may occur durmg the life of the cladding, fission product power ratio (MCPR). It is assumed that the plant operation is migration from this source is incrementally cumulative and controlled to the nominal protective setpoints via the continuously measurable. Fuel cladding perforations, however, instrumented variable, i.e., the operstmg domain. The current l
can result from thermal stresses which occur from reactor load line limit analysis contains the current operating domain operation significantly above design conditions and the map. The Safety Umit MCPR has sufficient conservatism to i
protection system safety settings. While fission product assure that in the event of an abnormal operational transient i
migration from cladding perforation is just as measurable as initiated from the MCPR operating limit in the Core Operating that from use related cracking, the thermally caused cladding Umits Report, more than 99.9% of the fuel rods in the core are
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perforations signal a threshold, beyond which still greater expected to avoid boiling transition. The MCPR fuel cladding I
thermal stresses may cause gross rather than incremental safety limit is increased by 0.01.cr single-loop operation as
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cladding deterioration. Therefore, the fuel cladding Safety discussed in Reference 2. The margin between MCPR of 1.0
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Umit is defined with margin to the conditions which would (onset of transition boiling) and the Safety Umit is derived from f
produce onset of transition bodmg, IMCPR of 1.0). These a detailed statistical analysis considering all of the uncertainties
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conditions represent a significant departure from the condition in monitoring the core operating state including the uncertainty l
intended by desgn for planned operation.
in the boiling transition correlation. The method of determining l
the Safety Umit is described in Reference 1. The boiling transition correlation and the uncertainties employed in deriving j
the Safety Umit are i
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Amendment No. 11,18, 21, 30,13, 72, 96,*'?.157,162, 238 f
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.1 1.1 (cont'd) t l
provided in Reference 3. Because the boiling transition At 100% power, this limit is reached with a maxwnum correlation is based on a large quantity of full scale data there fraction of limiting power density (MFLPD) equal to is a very high confidence that operation of fuel assembly at 1.00. In the event of operation with MFLPD greater the Safety Limit would not produce boiling transition. Thus, than the fraction of rated power IFRP), the APRM scram j
although it is not required to establish the safety limit, and rod block settings shall be adjusted as specified in additional margin exists between the Safety Limit and the Tables 3.1-1 and 3.2-3 respectively.
i actual occurrence of loss of cladding integrity.
l B.
Core Thermal Power Limit (Reactor Pressure <785 nsin)
However, if boilmg transition were to occur, clad perforation l
would not be expected. Cladding temperatures would At pressures below 785 psig the core elevation pressure l
increase to approximately 1100*F which is below the drop is greater than 4.56 psi for no boiling in the bypass l
perforation temperature of the cladding material. This has region. At low powers and flows, this pressure drop is l
been verified by tests in the General Electric Test Reactor due to the elovation pressure of the bypass region of the (GETR) where fuel similar in design to FitzPatrick operated core. Analysis shows that for bundle power in the l
above the critical heat flux for a significant period of time (30 range of 1-5 MWt, the channel flow will never go below l
minutes) without clad perforation.
28 x 108 lb/hr. This flow results from the pressure 4
differential between the bypass region and the fuel l
if reactor pressure should ever exceed 1400 psia dunng channel. The pressure differential is primarily a result of normal power operation (the limit of applicability of the boiling changes in the elevation pressure drop due to the i
transition correlation) it would be assumed that the fuel density difference between the boiling water in the fuel l
cladding integrity Safety Limit has been violated.
channel and the non-boiling water in the bypass region.
Full scale ATLAS test data taken at pressures from 0 to In addition to the bodmg transition limit (Safety Limit),
785 psig indicate that the fuel assembly critical power i
operation is constrained by the maximum LHGR identified in at 28 x 108 lb/hr is approximately 3.35 MWt. With the the Core Operating Limits Report.
design peaking factors, this corresponds to a core thermal power of more than 50%. Thus, a core thermal power limit of 25% for reactor pressures below 785
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psig is conservative.
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Amendment No. 'i, 21, 20,13, Si, 71, 100,l'7,157,162, 238
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t JAFNPP 1.1 BASES (Cent'd)
E.
References C.
Power Transient 1.
General Electric Standard Application for Reactor Fuel, Plant safety analyses have shown that the scrams NEDE-24011-P, latest approved revision and caused by exceedmg any safety system setting will amendments.
assure that the Safety Limit of 1.1.A or 1.1.B will not i
be exceeded. Scram times are checked periodically to 2.
FitzPatrick Nuclear Power Plant Single-Loop assure the insertion times are adequate. The thermal Operation, NEDO 24281, August 1980.
l power transeent resulting when a scram is accomplished other than by the expected scram segnal 3.
GE12 Compliance with Amendment 22 of (e.g., scram from neutron flux following closure of the NEDE-24011-P-A (GESTAR 111, NEDE-32417P, main turbme stop valves) does not necessarily cause December 1994.
l fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approachmg a Safety Limit provided scram segnals are operable is l
supported by the extensive plant safety analysis.
l D.
Reactor Water Level (Hot or Cold Shutdown Condition)
Dunng periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If i
reactor water level should drop below the top of the active fuel dunng this time, the ability to cool the core is reduced. This reduction in core cooling capability
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could lead to elevated cladding temperatures and clad i
perforation. The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds the core height. Establishment of the Safety Limit at 18 in above the top of the fuel L
provides adequate margin. This level will be i
continuously monitored whenever the recirculation j
pumps are not operating.
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Amendment No.
- i, SS, * *2, 238 14 I