ML20235C382

Proposed Tech Specs,Adding MAPLHGR Limits for Reload Fuel & Adjustment of Min Critical Power Ratio Limit to Reflect Results of Cycle 10 Analysis & Deleting Provisions for Single Loop Operation

ML20235C382
Person / Time
Site:	Quad Cities
Issue date:	09/18/1987
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML19304B493	List: ML20235C340 ML20235C382 ML20235C433 ML20235C487
References
NUDOCS 8709240434
Download: ML20235C382 (31)

Text

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DPR-29 D. Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time, any byproduct, source and special nuclear materials as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts required; Am. 38 2/03/77 E. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or l special nuclear materials without restriction to chemical or physical form, for sample analysis or instrument and equipment j calibration or associated with radioactive apparatus or components; Am. 43 1/30/78 F. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Quad Cities Nuclear Power Station, Unit nos. I and 2.

3. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. Maximum Power Level Commonwealth Edison is authorized to operated Quad Cites unit No.

I at power levels not in excess of 2511 megawatts (thermal).

Am. 101 B. Technical Specifications 2/20/87 The Technical Specifications contained in Appendices A and B as revised through Amendment No. are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

07838 8709240434 870918 PDR ADOCK 05000254 P PDR

SUMMARY

OF PROPOSED _TECHN! CAL' SPECIFICATION CHANGES QUAD _ CITIES UNIT 1 CYCLE 10 ,

License ~page 3:

Remove requirements.for coastdown and off-normal feedwater heating. This }

.is because Reference ll analyzes operation to 20 percent power and up to a-

.100*F reduction in.feedwater temperature, thereby bounding the license. .

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License pages 5 and 6:

Remove the Single Loop Operation Provisions from the license and place them in Technical Specifications.Section 3.6/4.6.H.

-1 Table of Contents-page 11:

Change 3.6/4.6.H from " Recirculation Pump Flow Mismatch" to " Recirculation Pump Flow limitations" to reflect the addition of single loop operation to this section.

Technical Specification page 1.0-5: l Add definitions for Dual Loop Operation and Single Loop Operation.

Technical Specification page 1.1/2.1-7:  ;

l Change analyzed conditions from up to 2511 MHth to be in compliance with j Regulatory Guide 1.49 which states that transients must be analyzed at 102% of rated core thermal power.

Technical Specification figure 2.1-1:

Add Single Loop SCRAM and Rod Block lines.

Technical' Specification figure 2.1-3:

Add operating region as defined by Increased Core Flow Analysis (Reference i l l) . Also, to make the Unit 1 Specifications match those implemented for Unit 2.

Technical Specification page 3.2/4.2-14:

Change Rod Block Monitor intercept from 42 to 43 to take advantage of the

.yete specific Analysis.

j Technical Specification page 3.3/4.3-5:

Change 3.3.C.5 to 0.71 seconds due to a change in the computer modeling of the MCPR transient.

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SUMMARY

OF PROPOSED TECHNICAL SPECIFICATION CHANGES

. QUAD CITIES UNIT 1--CYCLE 10 (Continued)

Technical Specification page 3.5/4.5-5:

All changes on page are.the result of the analysis for one relief. valve out of service.

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Technical Specification page 3.5/4.5-10:

Add maximum LHGR of 14.4 kw/ft for GE8X8E and GE8X8EB fuel this is due to the advanced design of the new GE8 fuel.

= Change Minimum Critical Power Ratio limits and associated 20% scram insertion times. 'This change is a result of a new computer model for the MCPR transient.

Technical Specification page 3.5/4.5-11:

Change referenced report. This is a result of the change to new computer models.

Change the pipe break sizes. This is also a result of the model changes.

Technical Specification page 3.5/4.5-12:

Change "or" to "and" for clarification because the Automatic Pressure Relief Valves enable both Core Spray and LPCI mode of RHR.

Add insert due to Relief Valve out of service analysis.

Technical Specification page 3. 5 / 4. 5-13 a:

Add due to new fuel design and the improvement in computer modeling of accidents.

Technical Specification page 3.5/4.5-14:

Add clarification of MCPR Operating Limit due to ability to operate with a j feedwater heater out of service and the new computer models.

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i Technical Specification page 3.5/4.5-14a:  !

Delete section on the statistical methods used in determining the required l i

mean 20% SCRAM insertion times for clarity. This information is contained {

in Reference 4 of Technical Specification 3.5/4.5-15. '

Technical Specification page 3.5/4.5-15:

Change Reference 1 to reflect new computer model. ,

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SUMMARY

OF PROPOSE 0 TECHN! CAL SPECIF! CATION CHANGES

_UAD Q CITIES UNIT'l CYCLE 10 (Continued)

.. Technical Specification figures 3.5-1:

Change to reflect fuel to be loaded for cycle 10'.

Technical. Specification page 3.6/4.6-5:

Change 4.6.G.I.b from power-flow relationship to core plate 4P-core flow relationship. This is because it has been shown that the core plate AP-flow relationship is more accurate than the power-core flow relationship. -Unit 2 has'already had this change approved.

-Change titles in section H to reflect movement of Single Loop Operation requirements from license to section H.

. Change 3.6.H.3 to reflect movement of Single Loop Operation requirements from license to section H.

Technical Specification page 3.6/4.6-Sa:

Delete remainder of original section 3.6.H.3.

Change page number to 3.6/4.6-5b to reflect movement of Single Loop Operation requirements from license to section H.

Technical Specification page 3.6/4.6-13:

Change from power-flow relationship to core plateAP-core flow relationship, This is because it has been shown that the core plate i aP-flow relationship is more accurate than the power-core flow relationship. Unit 2 has already had this change approved. {

Change Section H " Recirculation Pump Flow Hismatch" to " Recirculation Pump (

Flow Limitations" to reflect the addition of single loop operation to this {

section. '

Change reference to operation in Single Loop Operation not analyzed to a discussion of the analysis supporting Single Loop Operation. This section is identical to the one' approved for Unit 2 Cycle 9.

