ML19326B715

Proposed Tech Specs Change for Safety Limits & Limiting Safety Sys Settings,Limiting Operating Conditions & Min Equipment Test Frequency

ML19326B715
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/01/1976
From:	ARKANSAS POWER & LIGHT CO.
To:
References
NUDOCS 8004170551
Download: ML19326B715 (27)

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ARKANSAS 50-313 PROPOSED TECHNICAL SPECIFICATIONS CHtWGE

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p 'S,dETY LIMITS AND- LIMITING SAFETY SYSTEM SETTINGS gg M.1 '. SAFETY LIMITS, ' REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolsnt

system pressure, coolant temperature, and coolant flow during power operation of the plant..

j Ob'ective ,

_To maintain the integrity of the fuel cladding.

. Specification

. 2.1.1 The combination of .the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in Figure 2.1-1. If the actual pressure / temperature point is below and to the right of the pressure / temperature line the safety limit-is exceeded. .

2.1.2 The' combination or reactor thermal power and reactor power imbalance (power in the top half of the core minus the power in the bottom half of the core _ expressed as a percentage of the rated power) shall not exceed the safety limit as defined by the locus of points for O the.specified flow set forth in Figure 2.1-2. If the actual-reactor-

- thermal-power / reactor-power-imbalance point is above the line for the specified flow, the safety limit is exceeded. --

Bases To maintain the integrity.of the fuel. cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating. conditions. This is accomplished by operating within the nucleate boiling regime of heat transfei, wherein the heat transfer coefficient is large enough so _ that ' the clad surface temperature is only slightly greater than the coolant temperature. He upper boundary of the nucleate boiling regime is termed, departure from nucleate boiling (DNB). At this point there is a sharp reduction of the heat : ansfer coefficient, which would result in high '

-cladding temperatures and the possibility of cladding failure. Although DNB is not an observable parameter .during reactor operation,- the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of the BAW correlation. (1) The BAW-2 Lcorrelation-_has been developed to predict DNB and the location of DNB for axially . uniform and ' non-uniform _ heat . flux dist ributions. 'Ihe local DNB ratio  !

-(DNBR),. defined as the radio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. '!he minimum value of the DNBR, during steady-state operation, normal j operational 3 transients, and anticipated transients is limited to 1.3. A p . 1

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DNBR of 1.3 corresponds to a 95 percent probability at a 95 percent confi-ence level that DNB will not occur; this is considered a conservative mar-in to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has b;en considered in determining the core prctection safety limits. He differ-cnce in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip set points to correspond to the olevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.3 is predicted for the maximum possible thermal power (112 percent) when the reactor coolant flow is 374,880 gpm, which is 106.5% of the design flow rate for four operating reactor coolant pumps. This curve is based on the following nuclear power pe.sking factors (2) with potential fuel densification effects; N N N F = 2.67; F = 1.78; F = 1.50 g gi  :

These design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion, and form the core DNBR design basis.

The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod bowing:

p 1. Tge 1.3 DNBR limit produced by a nuclear power peaking factor of

(

F = 2.67 or the combination of the radial peak, axial peak and p8sition of the axial peak that yields no less than 1.3 DNBR.

2. The combination of radial and axial peak that prevents central fuel melting at the hot spot. The limit is 19.4 kW/ft.

Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.

l ne specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 correspond  ;

to the expected minimum flow rates with four pumps, three pumps, and one pump  ;

in each loop, respectively.  !

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations 'iown in Figure 2.1-3. The l curves of Figure 2.1-3 represent the conditions -t which a minimum DNBR of 1.3 is predicted at the maximum possible thermal power for the number of i reactor coolant pumps in operation or the local quality at the point of mini-tum DNBR is equal co 22 percent (1), whichever condition is more restrictive.

Using a local quality 1?mit of 22 percent at the point of minimum DNBR as a basis for curve 3 of Figure 2.1-3 is a conservative criterion even though the ,

quality at the exit is higher than the quality at the point of minimum DNBR.

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.3' r 1The DNRR an calculated by the BAW-2 correlation continually increases from point

(~'7t' minimum DNHR,-so that the er.it DNBR is always higher and is a function of the

}s_,)ressure. ,

Vie maximum thermal power for three pump operation is 86.4 percent due to a i power level trip produced by the flux-flow ratio (74.7 percent flow x 1.07 = 79.9 l percent power) plus the maximum calibration and instrumentation error. The taximum thermal power for other reactor coolant pump conditions is produced in'a similar manner.

