Text

Proposed Changes to Tech Specs & Justifications to Allow Single Loop Operation
	ML20151P252
Person / Time
Site:	Fermi
Issue date:	08/04/1988
From:	; DETROIT EDISON CO.
To:	;
Shared Package
ML20151P196	List: ML20151P193; ML20151P233; NRC-87-0132, Proposed Changes to Tech Specs & Justifications to Allow Single Loop Operation;
References
	CON-NRC-87-0132, CON-NRC-87-132 NUDOCS 8808090279
	Download: ML20151P252 (38)
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l Enclosure 3 Technical specific 8tiOD8 PaBe ChanBe8 with Justification l

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l 8808090279 880804 PDR ADOCK 05000341 PDC P

Encloruro 3 to NRC-87-0132 Page 1 A description of the Technical Specification changes to allow Fermi 2 single loop operation and their respective justification and reasons are given below:

1. Page 2-1, Specification 2.1.2 Increase in MCPR Safety Limit from 1.06 to 1.07 for Single Loop Operation (SLO) to account for additional uncertainty in establishing Safety Limits. With one recirculation loop only in operation, there is an increase in core flow measurement uncertainty. Additional uncertainty is experienced also with TIP measurements.

2. Page 2-3 Specification 2.2.1 Added footnote to specify time allowance to co:: ply with SLO requirements consistent with Specification 3 4.1.1.

3 Page 2-4, Table 2.2.1-1 Change of Functional Unit 2.b setpoint to include SLO.

Provision is included to adjust APRM gains in lieu of changing setpoints. This provides an expeditious method of complying with single loop requirements.

4. Page B2-1, Bases 2.0 Bases correction consistent with Specification 2.1.2.

5. Page B2-3, Bases Table B2.1.2-1 Bases correction consistent with Specification 2.1.2.

6. Page B 2-7, Bases 2.2.1 Bases addition to include 4W value for SLO and explanation of the non-applicability of the High Flow Clamped Flow Biased Neutron Flux-High setpoint in SLO.

7 Page 3/4 2-1, Specification 3.2.1 A MAPLHGR multiplier is provided for SLO. This multiplier is used prir.arily to allow for the marginal difference in uncovered time and reflood time associated with a LOCA during SLO.

Enclosurs 3 to NRC-87-0132 Page 2

8. Page 3/4 2-5, specification 3/4.3.2 Change of APRM Scram and Rod Block setpoint to include SLO.

Footnote # modified and # added to include provisions for APRM gain adjustment as discussed above.

9. Page 3/4 3-41 Specification 3 3.6 Added footnote to specify time allowance to comply with SLO requirements consistent with Specification 3 4.1.1.

10. Page 3/4 3-44, Table 3 3.6-2 Change of Rod Block Monitor and APRM Rod Block setpoint to include SLO. Provision for adjusting APRM gains included.

11. New pages 3/4 3-90, 3/4 3-91 and 3/4 3-92; New Specification

,3 3 10.

Consolidated stability related requirements in single specification. Additional requirements for single loop operation are included.

12. Page 3/4 4-1 and Two Additional Pages, Specification 3.4.1.1 Included actions for SLO including stratification limits.

Stability requirements moved to new Specification 3 3 10.

l 13 Page 3/4 4-2 Specification 4.4.1.1.1, 4.4.1.1.2, 4.4.1.1 3, 4.4.1.1.4.

Additional surveillance requirements added to reflect limits imposed for SLO, and to reflect stratification limits.

Stability surveillances moved to new Specification 3 3 10.

l 14. Page 3/4 4-4 Specification 4.4.1.2 Change of surveillance requirement 4.4.1.2 to reflect SLO.

Added Specification 4.0.4 allowance to overcome difficulty in perforning this surveillance below 25% THERMAL POWER.

Enclosura 3 to NRC-87-0132 Page 3

15. Page 3/4 4-5 Specification 3.4.1 3

a. Change APPLICABILITY to allow SLO.

b. Change to ACTION b. to improve flexibility Jn operation.

c. Inclusion of ACTION to reach HOT SHUTDOWN.

16. Page B 3/4 1-2, Bases 3/4.1.S Corrections made to properly reflect safety limit which is 1.07 for SLO.

17 Pages B 3/4 2-1 Bases 3/4 2.1 Bases of reduced MAPLHGR limits for SLO included.

18. Page B 3/4 2-2, Bases 3/4.2.2 Lc cection made to paragraph 3/4.2.2 to properly reflect the safety limit which is 1.07 for SLO.

19 Page B 3/4 2-3 Bases Table B 3 2.1-1 Footnote added to indicate conservat!sm introduced for SLO.

20. Page B 3/4 2-4 Bases 3/4.2 3

a. Reference to 1.06 removed to properly reflect the safety limit which is 1.07 for SLO.

b. Deletion of last two senter: es of second paragraph to properly reflect safety limit of 1.07 and remove misleading statement on determination of MCPR operating limit.

21. Pages B 3/4 3-8, Bases 3/4.3.10 Add Bases for new Specification 3/4.3 10.

22. Page B 3/4 4-1, Bases 3/4.4.1

a. Deletion of first sentence - no longer applicable.

Replace with "Insert G," justifying and describing new conditions for SLO. Also provides reasons for coolant stratification and vibration limits.

Enclosure 3 to NRC-87-0132 Page 4

b. Extra words added to third pa"agraph plus "Insert H."

Included to amplify reason for flow mismatch requirement.

c. Last sentence of fourth paragraph deleted and "In.9ert I" added to further describe stratification limit and reason for concern.