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REFERENCRS

1. GE document 23A5831 " Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station Unit 1 Reload 9," dated June 1987 (Attached).

2. NRC letter, MFN-148-85, H. N. Berkow to J. S. Charnley, " Acceptance for Approval of Puel Designs Described in Licensing Topic Report NEDE-240ll-P-A-6, Amendment 10 for Extended Burnup Operation," dated December 3, 1985 (Attached).

3. NRC letter MFN-082-85, C. O. Thomas to J. S. Charnley, " Acceptance for Referencing of Licensing Topical Report NEDE-240ll-P-A-6, Amendment 10,

' General Electric Standard Application for Reactor Fuel,'" dated May 28, 1985 (Attached).

4. GE letter JSC-058-84, J. S. Charnley to C. O. Thomas, " Submittal of Proposed Amendment 10 to GE LTR NEDE-24011-P-A-6," dated November 30, 1984.

5. NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel",

Revision 8.

6. GE document NEDO-24807, "Dresden Nuclear Power Station Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 Single Loop Operation,"

dated December 1980 (Attached).

7. NRC Generic Letter No. 86-09, " Technical Resolution of generic Issue No.

B-59 (N-1) Loop Operation in BWRs and PWRs," dated March 31, 1986 (Attached).

8. NRC letter C. O. Thomas to H. C. Pfefferlen, " Acceptance for Referencing of Licensing Topical Report NEDE-240ll, Rev. 6, Amendment 8, ' Thermal Hydraulic Stability Amendment to GESTAR II,'" dated April 24, 1985 (Attached).

9. GE document NEDC-31345P, " Quad Cities Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," dated July 1987 (Attached).

10. NEDC-23785P, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident," Volume I, II, and III.

11. GE document NEDE-31449, " Extended Operating Domain and Equipment Out-of-Service for Quad Cities Nuclear Power Station Units 1 and 2,"

dated June 1987 (Attached).

12. GE letter G-EBO-7-190, J. A. Miller to H. E. Bliss, "GE PRC 86-07 Limiting Control Rod Sequence for CRDA", May 6, 1987 (Attached).

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ATTACHMENT 2 l OUAD CITIES UNIT 1 CYCLE 10 RELOAD SIGNIFICANT HAZARDS EVALUATION Commonwealth Edison proposes to amend Facility Operating License DPR-29 for Quad Cities Unit 1 to support the Cycle 10 core reload. The proposed revisions include two basic types of changes: (a) changes specific to the cycle 10 reload fuel and (b) analyses and changes resulting from analyses performed to expand the operating region.

Description of Amendment Request The changes specific to the Cycle 10 reload fuel and analyses include:

1) Incorporation of the Cycle 10 Minimum Critical Power Ratio (MCPR) limit and new T AVE values and references resulting from the new ODYN methods.

2) Addition of Maximum Average Planar Linear Heat Generation Mate (MAPLHGR) limits for the reload fuel.

3) Addition of an LHGR limit specific to the GE8X8EB fuel.

4) 2ncreasing the rod block monitor setpoint.

The Technical Specification changes resulting from analysis performed i to expand the operating region and to allow operation with certain equipment out-of-service include:

5) Deletion of existing License Condition addressing Single Loop Operation (SLO) and incorporation of SLO in the body of the Technical Specifications.

E) cnanges to the analyzed operating region to include incresaed core flow (ICF) and feedwater temperature reduction (FTR).

7) Revision of the Automatic Pressure Relief Subsystem Technical Specification to require action only after two or more relief valves are found to be inoperable.

8) Deletion of the license operating restriction for coastdown to 40%

power and coastdown with off-normal FW heating.

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OPR-29 Issued 7/29/79 in addition, the licensee shall submit the additional information identified in Table 3.2 of this SE in accordance with the  !

schedule contained.herein. In the event these dates for

. submittal'cannot be met, the licensee shall submit a report explaining the circumstances, together with a revtsed schedule.

The licensee is required to implement the administrative controls  !

identified in Section 6 of the SE. The administrative controls shall be in effect immediately, except for those modifications indicated in Section 3.1 of the SE, which shall become effective  ;

on the dates indicated in Table 3.1 of the SE.

Am.'61 G. Systems Integrity 2/06/81 The licensee shall implement a program to reduce leakage from

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systems outside containment that would er could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the

-following.  ;

1.

Provisions establishing preventive maintenance and periodic ,

. visual inspection requirements, and '

4 2.

Leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

Am. 61 J. Iodine Monttorino 2/06/81 The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

l '. Training of personnel;

2. Procedures fer monitoring, and

3. Provisions for maintenance of sampling and analysis equipment.

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f' OPR-29 k F Am. 94 L. Post Accident Samplina 6/10/86 i

A program will be established, implemented, and maintained which will ensure the capability to obtain and analyze reactor coolant, radioactive lodines and particulate in plant chimney effluents, and containment atmosphere samples under accident conditions.

The program shall include the following:

1. Training of personnel.

2. Procedures for sampling and analysis, and 3.

Provisions for maintenance of sampling and analysts equipment.

