ML19312B838

Proposed Tech Specs 2.1,2.3 & 4.2.12 Re Flux/Flow Trip Setpoint for Unit 1 & Incorporation of Surveillance Testing Requirement for Internals Vent Valves

ML19312B838
Person / Time
Site:	Oconee
Issue date:	06/11/1976
From:	DUKE POWER CO.
To:
Shared Package
ML19312B833	List: ML19312B832 ML19312B838
References
NUDOCS 7911250049
Download: ML19312B838 (8)

Text

,

t can be related to DNB through the use of the BAW-2 correlation (1). The BAW-2 correlation has been developed to predict DNB and the location of DN3 for axially uniform and non-uniform heat flux distributions. The local DN3 ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a i particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. A DNBR of 1.30 corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to

, DNB for all operating conditions. The dif ference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setponts to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1A represents the conditions at which a minimum DNBR of 1.30 is predicted for the maxi =um possible thermal power (112 percent)whenfourreactorcoolantpumgsareoperating(minimumreactor coolant flow is 107.6 percent of 131.3 x 10 lbs/hr.). This curve is based on the combination of nuclear power peaking factors, with potential ef fects of fuel densification and rod bowing, which result in a more conservative DNBR than any other shape that exists during normal operation.

The curves of Figare 2.1-2A are based on the more restrictive of two thermal limits and include the ef fects of potential fuel densification and rod bowing:

1. The 1.30 DNBR limit produced by the combination of the radial peak, axial peak and position of the axial peak that yields no less than a 1.30 DNBR.

2. The combination of radial and axial peak that causes central fuel =elting at the hot spot. The limit is 20.15 kw/ft for Unit 1.

Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the reactor power i= balance produced by the power peaking.

The specified flow rates for Curves 1, 2, 3 and 4 of Figure 2.1-2A correspond to the expected minimum flow rates with four pumps, three pumps, one pu=p in each loop and two pumps in one loop, respectively.

The curve of Figure 2.1-1A is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3A.

The maximum thermal power for three-pump operation is 86.4 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.07 =

79.9 percent power plus the maximum calibration and instrument error. The maxi =um thermal power for other coolant pump conditions are produced in a similar manner.

2.1-2 ggg0h

~ - . - - - . 4 --4+i 3--y.-g--vy g e - 3- = * * +-

--t-- 9

. l 4.2.10 For Unit 1, Cycle 3 operation, the surveillance capsules will be removed from the reactor vessel and the provisions of Specification 4.2.9 will be revised prior to Cycle 4 operation.

For Unit 2, Cycle 2 operation, the surveillance capsules will be removed from the reactor vessel and the provisions of Specifica- ,

tion 4.2.9 will be revised prior to Cycle 3 operation. For Unit 3, Cycle 1 operation, the surveillance capsules will be removed from the reactor vessel for a portion of the cycle and the pro-visions of Specification 4.2.9 will be revised prior to Cycle 2 operation.

4.2.11 During the first two refueling periods, two reactor coolant system piping elbows shall be ultrasonically inspected along their longitudinal welds (4 inches beyond each side) for clad bonding and for cracks in both the clad and base metal. The elbows to be inspected are identified in B&W Report 1364 dated December 1970.

4.2.12 To assure that reactor internals vent valves are not opening during operation, all vent valves will be inspected during each refueling outage to confirm that no vent valve is stuck open and that each valve operates freely.

Bases The surveillance program has been developed to comply with Section XI of the ASME Boiler and Pressure Vessel Code, Inservice Inspection of Nuclear Reactor Coolant Systems, 1970, including 1970 winter addenda, edition. The program places major emphasis on the area of highest stress concentrations and on areas where fast neutron irradiation might be sufficient to change material properties.

The reactor vessel specimen surveillance program for Unit 1 and Unit 2 is based on equivalent exposure times of 1.8, 19.8, 30.6 and 39.6 years. The contents of the different type of capsules are defined below.

A Type B Type Weld Material RAZ Material HAZ Material Baseline Material Baseline Material

~

For Unit 3, the Reactor Vessel Surveillance Program is based on equivalent exposure times of 1.8, 13.3, 26.7, and 30.0 years. The specimens have been selected and fabricated as specified in ASTM-E-185-72.