~

.For each curve of Figure 2.1-3, a pressure-temperature point above .and to the left of' the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR. less than 22 percent for that particular reactor coolant Jnmap situation. Curves .lf,2 of Figure 2.1-3 is the most restrictive because any pressure / temperature point above and to the left of this curve will be above and to the left of the other curve.

REFERENCES (1) Correlation of Critical Heat Flux in a Bundle Cooned by Pressurized Water,

• BAW-10000A, May, 1976. '

-(2). FSAR, Section 3.2.3.1.1.c I

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. 2000 0 UnacceptaDie Operation 1800 7 1600 580 600 620 640 660 560 Reactor Outlet Temperature. *F ARKANSAS POWER & LIGHT CO. CORE PROTECTION SAFETY LIMIT ARKANSAS NUCLEAR ONE-UNIT I Figure 2.1-1 CYCLE 2 O

l 98

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/ Acceptacle 3 & 4 Pump Operation 80 a

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-60 -40 20 Reactor Power imbalance Rc. ACTOR COOLANT FLOW (gpm)

_ CURVE 1 374.880 2 280.035 3 184.441 ARKANSAS POWER & LIGHT CO. CORE PROTECTION SAFJTY LIMITS Figure 2.1 -2 ARKANTAS NUCLEAR ONE-UNIT 1 CYCLE 2 9b

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$60 580 600 620. 640 660 Reactor Outlet Temperature,*F REACTOR COOLANT FLOW LilH VE GPM POWER PUMPS OPERATING (iYPE OF LIMIT) 374.880 (1005)

• 1127. FOUR PUMPS (ONBR LIMIT) 2 280.035 (74.7%) 86.4% THREE PUMPS (ON8R LlulT) 3 184,441 (49.2%) 59.1% ONE PUMP IN EACH LOOP (OUALITY LIMIT)

• 106.5% OF CYCLE I OESIGN FLOW ARhANSAS POWER & LIGHT CO. CORE PROTECTION SAFETY LIMITS ARMANSAS NUCLEAR ONE UNIT i Figure 2.1 3 CYCLE 2 O

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The power level trip set point produced by the power-to'-flow ratio provides both high power level and low flow protection in the event

(_' the reactor power level increases or the reactor coolant flow rate V) decreases. 'Ihe power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations ,

for the pump situations of Table 2.3-1 are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 107 percent and reactor flow rate is 100 percent or flow rate is 93.5 percent and power level is 100 percent.

2. Trip would occur when three reactor coolant pumps are operating if power is 79.9 percent and reactor flow rate is 74.7 percent or flow rate is 70.1 percent and power level is 75 percent.

3. Trip would occur when one reactor coolant pump is operating in i each loop (total of two pumps operating) if the power is 52.4 percent and reactor flow rate is 49.0 percent or flow rate is 45.8 percent and the power level is 49.0 percent. #

The flux / flow ratios account for the maximum c'alibraiion and instrumentation ~

crrors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indica-tion of the RC flow. _

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(Go penalty in reactor coolant flow through the core was taken for an open core vsnt valve because of the core vent valve surveillance program during each rsfueling outage. For safety analysis calculations the maximum calibration and instrumentation errors for the power leve.1 were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kW/ft limits or DNBR limits. The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power level trip associated reactor power-to-reactor power imbalance boundaries by 1.07 parcent for a 1 percent flow reduction.

B. Pump monitors - -

In conjunction with the power / imbalance / flow trip, the pump moni-tors prevent the minimum core DNBR from decreasing below 1. 3 by t rip-ping the reactor. due to the loss of reactor coolant pump (s). The pump monitors also restrict the power level- for the number of pumps in operation.

C. kaactor coolant system pressure ,

p During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nucicar overpower trip set point. The trip setting limit shown in Figure 2.3-1 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient.(2) 12

f^ 7 The low pr:ssure (180u psig) and variable low pressure (11.72 T

-5114.7) trip setpoint shown in Figure 2.3-1 have been establish $t to maintain the DNB rat 10 greater than or equal to 1.3 for those g

design accidents that result in a pressure reduction. ( .' . 3 )

(v)

Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (11.72 T out ' 5154.7).

D. Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (619 i, shown in Figure 2.3-1 has been established to prevent es-cessive core coolant temperatures in the operating range. Due to calibration and inst rumentation errors, the safety analysis useil

- a trip set point of 6.'u F.

ii. Reactor building pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident , e/en in the absence of a low reactor coolant system pressure t rip.

F. Shutdown bypass In order to provide for control rod drive tests, ero power

[] physics testing, and startup procedures, there is provision for V bypassing certain segments of the reactor protection system.

The reactor protection system segments which can be bypassed are shown in Table 2.3-1. Two conditions are imposed when the bypass is used:

1. A nuclear overpewer trip set point of 5 5.0 percent of rated

-power is automatically imposed during reactor shutf wn.

2. A high reactor coolant system pressure trip set point of 1720 psig is automatically imposed.

The purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of tt.c reactor protection system bypassed. This high pressure trip set point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated. The overpower trip set point of s5.0 percent prevents any significant reactor power from being produced when performing the physics tests. Sufficient natural circulation (5) would be available to remove 5.0 percent of rated power if none of the reactor coolant pumps were operating.

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1900 1 ix Unacceptable  !

Operation P = 1800 psig 1100 1500 560 580 600 '620 640 660 Reactor Outlet Temperature. *F ARKANSAS POWER & LIGHT CO. PROTECTIVE SYSTEN MAXINUN ARKANSAS NUCLEAR ONE-UNIT I ALLOWABLE SET PolNT CYCL 2 2 Figure 2.3-1 O

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AHKANSAS POWER & LIGHT CD. PROTECTIVE SYSTEM MAXIMUM ARKANSAS NUCLEAR ONE UNIT 1 ALLOWABLE SETPOINTS CYCLE 2 Figure 2.3-2 lO .

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_ x,,/. Tabis 2.3-4 Reactor Pratsetien Systin Tria Setm..Jg Licits fm) Qf One Reactor Coolant Pump Four Reactor Cociat Pumps Three Reactor Coolant Purps Operating in Each tmop Operating (No: inst Operating (No:ninst (Nominst Operating Shutdown Operat ing Power - It:J'.) Operating Power - 75%) Power - 49f.) typas s suc1ccr power, % of 105.5 - 105.5 105.5 ctnd, mas 5.0(3) ructorr power based on 1.07 times flow minus 1.07 times flow minus

'Itw(2) tnd imbalance, 1.07 times flow minus typassed reduction due to reduction due to reduction due to i cf ratsd, max i= balance (s) imbalance (s) imbalance (s)

,uclear power based on NA NA 55i. Bypassed h sucpcionitors}%of

ted, rzx I4 ligh reactor coolant 2355 2355 2355 1720(3) ystco presst.re, psig, ax

.ow reactsr coolant sys- 1800 1800 1800 Bypassed 2a pressure, psig, min Driable low reactor (11.72 Tout -5114.7)(1) (11.72 Tout-5114.7) (1) (11.72 Tout-5114. 7) (1) sypassed solant system pressure, aig, :nin ascter crolant tesy, 619 619 619 619

, nsx -

igh reacter building 4 (13.7 psia) 4(18.7 psis) 4 (18.7 psia) 4 (18. 7 psia) 7 essure, psig, rax

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(1) Tout is in degrees .ahrenheit (F). (3) Autoestically set when other se;;nents of the RPS (as specified) are byysssed. "

(2) Reactor coolar.t system flow, 1 14) The pump conitors also produce a trip on: (a) loss of two reactor coolant pur.ps in one reactor coolant loop, and (b) loss of one or two reactor coolant 2 pumps during two-punp operation.

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3. 'l.lMITING CONDITIONS l'OR OPERATION.

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t E3. I Ri!ACI'Olt C001.ANI' SYSTEM v

'Appilcability Applies to th) operating status of the reactor coolant system.

Obj ective To specify those limiting conditions for operation of the reactor coolant sys-tem which must be met to ensure safe reactor operations.

3.1.1 Operational Components Specification 3.1.1.1 Reactor coolant Pumps A. Pump combinations permissible for given power levels chall be ,

as shown in Table 2.3-1. I 1

l B. The boron concentration in the reactor coolan.t . system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.