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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:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor ~

vessel steam dome pressure less than 785 psig or core flow less than 10%

..of_r.ated flow, be in at least HOT SHUTOOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - and comply with the requirements of Specification 6.7.1.ff e two r e c u c u lation re e. fr,o n amt s ha tt no t b e. le n % /10 7 fo r loop op'yle

.s i n~ 8 N U0 '1 THERMAL POWER, High Pressure and High Flow '

2.1.2 TheMINIMUMCRITICALPOWERRATIO(MCPR)shallnotbeles with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

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APPLICABILITY: OP TIONAL CONDITIONS 1 arl d l --

1~or ho recir c ulation lo op of err

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M*n I' W ACTION: fo r ,A /e /oc g n With MCPR less than 1.06 an(( e reactcr vessel steam dome pressure greater than 785 psig and core flow greater then 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply witn 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 Sr;.JTDOWN with reactor coolant system pressure less than or equal to 3325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and. comply with the requirements of Specification 6.7.1.

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2-1 FERMI - UNIT 2

SAFETY LIM) 6 AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1.

APPLICABILITY: As shown in Table 3.5.1-1.

ACTION:

4 With a reJctor protection system instrementation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to GPERABLE status with -

- -4+c sat.nnint adiusted consistent with the Trip Setpoint value.

+ The APRM Flow bin,e.( ow trwenta fron neeg ,,,+ j,e gag ,,g incpeMIt opo1 entee!g sfyle re c;,c u ja y,,, gcp n,,,4,,

D'ECC' l$1t- 66fjooll1f.$ Qto acGusfec/ WEf/tfit y }ou ry pen Specific.affon 3, y , l . I ,

iiRP,1 - LIN]l ? ?- 3 l

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TABLE 2.2.1 1 REACTOR PROTECTION SYSIEM INSTRUPTNTATION SETPOINTS A

E

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. ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES

.E_. 1. Intermediate Range Monitor, Neutron Flux-High i 120/125 divisions of i 122/125 divisions

-d full scale of full scale N

2. Average Power Range Monitor:

a. Neutron Flux-Upscale, Setdown i 15% of RATED THERMAL POWER 1 20% of RATED THERMAL POWER

b. Flow Biased Simulated Th'ersal Power-Upscale g9y

/ D M +ased__ i 1 0.66 W+51%, with c,.

5 0 fe_i"-M%,- wTIli' g -

u m - a maximum of n,g g 2) High T1cw Clamped i 11.5.5% 6 Trn i 115.5% of RATED THERMAL POWER i;;C"."'. DnWf&_

c. Fixed Neutron Flux-Upscale i 118% of RATED 1 120% of RfTED THERMAL POWER THERMAL POWER

d. Inoperative N.A. N.A.

. 3. Reactor Vessel Steam Dome Pressure - High 1 1068 psig i 1088 psig ,

4. Reactor @ ssel low Water Level - Level 3 > 173.4 inches *

> 171.9 inches

5. Main Steam Line Isolation Valve - Closure i 8% closed i 12% closed

6. Main Steam Line Radiation - High 1 3.0 x full power background i 3.6 x full power background

7. Drywell Pressure - High 1 1.68 psig i 1.88 psig

8. Scram Discharge Volume Water level - High

a. Float Switch < 594'8" < 596'0"

b. Level Transmitter [592'6" 5596'0"

9. Turbine Stop Valve - Closure 1 5% closed 1 7% closed

10. Turbine Control Valve Fast Closure Initiation of fast closure M.A.

11. Reactor Mode Switch Shutdown Position N.A. N.A.

12. Manual Scram N.A. N.A.

13. Backup Manual Scram N.A. N.A.

See Bases Figure B 3/4 3-1.

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Insert A TRIP SETPOINT ALLOWABLE VALUE

1) During two recirculation loop operation:

a. Flow Biased 10.66 W+51%, with 50.66W+54%,with a maximum of a maximum of

b. High Flow Clamped 5113 5% RATED THERMAL $115.5% RATED THERMAL POWER POWE3

2) During single recirculation loop operation:

a. Flow Biased 10.66W+45.7%,** 10.66W+48.7%,**

b. High Flow Clamped NA NA

- During single recirculation loop aperation, rather than cdjusting the APRM Flow Biased Serpoints to comp'.y with the single loop values, the gain of the APRMs may be adjusted for a period not to exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .3cch that the final APRM readings are at least 5 3% of rated power greatar than 100%

times FRTP, provided that the adjusted APRM readings do not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

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2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

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The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the &*h 4

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 d'2 yy o

to occur if the limit is not violated. Because fuel damage is not directly e observable, a step-back approach is used to establish a Safety Limit such that s E-l the MCPR is not less than 1.06.+, MCPR areater than 1.06# represents a conser- 4

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vative margin relative to the conditions required to maintain fuel cladding ! #p integrity. The fuel cladding is one of t,he physical barriers which separate I p2 i a

the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. ['y ob o

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however,

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can result from thermal stresses which occur from reactor operation signifi-cantly above design conditions and the Limiting Safety System Settings. While '

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fission product migration from cladding perforation is just as measurable as io3 that from use related cracking, the thermally caused cladding perforations h sig al a threshold beyond which still greater thermal stresses may cause gross Q {g, rather than incremental cladding deterioration. Therefore, the fuel cladding t, j

l Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a signi- -#.

E' ficant departure from the condition intended by design for planned operation. 1

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- 2.1.1 THERMAL POWER. Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calculations et 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 elevation 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 f rom 14.7 psia to 800 psia indicate that the fuel assembly criti-cal 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.

