ML20215D023

Single Loop Operation Analysis

ML20215D023
Person / Time
Site:	Clinton
Issue date:	03/31/1986
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML20215D013	List: ML20215D023 U-600679, Forwards Clinton Single Loop Operation Analysis. Proposed Tech Spec Changes,Necessary to Implement Operation,Also Encl
References
MDE-45-0385, MDE-45-385, NUDOCS 8610100583
Download: ML20215D023 (64)

Text

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MDE-45-0386 DRF No. A00-02665 CLINTON SINGLE' LOOP OPERATION ANALYSIS

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MARCH 1986 Prepared for ILLIN0IS POWER COMPANY CLINTON POWER STATION 1

Prepared by GENERAL ELEClHIC COMPANY NUCLEAR EflERGY BUSIhESS OPERATI0flS SAN JOSE, CALIFORNIA 95125 l

8610100583 861007 PDR ADOCK 0500 1 i

CPS APPENDII-15.B

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TABLE OF CONTENTS P_ age 15.B RECIRCULATION SYSTEM SINGLE-LOOP OPERATION 15.B.1-1 15.B.1- INTRODUCTION AND.

SUMMARY

' 15.B.1-1~

15.B.2 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT ,

15.B.2-1 15.B.2.1 Core Flow Uncertain.ty 15.B.2-1 15.8.2.1.1 Core Flow Measurement During Single-Loop 15.B.2-1 Operation 15.8.2.1.2 Core Flow Uncertainty Analysis 15 B.2-2 15.B.2.2 TIP Reading Uncertainty 15.B.2-4 15.B.3 MCPR OPERATING LIMIT 15.B.3-1 15.B.3.1 Abnormal Operational Transients 15.B.3-1 15.B.3.1.1 Feedwater Controller Failure - Maximum Demand 15.B.3-2 15.B.3.1.2 Generator Load Rejection With Bypass Failure 15.B.3-3

15. B . 3.- l . 3

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Summary and Conclusions 15.B.3-3 15.B.3.2 Red Withdrawal Error 15.B.3-4 15.B.3.3 Operating MCPR Limit 15.B.3-5 15.B.4 STABILITY ANALYSIS 15.B.4-1 15.B.4.1 Phenomena 15.B.4-1 15.B.4.2 Compl.iance to Stability Criteria 15.B.4-2 15.B.5 LOSS-0F-COOLANT ACCIDENT ANALYSIS 15.B.5-1 15.B.5.1 Break Spectrum Analysis 15.B.5-1 15.B.5 2 Single-Loop MAPLHGR Determination 15.B.5-1 15.B.5.3 Small Break Peak Cladding Temperature 15.B.5-2 6

15.B-i

s.

CPS ~

TABLE OF CONTEhTS (Continued)

- Page 15.B.6 CONTAINMENT ANALYSIS' - 15.B.6-1 15.B.7 f1ISCELLANE005 IMPACT EVALUATION 15.B.7-1 15.B.7.1' -Anticipated. Transient Without Scram Impact- 15.B.7-1 15.B.7.2 Fuel Mechanical Performance 15.B.7-1 15.B.7.3 Vessel-Internal Vibration 15.B.7-2 15.B.8 REFERENCES 15.B.8-1 ,

4 9

e 4

e 15.B-11

CPS LIST OF-TABLES f! UMBER TITLE PAGE 15.8.3-1 Input Parameters and Initial Conditions 15.B.3-7,8,9 15.B.3-2 Summary of Transient Peak Value and CPR 15.8.3-10 Results 1

15.8-iii

' CPS LIST OF FIGURES- l NUMBER TITLE PAGE

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15.B.2-1 -Illustration.of Single Recirculation Loop 15.B.2 Operation Flows 15.B.3-1 Feedwater Controller Failure - Maximum 15.B.3-ll, 12 Demand, 70.2", Power, 53.6% Flow-15.B.3-2 Generator Load Rejection with Bypass 15.B.3-13, 14 Failure, 70.2". Power, 53.6% Flow

'15.B.5-l' Uncovered Time vs. Break Area Suction 15.B.5-3

. Break, LPCI Diesel Generator Failure I

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.CFS

-15.BRECIRCULATIONSYSTEMSSINGLE-i_60POPERATION

- 15.B.1 INTRODUCTION AND

SUMMARY

- Single-loop operation (SLO) at reduced power is highly resirable in the event recirculation pump or other component maintenance renders one loop

- inoperative. To justify single-loop operation, accidents and abnormal operational transients associated with power operations, as presented in Sections 6.2 and 6.3 and the main text of Chapter 15.0, were reviewed for the single-loop case with only one pump in operation. This appendix presents the results of this safety evaluation for the operation of the Clinton Power Station (CPS) with single recirculation loop _ inoperable.

This evaluation is performed for PSX8R fueled core on an equilibrium cycle basis and is applicable to both initial and reload cycles oper-ation. The conditions are those of continued operation in the operating

-domain currently defined in Figure 4.4.5 of Chapter 4 up to maximum ~

power of approximately 70',' of rated.

Increased uncertainties in the core total flow and Traversing In-Core Probe (TIP) readings resulted in a 0.01 incremental increase in the Minimum Critical Power Ratio (MCPR) fuel cladding integrity safety limit during single-loop operation. No increase in rated MCPR operating limit and no change in the power dependent and flow dependent MCPR limit (MCPR p and MCPR 7

) are required because all abnormal operational transi-ents analyzed for single-loop operation indicated that there is more than enough i&PR margin to compensate for this increase in MCPR safety limit. The recirculation flow rate dependent rod block and scram set-point equation given in the CPS Technical Soecifications are adjusted for one-loop operation.

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Thermal-hydraulic stability was evaluated for its acequa'cy with respect to Gen" al Design Criteria 12 (10CFR50, Appendix A). It is shown that SLO satisfies this stability criterion. It is further shown that the

increase in neutron noise observed during SLO is independent of system stability margin.

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15.B.1-1 f

CPS.

'To prevent potential' control oscil'1ations from occurring in the recircu-lation. flow control syst2m, the flow, control in-the active' loop should

be in manual ~ mode for single-loop operation.

The liatting Maximum Average' Planar Linear Heat Generattbn' Rate.

. U%PLHGR) reduction factor for single-loop operation is calculated to be 0.85.

The containment response for a Design Basis Accident (DBA) recirculation line break with single-loop operation is bounded by the rated power two-loop operation analysis. presented in Section 6.2. This conclusion covers all single-loop operation power / flow conditions. .

LA generic assessment was made to determine-the impact'of. single-locp operation on.the Anticipated Transient Without Scram (ATUS). It was found that consequences of ATUS events postulated during single-loop ~

operation would be bounded by dansequences of ATWS events during two-loop operation.

.The fuel thermal and mechanical duty for transient events occurring during SLO is found to be bounded by the fuel design bases. The Average Power Range Monitor (APRM) fluctuation,should not exceed a flux ampli-tude of 15" of rated and the core plate differential pressure fluctua-tion should not exceed 3.2 psi peak to peak .to be consistent with_the fuel rod and assembly. design bases.

A recirculation pump drive ficw limit is imposed for SLO. The highest drive flow that meets acceptable vessel internal vibration criteria is the drive flow limit for SLO. Actual drive flow limit'in SLO was determined at Kuo Sheng 1, the BWR6/218 prototype plant and is about

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33,000 gpm. '

h 15.s.1-2 1

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CPS

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15.B.2. MCPR FUEL CLADDING IllTEGRfi7 SAFETY LIMIT-

..Except:for. core total flow and TIP reading, the uncertainties used-in the statistical. analysis to determine the MCPR fuel cladding integrity safety 1.imit are not dependent on whether coolant flow'tt*provided by one:or-two recirculation. pumps. Uncertainties used in the two-loop operation analysis are. documented in the FSAR. A 6% core flow measure-

. ment uncertainty has-been' established for single-1 cop operation (ccm-

. pared to 2.5% for two-. loop operation). As shcwn below, this value conservatively reflects the one standard deviation (one sigma) accuracy of the core flow measurement system documented in Reference'15.B.8-1.

The random noise compcnent of the TIP reading uncertainty was revised for single recirculation loop cperation to reflect the operating plant test results given in Subsection.15.8.2.2. This revision resulted in a.

single-loop operation process computer effective TIP uncertainty of 6.8%

for initial cores and 9.1% for reload cores. Comparable two-loop ' '

process computer uncertainty values are 6.3%-for.' initial cores and 8.7%-

for reload cores. The net effect of these two revised uncertainties'is a 0.01 increase in the required MCPR fuel cladding. integrity safety limit.

15.B.2.1 Core Flow Uncertainty 15.B.2.1.1 Core Flow Measurement During Single-Loco Ooeration The jet punp core flow measurement system is calibrated to measure core-flow when both sets of jet pumps are in forward flow; total core flow is the sum of the indicated locp flows. For single-1cep operation, how-ever,someinactivejetpumpswillbebackficwinglatactivepumpf. low above approximately 35%). 'Therefore, the measured ficw in the backficw-

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ing jet pumps must be subtracted from the measured f1cw in the active loop to obtain the total core f' low. ~

In addition, the fet pump coeffi-cient is different for reverse flow than for forward flow, and the measurement of reverse flow must becm'dified to account for this i

difference.