07838

DPR-29 TABLE OF CONTENTS (Cont'd)

Page 3.5/4.5 CORE CONTAINMENT COOLING SYSTEMS 3.5/4.5-1 A. ' Core Spray Subsystems and the LPCI Mode of the RHR System 3.5/4.5-1 B. Containment Cooling Mode of the RHR System 3.5/4.5-3 C. HPCI Subsystem 3.5/4.5-4 D. Automatic Pressure Relief Subsystems 3.5/4.5-5 E. Reactor Core Isolation Cooling System 3.5/4.5-6 F. Minimum Core and Containment Cooling System Availability 3.5/4.5-6 G. Maintenance of filled Discharge Pipe 3.5/4.5-7 H. Condensate Pump Room Flood Protection 3.5/4.5-8 I. Average Planar Linear Heat Generation Rate (APLHGR) 3.5/4.5-9 J. Local (HGR 3.5/4.5-9 K. Minimum Critical Power Ratio (MCPR) 3.5/4.5-10

.3.5 Limiting Conditions for Operation Bases 3.5/4.5-11 4.5 Surveillance Requirements Bases 3.5/4.5-16 3.6/4.6 PRIMARY SYSTEM BOUNDARY 3.6/4.6-1 A. Thermal Limitations 3.6/4.6-1 B. Pressurization Temperature 3.6/4.6-1 C. Coolant Chemistry 3 . 6 / 4' . 6-2 D. Coolant Leakage 3.6/4.6-3 E. Safety and Relief Valves 3.6/4.6-4 F. Structural Integrity 3.6/4.6-4 G. Jet Pumps 3.6/4.6-5 H. Recirculation, Pump Flow limitations 3.6/4.6-5 l I. Shock Suppressors (Snubbers) 3.6/4.6-5a 3.6 Limiting Conditions for Operation Bases 3.6/4.6-8 3.7/4.7 CONTAINMENT SYSTEMS 3.7/4.7-1 A. Primary Containment 3.7/4.7-1 B. Standby Gas Treatment System 3.7/4.7-7 C. Secondary Containment 3.7/4.7-8 D. Primary Containment Isolation Valves 3.7/4.7-9 3.7 Limiting Conditions for Operation Bases 3.7/4.7-11 4.7 Surveillance Requirements Bases 3.7/4.7-15 3.8/4.8 RADI0 ACTIVE EFFLUENTS 3.8/4.8-1 A. Gaseous Effluents 3.8/4.8-1 B. Liquic Effluents 3.8/4.8-6a C. Mechanical Vacuum Pump 3.8/4.8-9 D. Environmental Monitoring Program 3.8/4.8-10 E. Solid Radioactive Waste 3.8/4.8-13 F. Miscellaneous Radioactive Materials Sources 3.8/4.8-14 H. Control Room Emergency Filtration System 3.8/4.8-14a 3.8/4.8.A Limiting Conditions for Operation and Surveillance Req. Bases 3.8/4.8-15 0783B 11

• QUA0 CZTXES-OPR-29 II. Dose Equivalent.I-131 - That concentration of I-131 (microcurie /

gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132. I-133, 1-134, and I-135 actually present. The thyroid dose conversion factors used ,

for this calculation shall be those listed in Table III'of-TIO-14844, " Calculation of Olstance Factors for Power and Test Reactor Sites." )

LJJ. Process Control Program (PCP) - Contains the sampling, analysis, and formulation determination by which solidification of radioactive i l wastes from liquid systems is assured.

KK. Offsite Dose Calculation Manual (00CM) - Contains the.methodo_ logy  ;

and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, and in the calculation of _,'

gaseous and 11guld effluent monitor alarm / trip setpoints. 1_

LL. Channel . Functional Test (Radiation Monitor) - Shall be the irijsction of a simulated signal into the channel as close to the sensos as

. practicable to verify operability including alarm and/ or. trip y functions.

i MM. Source Check - The qualitative' assessment of instrument response when the sensor is exposed to a radioactive source.

NN. Member (s) of the Public 'Shall include all persons who are not l

occupationally associated with the plant. This category does not include employees of the utility, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

00. DUAL LOOP OPERATION (DLO) - Reactor power operation with both recirculation pumps running.

PP. SINGLE LOOP OPERATION (SLO) - Reactor power operation with one recirculation pump running.

1.0-5 ' Amendment No.89 07838

. . QUAD C1 TIES DPR-29 2.1 LIMITING SAFETY SYSTEM SETTING BASES The abnormal operational transients applicable to operation of the units have ,

been analyzed throughout the spectrum of planned operating conditions in '

accordance with Regulatory Guide 1.49. In adottion, 2511 MWt is the licensed  !

maximum steady-state power level of the units. This maximum steady-state power level will never knowingly be exceeded.

Conservatism incorpora#ed into the transient analysis is documented in References 1 and 2. Transient analyses are initiated at the conditions given in these References.

The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by technical specifications. The effects of scram worth, scram delay time, and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of negative reactivity is assured by the time requirements for 5% and 20% insertion. By the time the rods are 60% inserted, approximately 4 dollars of negative reactivity have been inserted, which strongly turns the transient and accomplishes the desired effect. The times for 50% and 90%

insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.

1 The MCPR operating limit is, however, adjusted to account for the statistical variation of measured scram times as discussed in Reference 2 and the bases of Specification 3.5.K.

Steady-state operation without forced recirculation will not be permitted except 4 during startup testing. The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.

The bases for individual trip settings are discussed in the following paragraphs.

For analyses of the thermal consequences of the transients, the MCPR's stated in .

Paragraph 3.5.K as the limiting condition of operation bound those which are  !

conservatively assumed to exist prior to initiation of the transients.

A. Neutron Flux Trip Settings

1. APRM Flux Scram trip Setting (Run Mode) {

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated thermal power. Because fission chambers provide the basis input signals, the i APRM system responds directly to average neutron flux. During transients the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fucl.

I 07838 1.1/2.1 7 Amendment No.83 {

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1. RMFlowReferenceScram and @ RM Rod Block Settings 1:2 ,

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lil-lil-g 98- .

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81- +- SLO R00 SLK 1

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-+- SLO SCRM d d' - DL0 SCRm 8 SI-i  :

-*- DLO R00 BLK 4 ti-31-20-li-0 i i i , , , , , , , ,

t il 21 38 41 50 bl 71 il 90 ill ill 121 Recirculation loop Flow (I of rated)

Tigure 2.1-1 l

i Amendment No.