FERMI - UNIT 2 B 2-1

1 Bases Table 82.1.2-1 1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT

Standard Deviation Quantity (% of Point)

Feedwater Flow 1.76 Feedwater Temperature 0.76 Reactor Pressure 0.5

. Core Inlet Temperature 0.2 Core Total Flow G4 Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel Friction Factor Multiplier 5.0 TIP Readings 6-F s

R Factor 1.5 Critical Power 3.6 Two recteeu latio,, / cp open+;o n 2E Si9)e recheulcJ:an foep ey>Ha h n do O Two r e co'se ule. Hen loop opeeatton G3 Siny lt rec kc u le. Ho.1 loep operaflo n 4S

The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core.

7he vako heeein apply fu fofA ko rec &e u is tioas lo op operat on c<w/ stste. re che u la+to,1 /oop opera f,'o n , exee,ot ca riotec/.

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FERH1 - UNIT 2 B 2-3

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LIMITING SAFETY SYSTEM SETTINGS m BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)-

2,verage Power Ranae Monitor (Continued)

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assura shutdown before the power could exceed the Safety Limit.

The 15% neutron flux trip remains active until the mode switch is placed in ~

the Run position.

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The APRM trip system is calibrated using heat balance data taken during steady. state conditions. Fission chambers provide the basic input to the 2 e. system and.therefore the monitors respond directly and quick 1v to changes due to transient operation for the case of the Fixed Neutron Flux-Upscale setpoint; i.e, for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel. For the Flow Biased Neutron Flux-High setpoint, a time constant of 6 i 1 seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown. The flow referenced trip setpoint must be adjusted by the

! specified formula in Specification 3.2.2 in order to maintain these margins when MFLPD is greater than or equal to FRTP. Adcl 7m e r f.

3. Reactor Vessel Steam Dome Pressure-High .

High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure ir. crease while operating wi?1 also tend to increase the power of the reactor by compressing voids thus addirg reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design t

pressure and takes into account the location of the pressure measurement compared

( to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure trip is bypassed. For a turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

k FERMI - UNIT 2 8 2-7

INSERT PAGE B 2-7 For single recirculation loop operation, the reduced APRM setpoints are based on a A W value of 8%. The ,oW value corrects for the difference in indicated drive flow (in percentage of drive flow which produces rated core flow) between two loop and single loop operation of the same core flow. The 5 3% decrease in setpoint is derived from 0.66 x 8%. The High Flow Clamped Flow Biased Neutron Flux-High setpoint is not applicable to single loop operation as core power levels which would require this limit are not achievable in a single loop configuration.

1 3/4.2 POWER DISTRIBUTION LIMITS 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 dw/9 6. r e e n c u /, /, . . /o Jha/

or /&

operation . 7A t h m ity of Fig u e ej s . 2,1 -1, a .2 e / -2 i eo<c/ 1.2o/-2 APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1 1, 3.2.1-2, 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

be reduceef 1'o e valu e o f 0. 90 M es the fts. re e k e u fa to'o n loop operation h% H When k sky /c fe c&c o(c ficat lo ojo ope ra h oo< .

SURVEILLANCE REQUIREMENTS 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:

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

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

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

d. The provisions of Specification 4.0.4 are not applicable.

FERMI - UNIT 2 3/4 2-1

( POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased neutron flux-high scram trip setpoint (S) and flow biased neutron flux-high control rod block trip setpoint (Sgg) shall be established according to the following relationships: ,

TRIP SETPOINT ALLOWABLE VALUE S < (0. ~ ^ unT e W .66W T E4%)T R'8I " * "ON L 5"1 B S y Y G ): 5fi(0.'~^"0T where: S and Sag are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million Ibs/hr, at 100% of RATED THERMAL POWER T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY. T 1s -

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 neutron flux-high scram trip setpoint and/or the A flow biased neutron flux-high control rod block trip setpoint less conservative AJ than the value shown in the Allowable Value column for S or SRB, as above determined, initiate corrective action within 15 miputes and adjust 5 and/or S

RB to be consistent with the Trip Setpoint value*'Trithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or reduce ll THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.2 The FRTP and the MFLPD for each class of fuel shall be deternined, the

, value of T calculated, and the sost recent actual APRM flow biased neutron y flux-high scram and flow biased neutron flux-high control rod block trip

' setpoints verified to be within the at,0ve limits or Adjusted or.the APRM gain readirigs shall be verified as indicated below,**hs required: l

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

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at j

least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with MFLP0 greater than or equal to FRTP.

d. The provisions of Specification 4.0.4 are not applicable, fep/au, Mtith-NELEILgreater than the FRTP during power ascension up t WK THERMAL POWE , adjusting the APRM er than or e ual to 100%times

, APRM gainNFLPD, may be fooN te, adjusted such that APRM rea n

(gg provided that t .M reading does 3JtWEb end T notice of adjustment is posted on the reactor con P

00% of RATED THERMAL l FERMI - UNIT 2 3/4 2-5 Amendment No. 9

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

_.___,_,,....,-...._mm

Insert B

1. During two recirculation loop operation:

S<(0.66W+51%)T S<(0.66W+54%)T SjB5(0.66W+42%)T SjB5(0.66W+45%)T

2. During single recirculation loop operation:

S<(0.66W+45.7%)T S<(0.66W+48.7%)T SjB5(0.66W+36.7%)T Sig5(0.66W+39.7%)T Footnote Insert Page 3/4 2-5

With MFLPD greater than the FRTP during power ascension up to 90%

of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100% times MFLPD provided that the adjusted APRM readings do not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactsa control panel. With MFLPD greater than FRTP and a single recirculation loop in operation, if the APRM flow biased setpoints have not been adjusted to their single loop values then the minimum required APRM reading must be increased by an additiont.15 3% of rated power.