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15.B.2-1

CPS y

In single-loop operation, the total' core flow is derived by the follow-ing formula:

f

- Total Core , Active Loop 1 f Inactive Loop 1

-C Flow Indicated Flow Indicated Flow

- i .i L A Where C (= 0.95) is defined as the ratio of " Inactive Loop True Flow" to

" Inactive Loop Indicated Flow". " Loop Indicated Flow" is the flow measured by the jet pump " single-tap" loop flow summers and indicators, which are set to read forward flow correctly.

The 0.95 factor was the result of a conservative analysis to appropri-ately modify the single-tap flow coefficient for reverse flow.* If a more exact, less conservative core flow is required, special in-reactor calibration tests would have to be made. Such calibration tests would involve: calibrating core support plate AP versus core flow during one-pump and two-pump operation along with 100*J flow control line and calculating the correct value of C based on the core support plate AP and the loop flow indicator readings.

15.B.2.1.2 Core Flow Uncertainty Analysis The uncertainty analysis procedure used'to establish the core flow uncertainty for one-pump operation is essentially the same as for two-pump operation, with some exceptions. The core flow uncertainty analysis is described in Reference 15.B.8-1. The analysis of one-pump core flow uncertainty is summarized below.

For single-loop operation, the total core flow can be expressed as follows (refer to Figure 15.B.2-1):

W "

C A-NI l .

• The analytical expected value of the "C" coefficient for CPS is s0.82.

i 15.B.2-2 l

. - J

CPS where:

W C

= total core flow, Wg = active loop flow, and -

W.

7

= inactive _ loop (true) ficw.

By applying the " propagation of errors" method to the above equation, the variance of the total ficw uncertainty can be approximated by:

"h a c sys A I rand rand where: '

og = uncertainty of total core ficw; og = uncertainty systemic to both locps; og =randemuncertaintfofactivelooponly; og = randem uncertainty of inactive locp only; o

c

= uncertainty of "C" coefficient; and a = ratio of inactive loop ficw (W T 7 to ' active icep ficw (Ug ).

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CPS From an uncertainty analysis, the c6nservative, bounding values of

' c, ,o g og and e C are 1.6%, 2.6%, 3.5%, and 2.8%.

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sys A I rand rand ,

respectively. Based on the above uncertainties and~a bounding value of 0.36* 'for "a", the. variance of the total flow uncertainty is approximately:

o (2.6)2 + -(l-

= (1.6)2 + (1-0 36) - )2 ((3.5)2+(2.8)2

= (5.0%)2 When the effect of 4.1% core bypass flow split uncertainty at 12%

(bounding case) bypass flow fraction is added to the total core flow uncertainty, the active coolant flow uncertainty is:

c (4.1%)2 = (5.1%)2 active = (5.0%)2 + (1- 1) coolant which is less than the 6% flow uncertainty assumed in the statistical analysis.

In summary, core flow during one-pump operation is measured in a con-servative way and its uncertainty has been conservatively evaluated.

15.8.2.2 TIP READING UNCERTAINTY To ascertain the TIP noise uncertainty for single recirculation loop operation, a test was performed at an operating BWR. The test was 4

performed at a power level 59.3% of rated with a single recirculation' pump in operation (core flow 46.3% of rated). A rotationally symmetric control rod pattern existed during the test.

• This flow split ratio varies from about 0 13 to 0 36 The 0.36 value . . .

is a conservative bounding value. The analytical expected value of the l flow split ratio for CPS is s 0.28. I i

j 15.8.2-4

, ,._ . _ . _ . . _ . _ . -_ __. _-- -- _ _ _ _ _ .- . -~ -

g-l L CPS t-Five consecutive traverses were made' with each of five TIP machines,.

-giving a total of 25 traverses. Analysis of this data resulted in a .

nodal TIP noise of 2.85%. Use of this TIP noise va'lue as a component of the process. computer total uncertainty results in a one-sigma process computer total effective TIP uncertainty value .for single-loop operation of 6.8% for ' initial cores- and 9.1% for reload corcs.

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W = Total Core Flow C

tl = Active Loop Flow A

W; = Inactive Loop Flow filinois Illustration of Single Recirculation Loop Figure Power Company Operation Flows 15.3.2-1 ,

15eRm2-6

CPS 15.B.3 MCPR OPERATING LIMIT 15.B.3.1 ABNORMAL OPERATING ~ TRANSIENTS

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Operating with one recirculation loop results in a maximum power output which-is about 30% below that which is attainable for two-pump oper-ation. Therefore, the consequences of abnormal operation transients from

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one-loop operation will be considerably less severe than those analyzed from a two-loop operational mode. For pressurization, flow increase, I

flow decrease, and cold water infection transients, the results pre-sented in Chapter 15 bound both'the thermal and overpressure conse-ouences of one-loop operation.

The consequences of flow decrease transients are also bounded by the 4

full power analysis. A single pump trip front one-loop operation is less severe than_a two-pump trip from full power because of the reduced initial power _-level.

The worst flow increase transient results from recirculation flow con-troller failure, and the worst cold water injection transient results from the loss of feedwater heater. For the former, the MCPR curve is f

-derived from.a postulated runout of both recirculation loops. This.

condition produces the maximum possible' power increase and hence maximum ACPR for transients initiated fron less than rated power and flow. When operating with only one recirculation loop, the flow and power increase associated with this failure with only one loop will be less than that associated with both loops; therefore, the MCPR curve derived with the f

two-pump assumption is conservative for single-loco operation. The latter event, loss of feedwater heating, is generally the most severe.

cold water increase event with respect to increase in core power. This event is caused by positive reactivity insertion from core inlet sub-cooling and it is relatively insensitive to init'al i power level. A generic statistical loss of feedwater heater analysis using different initial I

power levels and other core design parameters concluded one-pump operation with lower irittial power level is conservatively bounded by

the full power two-pump analysis. Inadvertent restart of the idle 15.B.3-1

CPS recirculation purp has been analyzed in the FSAR (Chapter 15.4.4) and is still applicable for single-loop operation. .

'From the above discussions, it is concluded that the transient conse-quence'from one-loop operation is bounded by previously submitted full pcwer analyses. .The maximum power level that can be attained with one-loop _ operation is only restricted by the MCPR and overpressure limits established frcm a full-power analysis.

In the following sections, the results of two of the most limiting transients analyzed for single-loop operation are presented. They are, respectively:

a. feedwater flow controller failure (maximum demand), (FWCF)

b. generator load rejection with bypass failure, (LRBPF).

The plant initial ccnditions are given in Table 15.B.3-1.

15.B.3.1.1 Feedwater Controller Failure - Maximum Demand The computer model described in Reference 15.B.8-2 was used to simulate this event. '

The analysis has been performed with the plant' conditions tabulated in Table 15.B.3-1, except the initial vessel water level at level setpoint l L4 for conservatism. By lowering the initial water level, more cold feedwater will be injected before Level 8 is reached resulting in higher heat fluxes.

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The simulated feedwater controller failure transient is shcwn in Figure 15.B.3-1.

Table 15.B.3-2 gives a summary of the transient analysis results. The ca'lculated MCPR is 1.27, which is well above the safety limit MCPR of 1.07 so no fuel failure due to boiling transition is predicted. The

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CPS peak vessel pressure predicted is 1065 psig and is well below the ASME limit of 1375 psig.

15.8.3.1.2 Generator Load Rejection With Bypass Failure The computer medel described in Reference 15.B.8-2 was used1 to simulate this event.

These analyses have been performed, unless otherwise noted, with the plant conditions tabulated in Table 15.B.3-1.

The simulated generator load rejection with bypass failure is shown in Figure 15.B.3-2. ,

Table 15.B.3-2 summarizes the transient analysis results. The peak vessel pressure predicted is 1181 psig and well below the ASME limit of 1375 psig. The calculated MCPR is 1.34 which-is considerably above the safety limit MCPR of 1.07.

15.B.3.1.3 Summary and Conclusions The transient peak value results and the Critical Power Ratio (CPR) results are summarized in Table 15.B.3-2. This table indicates that for the transient events analyzed here, the MCPRs for all transients are above the single-loop operation safety limit v'alue of 1.07 It is concluded the operating limit MCPRs established for two-pump operation are also applicable to single-loop operation conditions.

For pressurization, Table 15.B.3-2 indicates the peak pressures are below the ASME code value of-1375 psig. Hence, it is concluded the pressure barrier integrity is maintained under single-loop operation conditions.