.3.1.1.2 Steam Generator

-(O) A. One steam generator shall be operable whenever the reactor coolant average temperature is above 280 F.

3.1.1. 3 Pressurizer Safety Valves A. The reactor shall not remain critical unless both pressurizer code safety valves are operable.

t B. When the reactor is subcritical, at least one pressurizer code safety valve shall be operabic if.all reactor coolant system openings are closed, except for hydrostatic tests in accord-ance with ASME Boiler and Pressure Vessel Code,Section III.

3.1.1. 4 Reactor Internals Vent Valves The stcuctural integrity and o'perability of the reactor internals vent valves shall be maintained at a level consistent with the acceptance criteria in Specification 4.1.

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,n q Bases A reactor coolant pump or decay heat removal pump is required to be in opera-tion before the boron concentration is reduced by dilution with makeup water.

[ ] Either pump will provide mixing which will prevent sudden positive reactivity

(/ changes caused by dilute coolant reaching the reactor. One decay heat removal pump will circulate the equivalent of the reactor coolant system volume in one half hour or less. (1)

The decay heat removal system suction piping 'is designed for 300 F thus, the system can. remove decay heat when the reactor coolant system is below this temperature. (2,3)

One pressurizer code safety vaive is capable of preventing overpressurizatio s when the reactor is not critical since its relieving capacity is greater th in that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat. (4) Both pressuri er code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities. The code safety valves prevent overpres-sure for a rod withdrawal accident.(5) The pressurizer code safety valve lift set point shall be set at 2500 psig + 1 percent allowance for error and each valve shall be capable of relieving 300,000 lb/h of saturated steam at a pressure not greater than 3 percent above the set pressur(. .

The internals vent valves ' are pr'ovided to relieve the pressure generated by steaming in the core following a- LOCA so that the core remains sufficiently covered. Inspection and manual actuation of the internals vent valves (1) ensure Operability, (2) ensure that the valves are not open during normal

^

operation, and (3) demonstrate that the valves begin to open and are fully open

\ at the forces equivalent to the differential pressures assumed in the safety

[ Q analysis.

REFERENCES (1) FSAR, Tables 9-10 and 4-3 through 4-7.

(2) FSAR, Section 4.2.5.1 and 9.5.2.3.

(3) FSAR, Section 4.2.5.4.

(4) FSAR, Section 4. 3.10.4 and 4.2. 4.

(5) FSAR, Section 4.3.7.

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• 17

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I If a cont rol rod in the regulating or axial power shaping groups. 1

- is declared inoperable per Specification 4.7.1.2 operation above {

' 60 percent of the themal power allowable for the reactor coolant (m)

' k/ pump combination may continue provided the rods in the group are positioned such that the rod that was delcared inoperable is con- .

tained within allowable group average position limits of Specifica-tion 4.7.1.2 and the withdrawal limits of Specificat ion 3.5.2.5.3.

3.-5.2.3 The worth of single inserted control rods during criticality are limited by _ the restrictions of Specification 3.1.3.5 and the Control 6

Rod Position Limits defined _in Specification 3.5.2.5.

~

'3.5.2.4 Quadrant tilt:

1.

Except-for physics tests, if quadrant tilt exceeds 3.41*., power shall be reduced immediately to below the power level eutoff (see Figures 3.5.2-1A and 3.5.2-18). Moreover, the power level cutoff value shall be reduced 2*. for each it tilt in excess of 3.11". t ilt . For less than 4 pump operation, thermal power sha11 he reduced 2'e of the thermal power allowable for the reactor coolant pump combin-ation for each It tilt in excess of 3.41%.

,.- 2. Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the. quadrant power tilt shall be reduced to lessinthan ments 3.41'.and setpoints except limitsfor physics shall tests, or the following adjust- l be made:

a.

The protection system maximum allowable setpoints (Figure

2. 3-2) shall be - reduced 2*. i f power of each 1 *, t i l t .

b. The control rod group withdrawal limits (Figures 3.5.2-1A, 3.5.2-1B and 3.5.2-lC shall be reduced 2% in power for each 1**

tilt in excess of 3.41*..

c. The opera.:ional imbalance limits (Figures 3.5.2-3A 3.5.2-3B and 3.5.2-3C) shall be reduced 2% in power for each l'. tilt in excess of 3.41"..