During single recirculation loop operation with FRTP-greater than or equal to MFLPD, rather than adjusting the APRM setpoints to comply with the single loop values, the APRM gain may be adjusted for a period not to exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months such that the final APRM readings are at least 5 3% of rated power greater than 100% times FRTP, provided that the adjusted APRM readings do not exceed 100%

RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

5

. . . . ..-. -..--....a..-

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INSTRUMENTATION 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6. The control rod block instrumentation channels shown in Table 3.3.6-1 ihall be OPERABLE with their trip setpoints set consistent wii.h the values snown in the Trip Setpoint column of Table 3.3.6-2.

APPLICABILITY: As shown in Table 3.3.6-1.

ACTION:

M

a. With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is .

restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trjp Function requirement, take the ACTION required by Table 3.3.6-1.

SURVEILLANCE REQUIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentation cha'nnels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS ana at tne frequencies snown in Table 4.3.6-1.

Y 7he MAAA Flow Blusec/ Hea1co n Flur - H(9h anel I?od Bleek M dier in s tru m e nta H on necol not he c/e cla te c/

lnop erable. upoor, enfee,%3 s tag le re acto e r e c o'ec.u lo ko n loop operation pmWed the sefpoinh, ate cloftunec/

Within h ou t.s y er 6peetWc.c<fton B. % h / ,

FERH1 - UNil 2 3/4 3-41

9 TABLE 3..s.6-2 -

CONTROL'R00 BLOCK INSTRUMrNTATION SETPOINTS m TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE_

]1. R0D BLOCK MONITOR h a d/4 s . Up;c:10 ' O. SS U ^ '"*' ' " . 00 " '% J,uerf c l c- b. Inoperative RA NA -

(

m 2.

c.

APRM Downscale 1 5% of RATED THERMAL POWER 1 3% of RATED THERMAL POWER

'"""" '" ' " ~ " ""

c. cle.: Bia::d Mcutrar. Thx - High r l

b. Inoperative NA NA :

c. Dcwnscale > 5% of RATED THERMAL. POWER > 3% of RATED THERMAL POWER

d. Neutron Flux - Upscale, Setdown 312%ofRATEDTHERMALPOWER 314%ofRATEDTHERMALPOWER

3. SOURCE RANGE MONITORS

a. Detector not full in NA NA ,

b. Upscale < 1.0 x 105 cps '.

< 1.6 x 105 cps i

c. Inoperative NA NA j-

d. Downscale 1 3 cps ** 1 2 cps ** '-

D 4. INTERMEDIATE RANGE MONITORS .

Y a. Detector not full in NA NA !

$ b. Upscale < 108/125 divisions of < 110/125 divisions of Tull scale Tull scale i c. Inoperative NA NA

d. Downscale > 5/125 divisions of > 3/125 divisions of Tull scale Tuli scale

5. SCRAM DISCHARGE VOLUME

a. Water Level-High < 589'11\" < 591'0"

b. Scram Trip Bypass HA NA

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW 3I a. Upscale < 108/125% of rated flow < 111/125% of rated flow

$ b. Inoperative . HA NA l g c. Comparator i 10% flow deviation i 11% flow deviation S 7.

REACTOR MODE SWITCH SHUTDOWN POSITION NA NA W

The APRM rod block function is varied as a function of recirculation loop drive flow (W). The trip setting l l h of this function must be maintained in accordance with Specification 3.2.2.

l * **The downscale rodblock setpoint count rate may be re-fuced' to 0.3 cps prior to achieving a burnup of 2000 MWD /T on the first core provided the signal-to-noise ratio is >2. After a burnup of 2000 MWD /T cn the first core, the count rate may be reduced te 0.7 cps provide'd the signal-to-noise ratio is >2.

S 9 1 -

l i

l l

l Insert C:

a. Upscale

1) During two recirculation 10.66W+40% 50.66W+43%

loop operation 8 #

2) During single recirculation 10.66W+34.7% 50.66W+37 7%

loop operation Insert D:

a. Flow Biased Neutron Flux-High

1) During two recirculation 10.66W+42%* 1066W+45%*

loop operation

2) During single recirculation 50.66W+36.7%#* $0.66W+39 7%#*

loop operation

During single recirculation loop operation, rather than adjusting the APRM and RBM Flow Biased Setpoints to comply with the single loop values, the gain of the APRMs may be adjuste j for a period not to exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months such that the final APRM readings are at least 5 3% of rated power greater than 100% times FRTP, provided that the adjusted APRM readings do not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

INSTRUMENTATION 3/4 3 10 NEUTnun FLUI :!0NITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3 3 10 The APRM and LPRM* neutron flux noise levels shall not exceed three (3) times their established baseline value.

APELICABILITY: OPERATIONAL CONDITION 1 with THERMAL POWER greater than the limit specified in Figure 3 3101 and total core flow less than 45% of rated total core flow.

ACTION: With the APRM or LPRM# neutron flux noise level greater than three (3) times their established baseline noise levels, immediately initiate corrective action to restore the noise levels to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than or equal to the limit specified in Figuro 3 3 10-1 within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

SURVEILLANCE REQUIREMENTS 4 3 10.1 The provisions of Specification 4.0.4 are not applicable.