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CPS 15.B.3.2 R0D WITHDRAWAL ERROR -

The rod withdrawal error'(RWE) transient for two-lo'op operation docu-mented -in Chapter 15 employs a statistical evaluation of the minimum critical power ratio (MCPR) and linear heat generation rate (LHGR) response to the withdrawal of ganged control rods for both rated and off-rated conditions. The required MCPR limit protection for the event is provided by the rod withdrawal limits (RWL) system. Since this analysis covered all off-rated condition in the power / flow operating map, single-loop operation is bounded by the current technical specification.

The Average Power Range Monitor (APRM) rod block system provides addi-tional alarms and rod blocks when power levels are grossly exceeded.

Modification of the APRM rod block equation (below) is required to maintain the two loop rod block versus power relationship when in one loop operation.

One-pump operation results in backflow through 10 of the 20 jet pumps while the flow is being supplied into the lower plenum from the 10 active jet pumps. Because of the backflow through the inactive jet pumps, the present rod block equation was conservatively modified for use during one-pump operation because the direct active-loop flow measurement may not indicate actual flow above about 35% core flow without correction.

A procedure has been established for correcting the APRM rod block equation to account for the discrepancy between actual ficw and indi-cated flow in the active loop. This preserves the original relationship between APRM rod block and actual effective. drive flow when operating with a single loop.

The two-pump rod block equation is:

RB = mW + RB a m(100) 100 1

15.B.3-4 -

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CPS The one-pump equation becomes: '

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- R B = rM + RB100 - m(100) - maw where AW =

difference between two-loop and single-loop effective 4 drive ficw at the same core flow. This value is' expected tc be 8% of rated (to be determined by the Illinois Power Company during startup testing);

RB =- pcwcr at rod block in %;

m= flow reference slope W= drive flow in % of rated.

RB 100 = top level rod block at 100% flow.

If the rod block setpoint (RB100) is changed, the equation must be recalculated using the new value.

The APRM sqran trip settings are flow biased in the same manner as the APRM rod block setting. Therefore, the APRM scram trip settings are subject to the same procedural changes as the rod block settings dis-cussed above.

15.B.3.3 OPERATING MCPR LIMIT For single-loop operation, the operating MCPR limit remains unchanged from the normal two-loop operation limit. Although the increased uncer-tainties in core total flow and TiP readings resulted in a 0.01 increase in MCPR fuel cladding integrity safety' limit during single-loop oper-ation (Section 15.8.2), the limiting transients have been analyzed to indicate that there is'more than enough MCPR margin during single-loop operation to compensate for this increase in safety limit. For single 15.B.3-5 1

CPS loop operation at off-rated condition's, the steady-state operating MCPR limit.is established by the MCPRp and MCPR f curves. This ensures the .

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99.9% statistical limit requirement is always sat.fsfied for any postu-g lated abnormal operational occurrence. The abnormal operating transi-ents analyzed concluded that current pcwer dependent MCPR limits are p

bounding for single loop operation. Since the maximum core flow runout during single loop operation is only about 54% of rated, the current flow dependent MCPR f limits which are generated based on the flow runaut up to rated core flow are also adequate to protect the flow runout events during single loop operation.

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l CPS d

, TABLE 15.B.3-1 INPUT PARAMETERS AfiD IflITIAL C0flDITI0flS 1.' T,hermal Pcwer Level, MWt Analysis Value 2032

2. ' Steam Flow, lb per see Analysis Value 2304

3. Core F.lew, lb/hr 4.53 x 107

4. Feedwater Flow Rate, 1b/sec Analysis Value 2304

5. Feedwater Temperature, 'F 383

6. Vessel Dome Pressure, psig 978

7. _Cere' Pressure, psig 983 .

,_ 8. Turbine Bypass Capacity..% tlSR 35

9. Core Coolant. Inlet Enthalpy, Btu /lb 509 -

4 10..-Turbine Inlet Pressure, psig 943

11. Fuel Lattice P8x8R

12. Core Average Gap Conductance, Btu /sec-ft*- F 0.189

13. Core Leakage Flow, % 12.7

-14.

Required MCPR Operating Limit First Core 1.39

15. MCPR Safety Limit for Incident of Modera'te Frequency First Core 1.07 Reload Core 1.08

16. Doppler Coefficient c/*F **

, 17. Void Ccefficient c/% Rated Voids **

18. Scram Reactivity, $aK **

1 1-

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• • These values are calculated within the Computer code (Reference

15.B.8-2) based on the input from CRUNCH tape. '

15.B.3-7

_ _ _ = -_ _

CPS TABLE 15.3.3-l'(Continued)

IllPUT PARAMETERS ATID' INITIAL C0tIDITI0tlS

19. Control Red Drive Speed See FSAR Figure 15.0-1 Position Versus Time

20. Core Average Void Fraction, % 45.7

21. Jet Pump Ratio, M 3.71

22. Safety / Relief Valve Capacity, " NBR at 1210 psig 115.1 Manufacturer Dikker Quantity Installed 16

23. Relief Function Delay, sec. ~0.15

24. Relief Function Response Time Constant, sec. 0.10

25. Safety Function Delay, sec. 0.0

26. Safety Function Response Time Constant, sec. 0.2

27. Setpoints for Safety / Relief Valves Safety Function, psig 1175,1185,1195,1205,1215 Relief Function, psig 1145,1155,1165,1175

28. Number of Valve Groupings Simulated Safety Function 5 Relief Function 4

29. SRV Reclosure Setpoint - Both Mode

(",of Setpoint)

Maximum Safety Limit 98 Minimum Operational Limit 89

30. High Flux Trip, ". NBR Analysis Setpoint 127.2 (122x1.042),".NBR

31. High Pressure Scram Setpoint, psig' 1095

32. Vessel Level Trips, Feet Above Separator Skirt Bottom .

Level 8 - (L8), Feet 6.00 Level 4 - (L4), Feet 3.87 Level 3 - (L3), Feet 1.94 Level 2 - (L2), Feet -2.86 15.B.3-8

CPS

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TABLE 15.B.3-1 (Continued)

INPUT PAPAMETERS AND INITIAL C0t!DITIONS

33. APRM Thermal Trip Setpoint, 5 NBR 118.8 (114x1.042)

34. TPM Time Constant, sec. 7.0

35. RPT Delay one load rejection or turbine trip, sec. 0.14

36. RPT Inertia Time Constant for Analysis, sec.

5.0

37. Total Steamline Volume, ft3 3275

38. Pressure Setpoint of Recirculation Pump 1135 Trip - psig (Nominal)

Note: The transient analyses evaluated with the ODYN 06 code use slightly more conservative values for some parameters than described elsewhere in the FSAR.

15.B.3-9

CPS TABLE 15.B.3-2 '

SUMMARY

OF TRANSIENT PEAK VALUE AND CPR RESULTS LRBPF FWCF Initial Power / Flow (% Rated) 70.2/53.6 70.2/53.6 Peak Neutron Flux (% NBR) 70.3 83.6 Peak Heat Flux (%' Initial) 100.3 106.5 Peak Dome Pressure (psig) 1167 1051 Peak Vessel Bottom Pressure (psig) 1181 1065 Required Two Loop Initial ~MCPR- .

Operating Limit at SLO Condition 1.39 1.39 ACPR 0.05 0.12

. Transient MCPR 1.34 1.27 SLMCPR at SLO 1.07 1.07 S'

4 e

4 15.B.3-10

1 NEUTRON FLUX 2 PEAK FUEL CENTER TEM?

,, 3 AVE SURFF CE HEAT FLUX 150' _' 4 FEE 0WATEF, FLOW 5 VESSEL SlERM FLOW c 100.

o s'

W C

C m _L.A ,,

o e K_3 w c.0.

w w

w,-\,%- - 5 1

c- 'N x .

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O. 5. 10. IS. 20.

TIME ISEC) 1 LEVEL (INdM-REF-SEP-54I:.T 2 W R SENSED LEVELiINC-E5]

- 3 N R SENSEJ LEVELIIN:-E5:

150*

4 CORE INLET FL0n (FCT!

5-BTFASS STERM FLOWIFCT) 100.

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mu ' - '

50 y- c a 'N

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' 0.

O. 5. 10. 15. 20.

TIME (SEC1 Illinois

! Feedwater Controller Failure - Maximum Figure Power Company Demand, 70.2", Power, 53.6", Flow 15.B.3-1

! 15.B.3-11 .- .-

., 1 VESSEL PF ES RISE (PSI) 2 STM LINE FRES RISE (PSI) 3 TURBINE F RES RISE (PSI) 200* ,

4 CORE INLET SUB ISTU/LB)

S RELIEF VALVE FLOW (PCT)

. 6 TURS STEF M FLOW [ PCT)

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100. y.m

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0. J2 - -l"

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O. 5. 10. 15. 20.-

TIME (SEC) 1 VOID REAdilVITT 2 DOPPLER F EA:TIVITT 3 SCRAM RE;:TIVITY t-g RE: :mm

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0. -

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- f r

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O. 5. 10. 15. 20.