. . DPR.. y 160 120 - - AFRM - - BACKUP

- SCRAM AFRM 100 SLOCK ,

LINE (0.58WD + 50) 100 - ,

(100,87 (100,108)**

APRM 3 CRAM e'

LINE (0.58WD + 62 ,e

/

o' 8 0 ,,,,

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• NATURAL CIRCL*LATION f .

gggg 60 -

NOMINAL, CONSTANT XENON 100/100 POWER / FLOW LINE QOperatingRegionsupported ly N.E.D.0: -

24167 and I

N.E.D.0, 22192 40W *0perating on Single Loop or

. Natural Circulation is Lietted per Tech. Specs.

3.6.N.3 and 2.1.A.4 20E PUMP SPEED LINE

• • 0peration at greater than rated core flow is supported by NEDC-31449 RATED CONDITIONS POWER )

2511 MWeh CORE FLOW 98 M1bs/NR 0 .

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O 20 40 60 80 100 V 120 T CORE FLOW RATE (E OF RATED)

FICURE 2.1-3 (SCNEMATIC)

Amendment No.

APRM FLOW SIAS SCRAM RELATICNSNIP TO NORMAL CPERATING CONDITIONS i

l OPR-29 TABLE-3 3-3 IN57RUMENTA??ON THAT fNITIAVES ROD BLOCK Minimum Number of Operable or Tripped Instrument channels per Trio System I.11 Mstrument Trio tevel tettina 2 'APRM upscale (flow blas)[73 FeP W 1(0.58WD + 50]

HFLPQ i 2 APRM upscale (Refuel and 112/125 full scale l Startup/ Hot Standby mode) 2 APRM downscale(Il 13/125 full scale 1 Rod bloc nitor upscale (flow 10.65WD + 43( ]

bias) I 1 Rod block tronitor downscale(73 13/125 full scale l 3 IRM downscale(3] [8] 13/125 full scale 3 IRM upscale'(8) 1108/125 full scale 2(5) 12 feet below core centerline

$RM detector [ t in Startup position 3 IRM detector t in Startup 12 feet below core centerline position 2(5] (6) $RM upscale 1105 counts /sec 2(5) SRM downscale I93 1102 counts /sec 1 (per bank) High water level in scram i 25 gallons (per bank) discharge volume (50V) 1 50V high water level scram NA

! trip bypassed l

Notes

1. For the Startup/ Hot Standby and Run positions of the reactor mode selector switch, there shall be two operable or tripped trip systfyns for each function except the SRM rod blocks. IRM upscale and IRM downscale need not be operable in the Run position.

APRM downscale. APRM upscale (flow biased) and RBM downscale need not be operable in the Startup/ Hot Standby mode. The RBM upscale need not be operable at less than 307, rated thermal power. One channel may be bypassed above 30% rated thermal power provided that a Umtting control rod pattern does not exist. For systems with more than one channel per trip system if the first column cannot be met for one of the two trip systems, this condition may exist for up to 7 days provided that during that time the operable system is functionally tested intnediately and daily thereafter; if this condition lasts longer than 7 days the system shall be tripped.

If the first column cannot be met for both trip systems. the systems sna11 De tripped.

2. Wg is the percent of drive flow required to produce a rated core flow of 98 million Ib/hr. Trip level setting is In percent of rated power (2511 MWt).

3. IRM downscale may be bypassed when it is on its lowest range.

a. This function is bypassed when the count rate is GT/E 100CP5.

5. One of the four SRM inputs may be bypassed.

6. This SRM function may be bypassed in the higher IRM ranges (ranges 8. 9. and 10) when the IRM upscale rod block is operable.

7. Not reouired to be operable while performing b w power physics tests at atmrspm 4 pressure during or after refueling at power levels not to exceed SMWt.

8. This IRM function occurs when the reactor mode switch is in the Refuel or Startup/ Hot Standby position.

9. Thts trip 1s bypassed when the SRM is fully inserted.

07838 3.2/4.2-14 Amen @nent No. 90

Qua0-CITIES D7R-29 sidered inoperable, fully

• insertGd into the core, and p. ovide reasonable assurance electrically disarmed. that proper control rod drive performance is being

5. If the overall average of the ma ttita ined. The results of 20% insertion scram time data measurements performed on the )

generated to date in the current control rod drives shall ee cycle exceeds 0.71 seconds, the submitted in the annual MCPR operating limit must be operating report to the NRC.

modified as required by 3.

specification 3.5.K. The cycle cumulative mean scram time for 20% insertion will be determined intnedtately following the testing required in Specift-cationd 4.3.C.1 and 4.3.C.2 and the MCPR operating limit ad-justed, if necessary, as re- I quired by Specification 3.5.K.

D. Control Rod Accumulators D.

Control Rod Accumulators At all reactor operating pressures, a rod accumulator may be inoperable Once a shift, check the status of the provided that no other control rod in pressure and level alarms for each the nine-rod Square array around that accumulator.

rod has: ,

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1. An inoperable accumulated,

2. A directional control valve electrically disarmed wntle in a nonfully inserted pos*tton, or

3. A scram insertion greater than -

maximum permisstt,le insertion time.

If a control rod with an inoperable accumulator is inserted full-in and  !

its directional control valves are electrically disarmed. It shall not be constdered to have an inoperable accumulator, and the rod block asso-ctated with that inoperable accumu-l lator may be bypassed.

E. Reactivity Anomalies E. React h lif Nomalies The reactivity equivalent of the dif-ference between the actual critical During the startup test program and 1 rod configuration and the expected startups following refueling outages.

configuration during power operation the critical red configurations will shall not exceed 1% 0 k. If this be compared to the expected configur-attons at selected operating cond1- {

limit is exceeded, the reactor shall be shutdown untti the cause has been tions. These comparisons will be l determined and corrective actions used as base data for reactivity {

have been taken. In accordance with monitoring during subsequent power Specification 6.6. the NRC shall be operation throughout the fuel cycle.

notified of this reportable occur- At specific power operating condi- .

rence within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. tions, the critical rod configuration f will be compared to the configuration i expected based upon appropriately corrected past data. This comparison will be made at least every equiva-1ent full power month.

