ML20215A515

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Proposed Tech Specs,Adding MAPLHGR Limits for Reload Fuel & Adjusting Min Critical Power Ratio Limit to Reflect Results of Cycle 9 Transient Analyses
ML20215A515
Person / Time
Site: Quad Cities, 05000000
Issue date: 09/18/1986
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20215A501 List:
References
NUDOCS 8610060103
Download: ML20215A515 (25)
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Text

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ATTACHMENT 1 OUAD CITIES UNIT 2 CYCLE 9 PROPOSED TECHNICAL SPECIFICATIONS l

l l

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i 2072K i

8610060103 860718 PDR ADOCK 05000265 P PDR

QUAD CITIES DPR-30 TABLE OF CONTENTS (Cont'd)

Page 3.5/4.5 CORE CONTAihMENT 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

0. Automatic Pressure Relief Subsystems 3.5/4.5-5 E. Reactor Core Isolation Cooling System 3.5/4.5-6 i 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 LHGR 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 3.6/4.6-4 E. Safety and Relief Valves 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 Operatong Bases 3.6/4.6-8 3.7/4.7-1 3.7/4.7 - CONTAINMENT SYSTEMS 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 Limiting Conditions for Operation Bases 3.7/4.7-11 3.7 3.7/4.7-15 4.7 Surveillance Requirements Bases 3.8/4.8-1 3.8/4.8 RADIOACTIVE EFFLUENTS A. Gaseous Effluents 3.8/4.8-1 B. Liquid 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 i 0614H 11 Amendment No. 84 l J

QUAD CITIES DPR-30 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, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation snall be those listed in Table III of TID-14844, " Calculation of Distance Factors For Power and Test Reactor Sites."

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

KK. Offsite Dose Calculation Manual (0DCM) - Contains the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, and in the calculation of gaseous and liquid effluent monitor alarm / trip setpoints.

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

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

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 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.

0614H 1.0-5 Amendment No. 84

DPR-30 i

@RM Flow Reference Scram and # RM Rod Block Settings t

138 ,I 128-i  ;

ill-ill-i g 91 4 ,

I j gg_'  : SLOR0DBLK

)70- -+- SLO SCRM jbli . DLO SCRM I

fi 584 y  : -*- DLO R0D BLK  ;

& 48- l 30-f

. 20-10-O i i i i i i i i . . .

I il 21 30 41 50 bl 70 80 90 ill ill 128 Recirculation loop Flow G of rated) l Figure 2.1-1

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DPR-30 140 l

^"* "^'*E" 8'"^"

120 ---- -- - -

7-------- , -i APRM ROD BLOCK '

LINE (0.58WD + 50)-

l s'

100 _

/

APRM SCRAM ,'

LINE (0.58WD + 62 ,' ,

/

/

80 _ ,' ,

/

/

I

  • NATURAL ON NOMINAL, CONSTANT XENON b /

60 - 100/100 POWER / FLOW LINE

/

/

Operating Region Supported 40 -

By N.E.D.O. - 24167 and N.E.D.O. 22192 20% PUMP *0perating on Single Loop or SPEED LINE Natural Circulation is Limited per Tech. Specs.

3.6.H.3 and 2.1.A.4.

l RATED CONDITIONS POWER 2511 MWth CORE FLOW 98 Mlbs/HR I I I '

'O C 20 40 60 80 100 120 W CORE FLOW RATE (% OF RATED)

T FIGURE 2.1-3 (SCHEMATIC)

Amendment No. 64 APRM FLOW BIAS SCRAM RELATIONSHIP ~

TO NORMAL OPERATING CONDITIONS

'I QUAD-CITIES OPR-30 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 />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. Maximum allowable LHGR for all 8X8 fuel types is 13.4 ~

KW/ft.

K. Minimum Critical Power Ratio (MCPR) K. Minimum Critical Power Ratio (MCPR)

During steady-state operation at The MCPR shall be determined daily during rated core flow, MCPR shall be steady-state power operation above 25% of greater than or equal to: rated thermal power.

l 1.38 for tave 1 0.73 secs 1.43 for Tave 1 0.86 secs 0.385 Tave + 1.099 for 0.73 < tave < 0.86 secs where tave - mean 20% scram insertion time for all surveillance data from specification 4.3.C which has been generated in the current cycle.