TIME ISEC)

Illinois Feedwater Controller Failure - Maximum Figure Power Company Demand, 70.2',' Power, 53.6?; Flow 15.B.3-1 Cont'd. i 15.B.3-12 _ _

F 1 NEUTRON F LUX 2 PERK FUEL CENTER TEMP 3 AVE SURFRCE HEAT FLUX 150. 4 FEEDWRIEF FLOW 5 VESSEL Sl ERM FLOW c: ,

gt Lt. 11 y

to 50.

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O. 2. y- 6. 8-TIME (SEC) 1LEVELtINh-REF-SEPTI 5I 2 W R SENSED LEVEll!N-S il 200- 3 N R SENSE 3 LEVELIIIjUE0l 4 CUEg INtE Ft0L1FLi1 5 DRIVE FLN1 treil 100.

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h. ,

s m -

- 100 .

O.~ ' ' ' ' 12 . y- 6. 6-TIME (SEC)

Illinois  ! Load Rejection llith Bypass Failure Figure Power Company' 70.2",, Power, 53.6% Flow 15 B.3-2 15.B.3-13

1 VESSEL PF ES RISE (PSI) 3 * -

2 STM LINE PRES RISE (PSI) 200* --

3 TURBINE FRES RISE (PSI) 3' ,IV"' ' / y gli 3 4 RELIEF VE LVE FLOW (PCT) ll '

y 5 CORE @VE,V0I .: 3(PCT) 100.

21' f3t>/ u

.rf S E 0* E 5 6 5 ys R

~

b

-100. ~ '

O. 2. 4. 6. 8.

TIME (SEC1 1VOIDRERdT!V*TT .

2 DOPPLER FEACTIVITT 3* -

i N 3 SCROM RE;:TIVITT 4 TOTFi REf CTivlTi

? -

9 O.

I% . -

2 D -1. \

D  : _

8 -

e  :

2. ...'....A -

O. 1. 2. 3. 4.

TIME (SEC)

Illinois  ! Load Rejection with Bypass Failure Figure 70.2*.' Power, 53.6"; Flow

• 15.B.3-2 PowerCompanyl Cont'd.

l 15.B.3-14

CPS 15.B.4 STABILITY ANALYSIS .

~

15.B.4.1 phenomena The primary contributing factors to the stability performance with one recirculation loop-not in service are the power / flow ratio and the recirculation loop characteristics. At forced circulation with one recirculation loop not in operation, the reactor core stability is influenced by the inactive recirculation loop. As core flow increases in SLO, the inactive loop forward flow decreases because the driving head decreases with increasing core flow. The reduced flow in the inactive loop reduces the resistance that the recirculation loops impose on reactor core flow perturbatio.ns thereby adding a destabifizing.

effect. At the same time the increased core flow results in a lower power / flow ratio which is a stabilizing effect. These two countering effects may result in decreased stability margin (higher. decay ratio) initially as core flow is increased (from minimum) in SLO and then an increase in stability margin (lower decay ratio) as core flow is increased further and reverse flow in the inactive loop is established.

As core flow is increased further during SLO and substantial reverse flow is established in the inactive loop an increase in jet pump flow, core flow and neutron noise is observed. A cross flow is. established in the annula'r downcomer region near the jet pump suction entrance caused s

by the reverse flow of the inactive recirculation loop. This cross flow interacts with the jet pump suction flow of the active recirculation loop and increases the jet pump flow noise. This effect increases the-

-total ~ core flow noise which tends to drive the neutron flux noise.

To determine if the increased noise was being caused by reduced stability margin as SLO core flow was increased, an evaluation was performed which phenomenologically accounts for single-loop operation effects on stability (Reference 15.8.8-3). The model predictions were initially compared to test data and showed very good agreement for both two-loop and single-loop test conditions. An evaluation was performed to determine the effect of reverse flow on stability during SLO. With

, IS.B.4-1

CPS increasing reverse flow, SLO exhibited slightly lower decay ratios than two-locp operation. However, at low core flow conditions with no reverse flow, SLO was slightly less stable. This is consistent with observed behavior in stability tests at operating BWRs (Reference 15.B.8-4).

In addition to the above analyses, the cross flow established during reverse ficw conditions was simulated analytically and shown to cause an increase in the individual and total jet pump flow noise, which is consistent with tests data (Reference 15.B.8-3). The results of these analyses and tests indicate that the stability ch'aracteristics are not significantly different from two-loop operation. At low core flows, SLO may be slightly less stable than two-loop operation but as core ficw is increased and reverse ficw is established the stability performance is similar. At higher core ficws with substantial reverse flow in the inactive recirculation locp, the effect of . cross ficw on the flow noise results in an increase in system noise (jet pump, core flow and neutron -

flux ncise), but core thermal-hydraulic stability margin is very high, similar to two-loop operation.

15.B.4.2 Ccmoliance to Stability Criteria Consistent with the philosophy applied to two-loop operation, the stabilitycomplianceduringsingle-lodpoperationisdemonstratedona generic basis. Stability acceptance criteria have been established to demonstrate compliance with the requirements set fceth in 10CFR50, Appendix A, General Design Criterion (GDC) 12 (Reference 15.B.8-5). The generic stability analysis has been performed covering all licensed GE BWR initial core and reload core fuel designs including those fuels contained in the General Electric Standard Application for Relcad Euel (GESTAR, Reference 15.B.8-6-)*. The analysis demonstrates that in the event limit cycle neutron flux oscillations occur wi' thin' the bounds of safety system intervention, specified acceptab'le fuel design limits are not exceeded.

t

. *ine reioad fuel designs contained in GESTAR include fuel designs through the GEBX8E design (including barrier fuel).

15.B.4-2 1

CPS Since the reactor core is assumed to be in an oscillatory mode, the question of stability margin during SLO is not relevant from a safety standpoint (i.e. the analysis already assumes no st' ability margin).

The fuel performance during limit cycle oscillations is characteristic-ally dependent on fuel design and certain fixed system features (high neutron flux scram setpoint, channel inlet orifice diameter, etc.).

Therefore the acceptability of GE fuel designs independent of plant and cycle parameters has been established. Only those parameters unique to SLO which affect fuel performance need to be evaluated. The major consideration of SLO is the increased Minimum Critical Power Ratio (MCPR) safety limit caused by increased uncertainties in sys_ tem parame-ters during SLO. However, the increase in MCPR safety limit (0.01) is well within the margin of the limit cycle analyses (Reference 15.B.8-5) and therefore it is demonstrated that stability compliance criteria are satisfied during single-loop operation. Operationally, the effects of higher flow noise and neutron flux noise observed at high SLO core flows are evaluated.to determine if acceptable vessel internal vibration levels are met and to determine the effects on fuel and channel fatigue.

However, these are not considered in the compliance to stability criteria.

A Service Information Letter-380, Revision 1 (Reference 15.8.8-7) has been developed to inform plant operators how to recognize and suppress unanticipated oscillations when encountered during plant operation.

As a result of the above analysis and operator recommendations, the NRC staff has approved the generic stability analysis for application to single-loop operation (Reference 15.B.8-8) provided that the recommenda-tions of SIL-380 have been incorporated into the Plant Technical Specifications.

l .

15.8.4-3 1

CPS 15.B.5 LOSS-OF-COOLANT ACCIDEilT AilALYSIS An analysis of single recirculation loop operation.using the models' and assumptions dccumented in Reference 15.B.8-9 was performed for CPS. I Using'this method, SAFE /REFLOOD computer code runs were~m'ade for a full spectrum of large break sizes for only the recirculation suction line breaks (most limiting for CPS). Because the reflood minus uncovery time for the single-loop analysis is similar to the two-loop analysis, the maximum average planar linear heat generation rate (MAPLHGR) curves were modified by derived reduction factors for use during one recirculation

. pump operation.

15.B.5.1 BREAK SPECTRUM ANALYSIS SAFE /REFLOOD calculations were performed using assumptions given in

-Section II.A.7.3.1 of Reference 15.8.8-9. Hot node uncovered time (time between uncovery and reflood) for single-loop operation is compared to that for two-loop operation in Figure 15.B.5-1.

The total uncovered time for two-loop operation is 168 seconds for the-100% DBA suction break. This is the most limiting break for two-loop

. operation. For single-loop operation, the total uncovered time is 169

. seconds for the 100% DBA suction break. This is the most limiting break for single-loop operation.

15.B.5.2 SINGLE LOOP MAPLHGR DETERMINATION The small differences in uncove_ red time and reficod time for the limit-ing break size would result in a small change in the calculated peak.

~

cladding temperature. Therefore, as noted as Reference 15.B.8-9, the one and two-lcop SAFE /REFLOOD results can be considered similar and the

,  : generic alternate procedure described in Section I:.A.7.4. of this

reference was used to calculate the M;APLHGR reduction factors for

, single-loop operation. The most limiting single-loop operation MAPLMGR

reduction factor (i.e., yielding the lowest MAPLHGR) for P8X3R fuel

! is 0.85. Single-loop operation MAPLHGR values are derived i

i 3

!- 15.B.5-1

.- -a _ - , - - . - - -

CPS by multiplying the current two-loop MAPLHGR values by the reduction factor (0.85). As discussed in Reference 15.B.8-9, single-loop MAPLHGR values are conservative when calculated in this manner. For reload situations, the MAPLHGR must be assessed for each cycle to determine if it is still applicable because the single-loop MAPLHGR multiplier was, based on the calculated peak. cladding temperature from the two-loop analysis for the-initial core fuel.