4.3 10.2 With two reactor coolant system recirculation loops in operation, establish a baseline APRM and LPRM* neutron flux noise level value within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months upon entering the APPLICABLE OPERATIONAL CONDITION of Specification 3 310 provided that baselining has not been performed since the most recent CORE ALTERATION.

4 3 10 3 With one reactor coolant system recirculation loop not in operation, establish a baseline APRM and LPRM* neutron flux noise level value with THERMAL POWER less than or equal to the limit specified in Figure 3 3 10-1 prior to entering the APPLICABLE OPERATIONAL CONDITION of Specification 3 310 provided baselining has not been performed with one reactor coolant system recirculation loop not in operation since the most recent CORE ALTERATION.i

The baseline data obtained in Specification 4.3 10 3 is applicable to operation with one reactor coolant system recirculation loop not in operation and THERMAL POWER greater than the limits specified in Figure 3 3 10-1.

l # Detector levels A and C of one LPRM string per core octant plus detector

! levels A and C of one LPRM string in the center of the core should be

! monitored.

FERMI - UNIT 2 3/4 3-90 l

l L

SURVEILLANCE REQUIREMENTS ,

1 4 3 10.4 The APRM ara LPRM* neutron flux noise levels shall be determined to be less than or equal to the limit of Specification 3 310 when operating within the APPLICABLE OPERATIONAL CONDITION of Specification 3 3 10:

1

a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and l l

b. Within 30 minutes after completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.

t l

l # Detector levels A and C of one LPRM string per core octent plus detector levels A and C of one LPRP string in the center of the core should be monitored.

l f

I l

l FERMI UNIT 2 3/4 3-91

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N N=- 0 0 0 C

0 0 0 2 1 7 6 5 4 3 5 ayoA: a3=wzr weoo 5w g :oI*-

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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.

+ e +5 '

+ iI I i , b ih APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*,

ACTION'-

aMh_ohe feole.a reactor with coolant Zase f E.recirculation loop no system on, immediate y i rri& + a action to reduce T to less than or equal tn the limit specif T. u-a L 4.1.1-1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and initiate me ace the unit in at least, nGT !""TnnwN within s.

b. With no reactor coolant system recirculation loops in operation, immediately initiate action to reduce THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and initiate measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With two reactor coolant system recirculation loops in operation a D otal core flow less than 45% of rated core flow and THERMAL P gr er than the limit specified in Figure 3.4.1.1 the APRM and LPRMa* noise levels (Surv 4. 4.1.1. 3 ) :

1. Mon With hours of entry into this co n and at least a) 'in this condition and, once per hours thereafter wh Within 30 minut fter t completion of a THERMAL POWER b) i increase of at leas 'of RATED THERM % POWER in an hour l

by control rod so ent.

I RM** neutron flu oise levels greater than

2. With the APRM o e levels, imediately three times eir established baseline initiat orrective action to restore the no levels to within quired limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing re flowL POWER to 1

the l eater than 45% of rated core flow or by reducing to less than or equal to the limit specified in Figure 3. .1-1.

See Special Test Exception 3.10.4.

- Det+ctor levels A and C of one LPRM string per core octant plus detect ored when and C of one'LPRH-string in the center of the core s Only the center of the rn.

operating with a nonsymetrTiln onand C and two outer LMt!Latring detectors A and core LPRM strin n tored for operstions with a symetric cont J

4 1

FERMI - UNIT 2 1/4 4-1 f

l

Insart E Page 1

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

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

a) Place the individual recirculation pump flow controller for the operating recirculation pump in the Manual mode.

b) Reduce THERMAL POWER to less than or equal to 70%

of RATED THERMAL POWER.

c) Limit the speed of the operating recirculation pump to less than or equal to 75% of rated pump speed.

d) Increase the MINIMUM CRITICAL POWER RATIO (MCPR)

Safety Limit by 0.01 to 1.07 per Specificat, ion 2.1.2.

e) Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit to s value of 0 90 times the two recirculation loop operation limit per Specification 3 2.1.

f) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to thosg applicable for single recirculation loop operation per Specifications 2.2.1, 3 2.2, and 3 3 6.

l l g) With one reactor coolant system recirculation loop not in operation and THERMAL POWER greater than the limit specified Figure 3 4.1.1-1 and core flow less than 39%# of rated core flow, immediately initiate action to, within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , r2 duce THERMAL POWER to less than or equal to the limit specified in Figure 3 4.1.1-1 or increase core flow to greater than or equal to 39%f> of rated core flow.

APRM gain adjustments may be made in lieu of adjusting the APRM and RBM Flow Biased S9tpoints to comply with the single loop values for a period of up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Value to be determined during Startup Test Program (Core flow with both racirculation pumps at minimum speed). Final value to be provided within 90 days of completion of Startup Test Program.

Insart E Page 2 h) The provisions of Specification 4 3 10 3 must be satisfied unless THERMAL POWER is less than or equal to tne licit specified in Figure 3 4.1.1-1 or total core flow is greater than or equal to 45% of rated core flow. With one reactor coolant system recirculation loop not in operation and with THERMAL POWER greater than the limit specified in Figure 3.4.1.1-1, and total core flow less than 45%

of rated core flow, and the provisions of Specification 4.3 10 3 having not been satisfied, immediately initiate action to, within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , reduce THERHAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1 or increase total core flow to greater than or equal to 45% of rated core flow.