For cdre flows other than rated, these nominal values of MCPR shall be increased by a factor of kr where kr is as shown in Figure 3.5.2. If any time during operation it is de-termined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore opera-tion to within the prescribed i

limits. If the steady-state MCPR is t 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 />. Surveil-lance and corresponding action shall continue until reactor. operation is within the prescribed limits. _.

3.5/4.5-10 Amendment No. 51,69,79,80,85 0614H

MAPLHGR VS. Average Planar Exposure Fuel Type P8DRB239 12.5 12.s 7--

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s to.ess 2s.ess ss.ess 4s ess $s.ess everase Planar Exposure (redd/St)

NAPLHGR VS. Average Planar Exposure <

Fuel Type P8DGB263L 12.5 ,

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Figure 3 5-1 Sheet 1 of 6

MAPLHGR Vs. Auerage Pianar Exposure Fuel Types P8DGB265L. P8DRB265L 12.s ,

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MAPLHGR VS, Average Planar Exposure Fuel Type P8DRB2b5H BP8DRB265H s 2. 5

12. e ._ _

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Figure 3.5-1 Sheet 2 of 6

f MAPLHGR VS. Average Planar Exposure Fuei Type P8DRB282 12.5 12.s ,. -,

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NAPLHGR VS Average Planar Exposure Fuel Type BP8DRB283H 12.5 12.s ,4: --,_,

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MAPLHGR VS. Average Planar Exposure Fuel Type P8DGB284 12.5 , ,

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r MAPLHGR VS. Average Planar Exposure Fuel Type P8DGB298 12.5 , , ,

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MAPLHGR VS. Average Planar Exposure Fuel Type BP8DRB299L 12.5 D ---~.

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1 QUAD-CITIES DPR-30 G. Jet Pumps G. Jet Pumps

1. Whenever the reactor is in the 1. Whenever there is recircu-Startup/ Hot Standby or Run lation flow with the reactor in modes, all jet pumps shall be the Startup/ Hot Standby or Run intact, and all operating jet modes, jet pump integrity and 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 be in a cold shutdown condition within 24 a. The recirculation pump flow hours. differs by more than 10%

from the established

2. Flow indication from each of the speed-flow characteristics.

20 jet pumps shall be verified prior to initiation of reactor b. The indicated total core startup from a cold shutdown flow is more than 10%

condition. greater than the core flow value derived from estab-

3. The Indicated core flow is the lished core plate DP-core sum of the flow indication from flow relationships, each of the 20 jet pumps. If flow indication failure occurs 2. Additionally, when operating for two or more jet pumps, with one recirculation pump with immediate corrective action' the equalizer valves closed, the shall be taken. If flow diffuser to lower plenum indication for all but one jet differential pressure shall be pump cannot be obtained within checked daily, and the dif-12 hours, an orderly shutdown ferential pressure of any jet shall be initiated and the pump in the idle loop shall not reactor shall be in a cold vary by more than 10% from shutdown condition within 24 established patterns.

hours.

3. The caseline data required to H. Recirculation Pump Flow limitations l evaluate the conditions in Specifications 4.6.G.1 and l
1. Whenever both recirculation 4.6.G.2 will be acquired each pumps are in operation, pump operating cycle.

speeds shall be maintained within 10% of each other when H. Recirculation Pump Flow limitations [

l power level is greater than 80%

l Recirculation pumps speed shall be l and within 15% of each other l when power level is less than checked daily for mismatch.

! 80%.

l

2. If Specification 3.6.H.1 cannot be met, one recirculation pump shall be tripped.

l 0614H 3.6/4.6-5 Amendment No. 22 l

l _ _ -

QUAD-CITIES DPR-30

3. During Single Loop Operation for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the following restrictions are required:
a. The MCPR Safety Limit shall be increased by 0.01. (T.S.