F. Economic Generation Control System F. Economte Generation Control System Operation of the unit with the eco-nomic generation control system with Prior to entering EGC and once per automatic flow control shall be per- shift while operating in EGC, the EGC operating parameters will be reviewed j l

missible only in the range of 65% to for acceptability. l 100% of rated core flow, with reactor 1 h power above 20%.

07838 3.3/4.3-5 Amenonent No. 91 f ,

QUAD-CITIE S

> ~ DPR-29

• operable innedtately. The RCIC system shall b2 demonstrated to be operable daily thereafter.

Daily demonstration of the auto-matic pressure relief subsystem operability is not required provided that two feeawater pumps are operating at levels above 300 MWe; and one feeawater provided that during such 7 days pump is operating as normally all active comoonents of the required with one additional automatic pressure relief sub- feedwater pump operable at power

. systems, the core spray sub- levels less than 300 MWe.

systems. LPCI mode of the RHR system, and the RCIC system are operable.

3. If the requirements of Specification 3.5.C cannot be met, an orderly shutdiwn shall be tnitiated. and the reactor pressure shall be reduced to 90 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. D.

Automatic Pressure Relief Subsystems Automatic Pressure Reitef Subsystems Surveillance of the autonatic pressure relief subsystem shall be performed as follows:

1. The automatic pressure relief 1. The following surveillance shall subsystem shall be operable be carried out on a six-month whenever the reactor pressure is surveillance interval:

greater than 90 psig, irradtated a.

fuel is in the reactor vessel With the reactor at pressure and prior to reactor startup each reitef valve shall be from a cold condition. manually opened. Relief valve opening shall be verified by a compensating turbine bypass valve or control valve closure.

2 2. A logic system functional test From and after the date that two shall be performed each or more of the five relief valves of the automatic pressure refueling outage.

relief subsysten are made or found to be inoperable when the 1.

A simulated automatic initiation reactor is pressurized above 90 whicn opens all pilot valves psig with Irradiated fuel in the shall be performed each re-reactor vessel, reactor fueling outage.

operation is permissible only

.during the succeeding 7 days

• Jniess repatrs are made and 4

'orovided that during such time When it is determined that two 2.he HPCI subsystem is operable. or more valves of the automatic l 4 pressure relief subsystem are

3. If the requirements of Specifi- inoperable, the HPCI shall be cation 3.5.0 cannot be met, an demonstrated to be operable orderly shutdown shall be intti- seneciately.

41ed and the reactor pressure stall be reduced to 90 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

0783B 3.5/4.5-5 Amendment No. 80 1

-R QUAD-CITIES

~ ~ DDR-29 .

b-

• withinthOprescribedlimitswithth2

' hours, the reactor Bhall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Survet11ance and corresponding action shall continue until reactor operation is within the prescribed limits. Maximum allowable LHCR 1s 13.4 kW/ft for fuel types P8X8A and 8P8X8R. For fuel types CE8X8E and GE8X8E8 the maximum allowable LHGR 1s 14.4 kW/f t.

K.

Minimum Critical Power Ratto (MCPR) K.

During steady-state coeration at Minimum Critical Power Ratto (MCPR) rated core flow. MCPR shall be The MCPR shall steady-state be odetermined daily during power greater than or equal to:

rated thermal power.peration above 251 of 1.33 for Tayg 1 0.71 see 1.37 for TAVE 1 0.86 sec 0.278 TAVE

• I'I31 for 0.71 see i Tayg 10.86 see where TAVE = mean 20% scram insertion time for all survet11ance data from specification 4.3.C which has been generated in the current cycle.

For core flows other than rated, these nominal values of HCPR shall be increased by a factor of kg where kr is as shown in Figure 3.5.2. If any time duri4g operatton it 15 determined by norpul survat11ance that the limiting value for NCPR ts being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady-state MCPR is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surve111anc1 and corresponding action shall continue untti reactor oparation is within the prescribed limits.

07838 3.5/4.5-10 Amendment No. 97

' QUAD-CZT!ES DPR-29 3.5 LIMITING CONDITION FOR OPERATION BASES A. Core Spray and LPCI Mode of the RHR System This specification assures that adequate emergency cooling capability }

is available whenever irradiated fuel is in the reactor vessel.

Based on the loss-of-coolant analytical methods described in General Electric Topical Report NEDC-31345P core cooling systems provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident, to limit calculated fuel cladding temperature to less than 2200*F, to assure that core geometry remains ,

intact, to limit cladding metal-water reaction to less than 1%, and to

{

limit the calculated local metal-water reaction to less than 17%.

The limiting conditions of operation in Specifications 3.5.A.1 through 3.5.A.6 specify the combinations of operable subsystems to assure the availability of the minimum cooling systems noted above. Under these

- Limiting Conditions of operation, l.ncreased surveillance testing of the remaining ECCS systems provides assurance that adequate cooling of the core will be provided during a loss-of-coolant accident.

Core spray distribution has been shown, in full-scale tests of systems similar in design to that of Quad-Cities I and 2, to exceed the minimum requirements by at least 25%. In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel. The accident analysis is additional conservative in that no credit is taken for spray cooling of the reactor core before the internal pressure has fallen to 90 psig.

The LPCI mode ;' the RHR system is designed to provide emergency coolingsto the core by flooding in the event of a loss-of-coolant accident. This system functions in combination with the core spray system to prevent excessive fuel cladding temperature. The LPCI mode of the RHR system in combination with the core spray subsystem provides adequate cooling2for break areas of approximately 0.05 ft2 up to and including 4.26 ft , the latter being the double-ended recirculation line break with the equalizer line between the recirculation loops closed without assistance from the high-pressure emergency core cooling subsystems.

