ML20112D153

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Proposed Tech Specs,Revising Minimum Critical Power Ratio Safety Limit & Associated Basis
ML20112D153
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 05/30/1996
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20112D141 List:
References
NUDOCS 9606040106
Download: ML20112D153 (15)
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l Att:chment I to JPN-96-025 I REVISED TECHNICAL SPECIFICATION PAGES PROPOSED TECHNICAL SPECIFICATION CHANGES  :

REGARDING MINIMUM CRITICAL POWER RATIO SAFETY LIMIT l l

J New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT.

Docket No. 50-333 DPR-59 1

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1.1 FUEL CLADDING INTEGRITY '2.1 FUEL CLADDING INTEGRITY _

Acolicability: Aoolicability:

The Safety Umits established to preserve the fuel cladding integrity The Umiting Safety System Settings apply to trip settings of the '

apply to those variables which monitor the fuel thermal behavior. instruments and devices which are provided to prevent the fuel t

cladding integrity Safety Umits from being exceeded.

Otnective:  ;

Obiective:

The objective of the Safety Umits is to establish limits below which l

the integrity of the fuel cladding is preserved. The objective of the Umiting Safety System Settings is to define the level of the process variables at which automatic protective action is i initiated to prevent the fuel cladding integrity Safety Umits from '

. being exceeded. i t

Soecifications: Soecifications: .

A. Reactor Pressure >785 osig and Core Flow >10% of Rated A. ' Trio Settinas - ,

I The existence of a minimum critical power ratio (MCPR) less The limiting safety system trip settings shall be as specified i

.[ than 1.09 shall constitute violation of the fuel cladding below-  !

integrity safety limit, hereafter called the Safety Umit. An ,

.l MCPR Safety Umit of 1.10 shall apply during single-loop '

1. Neutron Flux Trio Settinas oparation. ,
a. IRM - The IRM flux scram setting shall be set at s120/125 of full scale. l E

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Amendment No. 14,21,00,40,^0,117,157,  !

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1.1 BASES . _.

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- A. Reactor Pressure >785 osia and Core Flow >10% of Rated 1.1 FUEL CLADDING INTEGRITY Onset of transition boiling results in a decrease in heat transfer The fuel cladding integrity 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 transient. 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.

Umit minimum critical power ratio (MCPR). This Safety Umit boiling transition is calculated from plant operating parameters i 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 t is one of the physical barriers which separate racioactive characterized by the critical power ratio (CPR) which is the ratio  ;

materials from the environs. The integrity of this cladding of the bundle power which would produce onset of transition 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 .:

may occur during the life of the cladding, fission product power ratio (MCPR). It is assumed that the plant operation is _ t 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 operating domain. The current- i 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  ;

protection system safety settings. While fission product assure that in the event of an abnormal operational transient 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  !

perforations signal a threshold, beyond which still greater expected to avoid boiling transition. The MCPR fuel cladding thermal stresses may cause gross rather than incremental safety limit is increased by 0.01 for single-loop operation as . I cladding deterioration. Therefore, the fuel cladding Safety discussed in Reference 2. The margin between MCPR of _1.0 i

Umit is defined with margin to the conditions which would (onset of transition boiling) and the Safety Umit is derived from produce onset of transition boiling, (MCPR of 1.0). These a detailed statistical analysis considering all of the uncertainties conditions represent a significant departure from the condition in monitoring the core operating state including the uncertainty ,

intended by design for planned operation. in the boiling transition correlation. The method of determining the Safety Umit is described in Reference 1. The boiling '

transition correlation and the uncertainties employed in deriving the Safety Umit are

, a Amendment No.

  • i, 18,21,30,13,72,9S,7,157,1S2, t

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1.1 (cont'd) provided in Reference 3. Because the boiling transition At 100% power, this limit is reached with a maximum l

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 j '

l the Safety Umit would not produce boiling transition. . Thus, ' than the fraction of rated power (FRP), the APRM scram i although it is not required to establish the safety limit, and rod block settings shall be adjusted as specified in I additional margin exists between the Safety Limit and the Tables 3.1-1 and 3.2-3 respectively. ,

actual occurrence of loss of cladding integrity. .{

B. . Core Thermal Power Limit (Reactor Pressure <785 osia) i However, if boiling transition were to occur, clad perforation l would not be expected. Cladding temperatures would At pressures below 785 psig the core elevation pressure increase to approximately 1100'F which is below the drop is greater than 4.56 psi for no boiling in the bypass

perforation temperature of the cladding material. This has region. At low powers and flows, this pressure drop is been verified by tests in the General Electric Test Reactor due to the elevation 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 f above the critical heat flux for a significant period of time (30 range of 1-5 MWt, the channel flow will never go below  !