1.1A);

b. The MCPR Operating Limit shall be increased by 0.01.

(T.S. 3.5.K);

c. The MAPLHGR Operating Limit shall be reduced by a multipitcative factor of 0.84. (T.S. 3.5.I);
d. The flow biased APRM Scram and Rod Block Setpoints shall_be reduced by 3.5% to read as follows:

T.S. 2.1.A.1; S 1 58HD + 58.5 T.S. 2.1.A.1;

  • S 1 (.58HD + 58.5) FRP/MFLPD T.S. 2.1.B:

S 1 58WD + 46.5

- T.S. 2.1.B;*

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

APRM upscale 1 (.5PWD +

46.5) FRP/MFLPD In the event that MFLPD exceeds FRP.

e. 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);

RBM Upscale 1 65WD + 38

. f. 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. Amendment No. 22 0614H 3.6/4.6-Sa

QUAD-C! TIES OPR-30 I. Shock Suppressors (Snubbers) I. Shock Suppressors (Snubbers)

1. During all modes of operation The following surveillance except Shutdown and Refuel, all. requirements apply to all snubbers snubbers listed in Table 3.6-1 listed in Table 3.6-1.

shall be operable exceptLas noted in 3.6.1.2 following. 1. Visual inspections shall be performed in accordance with the

2. From and after the time that a following schedule utilizing the snubber is determined to be acceptance criteria given by inoperable, continued reactor Specification 4.6.1.2.

operation is permissible during the succeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> only if Number of Snubbers the snubber is sooner made Found Inoperable Next operable. During Inspection Required or During Inspec- Inspection

3. If the requirements of 3.6.1.1 tion Interval Interval and 3.6.1.2 cannot be met, and orderly shutdown shall be 0 18 months initiated and the reactor shall  : 25%

be in a cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. I 12 months

25%
4. If a snubber is determined to be inoperable while the reactor is 2 6 months in the Shutdwon or Refuel mode, t 25%

the snubber shall be made operable prior to reactor 3, 4 124 days start-up. 25%

5. Snubbers may be added to 5,6,7 62 days safety-related systems without i 25%

. prior license Amendment to Table 3.6-1 provided that a revision >8 31 days to Table 3.6-1 is included with t 25%

the next license amendment request. The required inspection interval shall not be lengthened more than one step at a time.

Snubbers may be categorized in two groups, ' accessible' or

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

1 0614H 3.6/4.6-5b Amendment No. 22 I

QUAD-CITIES OPR-30 G. Jet Pumps Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly, and/or riser increases the cross-sectional flow area for blowdown following 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 higher 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 readings will be used to further evaluate jet pump operability in the event that the jet pumps fall the tests in Sections 4.6.G.1 and 2.

Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through inactive or broken jet pumps. This bypass flow is reverse with respect to normal jet flow. The indicated total core flow is a summation 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.

l 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 Limitation l The LPCI loop selection logic is described in the SAR, Section 6.2.4.2.5. For some limited low probability accidents with the recirculation loop operating 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 l prevent fuel temperatures from exceeding allowable limits. however, to limit the i probability even further, a procedural limitation has been placed on the allowable variation in speed between the recirculation pumps.

0614H 3.6/4.6-13 Amendment No. 22

QUAD-CITIES DPR-30 The licensee's analyses indicate that above 80% power the loop select logic could not be expected to function at a speed differential 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 of 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 making the loop selection.

Analyses have been performed which support indefinite single loop operation provided the appropriate restrictions are implemented within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-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 Loop 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 is 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 multiplicative 0.84 reduction of the MAPLHGR Operating Limit accounts for more rapid loss of core flow during some LOCA events when operating in Single !.oop than during Dual Loop. The closure of the suction valve in the idle loop prevents the loss of LPCI flow through the idle recirculation pump into the downcomer.

l l

l 3.6/4.6-13a Amendment No. 22 0614H

T l

l ATTACHMENT 2 l l

SUISEARY OF OUAD CITIES UNIT 2 CYCLE 9 RELDAD AND PROPOSED TECHNICAL SPECIFICATION CHANGES I. BACKGROUND Quad Cities Unit 2 Cycle 9 will use 88 BP8DRB299L and 64 BP8DRB299 reload fuel bundles. Both reload fuel types are prepressurized barrier fuel. Additional information on the Cycle 9 reload may be found in the

" Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station Unit 2, Reload 8 (Cycle 9)", 23A4758 (Attachment 3).