1) Perform Surveillance Requirement 4.4.1.1.4 if THERMAL POWER is less than or equal to 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is less than or equal to 50% of rated loop flow.

2. The provisions of Specification 3 0.4 are not applicable.

3 Otherwise, in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

V REACTOR COOLANT SYSTEM i

y SURVEILLANCE REQUIREMENTS 4.4.1.1.1 Each pump discharge valve shall be demonstrated OPErw by cycling each valve through at least one complete cycle of full travel du !!g each STARTUP* prior to THERMAL POWER exceeding 25% of RATED THERMAL e #2R.

4.4.1.1.2 Each pump MG set scoop tube mechanical and ele::,'ical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 105% and 102.5%, respectively, of rated core flow, at least once per 18 months.,

4.'.1 1 1 Establish a baseline APRM and LPRM** neutron flux noise valu in the regions for wnicii - s required (Specifi . , , ACTION c) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of entering the re io pin is required unless baselining has previ er ormed in the region s1nMheiastaafueling_

out Rict .Tnurt F

, 1 O'

If not performed within the previous 31 days.

"Octector 10 /e!! a and c of one LPRM strina ear ene ector,; piu, Jetectors A c and r ^' ene LIRM icing in tfie center of T.he surc shovid t.,e-monitorad

)

FERMI - UNIT 2 3/4 4-2

Insert F:

4.4.1.1 3 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that:

a. THERHAL POWER is less than or equal to 70% of RATED THERMAL POWER, and

b. The individual recirculation pump flow controller for the operating recirculation pump is in the Manual mode, and

c. The speed of the operating recirculation pump is less than or equal to 75% of rated pump sp d.Coreflowisgreaterthan39%ged,and of rated core flow when THERMAL POWER is greater than the limit specified in Figure 3 4.1.1-1.

4.4.1.1.4 With one reactor coolant system loop not in operation with THERMAL POWER less than or equal to 30% of RATED THERMAL POWER or with recirculation loop flow in the operating loop less than or equal to 50% of rated loop flow, verify the following differential temperature requirements are met within no more than 15 minutes prior to either THERMAL POWER increase or recirculation flow increase:

a. Less than or equal to 145 F between reactor vessel steam space coolant and bottom head drain line coolant, and

b. Less than or equal to 50 F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel ***, and

c. Less than or equal to 50 F between the reactor coolant within the loop not in operation and the operating loop.***

1. 1. Requirement does not apply when the recirculation loop not in operation is isolated from the reactor pressure vessel.

Value to be established during Startup Test Program (coreflow with both recirculation pumps at minimum speed). Final Value to be provided with 90 days of completion of Startup Test Program.

i

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_ __ - . . . . = C.,

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! (031Y8 %) 83M0d 7m83H13800 FERMI - UNIT 2 3/4 4-3

REACTOR COOLANT SYSTEM JET PUMPS ,

LIMITING CONDITION FOR OPERATION ____

3.4.1.2 All jet pumps shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

_ ACTION:

With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i SURVEILLANCE REQUIREMENTS l

4.4.1.2 Each of the above required jet pumps shall be demonstrated OPERABLE prier te vugpual P0"!P ::::: din- 25% ;f PATCO T" R"AL P0il:R end at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months shby determinin3@e!YiEulation loop flo#, total core flow, and diffuser-to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur;whec the recircuhtica pu p: re p ting :t th: :: : :p::1 h_ ilk THE/*M A L 60 e R. ope, at,'y is)

a. The indicatedArecirculation loop flow' differs by more than 10% from

@ e**b IA1 the established pump speed-loop flow characteristics.

psro ,( /2,f7ED

[WL b. The indicated total core flow differs by more than 10% from the established total coee flow value derived from recirculation loop Pbt0E R flow measurements.

c. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from the mean of all jet pump differential pressures in the same locp by more than 20% deviation from its normal

deviation.

% pro vi.s to ra of Sp eo,Re a1lon 4',0 4 ore no t c<pplic < bie l $ b osa r3; f 0 V'YY N1CtY h 5 Su r y e ,'ll4 s4 ce l'$ f et fot'M eV ytr'fh t n affe e exceedr9 525% dR ATED TH ERMAL. PotJ ER .

During the star 4 P t progran, data shall be recorded for the parameters listed to prov w;

sis for establishing the specified relationships M Comparisons of the .tual data in accordance with the criteria listed shall commence upon the conclusion of the startup test program.

- O 3/4 4-4 f* " ' ' 'MC ## " I' I' # 1 FERMI - UNIT 2 lo op aa wo rec,yeuta,t;.

Ic op of e ra1 inn.

- REACTOR COOLANT SYSTEM i

RECIRCULATION PUMPS ,

LIMITING CONDITION FOR OPERATION 3.4.1.3 Recirculation pump speed shall be maintained within:

a. 5% of each other with core flow greater than or equal to 70% of rated core flow.

b. 10% of each other with core flow less than 70%'of rated core flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*x c/wr'qp /wro ree/ecu/c.//en lo *t* 0/ *! *

YI* * '

ACTION:

With the recirculation pump speeds different by more than the specified limits, either:

a. Restore the recirculation pump speeds to within the specified limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

.w. 2 ...

b. M. N. . . F. P.u.!. 'L. .. .*.I. ./. 3.L. .f. 7. 'I.C,#",'*d.

. . . . .. , '$.*2, _Iid.i.u.

in pc?:'.icn and take the ACTION required by Specification 3.4.1.1.

Hor SHurpotw *vulsin IR- Acu rs .