minutes) without clad perforation. 28 x 108 lb/hr. This flow results from the pressure i differential between the bypass region and the fuel j If reactor pressure should ever exceed 1400 psia during 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 i I 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 boiling transition limit (Safety Limit), 785 psig indicate that the fuel assembly critical power '  ;

operation is constrained by the maximum LHGR identified in at 28 x 108 lb/hr is approximately 3.35 MWt. With the i the Core Operating Limits Report. design peaking factors, this corresponds to a core thermal power of more than 50%.- Thus, a core thermal i power limit of 25% for reactor pressures below 785 - I psig is conservative. I i

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Amendment No. li, 21,30,13,Si,74,100,7,157,1S2,- [

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1.1 BASES (Cont'd) ,

E. References C. Power Transient .. _

1. General Electric Standard Application for Reae. tor Fuel, Plant safety analyses have shown that the scrams NEDE-24011-P, latest approved revision and '

caused by exceeding any safety system setting will amendments. i assure that the Safety Umit of 1.1.A or 1.1.B will not be exceeded. Scram times are checked periodically to 2. . FitzPatrick Nuclear Power Plant Sin 0 le-Loop assure the insertion times are adequate. The thermal Operation, NEDO 24281, August 1980.

power transient resulting when a scram is accomplished other than by the expected scram signal 3. GE12 Compliance with Amendment 22 of ,

(e.g., scram from neutron flux following closure of the NEDE-24011-P-A (GESTAR 11), NEDE-32417P,  ;

main turbine stop valves) does not necessarily cause December 1994. j fuel damage. However, for this specification a Safety  :

Umit violation will be assumed when a' scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a  ;

Safety Umit provided scram signals are operable is "t supported by the extensive plant safety analysis.  !

D. Reactor Water Level (Hot or Cold Shutdown Condition) j During periods when the reactor is shut down, ['

consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core ,

is reduced. This reduction in core cooling capability j could lead to elevated cladding temperatures and clad  ;

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 Umit at 18 in. above the top of the fuel i provides adequate margin. This level will be  ;

continuously monitored whenever the recirculation '

pumps are not operating.

Amendment No. 'i, 98,162, 14

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3 Attachment 11 to JPN-96-025 SAFETY EVALUATIL3 FOR PROPOSED TECHNICAL SPECIFICATION CHANGES 3 REGARDING MINIMUM CRITICAL POWER RATIO SAFETY LIMIT l

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333

'. Attachment 11 to JPN-96-025 SAFETY EVALUATION PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING MINIMUM CRITICAL POWER RATIO SAFETY LIMIT Page 1 of 4

1. DESCRIPTION OF THE PROPOSED CHANGES The following proposed changes to the James A. FitzPatrick Technical Specifications establish a revised Minimum Critical Power Ratio (MCPR) safety limit and associated basis. The changes are required to support introduction of GE12, 10x10 fuel into the Cycle 13 core.

Paae 7 Change "1.07" in specification 1,1.A to "1.09." Change "1.08" in specification 1.1. A to " 1.10."

Paoes 12 and 13 Change "The margin between MCPR of 1.0 (onset of transition boiling) and the Safety Umit is derived from a detailed statistical analysis considering all of the uncertainties in monitoring the core operating state including the uncertainty in the boiling transition correlation as described in Reference 1. The uncertainties employed in deriving the Safety Limit are provided in Reference 1." to "The margin ,

between MCPR of 1.0 (onset of transition boiling) and the Safety Limit is derived i from a detailed statistical analysis considering all of the uncertainties in monitoring the core operating state including the unceitainty in the boiling transition correlation. The method of determining the Safety Limit is described in Reference

1. The boiling transition correlation and the uncertainties employed in deriving the Safety Limit are provided in Reference 3."

Eaae 14 Add new reference,1.1. Bases.E.3:

"3. GE12 Compliance with Amendment 22 of NEDE-24011-P-A (GESTAR ll),

NEDE-32417P, December 1994."

11. PURPOSE OF THE PROPOSED CHANGES The purpose of the proposed changes is to provide the appropriate MCPR safety limit for the Reload 12 / Cycle 13 core. Reload 12 will consist of GE12 fuel, which utilizes a 10x10 mechanical design.

Ill. SAFETY IMPLICATIONS OF THE PROPOSED CHANQE_E lhe proposed changes revise the Safety Limit Minimum Critical Power Ratio (SLMCPR) to be 1.09 for two-loop operation and 1.10 for single-loop operation, and make associated changes to the bases. These :N7ges are required to support loading GE12 fuel bundles in the Cycle 13 cois,.