The following sections provide a discussion of the key features of the reload and a summary of the proposed Technical Specification changes for Unit 2 Cycle 9.

II. CORE WIDE TRANSIENTS A. Relief Valve Out-of-Service All Core Wide Transients were performed with the most restrictive relief valve i.e., the Target Rock S/RV, out-of-service. This assumption has been incorporated in recent Quad cities Transient Analyses to support a future Technical Specification amendment allowing operation with a relief valve out-of-service (currently under review within CECO).

I B. MCPR SAFETY LIMIT The current MCPR fuel cladding integrity safety limit of 1.07 is maintained for cycle 9 since no basic fuel design changes are being

' introduced (reload fuel is propressurized barrier 8x8 retrofit).

Minor enrichment or gadolinia variations do not affect the safety limit.

C. Limiting MCPR Transient i The Cycle 9 MCPR operating limit required to preclude violation of l the fuel cladding integrity limit is 1.37 for P8x8R and BP8x8R l fuel. This value is based on the Load Rejection without Bypass (LR w/o BP) event. The current MCPR LCO of 1.34 will therefore be increased to 1.38 to cover the larger CPR of this cycle and to accommodate potential future cycle increases in CPR (i.e.

restoring some margin for reload 10 CFR 50.59 application).

i l

l

_.._..._._..._.,_._,_.__,__.,_.m.._-.. - - . . _ - _ _ . _ _ . .

D. Compliance to ASME Pressure Vessel Code The results of the Q2C9 analyses for the postulated MSIV closure with indirect scram and no Relief Valve credit, provided in Attachment 3, indicate that the peak steamline pressure will be 1517 psig and the peak vessel pressure will be 1334 psig. These values are within the Technical Specification safety limit of 1345 psig for steam dome pressure and the ASME vessel over pressurization limit of 1375 psig (110% of design pressure).

E. ATWS Recirculation Pump Trip (RPT)

This reload analysis again includes the ATWS mitigating Recirculation Pump Trip (RPT) system with a trip setpoint of 1250 psig.

III. LOCAL TRANSIENTS A. Rod Withdrawal Error The Rod Withdrawal Error (RWE) event has been statistically analyzed on a generic basis and is no longer analyzed on a plant / cycle specific basis. The generic analysis provides assurance that the 1.07 MCPR 1 Lait will not be violated at the 95/95 probability /

confidence level. The results of the generic RWE analysis provide a CPR = 0.22 which, when added to the safety limit, provides an event LCO of 1.29. This is bounded by both the LR w/o BP event and the proposed technical specification operating limit of 1.38.

B. Fuel Loading Error Event The worst case bundle misorientation for Q2C9 results in MCPR equal to the 1.07 safety limit when its initial MCPR is less than or equal to 1.20 (1.18 + 0.02, GE calculation plus an NRC imposed l

variable water gap penalty). This is bounded by the initial CPRs required by the Q2C9 LR w/o BP analysis and is further bounded by l

the proposed Q2 Technical Specification operating limit of 1.38.

i IV. STABILITY ANALYSIS NRC approval of GE's amendment 8 to GESTAR II stated that a cycle specific stability analysis was not required for BWR 3's since they have l

been shown to have adequate stability margins. As a result, GE did not provide a stability analysis in the supplemental reload licensing submittal for Q2C9. However, GE was later requested and did provide a i stability analysis for Q2C9.

The Q2C9 decay ratio at the intersection of the natural recirculation line and the extrapolated rod block line power level is 0.56. Since the Technical Specifications do not allow continued operation in the natural circulation mode, combinations of low flow and high power sufficient to produce high decay ratios are not permitted. GE has also confirmed that the reduced slope of 0.58 for APRM Rod Block floor biasing was used in the Q2C9 stability analysis. This assures the continued acceptability of operating in the expanded power / flow region previously approved by the NRC.