Early inspection of Reactor Coolant System piping elbows is considered desirable in order to reconfirm the int egrity of the carbon steel base metal when explosively clad with sensitized s :ainless s teel. If no degradation is observed during the two annual inspections, surveillance requirements will revert to Section XI of the ASME Boiler and Pressure Vessel Code.

4.2-3 .

l

inermal P0cer Level. 9 UNACCEPTABLE OPERATION

.. .120

( -29.112) (l121 (+28.I12)

.. 110 1

ACCEPTABLE 4 PUNP

" 100 OPERATION I I

(+40.90 ) 1

- 90

( -37,90 )

(86.4) ( +2 8.8 6.4) l

(-29 86.4) i 80 ACCEPTABLE

-- 70  !

PUNP

( +u0.64. 4) 2 OPERATION

( -3 7, 64.4)

( 29 58.9) -- 60 (58.9) (+28. 58.9)

ACCEPTABLE

-- 50 2.3 & 4 PUMP OPERATION

'~ 1

( -3 7, 36.9) 3g4

.. 30

.. 20

_. 10 i

i i i

+20 +40 +60 60 40 -20 0 Reactor Power imoalance. 5 CURVE REACTOR COOLANT FLOW (10/hr) 1 141.3 x 10 6 2 105.6 x 10 6 3 69.3 x 10 6 4 64.7 x 10 6 tillT 1 CORE PROTECTION 9FETY LIMIT OCONEE NUCLEAR STATION 0sE*te 4

FIGURE 2.1-2A 2.1-7

During normal plant ope ' tion with all teactor coolant ips operating,

.r;acetor trip is iniciacca whsn ths retctor power level reaches 105.5% of

. rated powar. Adding to this tha possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is more conservative than the value used in the safety analysis. (4)

Overpower Trip Based on Flow and Imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to acco==odate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. Analysis has demonstrated that the specified power-to-flow ratio is adequate to prevent a DN3R of less than 1.3 should a low flow condition exist due to any electrical malfunction.

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 decreases. The power level trip set point produced by the power-to-flow ratio provides overpower DNB pro-tection for all modes of pump operation. For every flow rate there is a =axi-sum 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 situtations of Table 2.3-1A are as follows:

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

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

3. Trip would occur when two reactor coolant pumps are operating in a single loop if power is 52.4% and the operating loop flow rate is 54.5% or flow rate is 47.9% and power level is 46%.

4 Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.4% and reactor flow rate is 49.0" or flow rate is 45.8% and the power level is 49".

The flux-to-flow ratios for Units 1 and 2 account for the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.

For safety calculations the maximum calibration and instrumentacion errors for the power level trip 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 l'imits. The reactor power i= balance (pewer in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-t)-flow ratio such that the bouncaries of Figure 2.3-2A - Unit I are produced. The power-to-flow ratio reduces the power 2.3 Unit 2 2.3-2C -

Unit 3 2.3-2 1

i

n 1evel trip and associated reactor power / reactor power-imbalance boundaries by 1.07% .for a 1% flow reduction.

The power-to-flow reduction ratio is 0.961 during single loop operation. [

Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s). The circuitry monitoring pump operational status provides redundant trip protection for DNB by tripping the reactor on a signal diverse f rom that of the power-to-flow ratio. The pump monitors also restrict the power level for the number of pumps in operation.

Reactor Coolant Svstem Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure set point is reached before the nuclear over-power trip set point. The trip setting limit shown in Figure 2.3-1A - Unit 1 2.3-1B - Unit 2 2.3-1C - Unit 3 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. (1)

The low pressure (1800) psig and variable low pressure (11.14 T -4706) trip (1800) psig (10.79 T " -4539)

(1800) psig (16.25 r "'-7756) setpoints shown in Figure 2.3-1A have been established to maintaSn the DNB 2.3-1B l 2.3-1C

ratio greater than or equal to 1.3 for those design accidents that result in t

a pressure reduction. (2,3)

! Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (11.14 T g -4746)

(10.79 T -4579)

(16.25 Tout " -7796)

Coolant Outlet Temperature  !

The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1A has been established to prevent excessive core coolant 2.3-1B 2.3-1C temperatures in the operating range. Due to calibration and inscrumentation

< errors, the safety analysis used a trip set point of 620 F.

Reactor Building Pressure ,

1 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 loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

2.3-3

l l

Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Table 2.3-1A. Two conditions are imposed when 2.3-1B 2.3-1C the bypass is used:

1. By administrative control the nuclear overpower trip set point must be reduced to a value < 5.0% of rated power during reactor shutdown.