3. If quadrant-tilt is in excess of 25'., except for physics tests or diagnosti'e testing, the reactor will be placed in the hot shutdown condition. Diagnostic testing during power operation with a quad-rant power tilt is permitted provided the thermal power allowable for the reactor coolant pump combination is restricted as stated in 3.5.2.4.I above.

4.

Quadrant tilt shall be monitored on a minimum frequency of once every two hours' during power operation above 15% of rated power.

3.5.2.5' Control rod positions:

1.

Technical Specification 3.1.3.5 (safety rod withdrawal) does not-

. prohibit the exercising of individual safety rods as required by Table.4.1-2 or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.

n[ ;2.

Operating rod group overlap shallibe 25% +5 between two sequential groups. except for physics tests. ,

47

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3. Except for physics tests or exercising control rods, the control rod withdrawal limits are specified on Figures 3.5.2-1A, 3.5.2-1B

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. 3

) and 3.5.2-1C for four pump operation and on Figures 3.5.2-2A, 3.5.2-2B and 3.5.2-2C for three or two pump operation. If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within four hours.

4. Except for physics tests, power shall not be increased above the power level cutoff (see Figures 3.5.2-1) unless the xenon reactivity is within 10 percent of the equilibrium value for operation at rated power and asymptotically approaching stability.

3.5.2.6 Reactor Power Imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power. Except for physics tests, imbalance shall be maintained within the envelopes defined by Figures 3.5.2-3A, 3.5.2-3B and 3.5.2-3C. If the imbalance is not within the envelopes defined by Figures 3.52-3A, 3.5.2-3B and 3.5.2-3C corrective measures shall be taken to achieve an acceptable imbalance.

If an acceptable imbalance is not achieved within four hours, reactor ppwer 3

shall be reduced until imbalarae limits are met.

3.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.

B0ses e power-imbalance envelopes defined in Figures 3.5.2-3A, 3.5.2-3B and 3.5.2-3C is based on 1) LOCA analyses which have defined the maximum linear heat rate (See Fig.

3.5.2-4) such thr.t the maximum clad temperature will not exceed the final Acceptance Critsria and 2) the Protective System Maximum Allowable Setpoints (Figure 2.3-2). Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary.

Operation in a situation that would cause the final acceptance criteria to l ba approached should a LOCA occur is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and un-certainty factors are also at their limits.* Conservatism is introduced by i application of:

a. Nuclear uncertainty factors

b. Thermal calibration

c. Fuel densification effects

d. Ilot rod manufacturing tolerance factors

o. Fuel rod bowing The 25 +5 percent overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke. Control i rods are arranged in groups or banks defined as follows: l

" Actual operating limits depend on whether or not incore or excore detectors tre used and their respective instrument and calibration errors. The method used to define the operating limits is defined in plant opersting procedures.

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3.5.2.4 ARKANSAS NUCLEAR ONE-UNIT 1 48e-

l Table 4.1-2 (Continued)

/~~T Minimum Equipment Test Frequency

\ )

s. s

)

Item Test Frequency

12. Flow Limiting Annulus Verify, at normal One year, two years, on Main Feedwater operating conditions, three years, and every 4 I

Line at Reactor that a gap of at least five years thereafter Building Penetration 0.025 inches exists measured from date of between the pipe and initial test.

the annulus.

13. SLBIC Pressure Calibrate Each Refueling Period Sensors

14. Main Steam Isolation a. Excercise Through a. Quarterly Valves Approximately 10%

Travel

b. Cycle b. Each Refueling Shut-down. 4

15. Main Feedwater a. Exercise Through a. Quarterly Isolation Valves Approximately 5%

/S Travel b' b. Cycle b. Each Refueling Shut-down.

16. Reactor Internals Demonstrate Operability Each refueling shutdown.

Vent Valves By:

a. Conducting a remote visual inspection of visually accessible sur-faces of the valve body and disc sealing faces and evaluating any observed surface irregu-latities.

b. Verifying that the valve is not stuck in an open position, and

c. Verifying through manual actuation that the valve is fully open with a force of < 400 lbs (applied  !

vertically upward). l 7, '^ ; 1 NJ l 73a i

_