15.B.S.3 SMALL BREAK PEAK CLADDIflG TEMPERATURE Section II.A.7.4.4.2 of Reference i5.B.8-9 discusses the low sensitivity of the. calculated peak cladding temperature (PCT) to the assumptions used in the. single-loop operation analysis and the duration of nucleate boiling. As this slight increase (s 50 F) in PCT is overwhelmingly offset by the decreased MAPLHGR (equivalent to 300 F to 500 F PCT) for single-loop operation, the calculated PCT val'esu for small breaks will be well belcw the 1496*F small break PCT value previously reported for CPS, and significantly below the 2200 F 10CFR50.46 cladding temperature limit.

1 9

1 .

9 6

1' 15.B.5-2

- ,  : = .. - - - - - -

. ::0 TWO LOOPS 200 -- - --

- SIfiGLE LOOP 180 -

i

.=

.t C ~~., ,.-

5 isc - -,-

C

> s. g ,-

M  %

3 '

y ~..

~

2 sa0 - ~ . . . .

2 E 1.0 ft: BR E AK w

C C

e 1:0 -

w 5 7 m

C a.

= 100 -

=

9

=

M j a0 -

c C

m w '

E a so -

C 40 -

20 -

t 0

20 30 40 50 60 70 80 90 100 BREAK ARE A (% CF CBA) l Illinois Uncovered Time vs Break Area-Suction Break, Figure Power Company, LPCI Diesel Generator Failure 15.B.5-1

CPS 15.B.6 CONTAIfMEflT ANALYSIS . -

A single-loop operation containment analysis was performed for CPS. The peak drywell and containment pressure / temperature, peak suppression pool temper'ature, chugging loads, condensation oscillation and pool swell containment response were evaluated over the entire single-loop oper-ation power / flow region.

The analysis shcws that the peak drywell pressure during single-loop operation is 32.9 psia and occurs under recirculation line break at the maximum vessel subcooling condition in the power / flow map (52.6% power /

32.5% core ficw). This is below the drywell peak pressure of 33.6 psia

~

for the design basis accident main steam line break at the rated two-loop-operation reported in Chapter 6 of the FSAR.

The bounding event for the drywell temperature response is a double-ended break of a main steamline. The steam break flow is not affected by the increased vessel subcooling under SLO, but decreases due to the lower vessel pressure under SLO. It is concluded that the peak drywell temperature for SLO is bounded by that of the FSAR.

The peak containment pressure, containment temperature, and suppression potl temperature are longer term results than peak drywell pressure and are not affected by subcooled blowdown under SLO. These are governed mainly by the leng-term release of decay heat, emergency removal of the RHR service water, etc. Since the initial pcwer level is lower for SLO conditions compared to that of the FSAR, it is concluded that these parameters are bounded by those reported in the FSAR.

Finally, the chugging and pool swell loads evaluated at the maximum vessel subcooling power / flow condition during single-lcop operaticn are shown to be bounded by the peak values presented in the.FSAR. The corresponding condensation oscillation load increases slightly, but is adequately covered by existing load des.ign margin.

15.B.6-1

CPS 15.B.7 MISCELLAt:EOUS IMPACT E'lAlliATI0ft 15.B.7.1 -Anticioated Transient Without Scram (ATWS)

~

Imaact Evaluation 9 q

• The principal-difference between single-loop operation (SLO) and normal two-loop operation (TLO) affecting Anticipated Transient Without Scram

, (ATWS) performance is'that of initial reactor conditions. Since the SLO initial power / flow condition is less than the rated condition used for TLO ATHS analysis, the transient response is less severe and therefore

~

bounded by the TLO analyses.

It is concluded that if an-ATWS event were initiated at CPS from the SLO

~

i conditions, the results would be less severe than if it were initiated from rated conditions.

15.B.7.2 Fuel Mechanical Performance Evaluations were performed to determine the acceptability of CPS single-loop operation on P8X8R fuel rod and assembly thermal / mechanical performance. Component pressure differential and fuel rod overpower values were determined for anticipated operational occurrences initiated from SLO. conditions. These values were found to be bounded by those applied in the fuel rod and assembly design, bases.

It is observed that due to the substantial reverse flow established during SLO both the Average Power Range Monitor (APRM) noise and core j plate differential pressure noise are slightly incr, eased. An analysis has been carried out to determine that the APRM fluctuation should not.

exceed a flux amplitude of 15f. of rated and the core plate differential

' ~

4 pressure fluctuation should not exceed 3.2 psi peak ~to p*eak to be consistent with the fuel rod and assembly design bases.

15.B.7-1

CPS 15.B.7.3 Vessel Internal Vibration -

A recirculation pump drive flow limit is imposed for SLO. The highest drive flow that meets acceptable vessel internal vibration criteria is the drive flow limit for SLO. An assessment has been made for the expected reactor vibration level during SLO for CPS.

Before providing the results of the assessment, it is prudent to define the term " maximum flow" during balanced two-loop operation and single-loop operation. Maximum flow for two-loop operation is equal to rated volumetric core flow at normal reactor operating conditions. Maximum flow for single-loop operation is that flow obtained with the recircula-

~

tion pump drive flow equal to th.at required for maximum flow during two-loop balanced operation. For rated reactor water temperature and pressure, the maximum recirculation pump drive flow for CPS is about 33,000 gpm. This is a measured value from the Kuo Sheng I plant which is the prototype plant for Clinton.

Startup tests at the Kuo Sheng 1 plant showed all components, including the in-core guide tube during single-loop operation, to have vibration levels within acceptance limits. Since CPS is not a prototype plant, there is no reactor internal vibration monitoring program. Instead, the data from the Kuo Sheng 1 plant is used for CPS SLO assessment.

From the above, it can be inferred that the vibration levels of the reactor internal components for CPS would be expected to be within acceptance limits during single-loop operation with maximum flow as defined above.

O 15.8.7-2

CPS 15.B.8 REFERENCES -

-15.8.8-1 " General Electric BWR Thermal Analysis Ba' sis (GETAB); Data, Correlation, and Design Application", NED0-10958-A, January 1977.

15.8.8-2 " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors", NED0-24154, October 1978.

15.B.8-3 Letter, H. C. Pfefferlen ('GE) to C. O. Thomas (NRC), "Sub-mittal of Response to Stability Action Item from NRC Concern-ing Single-Loop Operation," September 1983.

15.B.8-4 S. F. Chen and R. O. Niemi, " Vermont Yankee Cycle 8 Stability and Recirculation Pump Trip Test Report", General Electric Company, August 1982 (NEDE-25445, Proprietary Information).

15.8.8-5 G. A. Watford, " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria",

General Electric Company, October 1984 (NEDE-22277-P-1, Proprietary Information).

15.8.8-6 " General Electric Standard Application for Reload Fuel",

General Electric Company, April 1983 (NEDE-24011-P-A-6).

15.B.8-7 "BWR Core Thermal Hydraulic Stability", General Electric Company, February 10,1984(ServiceInformationLetter-380, Revision 1).

15.B.8-8 Letter, C. O. Thomas (NRC) to H. C. Pfefferlen (GE), " Accept-ance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment 8,. Thermal Hydraulic Stability Amendment to GESTAR II," April 24, 1985.

15.B.8-1

CPS

' 15.B.8 REFERENCES-(Cont'd)

. 15.B.8-9 " General Electric Company Analytical Mode'l for Loss-of-Coolant

- Analysis in Accordance with .10CFR50 Appendix K Amendment No. 2

- One Recirculation Loop Out-of-Service"., NED0-20566-2 Revi-sion 1, July'1978, c

4 +

A G

i 15.B.8-2

. m 2.0 " SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow .

2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: -

WithTHERMALPOWERexceeding25%of[RATEDTHERMA.LPOWER'and.thereactorvessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, he in at least HOT SHUTOOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and_ comply with the requirements o.f.

Specification 6.7.1. . /tectua/Ah ~ .

THERMAL POWER, High Pressure and Hioh Flow & 9teY be b ud. l.07 Y g

2.1.2[TheMINIMUMCRITICALPOWERRATIO(MCPR)shallnotbelessthan1.06 with/the reactor vessel steam dome pressure greater than 785 psig,and core flow-greater than 10% of rated flow.

APPLICABILITY: ORERATIONAL CONDITIONS 1 and 2. - -

w GN T W  ?AL M ~ V OL hA4.

O ACTION: g7 g g WithMCPRlessthan1.06/andthereactorvesselsteamdomepressuregreater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE ,

2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig. ,

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION: .

