ML20138P376

Proposed Tech Specs 3/4.3.6,3/4.3.7.6 & 3/4.9.2,lowering source-range Monitor Downscale Rod Block Setpoint

ML20138P376
Person / Time
Site:	Fermi
Issue date:	12/23/1985
From:	DETROIT EDISON CO.
To:
Shared Package
ML20138P373	List: ML20138P372 ML20138P376
References
VP-85-0236, VP-85-236, NUDOCS 8512260078
Download: ML20138P376 (7)

Text

--

?

s m

_g.

Attcchesnt I.

Pega i of 4 VP-85-0236 Proposed Change ~to

. Detroit Edison Technical Specifications Regarding Lowering the Source Range Monitor Downscale Rod Block Setpoint ip Technical Specification'Affecte'd:

3/4.3.6 Instrumentation / Control Rod Block Instrumentation 3/4.3.7.6 Instrumentation / Source Range Monitors

-3/4.9.2 Refueling Operations / Instrumentation l Discussion The' surveillance requirements of 4.3.7.6.c require that prior to withdrawal.

of control rods,-the Source Range Monitor' count rate be verified to be at lea'st'0.7 counts per second.(cps) with the detector fully inserted. Also, the surveillance requirements.of 4.9.2.c require that. ' prior to and during core alterations, the SRM count rate be verified to be at least 0.7-cps.

In addition,~ Table. 3.3.6-2 requires a SRM downscale trip set point of 0.7

.-cps for.the, control rod block function to'be. considered operable'.

JBased on the current SRM count rate, Detroit Edison estimates that the antimony-beryllium source ~ strength may be insufficient to maintain O.7.

, cps by-February 15, 1986, due to normal decay of the sources.

~

The normal. control rod withdrawal' scheme is inefficient for source regeneration until after the low power setpoint is reached, which

corresponds.to about 20% core thermal = power. -For a " Group Notch Plant" like Fermi-2 the most conservative possible withdrawal scheme past the 50%

control rod density does not reliably regenerate these sources until the reactorspower level is in excess'of 20 to 25% core' thermal power.

Other.means of meeting the 0.7 cps requirement include installing new

-sources which would result in' delays in the startup test schedule. The

' delay is due mostly to the. fact that the plant would need to enter.into the refueling mode =and open the reactor vessel and remove part of the fuel in order.to replace the' sources. Therefore, Detroit Edison proposes to temporarily lower the downscale SRM rod block trip setpoint count rate from

-0.7 cps to 0.3 - cps. The normal decay of the antimony-beryllium source.

' strength to 0.3 cps allows Detroit Edison an additional 8 weeks before the rod block setpoint is reached. Operating with a setpoint of 0.3 cps would

.be allowed until a burnup-of 2000 MWD /T on the first core (which.is slightly less than.100 full power! days) is achieved. This burnup restriction.is proposed to allow sufficient flexibility in operating time andfpower levels to reliably-regenerate the asutron sources.

A B51226007e SS p.PDR ADOCK O 341 PDR M

P

?

Attachment I Page 2 of 4 VP-85-0236 The~followingfactorsjustifya:rodblocksetpointcountrateof0.3 cps:

1)

The SW(s are not required to perform any protective or mitigative

' safety-function in the transients or accidents analyzed in Chapter 15 of.the FSAR.

'2)! iThe SRMs have been demonstrated.to be capable of monitoring" count rates as low as 0.1 cps and maintain an acceptable signal-to-noise ratio.

~

Asidescribed in FSAR Section 7.6.1.12 the SRMs provide neutron flux -

information'during reactor startup at low level flux operations until.the L IRMscare well1up on scale (Range 3 of IRMs). The SRMs also provide an 5 -

upscale rod block at 10 cps and a downscale rod block trip setpoint at 0.7 cps. These rod blocks prevent control rod withdrawal until the cause of-high or low count 1 rates is determined by the operator. However, the SRMs are not required to perform any protective or mitigative safety function in the transients or accidents analyzed in Chapter 15 of the FSAR. General-Electric has reviewed this information and concurs that the SRM perform no mitigating. functions for accidents.

The only!important consideration in lowering the minimum count rate requirement and-the downscale rod block is that sufficient monitoring capability be maintained to detect positive reactivity insertions from the initial subcritical condition in a smooth and continuous fashion.

Important to the verification of this monitoring capability is providing

-assurance that:

(1) the' reading is well on scale and (2) the signal-to-noise' ratio is greater than 2 to 1.

