ML20207P901

Forwards Addl Info Re 861117 Application for Amend to License NPF-39,changing Tech Specs for Operation W/Partial Feedwater Heating & Increased Flow

ML20207P901
Person / Time
Site:	Limerick
Issue date:	12/22/1986
From:	Gallagher J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Butler W Office of Nuclear Reactor Regulation
References
NUDOCS 8701200387
Download: ML20207P901 (6)

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^/AA PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P.O. BOX 8699 PHILADELPHIA. PA.19101 12153041-4000 December 22, 1986 Docket No. 50-352 Mr. W. R. Butler, Director Project Directorate No. 4 Division of Boiling Water Reactor Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Butler:

The purpose of this letter is to provide additional information for Philadelphia Electric Company's November 17, 1986 Application for Amendment of Facility Operating License NPF-39 which requests changes to the Limerick Operating License and Technical Specifications relating to operation with partial feedwater heating and increased core flow.

The additional information, pertaining to items I and III of the Significant Hazards Consideration, was requested by Mr. Robert Martin of the NRC staff during a telephone conversation on December 12, 1986. Attachment A provides this additional information. Also included is Technical Specification page 3/4 2-9 which is provided to correct typographical errors in the original submittal. These errors were also the subject of a discussion with Mr. Martin.

If you have any questions or require additional information, please do not hesitate to contact us.

Very truly yours,  ;

J. W. Gallagher l Vice President 1 Nuclear Operacions Attachments 00{

cc: E. M. Kelly, Senior Resident Site Inspector 'g See Attached Service List f 8701200387 861222 PDR ADOCK 05000352 PDR g- _ _

r ATTACHMENT'A I. The following. changes "do not involve a significant increase in the probability or consequences of an-accident previously evaluated".

A.- The control rod block instrument setpoints for the reactor coolant system recirculation flow upscale condition are changed from 108:to 111% of-rated flow and 111 to 114% of rated flow for the trip setpoint and allowable value, respectively. The operational design basis for this trip is- -

4 to provide an operator warning signal in response to unexpected high core flow operation. In addition, another consideration is the alarm function provided to the. operator ,

on a hypothetical malfunction of the drive flow signal f~ conditioning equipment that would result in a false high '

signal. The Flow Unit upscale trip performs no. safety function and nas not been taken credit for by any LGS design basis transient or accident analysis. Therefore, the changes do not involve a significant increase in the probability or-consequences of an accident previously ,

evaluated.

B. The MG set scoop tube mechanical and electrical stop overspeed setpoints are changed from-105 and 102.5% to 109 and 107% of rated core flow, respectively. For operation without ICF, the limiters were set to 102.5 and 105% of rated core flow based on system analysis that was performed at 100% rated flow, including the effect of plant design transients with mrrimum core flow runout to 102.5% rated flow. For increased core flow justification the system was l re-evaluated at 105% rated flow,' including the effect of

! plant design transients with maximum core flow runout to 107% rated flow. The transients where recirculation flow increases from 105% to the limiter setting of 109%

(mechanical stop) have been reviewed and are covered by the application of the Kf curve (Technical Specification Figure 3.2.3-2) for 109%. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

C. A high flow clamping feature is added to the Rod Block Monitor (RBM) that ensures the 106 and 109% (trip setpoint and allowable value, respectively) rod blocks currently included in the Technical Specifications cannot be exceeded.

The rod withdrawal error transient was evaluated under ICF and/or PFH conditions. When TCP is employed, the RBM setpoint (which is flow biased) increases, giving higher l MCPR limit. Thus, the RBM should be clipped at flows ,

greater than 100% of rated so that the CPR values determined I without ICF apply. Since these changes only ensure the current RBM setpoints cannot be exceeded, it follows that the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

II. The following changes "do_not involve a significant reduction in a margin of safety."

A. The control rod block instrument setpoints for the reactor coolant system recirculation flow upscale condition are changed from 108 to 111% of rated flow and 111 to 114% of rated flow for the trip setpoint and allowable value, respectively. These setpoints are increased to maintain an adequate trip setpoint margin allowing for hardware uncertainties and signal noise, so as not to impact plant high flow operations. Whereas the licensed core flow range is being increased by 5% (from 100% to 105%) the high flow trip setpoint was increased only 3% (from 108% to 111%);

therefore, the function of the trip level as an indication /

alarm of unintended high flow operation is actually enhanced. Therefore, these changes do not involve a significant. reduction in a margin of safety.

B. The MG set scoop tube mechanical and electrical stop overspeed setpoints are changed from 105 and 102.5% to 109 and 107% of rated core flow, respectively. These setpoints are increased to allow operation at 105% of rated core flow.

Whereas the licensed core flow range is being increased by 5% (from 100% to 105%) the scoop tube mechanical and electrical stops are increased by 4% (from 105% to 109%) and 4.5% (from 102.5% to 107%), respectively. Therefore, the function of the stops to prevent unintended high flow operation is actually enhanced. Application of the Kf curve (Technical Specification Figure 3.2.3-2) for 109% assures the operating margin to the safety limic MCPR currently in the Technical Specifications is maintained. Therefore, the changes do not involve a significant reduction in a margin of safety.

C. A high flow clamping feature is added to the Rod Block Monitor (RBM) that ensures the 106 and 109% (trip setpoint and allowable value, respectively) rod blocks currently l included in the Technical Specifications cannot be exceeded, thus maintaining the margin of safety currently embodied in the Rod Withdrawal Error analysis. Therefore, these changes r do not involve a significant reduction in a margin of l safety.

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III. Operation at ICP and/or PFH with the requested MCPR changes "does not involve a significant reduction.in a margin of safety".