V. ACCIDENTS A. Loss of Coolant Accident Loss of coolant accidents have been analyzed and approved by the NRC for Quad Cities Unit 2 in accordance with 10 CPR 50.46 and 10 CFR 50 Appendix K. The LOCA Analysis report for Quad Cities has been amended to include LOCA results for fuel type BP8DRB299L.

MAPLHGR curves have been incorporated into the Quad Cities Unit 2 proposed Technical Specifications submittal to include the two new fuel types (BP8DRB299 and BP8DRB299L). The curve for fuel type BP8DRB282 will also be extended to 40000 MWD /ST. Attachment 4 contains the appropriate Errata and Addenda sheets to NEDO 24146A for the above changes.

B. Rod Drop Accident The Rod Drop Accident (RDA) event has been statistically analyzed on a generic basis and is no longer analyzed on a plant / cycle

! specific basis. The generic analysis provides assurance that the 280 cal / gram enthalpy deposition limit will not be violated. The highest deposition of enthalpy calculated was 158 cal / gram. This Provides confidence on the 95/95 level that the Technical Specification limit will not be violated in the unlikely event of the postulated Design Basis RDA.

VI. SINGLE LOOP OPERATION (SLO)

A. Quad-cities has operated for 5 years with SLO capability. In 1981, NRC approval was granted and an amendment was placed in the operating license to incorporate required SLO restrictions. At this time it is desired to delete the restrictions from the license and place them in the body of the Technical Specifications. The following sections summarize previous analyses performed in support of SLO at Quad Cities.

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B. Plant Transients during SLO In Section 3.1 of Reference 1 GE has summarized their review of abnormal operating transients during Single Loop Operation (SLO). In response to NRC questions raised during their review of Cooper Station's request for SLO Technical Specifications, GE completed specific analyses of numerous plant transients initiated during SLO. The results demonstrate the applicability of the Reference 1 analyses for the CBCo BWR3's.

The four most important aspects of plant transients and SLO are:

(1) They are initiated from less than rated power due to the lower core flow:

(2) The safety limit MCPR and LOO MCPR have been increased (0.01) to account for increased uncertainties in core flow and TIP readings during SLO; (3) The APRM Scram and Rod block and RBM flow biased setpoints are adjusted to preserve the relationship normally existent between the setpoints and operating points in the power / flow map during two loop operation; and (4) The idle recirculation loop is effectively isolated by closing and electrically disarming either the discharge or suction valve, and by closing of the crosstie (equalizer) line; the internal idle jet pump loop is either in forward

! or reverse flow, depending on the speed of the operating pump.

Because the highest power attainable during SLO is expected to be some 18 to 28% less than rated two loop thermal power, plant transients become less severe. As demonstrated in GE analyses in support of the improved power / flow map, the most limiting power / flow condition is the 100% power /100% core flow point.

The increase in the MCPR safety limit for SLO by 0.01 to account for increased uncertainties in core flow and TIP readings (see Section VI.E) necessitated an increase in the MCPR LOO for SLO.

This provides the same protection on SLO as on two loop operation against penetration of the safety limit. Because of the lower initial power on SLO, the resulting CPRs are less for transients; hence, the LOO determined for two loop transients only needs to be increased by the 0.01 uncertainties adder which was applied to the MCPR Safety Limit.

1

r During SLO the reverse flowing idle loop jet pumps alter the normal two loop drive flow to core flow relationship by diverting flow away from the core. Because the flow-biased setpoints are referenced to drive flow, the setpoints must be reduced during SLO to re-establish the two loop relationship between drive flow and core flow. Should this correction not be made and a transient initiate during SLO, the new drive flow to core flow relationship would result in a flow biased trip occurring at a higher neutron flux to core flow ratio than planned.