With the reactor coolant system pressure, as measured in'the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant I system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with I the requirements of Specification 6.7.1.

l l REACTOR VESSEL WATER LEVEL 2.1. 4 The reactor vessel water level shall be above the top of the active irradiated fuel.

I CLINTON - UNIT 1 2-1

~

. . J I AllLE 2.2.1-1 -

l REACTOR PROIICTION SYSTEM INSTRUMENTATION SETPOINTS c

FUNCIIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE i 1. Intermediate Range Monitor

! c .

l 5. a. Neutron Flux-liigh < 120/125 divisions < 122/125 divisions l

Ef full scale 8f full scale l

b. Inoperative NA NA

2. Average Power Range Monitor:

a. Neutron Flux-liigh, Setdown < 15% of RATED '

< 20% of RATED TilERMAL POWER THERMAL POWER

b. Flow Biased Simulated Thermal Power-liigh (W-dW) . (V/-dW)

1) Flow Biased 1 0.66 8% with < 0.66 1%, with m a maximum of a maximum of a 2) liigh Flow Clamped , < 111.0% of RATED -

< 113.0% o' fRATED TilERMAL POWER TilERMAL POWER .

c. Neutron Flux-liigh < 118% of RATED < 120% of RATED TilERMAL POWER -

TilERMAL POWER

d. Inoperative NA . NA

3. Reactor Vessel Steam Dome Pressure - liigh 1 1065 psig , i 1080 psig w

4. Reactor Vessel Water Level - Low, Level 3 > 8.9 inches above > 8.3 inches instrument zero* above instrument zero Z

S. ReactorVesselWaterLevel-liigh,hevel8 < 52.0 inches ab6ve '

< 52.6 inches above a Instrument zero* Instrument zero -

.g 6. Main Steam Line Isolation Valve - Closure 1 8% closed i 12% closed

ha 7. Main Steam Line Radiation - liigh < 3.0 x full pow 6r 1 3.6 x full power

'd liackground background

.] .

. )

TABLE 2.2.1-1(Continued)

~

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS sz e FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE

8. Drywell Pressure - High 5 1.68 psig i 1.88 psig

e. 9. Scram Discharge Volume Water Level - High

~

a. Level Transmitter IC11-N601A $ 30 in.t 5 39.in.

IC11-N601B $ 30 in.t $ 39 in.

• IC11-N601C $ 30 in.tt $ 39 in.

IC11-N6010 $ 30 in.ft < 39 in. $

l b. Float Switch IC11-N013A < 762 ft. 1.375 in.'as'1 < 763 ft. 2 in. ms1 y IC11-N0138 $ 762 f't. 1.125 in.,ms1 $ 763 ft. 2 in, usi IC11-N013C $ 762 ft. 0.75 in. ms1 5 763 ft. 2 in. ms1 l IC11-N0130 $ 762 ft 1.125 in. ms1 5 763 ft. 2 in. ms1 l

10. Turbine Stop Valve - Closure 1 5% closed ', 1 7% closed

11. Turbine Control Valve Fast Closure, Valve > 530 psig '

g 465 psig Trip System 011 Pressure - Low NA NA -

12. Reactor Mode Switch Shutdown Position NA NA Y

= ,,

13. Manual Scram 2

NA NA _f ]

• See Bases Figure B 3/4 3-1. (a)7M. b 22Il d 6M% M [a4.[" h tInstr'ument zero is 759 ft. 11 in. ms1 a M (g i ttInstrument zero is 759 ft. 10.5 in. ms1 gg '

8 i

(% MWg Wcl u %&4Ay e f & 14 m

& at ,

m m cne aw =0 frt / M

• *' = 8 % k N --- ---QT D--

3

/tflh I E

2.1 SAFETY LIMITS

' f- AAr adI07f i pc S

y u

a .

y l 1

BASES [ -

~

2.

1.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary system piping are the  !

principal barriers to the release.of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than ~

1.03 44GPR greater-the 1.00 fepresentsf a conservative margin relative to th'e conditions required to maintain fuel cladding integriFty. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative -

freedom from perforations or cracking.. Although some corrosion or use-related

-cracking may occur'during the li'fe of 'the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perfora

• tion is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.

~, Therefore, the fuel cladding Safety Limit is defined with a margin to the con-ditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POW'ER, low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calcula-tions at pressures below 785 psig or core flows less than 10% of rated flow.

Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all ele'ation v head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr. Full-scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.

Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. ,

CLINTON - UNIT 1 8 2-1

.EN 17166

. m j BASES TABLE 8 2.1.2-1 .

UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLA00ING SAFETY LIMIT STANDARD DEVIATION QUANTITY (% of Point)

Feedwater Flow 1.76 Feedwater Temperature .

0.76 -

' ' ~

Reactor Pressure 0.5 Core Inlet Temperature 0.2 CoreTotgF , ,,,3 7g .2-5 s..A ..o ~

t,'

g.o Channel FTow Area 3.0 Friction Factor Multiplier 10.0 s

Channel' Friction. Factor 5.0 Multiplier TIPRehdgs 04 9

6.3 *:*

RFactok f 1. 5 -

Critical Power 3.6 Note: The uncertainty analysis used to establish the corewide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core.

CLINTON - UNIT 1 8 2-3 2 17 186

. A 3/4.2 POWER DISTRIBUTION LIMITS

,N 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3.A APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION: -

~

With an APLHGs exceeding the lim'its of Figure 3'.2.1-1, 3.2.'l-2, or 3.2.1-3, initiate corrective action within 15 minutes and restore APLHGR to within the required limits 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 2.5% of RATED THERNAL POWER within the amxtJ . hours. -

SURVEILLANCE REOUIREMENTS

~

4.2.1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3:

1. 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,

c. Initially and at least once per 12~ hours when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR, and

d. The provisions of Specification 4.0.4 are not applicable.

1Xe & -

3. 2. /-1, 3. 2.1 -- 2,

& 3.2.I 3 Ad$ de .ARAus ck

% a vdur LXar-Au Mj 0.es -Ak tirpg h.

lAd

.bry _f.

A k -

sw &

D

( CLINTON - UNIT 1 3/4 2-1 s ms 5 9 n ac,

w -

W -p ~

A M Ii j

- ceili .-

POWER DISTRIBUTION LifiITS )/fe & d ~

/t 3/4.2.2 APRM SETP0INTS d'

• s %:& e tiu @ tir 7 B & 2.2.I-!_,

~~ ____ _

- - ^

LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow-biased simulated thermal power-high scram trip setpoint (S) and flow-biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:

TRIP SETPOINT ALLOWABLE _VALUE 1 (0.6 _ $ $ (0.6 Q )T S

RB i (0.66 42%)T. S R8 I -(0. 66b5%)T where: S and S R8 areinpercentof'RATIDTHERMLP0WER, i W = Loop recirculation flow as a percentage of the loop recirculation

( ficw which produces-a.cated -core flow of 84.5 million ibs/hr.

T = The ratio of FRACTION OF RATED THERMAL POWER (FRTP) divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD). T is applied only if less than or equal to 1.0.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With the APRM flow-biased simulated thermal power-high scram trip setpoint and/or the flow biased neutron flux-upscale contro1 rod block trip setpoint less con-servative than the va.lue shown in the Allowable Value column for S or'SR8, as determined above, initiate corrective action within 15 minutes and adjust S and/or S

gg to be consistent with the Trip Setpoint value* within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 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 4.2.2 The FRTP and MFLPD shall be determined, tne value of T calculated, and the most recent actual APRM flow-biased simulated thermal power-high scram and flow-biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

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 />,

• With MFLPD greater than the FRTP rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 100% times MFLPO, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of.the adjustment is posted.on the reactor control panel.

CLINTON - UNIT 1 3/4 2-5 mm s s inna

s

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TABLE 3.3.6-2 T[lIS PAGE OPEN PENDING RECEIPT OF b CONTROL R0D BLOCK INSTRUMENTATION S' 1 STION ROM M AWN "i

@ TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE

1. R00 PATTERN CONTROL SYSTEM ,

y a. Low Power Setpoint (*)% of RATED TilERMAL POWER (*)% of RATED TiiERMAL POWER r b. RWL liigh Power Setpoint (*)% of RATED TilERMAL POWER (*)% of RATED TilERMAL POWER

2. APRM g @l) ,

a. Flow Biased Neutron Flux '

- Upscale < 0.66 + 42%** < d. 66 W' + 45%"*

b. Inoperative HA HA.

c. Downscale -> 5% of RATED TilERMAL POWER > 3% of RATED TilERMAL POWER

d. Heutron Flux - Upscale

~~

t Startup $ 12% of RATED Tl!ERMAL POWER 1 14% of RATED TilERMAL POWER w 3. SOURCE RANGE MONITORS

a. Detector not full in NA NA T b. Upscale < 1 x 105 cps < 1.6 x 105 cps

/' g - v.