The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in Reference 3. Using the results developed in this reference, the repair period is found to be less than l

07838 3.5/4.5-11 Amendment No. 98 l

i i

,------ unma en sgg 0PR.29 r .- Should the loss of .one RHR pump occur, a nearly full complement of core and contatnnent cooling equipment is available.

• core spray subsystem will p;rform the core cooling function,Thrte RHR oumps in conjunction with the Because of the availability of the majority of the core cooling equipment, which will be demonstrated to be operable, a 30-day repatr period is justified. If the LPCI mode of thethe fulfill RHR systen is notcooltng containment available, at least two RHR pumps must be available to function.

basis. The 7-day repair period is set on this

8. RHR Service Water l

The containment cooling mode of the RHR system is provided to remove heat energy from the containment in the event of a loss-of-coolant accident. For the flow specif ted, the containment long-term pressure is limited to less than 8 psig and is therefore more than ample to provide the required heat-renoval capability (reference SAR Section 5.2.3.2).

The Containment Cooling mode of the RHR System consists of two loops. Each loop consists ofequipment, electrical 1 Heat Exchanger, 2 RHR Pumps, and the associated valves, piping, and instrumentation. The "B" loop on each unit contains 2 RHR Service "A" Water Pumps. During the period f loop on each unit may utilize the "A" and "B" rom Novereer 24 1981, to July 1, 1982, the  !

2 via a cross-tte line. After July 1, 1982, each RHR Service Water Pumps from Unit "A" loop will contain 2 RHR Servtce Water containment Pumps.

cooling Either set of equipment is capable of performing the function.

Loss of one RHR service water pump does not seriously jeopardize the containment cooling capability, as any one of the renaining three pumps can satisfy the cooling requirenants. $1nce there is sont redundancy left, a 30-day repair period is adequate. Loss of-one loop of the containment cooling mode of the RHR system leaves one renaining system to perform the containment cooling function. The operable system is demonstrated to be operable each day when the above condition occurs. Based on the fact that when one loop of the containment cooling mode of the RHR system becomes inoperable, only one system rematns, which is tested daily, a 7-day repair period was specified.

j C. High-Pressure Coolant Injection The high-pressure coolant injection subsystem 15 provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI mode of the RHR system or core spray subsystems can protect the core.

The HPCI meets this requirement without the use of offsite electrical power. For the pipe breaks for which the HPCI is intended to function, the core never uncovers and is continuously cooled, thus no cladding damage occurs (reference SAR Section 6.2.5.3). The repair times for the limiting condtttons of operation were set considering the use of the HPCI as part of the tsolation cooling system.

D. Automatic Pressure Relief The relief valves of the automatic pressure relief subsystems are a backup to the HPCI subsystem. They enable the core spray subsysten and LPCI mode of the RHR l systen to provide protection against the small pipe break in the event of HPC1 f ailure by depressurizing the reactor vessel rapidly enough to actuate the core l spray subsystems and LPCI mode of the RHR system. The core spray subsystem and/or the LPCI mode of the RHR system provide sufficient flow of coolant to itmtt fuel cladding temperatures to less than 2200 0F, to assure that core geometry rematns intact, to limit the core wide clad metal-water reaction to less than 1%, and to limit the calculated local metal-water reaction to less than 17%.

Analyses have shown that only four of the five valves in the automatic tepressurization systen are recutred to operate. Loss of one of the relief valves does not significantly affect the pressure-reltaving capability, therefore continued operation is acceptable provided the appropriate MAPLHGR reduction factor 15 applied to assure canpliance with the 2200*F PCT limit. Loss of more than one relief valve significantly reduces the pressure relief capability of the ADS: thus, a 7 day repair period is specified with the HPCI available, and a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> repair period with the NPCI unavailable.

E. RCIC The RCIC system is provided to supply continuous makeup water to the reactor core when the reactor is isolated from the turbine and when the feedwater system 15 not available. Under these conditions the pumping capacity of the RCIC system 15 sufficient to maintain the water level above the core without any other water system in operation. If the water level in the reactor vessel decreases to the RCIC initiation at initiated level, anythe system automatically starts. The system may also be nunually time.

( . . .

OUAO-CITIES OPR-29 H. Condensate Pump Room Flood T*otection See Specification 3.5.H I. Average Planar LHGR This specification assures that the peak cladding temperature following the postulated design-basis loss-of-coolant accident will not exceed the 2200'F limit specified in the 10 CFR 50. appendix K considering the postulated effects of fuel pellet densification.

The peck cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat-generation rate of all the rods of a fuel assembly at any axial location and is only secondarily dependent on the rod-to-rod power distribution within an assembly. $1nce expected local variations in power distribution within a fuel assembly affect the calculated peak cladding temperature by less than i 20*F relative to the peak temperature for a typical fuel design, the limit on the average planar Ll1GR is sufficient to assure that calculated temperatures are below the limit. The maximum average planar LNGR's shown in Figure 3.5-1 are based on calculations employing the models described in Reference 2.

The Average Planar Linear Heat Generation Rate (APLHGR) also serves a secondary function which is to assure fuel rod mechantcal integrity.

)

l I

07838 3.5/4.5-13a Amenoment No. 83

{ ..

Qua0-CETIES .

OPR-39 J. Local LHGR This specification assures that the maximum linear heat-generation rate in any rod is less than the design linear heat-generation rate even if fuel pellet denstfication is postulated. The power spike penalty is discussed in Reference 2 and assumes a linearly increasing-variation'in axial gaps Detween core bottom and top and assures with 95% confidence that no more than one fuel rod exceeds the design LHGR due to power-spiking. No penalty is required in .$ specification 3.5.L because it has

'been accounted for in the reload transient analyses by increasing the calculated peak LHGR by 2.2%.