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. Attachment 11 to JPN-96-025 SAFETY EVALUATION PROPOSED TECHNICAL SPECIFICATION CHANGES  ;

REGARDING MINIMUM CRITICAL POWER RATIO SAFETY LIMIT  :

Page 2 of 4 GE12 fuel was demonstrated to meet the fuel licensing acceptance criteria of amendment 22 of NEDE-24011-P-A (GESTAR ll), General Electric Standard ,

Application for Reactor Fuel as described in the reference 1 report. Changes to i GESTAR ll to address Gell, GE12 and GE13 fuel designs have been incorporated in NEDE-24011-P-A-11 (reference 2). The fuel licen ing acceptance criteria from reference 2 discussed below are the same as those oescribed in reference 1.  ;

i Subsection 1.1.5.A of reference 2 requires " Safety Limit MCPR shall be recalculated l following steps in 1.1.5.B or reconfirmed when a new fuel design or new critical l power correlation is introduced." Subsection 1.1.7.A of reference 2 requires "The I currently approved critical power correlation will be confirmed or a new correlation I will be established when there is a change in wetted parameters of the flow l geometry; this specifically includes fuel and water rod diameter, channel sizing and (

spacer design." Subsection 1.1.7B of reference 2 allows "A new correlation may  !

be established if significant new data exists for a fuel design (s)."

Reference 1 discusses the derivation, applicability and uncertainty of the GEXL10 critical power correlation as applied to GE12 fuel. This correlation was used as a basis for determination of the SLMCPR for GE12 fuel. The SLMCPR is also influenced by bundle design parameters which affect the bundle R-Factor distribution and core radial power distribution. These parameter.s include spacer i design, assembly dimensional geometry, bundle radial power distribution, and fuel ]

discharge exposure. The SLMCPR calculated for GE12 fuelis 1.09, which is being i conservatively applied to the entire core.

Reference 3 describes operation of the FitzPatrick Plant with a single Reactor Water  !

Recirculation loop in service (single-loop operation, SLO). This mode of operation l requires raising the SLMCPR by 0.01 to account for changes in core flow and Traversing incore Probe (TIP) uncertainties. Therefore, the SLMCPR for SLO j applicable to GE12 fuel is 1.10.

The changes to the bases are administrative in nature. They reflect relocation of details of the c itice! power correlation from the General Electric Standard )

Application for Reactor Fuel, NEDE-24011-P-A (GESTAR ll) to documents which report fuel bundle design compliance with the fuel licensing acceptance criteria of GESTAR I! (reference 2). Since this change does not affect the criteria to be satisfied by a ruw fuel design, there is no effect on nuclear safety.

I l Attachment il t2 JPN-96-025 '

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SAFETY EVALUATION I PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING MINIMUM CRITICAL POWER RATIO SAFETY LIMIT i

Page 3 of 4 IV. EVAL.UATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the proposed Amendment I would not involve a significant hazards consideration as defined in 10 CFR 50.92, l since it would not:

1. involve a significant increase in the probability or consequences of an accident previously evaluated because:  !

A change in the SLMCPR does not affect initiation of any accident.

Operation in accordance with the revised SLMCPR ensures the consequences i of previously analyzed accidents are not changed.

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2. create the possibility of a new or different kind of accident from nny accident previously evaluated because:

1 The SLMCPR establishes a performance limit for the fuel. Therefore l changing the limit will not initiate any accident. l l

3. involve a significant margin of safety because: l l

The analyses performed to determine the revised SLMCPR assure I maintenance of the same margin of safety as presently exists for the i prevention of onset of transition boiling.

V. IMPLEMENTATION OF THE PROPOSED CHANGES l 1

Implementation of the proposed changes will not adversely affect the ALARA or Fire Protection Program at the FitzPatrick plant, nor will the changes impact the environment.

VI. CONCLUSION Based on the discussions above,implementatiori of a SLMCPR of 1.09 (1.10 for SLO) does not involve a significant hazards consideration, or an unreviewed safety question, and will not endanger the health and safety of the public. The Plant Operating Review Committee and Safety Review Committee have reviewed this proposed Technical Specification change and agree with this conclusion.

, Attrchment il to JPN-96-025 SAFETY EVALUATION PROPOSED TECHNICAL SPECIFICATION CHANGES 1

REGARDING MINIMUM CRITICAL POWER RATIO SAFETY LIMIT l Page 4 of 4

] Vll. REFERENCES (1) GE12 Compliance With Amendment 22 of NEDE-24011-P-A (GESTAR 11),

NEDE-32417P, December 1994. i

4 (2) General Electric Standard Application for Reactor Fuel, (GESTAR ll), NEDE-1 24011-P-A, November 1995.

, (3) FitzPatrick Nuclear Power Plant Single-Loop Operation, NEDO-24281, August 1980. i 4

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Att chm:nt ll1 to JPN-96-025 MARKUP OF TECHNICAL SPECIFICATION PAGE CHANGES l

PROPOSED TECHNICAL SPECIFICATION CHANGES  ;

REGARDING MINIMUM CRITICAL POWER RATIO SAFETY LIMIT j l

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