DHere was s he in cat lees i SURVEILLANCE REQUIREMENTS 4.4.1.3 Recirculation pump speed shall be verified to be within the limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

"See Special Test Exception 3.10.4.

FERMI - UNIT 2 3/4 4-5 i

1 l

REACTIVITY CONTROL SYSTEMS BASES g

3/4.1.3 CONTROL RDOS The specifications of this section ensure that (1) the minimum SHUTDOWN MAh31N is maintained, (2) the control rod insertion times are consistent with those used in the safety analyses, arid (3) limit the potential effects of the rod drop accident. The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable 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 ooeration with a large number of inoperable control roos.

Control rods that are inoperable for other reasons are permitted to be taken out of se vice provided that those in the nonfully inserted position are consistent with the SHUT 00WN MARGIN requirements.

ine nuaber of control rods permitted to be inopera' ole 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 must be shut _do w for investigation and resolution of the problem. g Q fc/apd/g u f.f The control rod system it designed to bring the reactor subcri rate fast enough to prevent the MCPR from becoming less tha , during the limiting power transient analyzed in Section 15B of the FSAR This analysis shows that the negative teactivity rates resulting from the scram with the average response of all the drives as given in the speci cations, provide tw required protection and MCPR remains greater than 1,46 The occurrence of scram times lenger then those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

The scram discharge volume is required to be OPERABLE so that it will be available 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 Specification 3.1.3.1 then applies. 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.

O FERMI - UNIT 2 8 3/4 1-2

y i

3/4.2 POWER O!STRIBUTION LIMITS BASES 7 .

The specifications of this section assure that the peak cladding temperature following the oostulated design basis loss-of-coolant accident will not exceed the 2200*F limit specifica in 10 CFR 50.46.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location ar.d is dependent only secondarily on the rod to rod power distribution within an assembly. The pesk clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the oesign LHGR corrected for densification. This LHGR times 1.02 is used in the heatup coae along with the exposure cependent steady state gap conductance and rod-to rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod diviaed by its local peaking factor. The limiting value for APLHGR is shown in Figures 3.2.1-1, 3.2.1-2 and 3.2.1-3.

The calcult.tional procedure used to establish the APLHGR shown on Figu.%s 3.2.1-1, 3.2.1-2 and 3.2.1-3 is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 13 CFR 50.

A complace discussion of each code employed in the analysis is presented in Reference 1. Differences in this analysis compared to previous analyses can be broken down'as follows.

a. Input Changes

1. Corrected Vaporization Calculation - Coefficients in the vaporization correlation used in the REFLOOD cooe were corrected.

2. Incorporated more accurate bypass areas - The bypass areas in the top guide were recalculatea using a more accurate technique.

3. Corrected guide tube thermal resistance.

4. Correct heat capacity of reactor internals heat nodes.

Fe c plan t ope rc.f.% with n e tyle op e r a t.y rechoulah'o loops, fke MAPhH&ft limcb of F. pre 3.2 l-) 1 3.2 1-2,o d 3 2.1-3 ca'e mulhy lied b y O.90. The corn h.4 t%fer- ve 0.4o ,

l.s de riv eel bo,n Loc 4 analyds in ttwied fnm +/9e / loop openMn h accou.d de e w lie r dolling frans> Won ai

e limiltny hel node corqued 7b /Ae s tancia.d Mc4' FERMI - UNil -

4 2-3

POWER DISTRIBUTION LIMlls BASES 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.

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

A few of the changes affect the accident calculation irrespective of CCFL. These changes are listed below,

a. Input Change

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

b. Model Change

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 B 3.2.1-1.

3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based o.1 a power distribution which would yield the design LHGR at RATED THERMAL POWER. The flow biased simulated thermal power-upscale scram setting and flow fde fuef biased simulated thermal power-upscale control rod block functions of the APRM c$nJd[3 instruments must be adjusted to ensure that the MCPR does not become less shan eg ggTy b=ps or that > 1% plastic strain does not occur in the degraded situation. The . .

"* I scram settings and rod block settings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and MFLPD indicates a higher peaked power distribution to ensure that an LHGR transient woulo not be increased in the degraded condition.

FERM) - UNil ? E '4 2-2

1 BASES TABLE B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters:

Core THERMAL P0WER.................... 3430 MWt* which corresponds to 105% of rated steam flow Vessel Steam Output................... 14.86 x 10s 1bm/hr which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure............ 1055 psia Design Basis Recirculation Line Break Area for: ,

a. Large Breaks 4.1 ft2

b. Small Breaks 0.1 ft g 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 1.18 Y#

A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection 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.

94 for slyle re e ic e u fa1io n /o op op e rn fio n , /oss o 4 n u e.( u1<

boilig is a ss u ,u e el af o . I e e c one{ affe < L o c 4 reg a<cNe s.s of /n itial MC PR .

FERMI - UNIT 2 B 3/4 2-3

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POWER DISTRIBUTION LIMITS BASES O

3/4.2.3 MINIMUM CRITICAL POWER RATIO The 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 1.00, and an analysis of abnormal operational transients. For any abnormal operating transients 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 a0 normal 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. 'h: liniting trentient yield; th: 10 90;t

&lt: MCM. "her added t: the C:fety Limit MCPn of 1.05, th: requir:d r'nimum cptr: ting l'-it M P' Of Speci't::ticr 3.2.2 i: Obtained and presente'd " -

fige 2. 2. 21.