GE has analytically calculated the magnitude of the setpoint reduction to be 3.5% (Based upon a slope of 0.58) for D 2/3 and Q 1/2. This correction has been applied in the proposed Technical Specifications. The 3.5% reduction in APRM setpoints corresponds to a 6.1% reduction in core flow.

Of the flow biased trips, only the RBM is actually credited in safety analyses. Modifying all of the flow-biased trips is conservative, appropriate, and provides the same degree of protection during SLO as two loop operation. The steeper slope of the RBM requires a slightly larger reduction in the RBM set point (6.1%) x (.65) = 4%.

In summary, abnormal plant transients initiated from SLO are conservatively bounded by two loop analyses providing the above mentioned adjustments are made to the MCPR safety limit and LCO, APRM Scram and Rod Block and RBM flow-biased setpoints, and the idle loop is sufficiently well isolated.

C. Accidents During SLO The Q 1/2 PSAR identifies 4 categories of design basis events:

Rod Drop Accident (RDA), Main Steamline Break (MSLa), Refueling Accident, and Loss-of-Coolant Accidents (LOCA). In addition, GESTAR identifies the Puel Assembly Loading Error and Recirculation Pump Seizure as design basis events. The consequences should one of these accidents occur during single loop versus two loop operation have been addressed by GE in GESTAR and Reference 1. A discussion of the LOCA and Recirculation Pump seizure events follow. Dual Loop Operation events bound Single Loop Operation events in the other four categories.

1. Recirculation Pump Seizure during SLO This accident is defined as the "... instantaneous stoppage of one recirculation pump shaft while the reactor is operating at full power." GE has considered this accident to be mild by comparison to a LOCA but analyzed the event for SLO for Browns Ferry Unit 1. The results of this

1 I

analysis demonstrate that the MCPR Safety Limit will not be penetrated, hence, no fuel failures are expected. The analysis was performed at power / flow ratios up to 82%/56%.

Because BF1 is a large core, BWR-4, high pcwer density plant, this analysis is conservatively applicable to SLO on D 2/3 and Q 1/2.

I

2. Loss-of-Coolant Accident during SLO The most limiting (two loop operation) LOCA for D 2/3 and Q 1/2 is a complete severance of the 28 inch recirculation suction pipe. GE developed a new ECCS-LOCA methodology for

! SLO (Reference 2). GE has compared the two loop and single loop vessel reflood times and hot node uncovery times (Reference 1, Figures 5.1, 5.2, and 5.3) and found them to be nearly identical (to within a few seconds in most instances). Based upon this and their SLO ECCS-LOCA Licensing Topical Report (LTR), GE finds the limiting break during SLO is again the complete severance of the recirculation suction pipe. In accordance with the LTR they have calculated MAPLHGR reduction factors in lieu of a full ECCS-LOCA analysis for SLO. The single loop MAPLHGR limits will be the two loop values multiplied by 0.84 for 8x8R and P8x8R fuel types. This reduction is required due to a decreased time to On-Set of Transition Boiling (OTB) should a LOCA initiate from SLO as compared to two loop operation. The 0.84 MAPLHGR reduction factor has been incorporated in the Iroposed Technical Specifications in lieu of the previous, conservative valve of 0.7.

For SLO, GE tequires either the suction or discharge valve in the idle loop recirculation line be closed and disarmed to prevent LPCI flow back through the idle pump and into the annulus through the suction pipe should a LOCA occur and LPCI be injected in the idle loop. This provision has been included in the proposed Technical Specifications.

D. Core Stability during SLO 1

The USNRC has addressed the issue of reactor stability in SLO in Generic Letter 86-09.

The following is an excerpt from that letter:

"...In low flow operating regions, it has been necessary to develop special operating procedures to assure that General Design Criteria 10 and 12 are satisfied in regard to thermal-hydraulic instabilities. Technical Specifications consistent with these procedures have been accepted by the staff for reactors which are not demonstrably stable based on analyses using approved analytical methods;..."

The USNRC has approved GE's stability methodology. The acceptance criteria is 0.80 versus a QC2C9 calculated decay ratio of 0.56.

This decay ratio was calculated at the intersection of the natural circulation line and the extended APRM Rod Block line.