K. Minimum Critical Power Ratio (MCPR)

The steady state values for MCPR specified in this specification were selected to provide margin to acconenodate transients and uncertainties in monitoring the core operating-state as well as uncertainties In the critical power correlatio9 ttself. These values also assure that operation will be suct. that the initial condition assumed for the LOCA analysis plus two percer,t for uncertainty is satisfied. For any of the special set of transients or disturbances caused by single operator error or single equipment malfunction. It is reouired that design analyses initialized at tnis steady-state operating limit yield a MCPR of not less than that specified in Specification 1.1.A at any ting during the transient assuming instrument trip settings given in Specification 2.1. For analysis of the thermal consequences of these transtants. the value of MCPR stated in this specification for the limiting condition of. operation bounds the initial value of MCPR assumed to exist prior to the initiation of the transients. This initial condition which is used in the transient analyses will preclude violation of the fuel cladding integrity safety limit. , ,

Assumptions and methods used in calculating the required steady state i MCPR 11mit'for each reload cycle are documented in References 2 and 4.  !

The results apply with increased conservattse while operating with MCPR's greater than specified.

The most limiting transients with respect to MCPR are generally:

a) Rod withdrawal error b) Load rejection or turbine trip without bypass c) Loss of feedwater heater The MCPR Operating Limit reflects an increase of 0.03 over the most limiting transient to allow continued operation with one feedwater heater out of service.

Several factors influence which of these transients results in the l largest reduction in critical power ratio such as the specific fuel leading, exposure. and fuel type. The current cycle's reload licensing analyses specifies the limiting transients for a given exposure increment for each fuel type. The values specified as the Limiting Condition of Operation are conservatively chosen to bound the most restrictive over the entire cycle for each fuel type.

The need to adjust the MCPR operating limit as a function of scram time arises from the statistical approach used in the implementation of the 00YN computer code for analyzing rapid pressurttation events. Generic statistical analyses were performed fc? plant groupings of similar design which considered the statistical variation in several parameters (initial power level. CRO scram insertion time, and model uncertainty). These analyses (which are described further in Reference

4) produced generic Statistical adjustment Factors which have been applied to plant and cycle specific 00YN results to yield operating limits which provide a 95% probability with 95% confidence that the limiting pressurization event will not cause MCPR to fall below the fuel cladding integrity safety limit.

07838 3.5/a.5-la . Amenenent No. 83

QUAD-CITIES -

OPR-89 For core flow rates less'than rated..the steaGy state MCPR is increased by the formula given tn the specification. This ensures that tne MCPR will be maintained greater than that specified in Specification 1.1.A even in the event that the motor-generator set speed controller causes the scoop tube positioner for the fluid coupler to move to the maximum speed posttton.

References

1. " SAFER /GESTR-LOCA Loss of Coolant Analysis for QuadCities Nuclear Power Station Units 1 & 2" NEDC-31345P.*

2. " Generic Reload Fuel Application." NEDE-24011-P-A**

3. I. M. Jacobs and P. W. Marriott. GE Topical Report APED 5736. " Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards." April, 1969.

4

" Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors

• General Electric Co. Licensing Topical Report NEDO 24154 Vols. I and !! and MEDE-24154 vol. III as supplemented by letter dated Septemoer 5.

1980 f rom R.H. Suchholz (GE) to P. $. Check (NRC).

Approved revision at time of plant operation.

Approved reviston number at time reload fuel analyses are performed.

)

1 i

07838 3.5/4.5-15 Amenenent No. 83 l

l 1

l' l

GUAD CITIES DPA- 29

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GUA0 CITitS

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1

' QUAD-CITIES .

i OPR 29 e' . 'G. Jet rumps G. Jet Pumps l 1. Whenever the. reactor is in the 1. Whenever there is recirculation f Startup/ Hot Standby or Run modes, all jet pumps shall be- flow with the reactor in the Startup/ Hot Standby or Run intact, and 411' operating jet modes, jet pump integrity and 1 pumps shall be operable. If it operability shall be checked is determined that a jet pump is daily by verifying that the Inoperable an orderly shutdown following two conditions do not-shall be initiated and the occur simultaneously:

reactor shall De in a cold. j shutdown condition within 24 a. The recirculation pump flow hours, j differs by more than 10% j from the established  ;

speed-flow characteristics. j

b. The indicated total core flow is more than 10%

greater than the core flow -

value derived from  !

estab11she.1 core plate OP-core flow relationships. ]

2. Additionally, when operating 2.

with one recirculation pump with Flow indication from each of the the equalizer valves closed, the 20 jet pumps shall be verified diffuser to lower plenum prior to initiation of reactor differential pressure shall be I

startup from a cold shutdown checked daily, and the condition. i differential pressure of any jet )

pump in the idle loop shall not  !

vary by more than 10% from established patterns.

3. The indicated core flow is the 3. The baseline data required to sum of the flow indication from evaluate the conditions in each of the 20 jet pumps. If Specifications 4.6.G.1 and I flow indication failure occurs a.6 G.2 will be acquired each for two or more jet pumps, operating cycle.

inenediate corrective action {

shall be taken. If flow Indication for all but one jet pump cannot be obta?ned within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, an orderly shutdown shall be initiated and the reactor shall be in a cold '

shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

H. Recirculation Pump Flow Limitations H. Recirculation Pump Flow Limitations

1. Whensver both recirculation Recirculation pumps speed shall be pumps are in operatton. pump checked daily for mismatch, speeds shall be maintained i within 50% of each other when J power level is greater than 80%

and within 15% of each other i when power level is less than 80%.

2. If Specification 3.6.H.! cannot be met, one recirculation pump shall be tripped.

j I

I 07838 3.6/a.6-5 Amenenent No. 23

- - - - - - . - - - - _ 1

)

OUAD-C! TIES OPR-29

3. During Single Loop Operation for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the following restrictions are requtred:

4. The MCPR Safety Limit shall be increased by 0.01 (T.S.

,, 1.1A):

b. The MCPR Operating Limit shall be increased by 0.03 (T.S.~3.5.K);

c. The flow biased APRM Scram and Rod Block Setootnts shall be reduced by 3.5% to read as follows:

7.5. 2.1.A.1; 5 1 .58WD + 58.5 T.$. 2.1.A.1;

• 5 1 (.5860 + 58.5) FRP/MFLPD T.S. 2 1.8; 5 1 58WO + 46.5 T.S. 2.1.8;

• 5 1 (.58WO + 46.5) FRP/MFLPD T.S. 3.2.C (Table 3.2-3); * .