$ CPS I *

c. Inoperative HA -

HA ll d. Downscale 13 cps ***ll e.") ,- jef{, g 31.8 cps ***k

{' / 9"' '

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in NA 4 NA

b. Upscale *

< 108/125 division of full scal'h < 110/125 division of full scale

c. Inoperative HA HA'

d. Downscale 1 5/125 division of full scale 1 3/125 division of full scale

5. SCRAM DISCliARGE VOLUME -

a. . Water Level-liigh 5 12" # $ lb" #

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW g

a. Upscale ,

$ 108% of rated flow $ 111% of rated flow g 7. REACTOR MODE SWITCil g a. Shutdown Mode NA NA

$ b. Refuel Mode Nr. NA

. l TABLE 3.3.6-2 (Continued)

CONTROL R00 BLOCK INSTRUMENTATION SETPOINTS i

TABLE NOTATIONS To be determined during startup test program. The actual setpoints are l the corresponding values of the turbine first stage pressure for these l power levels.

The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W). The trip setting of this function must be ma'intained in accordance.with Specification 6 3.2.2 2. 3 aM wcfL@)p Ntf6L2.2

• • • May be reduced to 0.7 cps provided signal'to noise ratio is > 2.0.

1. Instrument zero is 758' 5" ms1.

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y S 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS ,

LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:

1 .

a. Total core flow greater than or.equa to 45'" of rated core flow, or

> b. 7hERAAL POWER & On , , /}ind if - Fp-- *3.4.1.1 - 1, et-jr.]APRMorLPRMtnoiselevelsuitninthecperatingregie-monitoring is-requ M ot larger than three times their established baseline noise levels.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a. Wita one reactor coolant system recirculation loop not in operation : '

Tnmediately'iniBate actiori to reduce THERMAL-POWER to less than or equai g ,p to the' limit specified in Figure 3.4.1.1-1 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ,and' initiate mea'sures to place'the unit in at 1, east HOT _. SHUTDOWN _ wi. thin the_ next L A

s. __

'12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sj g gf-g h A -fQ w M &11 Ale ~CnA s mJL of

b. With no reactor coolant system recircu aDon loops irr ope on, immediately

. initiate action to recuce THERM SQWERht; less then-or-equa! to the '4-itA M peci' fed 'n Ficjure 3.,4.1.1-1 within'4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and initiate measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and i 'OT SHUT _0_0 within e next 6 h /t

c. With o reactor coolant' system recirculation loops in operation and total core flow less than 4.5% of rated core flow and THERMAL POWE3ghE=4 e.mo"-

52-the-44MtMpec4 tied-M } Figure 3.4.1.1-1, and with the APRM or LMIMt neutron flux noise levels greater than three times their established baseline noise levels, immediately initiate corrective action to restore the noise levels to within the required limits wpiin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ey increasing core flow to-greateHMan-45%-ot-rated-cor; fic or by reducing THERMAL POWER.to 1 :: th r-' 't+ 'imi-%;i fi:d ' n "igur: 3.'.1.1-1 o--

r

- -?

rasERY

1. \/cdur test Md N Mh T

S '

ggf na% . (M okuM&a R C .- j

' "See Soecial Test Exception 3.10.4

?0etector levels A and C of one LPRM string per core octant plus detectors A and C of one LORM string in the canter of the core should be monitorec.

?NeeSurveillanceRecuirement4.4.1.1.2.

Y R f 3/4 4-1 V A -

b CLINTON - UNIT 1

INSERT A:

-1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a) Place the recirculation flow control system in the Local Manual (Position Control) mode, and .

b) Reduce THERMAL POWER TO f 70% of RATED THERMAL POWER, and c) Increase the MINIMUM CRITICAL PO%ER RATIO (MCPR) Safety Limit by 0.01 to 1.07 per Specification 2.1.2, .

and . . .

d) Reduce the Maxim 6m kver' age PlEnar" Linear Heat Generation Rate (MAPLHGR). limit to a value of 0.85 times the two-recirculation-loop operation limit per

..Specificarion 3.2..I, and e) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block Trip Setpoints and Allowable Values to those-applicable for single-recirculation-loop operation per Specifications 2.2.1, 3.2.2, and 3.3.6, and f) Reduce the volumetric flow rate of the operating recirculation loop to n 33,000gpm**,

and g) Perform Surveillance Requirement 4.4.1.1.2 if thermal power is f 30%*** of RATED THERMAL POWER or the recirculation loop flow in the operating loop is-f50%*** of rated loop flow.

2. The provisions of Specification 3.0.4 are not applicable.

In

3. Otherwise, place the unit richin HOT 3 SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• • This value represents the design volumetric recirculation loop flow which produces 100% core flow at 100% THERMAL POWER. The actual value to be applied will be determined.

during the Startup Test Program.

psg

• • • Initial values. Final values be determined during Startup Testing based upon the threshold THERMAL POWER and ,

recirculation loop flow which sweep the cold water from the l vessel bottom head preventing [ stratification.

l l

1

~

11

INSERT B: -

d. With one or two reactor coolant recirculation loops in operation, and total core flow less than or equal to (39)j%,

and THERMAL POWER within the restricted zone of Figure 3.4.1.1-1, within 15 minutes initiate corrective action to reduce THERMAL POWER to within the unrestricted zone of Figuye 3.4.1.1-1, or increase core flow to greater than (39) % within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

1. Value to be established during Startup Test Program. (Core -

flow with both recirculation pumps at rated speed and minimum control valve position.') . . . .

e .

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

REACTOR COOLANT SYSTEM

^ j RECIRCULATION LOOPS 8 8 "" = h-a SURVEILLANCE REQUIREMENTS 4.4.1.1'4Each reactor coolant system recir::ulation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by:

a.

Verifying that the control valve fails "as is" on loss of hydraulic-pressure at the hydraulic control unit, and

b. Verifying that the average rate of control valve movement is: '

1.

2. Less than or equal to 11% of stroke per. second o'pening, and Less than or equal to 11% of stroke per second closing.

4.4.1.1.2 When total care flow-is-less-than45% sf rated fisw with t.u cevient A J2.systas-rectretrietion-loops-in-operat4en end THERMAL-POWER-944mit--specified-in Figure 3.4.1.1-1, establish a baseline PRMreater-than the L_

and LPRM*

" neutron flux noise value within 2 ours of entering thi operating region unless baselining has previously been performed in the gion since the last CORE ALTERATION, and

[a. Determine the APRM and LPRM* noise levels at least onca per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and f b.

Determine the APRM and LPRM* noise levels within 30 minutes after the

.> completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.

hMSERT

• y C.

witG~ W b (kl u 22 5 I

W:"*)

trow >

J

• Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

CLINTON - UNIT 1 3/4 4-2

• *JG 2 01985

INSERT C:

4.4.1.1.3 With one reactor system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that:

a. Reactor THERMAL POWER 4 70% of RATED THERMAL POWER.

'~

~

b. The recirculation flow control system is in the Local Manual (Position Control) mode,

c. The volumetric flow wate of the operating loop is- 4 33,000 gpm

, and

~ " '

d. Core flow is' greater-than (39)#% when THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1.

14.4.1.'14. "With one reactor coolant system recirculation ~1oop not in operation, within no more than 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperaturg*;equirements are met if IEERMAL POWER is 4 30% of RATED THERMAL POWER or the recircWlgtion loop flow in the operating loop is 4 50% of rated loop flow:

a. 4100*F between reactor vessel steam space coolant and bottom head drain line coolant,

b. 4 50*F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c. 4 50*F between the reactor coolant within the loop not in operation and the operating a.J d A W a n

"" U M w*'*-

Value to be established during Startup Test Prog am TC551_:1ort Q*

p flow with both recirculation pumps at rated speed and 3 minimum control valve position.)

This value represents the design volumetric recirculation loop flow which produces 100% core flow at 100% THERMAL POWER.

Initial values Final values to be determined during Startup Testin based upon the threshold THERMAL POWER and recirculation oop flow which will sweep the cold water from the vessel bottom head preventing stratification.

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, REACTOR CCOLANT SYSTEM JET PUMPS LIMITING CONDITION FdR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With one or more }et pumps inoperable, be in at le.ast ff0T SHUTD0'aN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. , , ,

~

SURVEILLANCE REOUIREMENTS

. 2. AR H,(lWir2- #t of ~he< above E'ach M .atrequired au wnskb<t jet pumpsOPERA 6LE a.s-shall be demons trated Ws

• * ,5 prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and at least once OPERABLE per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by determining recirculation loop flow, total core flow and diffuser-to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur when both recirculation loop flows are operating at the same flow control valve position.

/, A; The indicated recirculation loop flow differs by more thar 10% from the established flow control valve position-loop flow chzracteristics.

2. ,k The indicated total core flow differs by more than 10% from the estab-lished total core- flow value derived from recirculation. loop flow measurements. . -

3. p. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established patterns by more than 1C%.