This represents a point which is less stable than the regions of allowed operation at Quad cities Unit 2 under the existing and proposed Technical Specifications which does not allow continuous operation on natural circulation.

Based upon the approved GE methodology and the NRC generic letter, no Technical Specifications incorporating stability monitoring are required for SLO at Quad Cities.

E. Core Monitoring Uncertainties During SLO GE has identified larger measurement uncertainties during SLO in core flow and TIP readings (Reference 1 Section 2). The overall effect of factoring the new values for these uncertainties (6%

and 2.85%, respectively) into the process computer uncertainty has resulted in an increase in the latter from 8.7% to 9.1% for SLO on reload cores. This in turn resulted in an increase of 0.01 in the MCPR Safety Limit. In order to maintain the two loop margin between the MCPR LCO and Safety Limit, the LCO should also be increased by 0.01 during SLO. This value is incorporated in the proposed Technical Specifications in lieu of the previous conservative valve of 0.03.

F. APRM Noise and Core Plate P Fluctuations During SLO Brown's Ferry Unit I ran in the single loop mode in the Fall of 1978. During this time the unit experienced some increase in APRM noise and core plate p fluctuations apparently driven by core flow oscillations. Reports indicate the APRM noise increased to 12% (of scale) peak-to-peak at 85% pump speed.

During several conference calls between CECO and GE, GE has emphasized that their engineers have analyzed the effects of such fluctuations on core materials. Using 2.9 psi peak-to-peak fluctuation in core plate p and 20% of scale peak-to peak noise in APRMs, such things as cladding and fuel channel duty and crack propagation in core materials are not adversely affected (throughout the design lifetime of the channels, cladding, or core materials).

Subsequent Single Loop Operation tests performed at Brown's Ferry Nuclear Power Plant on February 9, 1985 have conclusively demonstrated that the increased Core Plate P and APRM noise experienced during SLO is caused by increased flow noise associated with back flow through the idle loop jet pumps and i does not represent a less stable mode of operation. Thus, the Core Plate P and APRM noise surveillance required in the current Quad Cities Unit 2 SLO licensing restriction is no longer needed and has therefore been deleted in the proposed License Amendment.

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VII. Proposed Technical Specification Changes Based on the preceeding discussion, Commonwealth Edison believes that the proposed Technical Specifications for Quad-cities Unit 2 Cycle 9 are acceptable and request their approval. The proposed changes are summarized below.

T.S Page, Section Description License p. 5, 6; Sec. 3.J Delete Provisions for SLO T.S. p. 11 Change title for section 3.6/4.6H in Table of Contents

p. 1.0-5 Add definitions of SLO and DLO Figure 2.1-1 Revise to show SLO and DLO scram and rod block settings Figure 2.1-3 Revise to reflect extended load line limit analysis previously implemented at Quad cities
p. 3.5/4.5-10 Sec. 3.5.K Revise to incorporate Cycle 9 MCPR limit. Also deleted reference to LHGR waiver for barrier ramp test in 3.5.J.

Figure 3.5-1, Sheet 1-6 Replotted and rearranged all figures for clarity. In addition, deleted fuel types 8DRB265L, 8D250 and 8D262 (no longer used); added reload fuel types BP8DRB299L and BP8DRB299; extended BP8DRB282 to 40,000 MWD /T.

p. 3.6/4.6-5, Sa, Sec. H Changed title, incorporated provisions for SLO.
p. 3.6/4.6-Sb Pagination change
p. 3.6/4.6-13, 13a Incorporate SLO in bases.

! .i VIII References i 1. GE document NEDO-24807, "Dresden Nuclear Power Station Units 2 and 3 l

and Quad Cities Nuclear Power Station Units 1 and 2 Single Loop Operation," dated December 1980 s

2. GE document NEDO-20566-2, " General Electric Company Analytical Model for Loss-of-Coolant Analysis in accordance with 10 CFR 50 Appendix K

! Amendment No. 2 - One Recirculation Loop Out-of-Service," dated July l -

1978 I ~ .

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