APRM UPSCALE i (.58WO +

46.5) FRP/MFLPD

• In the event that MFLPD excedes FRP.

d. The flow biased RBM Rod Block setpoints shall be reduced by 4.0% to read as follows:

T.S. 3.2.C (Table 3.2-3);

R8M UPSCALE 1 65WO + 39

e. The suction valve in the idle loop shall be closed and electrically isolated except when the idle loop is being prepared for return to service, l

07838 3.6/4.6-Sa Amendment No.

0

, QUAD-CITIES 6' OPR-29

!. Shock Suppressors (Snubbers)

'!. Shock Suppressors (Snebbers)'

The fo11 ewing surveillance require-ments apply to all snubbers Itsted in Table 3.6-9.

1. Visual inspections shall be l 1. During all modes of operation performed in accordance with the except Shutdown and Refuel, all following schedule utilizing the

! snubbers listed in Table 3.6-1 acceptance criterta given by

! shall be operable except as Specification 4.6.1.2.

noted in 3.6.1.2 following.

Number of Shubbers l 2. From and after the time that a Found Inoperable Next snubber is determined to be Ouring Inspection Required inoperable, continued reactor or During Inspec- Inspection operation is permissible during tion Interval Interval the succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> only.if the snubber is sooner made 0 18 months.

operable. 125%

3. If the requirements of 3.6.1.1 1 12 months and 3.6.1.2 cannot be met, an 25%

orderly shutdown shall be-initiated and the reactor shall 2 6 months be in a cold shutdown condition 125%

within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

3.4 124 days

4. If a snubber is determined to be 125%

inoperable while the reactor is in the Shutdown or Refuel mode. 5.6.7 62 days the snubber shall be: made 125% j operable prior to reactor startup. 18 31 days 125%

5. Snubbers may be added to safety-related systems without The required inspection interval prior license amenenent to Table shall not be lengthened more 3.6-1 provided that a revision than one step at a time.

to Table 3.6-1 is included with the next license amenenent Snubbers may be categorized in request. two groups. ' accessible' or

' inaccessible' based on their accessibility for inspection during reactor operation. These two groups may be inspected j independently according to the j above schedule. ]

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l 07838 3.6/4.6-5b ,

knenenent No.76

ope-39 G. Jet Pumps

• Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly, and/or rtser. increases.the cross-sectional flow area for blowdown followtng the postulated design-basis double-ended recirculation line break.

Therefore if a failure occurs, repairs must be made to assure the validity of the calculated consequences.

'The following factors form the basis for the surveillance requirements:

1. A break in a jet pump decreases the. flow resistance Characteristic' of the external piping loop causing the recirculation pump to operate at a nighsr flow condition when compared to previous operation,

2. The change in flow rate of the failed jet pump produces a change in the indicated flow rate of that pump relative to the other pumps in that loop. Comparison of the data with a normal relationship or

' pattern provides the indication necessary to detect a failed jet pump.

3. The jet pump flow deviation pattern derived from the diffuser to lower plenum differential pressure readtngs will be used to further evaluate jet pump operability in the event that the jet pumps fail the tests in Sections 4.6.G.1 and 2.

Agreement of indicated core flow with established core plate DP-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through. tnactive or broken jet pumps. This bypass flow is reverse with respect.to normal jet pump flow. The indicated total core flow is a sunnation of the flow indications for the 20. individual jet pumps. The total core flow measuring instrumentation sums reverse jet pump flow as-though it were forward flow. Thus, the indicated flow is higher than actual core flow by at least twice the normal flow through any backflowing pump.

Reactivity inventory is known to a high degree of confidence so that even if a jet pump failure occurred during a shutdown period. subsequent power ascension would promptly demonstrate abnormal control rod withdrawal for any power-flow operating map point.

A nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the converse is not true. The lack of any substantial stress in the jet pump body makes failure impossible without an initial nozzle riser system failure.

H. Recirculation Pump Flow Limitations The tPCI Hop selection logic is described in the SAR, Section 6.2.4.2.5.

For some limited low probability accidents with the recirculation loop operatinJ with large speed differences. it is possible for the logic to select the wrong loop for injection. For these limited conditions, the core spray itself is adequate to prevent fuel temperatures from exceeding allowable limits. However, to limit the probability even further, a procedural limitation has been placed on the allowable variation in speed between the recirculation ptsaps.

The licensee's analyses indicate that above 80% power the loop select logic could not be expected to function at a speed dif ferential of 15%. Below 80%

power, the loop select logic would not be expected to function at a speed differential of 20%. This specification provides a margin af 5% in pump '

speed differential before a problem could arise. If the reactor is operating on one pump.' the loop select logic trips that pump before r4 king the loop selection.

'Analysty base been performed which support indefinite single loop operation provided the appropriate restrictions are implemented within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The MCPR Safety Limit has been increased by 0.01 to account for core flow and TIP reading (; uncertainties which are used in the statistical analysis of the safety limit. The MCPR Operating Limit has also been increased by 0.01 to maintain the same margin to the safety limit as during Dual toop operation.

The flow biased scram and rod block setpoints are reduced to account for )

uncertainties associated with backflow through the idle jet pumps when the operating recirculation pump 15 above 20-40% of rated speed. This assures that the flow biased trips and blocks occur at conservative neutron flux levels for a given core flow.

The closure of the suction valve in the idle loop prevents the loss of LPCI flow through the idle rectreulation pump into the downcomer.

07838 3.6/4.6-13 Amendment No.

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