NSE9i'-- -

CLINTCN - UNIT 1 3/4 4-4

^

INSERT D:

b. During single recirculation loop operation, each of the above required jet pumps shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that no two of the following conditions occur:

1. The indicated recirculation loop flow in the operaging loop differs by more than 10% from the established single recir~culation flow control valve position-loop flow characteristics.

2. .The indicated total gore flow differs by more than 10%

from the established total core'flo'w-value~ derived from single recirculation loop-flow measurements.

3. The indicated diffuser-to-lower-plenum differential pressure of,any individual jet pump differs from established single recirculation loop patterns by more than 10%.

c. The provisions of Specification 4.0.4 are not applicable provided that this surveillance is performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 25% of RATED THERMAL POWER.

To be determined during the startup test program.

t 81 c . .a. . . . . _ . _

qui 9 m qmme s-8 " '

s a ,'

e .

c:

y REACTOR COOL NT SYSTEM RECIRCULATION LOOP FLOW

' ~

LIMITING CONDITION FOR OPEIATION 3.4.1.3 Recircula loop flow mismatch I g maintained within:

a. 5% of rated [.rc#cu!:tiedflow with[ core flow

• greater than or equal to 70% of rated core flow.

cout

b. 10% of rated recirculatic(flow with c flowhessthan70%ofratec -

core flow.' ,

APPLICA8ILITY: OPERATIONALCONDITIONS1**and2h ~h AC. TION:

With recirculation loop flows different by more than the specified limits,

, either:

a. '

Restore the recirculation loop flows to within the specified limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />' or

^ Skufthe eclare dewa-anL recirculasion-icep

b. [and take the ACTION required by Specification 3.4.1.1.

k $ with the-tower iluw u v i. In Operativo a~

SURVEILLANCE REOUIREMENTS 4.4.1.3 Recirculation loop flow mismatch

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shall be verified to be within :ne limits at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 CONTROL RODS (Continued)

A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimu;m. The requirements for the various scram time measurements ensure that any indication of systematic pro-blems with rod drives will be investigated on a timely basis.

Damage within the control -rod. drive mechanism could be 'a-generic problem, there-fore with a control rod immovab1'e because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods. _

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are con-sistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor hut dow investigation and resolution of the problem. -

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The control rod system is designed to bring the reactor su Fcir at a rate' fast enough to prevent the MCPR from becoming less thanj uring the limit-ing power transient analyzed in Section 15.4 of the FSAR. This analysis .

shows that the negative reactivity rates resulting from the scram with the -

average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than MP The occurrence of scram times longer then those specified should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval isreducedinordertopreventoperationofthe(riactorforlongperiodsof time with a potentially serious problem. t g ,

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The scram discharge volume is required to be OPERABLE s'o that it will be avail-able when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inoperable and Specifi-cation 3.1.3.1 then appifes. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that 'a rod is properly coupled and there-fore this check must be performed prior to achieving criticality after completing CLINTON - UNIT 1 8 3/4 1-2 m $ e us

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. POWER DISTRIBUTION LIMITS BASES 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

b. Model Change

1. Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core. -

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2. Incorporate NRC ' pressure transfer assumption - The assubption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was, changed.

-A-few-ef the changes affect thelaccident' calculation irrespective of-CCFL.

These changes are listed below.

a. Inout Change

1. Break Areas - The DBA break area was calculated more accurately.

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1. Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for heatup calculation.

A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table 8 3.2.1-1. -

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3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on T ~

power distribution which would yield the design LHGR at RATED HERMAL POWER.

The flow biased simulated thermal power-high scram setting and he ficw biased neutr'on flux-upscale control rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than . r that > 1%

plastic strain does not occur in the degraded situation. -The scram and rod-block setpoints are adjusted in accordance with the formula in this specifica-tion when the combination of THERMAL POWER and MFLP0 indicates a peak power distribution to ensure than a LHGR transient would not be increased in degraded conditions. _- __

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SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS

• Plant Parameters: , ,

Core THERMAL POWER . . . . . . . . . . . . . . . . . . . . 3015 MWt** which corresponds to 105% of rated steam flow Vessel Steam Output ................... 13.08 x 108 lb,/hr which

, corresponds to 105% of-rated -

. steam flow -

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Vessel Steam Dome Pressure.'........'.... 1660 psia -

Design Basis Recirculation Line Break Area for:

a. Large Breaks 2.2 ft 2,

b. Small Breaks 0.09 ft 2, Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL -

FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft) FACTOR RATIO Initial Core 8x8 13.4 . 1. 4 1.17

• A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and Section 6.3 of the FSAR.

• • This power level meets the Appendix K requirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.

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POWER DISTRIBUTION LIMITS S

BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO Tb.e required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.06, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2;2. . . .

To assure that the fuel cladding integrity Safety- Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. .The-1-imiting-transient yields-the-largest-dal-ta-MCPR When-added to 2 -.

the-Safety-Limit MCPR of 1.06;-the-required-minimum-operating-limit-MCPR of -E. _

-5peef fication-372r3-is-obtained and presented-in-Figure-3rar3QThe power- - .

flow map of Figure B 3/4 2.3-1 gives operational limits. -

TEMA ad 15.6 5 - 1 The evaluation of a given transient begins with the system initial parameters shown in FSAR W 15.0-2 that are input to a GE-core dynamic behavior tran-sient computer program. The code used to evaluate pressurization events is described in NEDO-24154(3) and the program used in non pressurization events .. _

is described in NE00-10802(2) . The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic TASC cede described in NEDE-25149(4) The principal result of this evaluation is the4 reduction in MCPR caused by the transient.

The purpose of the MCPR f and MCPR p

of Figures 3.2.3-1 and 3.2.3-2 is to define.

operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power the required MCPR is the larger value of the MCPR and MCPR at the existing core flow and power state. The MCPRfs are f p established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured. .

The MCPR 7 s were calculated such that for the maximum core flow rate J.nd the corresponding THERMAL POWER along the 105% of rated st'eam flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPRs,were cal-culated at different points along the 105% of rated steam flow control line corresponding to different core flows. The calculated MCPR at a given point of core flow is defined as MCPRf .

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3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM hSY /

EfraTtun with bne reactor' core coolant recirculaf.fon loop inoperable isgo-E dibited'until an evaluation of-the performance of the ECCS'during one-leo pyera'tionhasbeen' performed,evaluatedanddetermined40beaccept'able An inoperable Jef. pump is not, in i,tself, a sufficient, reason to , declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown'are~a and reduce ^the' capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump. failure can be detected by monitoring jet pump per-formance on a prescribed schedule for significant degradation. Recirculation loco' flow mismatch limits are in compliance with ECCS LOCA analysis design

' criteri$ The limits will ensure an adequate core flow coastd wn from either ec ulation loop following a LOCA. D M'r Q. M M9 & % c- *

-In crdeFto prevent undue strets on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50 F of each other prior to startup of an idle loop. The loop temperature must also be within 50 F of the reactor pressure vessel coolant temperature to prevent thermal shock to the p circulation pump and recirculation nozzles. {$1nce the coolant J the bottom' of the vesseT is at a lower-temp _erature tha_n7ne coolantJn-the upper regions-2 Af'theiore, undue stress on the vejisel'would re,sult-if'the temperature' differ j genc'e was greatep-th'an 100*F.-l Q TNSERT G/

The objective of GE BWR plant and fuel design is to provide stable operation with margin over the normal operating domain. However, at the high. power / low flow corner of -the operatihg domain, a small probability of neutron flux limit cycle oscillations exists depending on combinations of . operating conditions (e.g., rod pattern, power shape). To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flux noise levels should be monitored while operating in this region.

Stability tests at operating BWRs were reviewed to determine a generic region .

of the power / flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio of 0.6 was chosen as the bases for determining the generic region for surveillance to account for the. plant to plant variability of decay ratio with core and fuel designs. This generic region has been determined to correspond to a core flow of less than or equal to 45% of rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1.

Plant specific calculations can be performed to determine an applicable region for monitoring neutron flux noise levels. In this case the degree of conserva-tism can be reducad since plant to plant variability would be eliminated. In this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.

CLINTON - UNIT 1 8 3/4 4-1

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The impact of single recirculation-loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpointo are adjusted as 'noted in Tables -

2.2.1-1 and 3.3.6-2, respectively, MAPLHGR limits are .  ;

decreased by the factor given in Specification 3.2.1, and

. MCPR operating limits are adjusted per Section 3/4.2.3.

Additionally, surveillance on the volumetric flow rate of -

the operating recirculation loop is imposed to exclude the possibility of e:Itcessive. core internals vibration. The

surveillance on differgntial temperatures below (30%)

THERMAL POWER or (50%) rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation-pump ,+and vessel bottom' head during'the extended operation of the single recirculation loop mode.

Initial Values. Final values to be determined during Startup Testing based on the threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.-

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In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.

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Sudden equilization of a temperature difference > 100*F between_the, reactor vessel _ bottom. head. coolant.and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the ,

reactor vessel bottom head.

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