ML19338C784

Proposed Revision to Tech Specs,Section 2 Re Safety Limits & Limiting Safety Sys Settings

ML19338C784
Person / Time
Site:	Rancho Seco
Issue date:	08/11/1980
From:	SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
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ML19338C783	List: ML19338C781 ML19338C784
References
NUDOCS 8009050241
Download: ML19338C784 (13)

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P.AflCHO SECO I TECilillCAL SPECIFICATl0!15 Safety Limits and Limiting Safety System Settings 2.

SAFETY LIMITS AtlD LlHITlflG SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies'to reactor thermal power, reactor. power imbalance, reactor coolant

~ system pressure, coolant temperature, and coolant flow during power operation i

of the plant.

Obj ecti ve To maintain the integrity of the fuel cladding.

Specifiestion 2.1.1 The combination of the reactor system pressure and c.colant temperaturc 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 within the restricted region the safety limit is exceeded.

2.1.2 The combination of reactor thermal power and reactor power imbalance (power in the top h'alf of the core minus the power in the bottom half

,of the core expressed as a percentage of the rated power) shall not a

l cxceed the safety limit as defined by the locus of points (solid 'line) for,the specified flow set forth in Figure 2.1-2.

If the actual-reactor-thermal power / reactor power-imbalance point is above the line l

for the specified flow, the safety limit is exceeded.

Bases The safety limits prpsented have been generated using BAW-2 critical heat flux and the actual measured flow rate (2).

This deve o ment is

.CilF) correlation \\I

(

l discussed in the Rancho Seco Unit 1, Cycle 2 Reload Report, reference 2 The flow rate utilized is 104.9 percent of the design flow (369600 gpm) based on i

four pump operation.

(2.3)

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i.

i 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 region of heat transfer, wherein the heat transfer coefficient is l

'large-enough so that the clad surface temperature is only.slightly greater than the coolant temperature.

The upper boundary of the nucleate boiling region is. termed " departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer 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

$909050'LM 2-1

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 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 DUB for axially uniform and non-uniform heat flux distributions.

The local DNB ratio (DNBR),

defined as the ratio 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.

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 difference 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 setpoints to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 104.9 percent of 369,600 gpm).

This curve is based on the combination of nuclear power peaking factors, with potential effects of fuel densification and rod bowing, which result in a more conservative DNBR than any other shape that exists during normal ope ra t ion.

The curves of Figure 2.1-2 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 melting at the hot spot.

The limit is 20.4 KV/ft.

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

The 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 shown in Figure 2.1-3 The curves of l

Figure 2.1-3 represent the conditions at which a minimum DNBR of 1.30 is predicted I

at the maximum possible thermal power for the assumed design flow, or the local 70 quality at the point of minimum DNBR is equal to 15 percent, whichever condition I

is more restrictive.

for Figure 2.1-3, a pressure-temperature point above and to the lef t of th 70 curve would result in a UNBR c a.ter than 1.30.

The 1.30 DNBR curve for four-2-2 Proposed Anendment No. 70

RAllCHO SECO Uf!!T 1 TECHillCAL SPECIFICATIO!!S Safety Limits and Limiting Safety System Settings pump operation is more restrictive than any other reactor coolant pump situation because any pressure /termperature point above and to the left of the four pump curve will be above and to the left of the other curves.

The maximum thermal power for three pump operation depicted in Figure 2.1-2 is 86.85 percent due to a power level trip produced by the flux-flow ratio 1.08 times 70 74.4 percent design flow = 80.35 percent power plus the maximum calibration and instrumentation error.

The maximum thermal power for other coolant pump conditions is produced in a similar manner.

The actual maximum power levels are calculated by the RPS and will be directly proportional to the actual flow during partial pump operation.

Re fe re nce s (1)

Correla tion of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, tiarch, 1970.

(2)

Rancho Seco Unit 1, Cycle 2 Reload Report BAW.

(3)

Rancho Seco Unit 1, Cycle 3 Reload Report BAW-lh99, September, 1978.

2-3 Proposed Anendnent Ib. 70

t e

. Figure 2.1-1.

C' ore Protection Safety Limit, Pressure Vs Temperature, 2400 co 2200 w

C1 sk U

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Region 70 o

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t 560 580 600 620 640 Reactor Outlet Temperature. F Proposed Anendment No. 70 i

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Proposed Amendment No'. 70

1 FIGURE 2.1-3.' Core Protective Safety Bases

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86.85 Three (DNBR Limit)

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3 187986 58.88 One in each loop 70 (quali ty 1imi t) i Proposed Change No. 70 RANCHO SECO UNIT 1 g.s,;U "g Q"

TECllNICAL SPECIFICATIONS SACRAMENTO MUtilCirAL UT!!.lTY DISTRICT

f1ANCHO SECO UNIT 1 TECHNICAL CPECIFICATIONS Safety Limits and Limiting

(_

Safety System Settings 2.3 LIMITING SAFETY SY{TDI SETTINGS, PROTECTIVE INSTRlR4ENTATION Applicability Applies to instruments monitorin5. reactor pnuer, reactor power imbalance, reactor coolant system pressure, reactor em lant outlet temperature, flow, number of pumps in operation, and high Reactor Building pressure.

Objective To provide automatic protection action to prevent any combination of process variatics from exceeding a safety limit.

Specification 2.3.1 The reactor protection system trip setting limits and the permissible

. bypasses for the instrumen't channels shall be as stated in tabic 2.3-1 and figure 2.3-2.

Bases The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if

\\.

any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.

The trip setting limits for protection system instrumentation are listed in table 2.3-1. The safety analysis has been based upon these protection system instrumentation trip set points plus calibration and instrumentation errors.

Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.

During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 105.5 percent i

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2-5

Ib4NCllO SECO UNIT 1 TECHNICAL S?ECIFICATIONS -

Safety Limits and Limiting Safety System Settings of rated power. Adding to this the possible variation in trip set points due

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to calibration and instrument errors, the maximum actual pcuer at which a trip would be actuated could be 112 percent, which was used in the safety analysis.(4)

A.

Overpower trip based on flow and imbalance The power Icyc1 trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been estab-lished to accommodate the most severe thermal transient' considered in the design, the loss-of-coolant flow accident from high power.

The analysis in section 14 demonstrates the adequacy of the specified power-to-flow ratio.

The power icvel trip set point produced by the power-to-flow ratio provides both high power Icyc1 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 D D 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 108 pctcent and reactor flow rate is 100 percent, or flow rate is 92.59 percent and power level is 100 percent.

2.

Trip would occur.when three reactor coolant pumps are operating if power is 80.35 percent and reactor _ flow rate is 74.4 percent 70 or flow rate is 69.44 percent and power level,is 75 percent.

3.

Trip would occur when one reactor coolant pump is operating in

'cach loop (total of two pumps operating) if the power is 52.38 percent and reactor flow rate is 48.5 percent or flow rate is 45.37 percent and the power level is 49 percent.

l.

l i

For safety analysis calculations the maximum calibration and instrumentation prrors for the power icvel were.used.

j 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 D:iBR limits. The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power Icvc1 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 icvel trip and associated reactor power reactor power-imbalance boundaries by 1.08 70 3

percent for a l' percent flow. reduction.

1 2-6 Proposed Amendment No. 70 y

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RAtlCl10 SECO UtilT l' TEClittlCAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings B.

Pump monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 L

by tripping the reactor dt.e to (a) the loss of two reactor coolant pumps In one reactor coolant icop, and (b) loss one one or two reactor ccolant pumps during two-pump operation.

The punp monitors also restrict the power level to 55 percent for one reactor coolant pump operation in each loop.

t C.

Reactor coolant system pressure 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 nuclear 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. (l i The low pressure (1900 psig) and variable' low pressure (12 96 T

- 5834) out trip set point shown in figure 2.3-1 have been established to main,tain the DNB ratio greater than or equal to 1.3 fbr those design accidents that result in a pre'ssure reduction. (2,3)

Due to the calibration and instrumentation errors the safety analy-sis used a variabic low reactor coolant system pressure trip value

'of (12.96 T

- 5884).

out D'

Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (619 F) shown in figure 2.3-1 has been established to prevent

' excessive core coolant temperatures in the operating range.

Due to calibration and instrumentation errors, the safety analysis used i

a trip set' point of 620 F.

1 E.

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, even in the absence of a low reactor coolant system pressure trip.

F.

Shutdown bypass In order to provide for control rod drive tests, zero power physics testing, and startu'p procedures, there is provision for bypassing certain segments of the reactor protection systen.

The reactor protection system segments which can be bypassed are shown in 2-7 e

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TECilillCAL SPECIFICATl0!iS Safety Limits and Limiting Safety' System Settings tabic 2.3 1.

Two conditio.

are imposed when 'the bypass is used:

1.

By administrative control the nucicar overpower trip set point must _be reduced to a value 15.0 percent of rated power during reactor shutdown.

A high reactor coolant system pressure trip set point of 1820 2.

psig is automatically imposed.

The purpose of the 1820 psig high pressure trip set point is to prevent nornal 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 overpower trip set point of 15.0 percent prevents any significant reactor power from being produced when performing the' physics Sufficient natural circulation (5) would be available to remove i

tests.

5 0 percent of rated power if none of the reactor coolant pumps were operating.

REFEREllCES (1)

FSAR, paragraph 14.1.2.2.

(2)

FSAR, paragraph 14.1.2.7 i

(3)

FSAR, paragraph 14.1.2.8 FSAR, paragraph 14.'i.'2.3 (4)

(5)

FSAR, paragraph 14.1.2.6 i

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TABLE 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS One Reactor Coolant Pums Tour Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop Shutdewn Opera ting (Nominal Operating (Nominal (Nominal Operating Bypass Operating Power - 100% )-

Operating Power - 75%)

Power - 49%)

1.

Nuclear power, % of rated, max.

105.5 105.5 105.5 5.0 I3)

I) 2.

Nuclear power based on flow 1.08 times flow minus 1.08 times flow minus 1.03 times flow minus Bypassed l}

and Imbalance, % of rated, nax.

reduction due to reduction due to

' reduction due to Imbalance (s)

Imbalance (s)

Imbalance (s)

3. ' Huc1 car power based on " pump NA NA 55 Bypassed monitors, t of rated, max.

4 High reactor coolant 2355 2355 2355 1820I4) system pressure, psig, max.

5.

Los reactor coolant system 1900 1900 1900 Bypassed pressure, psig, ntn.

6.

Variabic low reactor coolant 12.96 T

- 5834 12.96 T

- 5S34 12.96 T

- 5334 Bypassed out out out system pressure, psig, min.

7.

Reactor coolant temp. F., max.

619 619-619 619 8.

High Reactor Building 4

4 4

,4 pressure, psig, max.

(1)

T is in degrees Fahrenheit (F).

(2)

Reactor coolant system flow, %.

(3)

Administratively controlled reduction set only during reactor shutdown.

(4)

Aut6mitically set when other segments of the RPS (as specified) are bypassed.

(5)

The pump monitors also produce a trlp on: (a) loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two pump operation.

2-9 '

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