The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15B.0-1 that are input to a GE-core dynamic behavior transient computer program. The code used to evaluate pressurization events is described in NEDO-24154(3) and the program used in nonpressurization events is described in NEDO-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 code described in NEDE-25149(4) . The principal result of this evaluation is the reduction in MCPR caused by the transient. ,

The purpose of the fK factor of Figure 3.2.3-2 is to define operating limits at other than rated core flow conditions. At less than 100% of rated flow the required MCPR is the product of the MCPR and thef K factor. The K f factors assure that the Safety Lim 1t MCPR will not be violated during a flow increase transient resulting from a motor generator speed control failure. The Kf factors may be applied to both manual and automatic flow control modes.

The K, factor values shown in Figure 3.2.3-2 were developed generically and are aph11 cable to all BWR/2, BWR/3, and BWR/4 reactors. The Kf factors were l

I derived using the flow control line corresponding to RATED THERMAL POWER at j rated core flow.

For the manual flow control mode, the Kg factors were calculated such that for the maximum flow rate, as limittd by the pump scoop tube setpoint and the corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different l core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the Kf.

FERMI - UNIT 2 B 3/4 2-4

r INSTRUMENTATION BASES 3/4 3 10 NEUTRON FLUX MONITORING INSTRUMENTATION 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 operating domain, a small probability of limit cycle neutron flux 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 3 10-1 (Reference 1).

Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations. BWR cores typically operate with neutron flux noise caused by random boiling and flow noise. Typical neutron flux noise levels of 1-12% of rated power (peak-to-peak) have been reported for the range of low to high recirculation loop flow during both single and dual recirculation loop operation. Neutron flux noise levels which significantly bound these values are considered in the thermal / mechanical design of GE BWR fuel and are found to be of negligible consequence (Reference 2). In addition, stability tests at operating BWRs have demonstrated that when stability related neutton flux limit cycle oscillations occur they result in peak-to-peak neutron flux limit cycles of 5-10 times the typical values. Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations.

Baseline data with two reactor recirculation loops in operation should be taken near the maximum rod line at which the majority of operation will occur. However, baseline data taken at lower rod lines (i.e. lower power) will result in a conservative value since the neutron flux noise level is proportional to the power level at a given core flow. Because of the uncertainties involved in SLO at high reverse flows, baseline data should be taken at or below the power specified in Figure 3 3 10-1. This will result in approximately a 25% conservative baseline value if compared to baseline data taken near the rated rod line and will therefore not result in an overly restrictive baseline value, while providing sufficient margin to cover uncertainties associated with SLO.

Refeg nces j

(1) 'BWR Core Thermal-Hydraulic Stability" Service Information Letter 380, l

Revision 1, February 1984.

(2) G. A. Watford, "Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria," December 1982 (NEDE 2227i-P).

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l

(

i 3/4.4 REACTOR COOLANT SYSTEM .

BASES .-

3/4.4.1 RECIRCULATION SYSTEM i

ivvp invywiaLiv i, 0;;r;ti n with ;n; ce;cter ce.e ceelent .. .. vioilvn pr:hitfted unt4' ra ev-!ustier of the pe-fe--ence of the ECCS du 'a; cae 'eep eue!eeted, ead date- iaed +c be ecceptekle.

op+pa4*:n kes beca pe fe-~=d, Zerse t t G-An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of refloodir;g the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump f ailure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation pemp speed mismatcu limits are in compliance with the ECCS G> e d M LOCA analysis design criteria 3 (or No r e c, e c u /n y 'on */o ep ojoe r a Na a .

In order to prevent ur.due stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other The loop temperature must also be within prior to startup of an idle loop.50'F of the reactor pressure Sin :vessel th: :::!:ntcoolant

'" temperature t 7 shock to the recirculation pump and recirculation nozzles.

g -the bottom of th ;;;; 1 i: :t : ? r:r t::per:tur: th:r th: cee!:nt i- tthe 4' the u?per peratu-e

i

n

ef the cere, undue St-es ca the ueese! "eu!d -e5U": t diiivi ens. -d wi setic then 145 T. Juerf.t.

3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of A total of 11 OPERABLE safety /

1325 psig in accordance with the ASME Code.

relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.

Demonstration of the safety / relief valve lift sett;ngs will occur only during shutdown and will be performed in accordance witn the provisions of Specification 4.0.5.

The low-low set system ensures that a potentially high thrust load (der,is-nated as load case C.3.3) on the SRV discharge lines is eliminated during sab-sequent actuations. This is achieveo by automatically lowering the closing set-point of two valves and lowering the opening setpoint of two valves following the initial opening. Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced setpoint does not violate the design basis. .

FERMI - UNIT 2 B 3/4 4-1

r i

Insert G:

The impact of single recirculatica loop operation upon plant safety is assessed and shows that sindle-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 setpoints (or APRM gains) 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 32.1. A time period of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is allowed to make these adjustments following the establishment of single loop operation since the need for single loop operation often cannot be anticipated. MCPR operating limits adjustments in Specification 3 2 3 for different plant operating situations are applicable to both single and two recirculation loop operation.

Additionally, surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive core internals vibration. The surveillance on differential temperatures below 30% THERMAL POWER or 50% rated recirculation loop flow is to prevent undue thermal stress on lesse7 nozzles, recirculation pump and vessel Dottom head during the extended operation of the single recirculation loop mode.

Insert H:

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

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

Insert I:

Sudden equalization of a temperature difference >145 F between the l 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 botton head.

1 1

__j

NRC-87-0132, Proposed Changes to Tech Specs & Justifications to Allow Single Loop Operation

Text

2.0 INTRODUCTION

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