ML20008F605

Proposed Changes to Tech Specs 2.1 & 2.3 for Reactor Core Safety Limits & Limiting Safety Sys Settings Protective Instrumentation

ML20008F605
Person / Time
Site:	Rancho Seco
Issue date:	04/17/1981
From:	SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
Shared Package
ML20008F604	List: ML20008F603 ML20008F605
References
TAC-43369, NUDOCS 8104210368
Download: ML20008F605 (10)

Text

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Figu re 2.1-2 CORE PROTECTION SAFETY LIMITS, REACTOR POWER IMBALANCE (CYCLE 5)

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$SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT Proposec Amendment No. 76 Rev. 1

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FIGUP.E 2.1-3 Core Protective Safety Bases

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FM CHO SECO UNIT I TECHt:lCE SPECIF.lCATIC!!$

Safety I.inits and Limiting Safety Systen Settings pu o coeration is rere rettrictive than any other reactor ecelant cu p situation because any p essu e/te,erature coint aeeve and to the left of the four pump curve will be abo e anc to the left of the other curves.

The maximum ther al power for three pump cperation depicted in Figure 2.1-2 is 83.65 percent due to a pet.er level trip produced by the flux-fica ratio 1.06 tires 76 74.4 percant d sign fica = 78.86 percent po..er plus the v aximun calibratica and ins tes. en ta t ic, e rror.

The Nxir um thermal ec..er for other coolant pu o concitions Is produced in a similar canner.

The actual eaxieun ;caer. levels are calculated by the RD5 and will be directly prcpertional to the actual ficw curing partial pump c;eration.

References (1)

Correlation of Critical Heat Flux in a Bundle Coo!ad by Pressurized Vater, BAV-10000,tiarch, 1070.

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(2)

Rancho Secc Unit 1, Cycle 2 Relcad Repcrt BAV.

(3)

Rancho Seco Unit 1, Cycle 3 Reload Report BAV-14co, September, 1978.

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l Proposed A=endment No. 75,' Rev. 1 l

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS j

Safety Limits and Limiting Safety System Settings

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2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION AcolIcabilIty l

Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high Reactor Building pressure.

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

I Specification 2.3.1 The reactor protection system trip setting limits and the permissible bypass for the instrument channels shall be as stated in table 2.3-1 and figure 2.3-2.

I 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.

Tha trip setting limits for protection system instrument'ation 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 104.9 percent.

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2-5 Proposed Amendment No. 76, Rev. 1 I

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATICNS Safety Limits and Limiting Safety Systen Settings of rated power. Acding to this the possible variation in trip set points due to calibration and instrument errors, the maximum actuai power at wnich a trip wouldbeac{fatedcouldbe112 percent,whichwasusedinthesafety I

analysis.

A.

Overpower trip based on flow and imbalance The power level trip set point produced by the reactor coolant system flow is based en a power-to-flow ratio which has been estab-lished to accommcdate 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 level trip set point produced by the power-to-flow ratio provides both high power level and Icw 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 DNS protection for all modes of pump operation.

For every flow rate there is a maximum permissible low flow rate. Typical power level and Icw flew 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 106 percent and reactor flow rate is 100 percent, or flow rate is 9L.34 percent and power le el is 100 percent.

2.

Trip would occur when three reactor coolant pumps are operating if power is 78.8 percent and reactor flow rate is 74.4 percent or flow rate is 70.75 percent and poaer level is 75 percent.

3 Trip would occur when one reactor coolant pumo is operating in each loop (total of two pumps operating) if the power is St.k percent and reactor flow rate is 43.5 percent or flew rate is L6.22 percent and the pcwer level is 49 percent.

i For safety analysis calculat?ons the maximum calibration and instrumentat on errors for the power level were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These tFermal limits are either pover' peaking kW/ft limits or DNSR limits. The reactor power imbalance (power in the top half of core minus pcwer in the bottem half of core) reduces the sewer level tric produced by tne power-to-ficw ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the pcwer level trip and associated reactor-power reactor-power-imcalance boundaries by 1.06 l

percent for a 1 cercer.t ficw reduction.

2-6 Freccsed Amencment No. 76, Rev. I

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Safety Lini:s and Litteing Sof :y Sye:c Se::ings 3.

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The rue, : nit:ra :reven: :he =inimu= ::re :::37, f ro-decre: sire 5:1. 1.3 by trippit; :he ::::: r duc :: () the less Of :ve rose::: : c::n: punpn in one re:::Or : 01:n: loop. and (b) I:ss of car er two reac:or rol n:

pu ps during tve-pu=; Operati:n.

The pun; ::nt: Ors also res:rie: :ha power 1: eel '.o 55 ;cr:en: for onc rea::c: :: lan: pump oper::i:n in ench 100p.

C.

Reac:c: : ol a: cys:e= pressurs During a scar:up acciden: frc= icu peu:: or a si:v ::d vi:hdrawal frc:

high power, :he sys::= high pressure trip se: point is rasched bafore the nucicar overpcuet : rip se: poin:.

The : rip se::ing 11=i: shown in figure 2.3-1 for high rece:or coolan: systa= pressure (2300 psig) has been es:ablished :o esin:ain :he sys:c= pressure bel:w :he safe:y li=it (2730 psig) for any design :tansien: (1) and =ini=1:e :he challenges to the EH0V and code safe:iss.

The low pressure (1900 psig) and variable 1:v pressure (12.96 T

- 353')

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trip set point show= in figure 2.3-1 have been es:ablished :o

=aintain :he DNS ratio gr2a:cr chan or equal :o 1.3 for those design

' acciden:s :ha: resul: in a pressure redue:icn.

(2,3)

Due to the calibration and instru=ents: ion errers :he safety anclysis used a variabic lov reacror 00 clan: sys:c= pressure : rip value of (12.96 T

- 5834).

out D.

Coclant cu:le

c=pers:ure The high reae:ce c clant ou:le: :e:perature : rip set:ing 11=1: (618 F) shown in figure 2.3-1 has been es:ablished :o preven: encessive core coolant :c=peraturas in the oper::ing range. Due :o calibra:ica and instru=cn:ation errors, the safety analysis used a trip se: poin:

of 620 F.

j E.

React:r 3uilding p:2ssure The high Reactor Suilding pressure : rip se: in; 11:1:' (4 psis) provides l

pesi:ive assurance : hat a reactor : rip vill Oc:ur in :he unlikely event of a s:::= line f ailure in :he Rese:Or 3uilding or a less-of-l coolant acciden:, even in the assence of a lou reac:or : clan: system pressure : rip.

F.

Shutdown bypass In order :o provide for con:rol red drive ::s:s, :cro pcuer physics testing, and startu, precedures, :!. ore is provision for *,ypassing cer:nin scgeen:s of the rene:Or pro:e :icu sys:c=.

The rc:c:cr protect:en ysten segmen:s wnich can be '0ypassed ar shcun in 3i Proposed A=endmen: 30. 76, Rev. 1

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P00R ORIGINAL

7 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Design Features REFERENCES (1)

FSAR table 3 2-1 (2)

FSAR table 3.2-2 (3)

FSAR paragraph 3.2.4.2 (4)

FSAR paragraph 4.1.3 (5)

FSAR paragraph 4.1.2 9

5-5

.