A. MCPR - Results of safety evaluation show that the current Technical Specifications, with the exception of the MCPR operating limits, are adequate and preclude the violation of any safety. limits for ICF and/or PFH operation. .

Accordingly, the MCPR limits are revised to assure.that the margin of safety is' maintained as demonstrated in the analysis.provided by General Electric Report NEDC-31323,

" Increased. Core Flow and Partial Feedwater Heating Analyses

'for Limerick Generating' Station, Unit 1,' Cycle 1".

B. Overpressurization analysis was re-performed and found to be minimally affected by ICF operation. 'The maximum steam line pressure of 1235 psig compares to the FSAR Value of 1227, thus 140 psig margin exists to the ASME code upset limit of 1375 psig. ,

C. Rod Withdrawal Error Event requires that'the RBM be clipped:

at flows greater than 100% of rated, thus there is no change in margin for this event relative to the FSAR values.

D. Fuel Loading Error re-evaluated for a worst case scenario results in a maximum increase of 0.04 from the 0.11 value reported in the FSAR. Even with a delta-CPR of 0.15, this event does not become limiting or reduce overall MCPR margin. The MAPLHGR limits at rated conditions reported in

. the FSAR are also bounding for the Fuel Loading Error.

E. Since Limerick l'uses the banked position withdrawal.

sequence, RDA analysis need not be performed as previously approved by NRC in Amendment No. 9 to NEDE-240ll, Revision 6.

F. For the requested ICP and PFH operation, LOCA analyses result in less than 10 degrees Fahrenheit increase in PCT compared to rated conditions. Since the peak value at rated conditions reported in the FSAR is 2090 degrees Fahrenheit, the margin to the limit is minimally impacted and

-conformance to the required PCT limit.is maintained. Thus the MAPLHGR limits at rated conditions reported in the FSAR are applicable to operation with ICF and PFH.

i G. Specific thermal-hydraulic stability analyses are not l required as stated in NEDE-240ll-P-A-8-US, as amended.

H. Analyses performed to determine the impact on Anticipated Transients Without Scram (ATWS) show that ATWS performance with the most limiting condition of 100% power /87% flow with PFH is bounded by the results reported in the FSAR.

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I. Mechanical evaluation of reactor internals and fuel assemblies demonstrates that all variables are within allowable design values. The fuel bundle lift is very small (about 0.01 in.) for both ICF and rated flow conditions, compared to the design criterion of 0.52 in.

J. The vibration measurement program for Browns Ferry 1 includes data up to 113% core flow (" Assessment of Reactor Internals Vibration in BWR/4 and BWR/S Plants", NEDE-24057-P-A). These data show that tne reactor internals response to flow induced vibration is within acceptable limits up to 113% core flow.

K. The implementation of FFWTR and FWHOS will have a minor impact on the feedwater nozzle. The incremental fatigue usage due to FFWTR is 0.027 per cycle and incremental usage due to FWHOS for 12% per year is 0.013 per cycle. The analyses also show the feedwater sparger fatigue usage factor will not exceed 1.0 for the 40 year life of the plant.

L. The results of containment analysis show that for ICF and/or PFH, all containment paramet2rs are bounded by the results reported in the FSAR with the following exceptions:

1. The peak value for drywell deck downward differential pressure is 28.6 psid as compared to the FSAR value of 26.0 psid. It is, however, bounded by the appropriate design limit of 30 psid.

2. The pool swell loads result from coupled effects of initial drywell pressurization, vent clearing, and suppression chamber compression phenomena. The loadings in the first few seconds following the hypothetical pipe rupture include vent clearing, bubble expansion, pool swell and fallback pressure loads. All these loads are primarily dependent upon the drywell pressurization rate, and, to a lesser extent, the vent energy flow during the initial blowdown transient. The predicted drywell pressurization rate up to vent clearing time for ICF/PFH is bounded by that of the PSAR-reported value. The calculated LOCA bubble expansion rates, which are functions of tha vent energy flow, are also bounded by the corresponding FSAR values. Therefore, the pool swell loads under the limiting ICF/PFH conditions are bounded by the design values.

The conditions under which condensation oscillation (CO) and chugging loads occur during ICP/PFH are found to be bounded by the test conditions of the 4TCO tests

("4T condensation Oscillation Test Program Final Test Report", NEDE-24811-P) which were used to formulate the

Limerick CO and chugging design load definitions.

l Therefore, the CO and chugging loads with ICF/PFH are bounded by the design values.

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C. -l

-.- POhfER DISTRIBUTION LIMITS

. LIMITING CONDITION FOR. OPERATION (Continued) l l

l ACTION:

.a. 'With the end-of-cycle recirculation pump trip system inoperable '

per Specification 3.3.4.'2, operation may continue and the provisions of Specification 3.0.4 are not applicable.provided that, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, MCPR is determined to.be greater than or.

equal to the MCPR limit as a function.of.the' average scram time shown in the appropriate figure taken'from Table 3.2.3-1 for EOC-RPT inoperable curve times the K g shown in Figure 3.2.3-2.-

b.

~

With MCPR.less than the applicable MCPR limit as identified in ACTION'a above initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER '

within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

a.  ?'= 1.0 prior to performance of the initial scram time

-measurements for the cycle in accordance with Specification 4.1.3.2, or

b. t as defined in Specification 3.2.3 used to determine the limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance. test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR.

-limit. determined from the appropriate figure taken from Table 3.2.3-1 times the K g ~ shown in Figure 3.2.3-2.

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase'of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.

d. The provisions of Specification 4.0.4 are not applicable.

LIMERICK - UNIT 1 3/4 2-9