2. A high reactor coolant system pressure trip setpoint of 1720 psig is ,

automatically imposed.

T*.te purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the 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 over power trip set point of < 5.0% prevents any significant reactor power from being produced when performing the physics tests. Sufficient natural circulation (5) would be available to remove 5.0% of rated power if none of the reactor coolant pumps were operating.

Two Pump Operation A. Two Loop Operation Operation with one pump in each loop will be allowed only following reactor shutdown. After shutdown has occurred, reset the pump contact monitor power level trip setpoint to 55.0%.

B. Single Loop Operation Single loop operation is permitted only after the reactor har been

-tripped. After the pump contact moniter trip has occurred, the following actions will permit single loop ope-:cion:

1. Reset the pump _ contact monitor t ower level trip setpoint to 55.0%.

2. Trip one of the two protective channels receiving outlet temperature information from sensors in the Idle Loop.

3. Reset flux-flow setpoint to 0.961.

REFERENCES (1) FSAR, Section 14.1.2.2 (4) FSAR, Section 14.1.2.3 (2) FSAR, Section 14.1.2.7 (5) FSAR, Section 14.1.2.6 (3) FSAR, Section 14.1.2.8 2.3-4

~.

. . P0wer Level.

-- 120 UN ACC EP T A BL E OPERATiOM (107)

. +

• l -- 100 d g

' I ACCEPTABLE *

// 4 PUMP l a OPERATION #,

g 80 (79.9)

^

(-28,79) (31,79) l i I ACCEPTA3LE l 3& 4 PUMP -- 60 I OPERATION I -

I '"I

(-2d.51.9) l (34.51.9) l I

. 40 ACCEPTABLE 2.3 & 4 PUMP l

(-28,24.4) (31,24.4) l

-- 20 g

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n n l nI n i i

1. 1 #1 , e , ,

-60 -40 -20 0 +20 +40 +60 Reactor P0ner Imualance, %

2'3-8 UNIT 1 PROTECTION SYSTEM MAXIMUM bb oko(Ogjft,gQPgg y

FIGURE 2.3-2A l

l l

l

Table 2.3-1A

• limit 1 Reactor Protective System Trip Settina Limita Two Reactor One Beactor Four Reactor Three Reactor Coolant Pumpe Coolant Pump Coolant Pumme Coolant Pumpe Operating in A Operating in Operating Operating Single Loop Each Loop (Operating Power (Operating Power (Operating Power (Operating Power Shutdown RPS Segment -100Z Rated) -75% Rated) -461 nated) -491 Rated) syge=

1. Nuclear Power Max. 105.5 105.5 105.5 105.5 5.0(3) )

(I Rated)

2. Nuclear Power Max. Based ,1.07 times flow 1.07 times flow 0.961 times flow 1.07 times flow Bypassed on Flow (2) and labalance, minue reduction minus reduction minus reduction minua reduction (1 Rated) due to imbalance due to imbalance due to imbalance due to imbalance

3. Nuclear Power Max. Based NA NA 551 (5)(6) 551 (5) Bypassed on Pump Monitors. (Z, Rated)

4. iligh Reactor Coolant 2355 2355 2355 2355 1720(4)

System Pressure, peig, Max.

u y 5. Iow Reactor Coolant , 1800 1800 1800 1800 Bypassed l p System Pressure, pals, Min.

6. Variable 1.ow Reactor III ( 11.14 Tout - 4706 )( (ll .14 T,,g- 4706 )(l} (l l .14 T,,g- 4706 ) Bypassed Coolant System Pressure

( 11.14 7 t 470$

pois, Min. .

7. Reactor Coolant Temp. 619 619 619 (b) 619 619 F., Man.  ;

8. High Reactor Building 4 4 4 4 4 Pressure, paig, Max.

(1) T" " le in degrees Fahrenheit (*F). (5) keector power level trip set point produced by pump contact monitor reset to 55.01.

(2) Reactor Coolant System Flow, I.

(6) specification 3.1.8 applies. Trip one of the (3) Administratively controlled reduction set two protection channele receiving outlet temper-only during reactor shutdown. eture information'from sensors in the idle loop.

(4) Automatically set when other segments of the RFS are bypassed,

_ - _ _ - _ - _ -___ - -- - - - -