It has been experimentally verified that the SRMs are capable of measuring 0.1 cps. After calibration of the SRMs, the noise level is below 0.1 cps. Therefore, the minimum

~

-count rate of.0.3 cps is sufficient to maintain an SRM signal-to-noi'se ratio of at least 2 to 1.

The proposed value of 0.3 cps for the trip Lsetpoint while maintaining a signal-to-noise greater than 2 is on scale and

.provides adequate-neutron monitoring.

With regard to reactivity addition transients, the limiting fault at low power conditions is the Rod Drop Accident (RDA) which is analyzed by General Electric (GE) in NED0-10527 and its supplements and is described in Section 15B.4-of the FSAR..It should be noted that no credit is taken for the-SRMsLin the Fermi 2 RDA analysis.

In-addition to the above considerations, the known operational characteristics of the initial core indicate that the_ temporary reduction to'the SRM downscale rod block trip setpoint does not present a hazard.

The reactor has been brought critical several times since initial criticality on June.21, 1985 and the reactivity characteristics of the unaltered core are well known.

L l-s k

l

~

A'ttachment I Page 3 of 4 VP-85-0236 Significant Hazards Analysis

As' stated'in.10 CFR 50.92(c), a: proposed amendment-to an operating license

. involves no 'significant hazards consideration if operation of the facility

~

- in accordance with the proposed-amendment would not:

(1) involve a

~

significant-increase in the probability or consequences of an accident previously evaluated,((2)' create.the possibility of.a new-or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.'

The: proposed revisions to Fermi 2 Technical. Specifications 3/4.3.6, F

. 3/4.3.7.6 and 3/4'.9.2 would lower the SRM downscale rod block setpoint and allow operation with.the SRM's at a lower count rate.

Since the SRM's are

+

- not-required-to perform any protective'or mitigative safety function and

-are fully operable at the lower count rate, this proposed change does not

~

involve a significant increase in the probability or consequences of an accident previously evaluated.-

~_ The. proposed additions'do not create the possibility of a~new or different

- kind of accident from any accidents previously evaluated.. The proposed setpointfis well within'the demonstrated operating range of the SRM's to detect neutron levels in the' core. Since the SRM' channels with the lower setpoint will provide accurate. neutron flux information to the operator T

' 'during rod withdrawal.and have-no effect on RDA analysis results, new or

-different accident conditions are not created.

The-proposed SRM setpoint change does not -involve any reduction in margins of safety since the SRMs perform no mitigating function for accidents. In

- addition the SRMs have been demonstrated to reliably operate'at count rates

- lower'than the proposed change (as low as 0.1 cps).

Summary Based on the justifications.given above which show that SRMs have no safety

. function, and that the SRMs are ~ capable of monitoring count rates as low as

- 0.1 eps, Detroit Edison believes that lowering the-SRM rod. block trip setpoint' count rate from 0.7 cps to L.3 cps until the first core has been

~

irradiated to a burnup 2000 MWD /T is acceptable.

Given-the justifications stated above, it has been determined that these changes to the Technical Specifications do not involve a significant reduction in' safety margins since the SRMs are not required to perform any

- protective or mitigative safety function. Also, no increase in the probability.or consequences of an accident previously evaluated is involved nor is the possibility of a'new'or different kind of accident from any

~

- accident previously evaluated created. Thus the proposed changes to the

- Technical Specifications do not involve any significant hazards considerations.

y m,,.

~, -

.f...

Attachment I Page 4 of 4 VP-85-0236

Reduction inxSRH rod block setpoints to 0.7 cps has been reviewed and found

. acceptable.by the NRC infa prior case. In this case the NRC found that the

-lowering'the SRM setpoint was acceptable if it could be demonstrated the SRMs were operable at the new setpoint and did not involve a significant hazards consideration.

LSpecifically, a reduction to 0.3 cps was previously_ requested by'the

' Nebraska Public Power District by a letter dated April 4, 1974.- Nebraska Public~ Power District. (NPPD) requested a temporary waiver-of the Technical Specification minimum count rate to 0.3 cps for source range monitors during startup. NPPD requested the waiver to allow operational' flexibility

'during the latter stages of startup operations _since delays in the startup schedule resulted in decay of the neutron sources.- This had resulted in a concern that the minimum count rate requirement would not be met on: reactor restart..NPPD subsequently received approval for the requested Technical Specification amen' ment for Cooper Nuclear-Station (Moore to Pilant letter, d

' dated; April 17, 1974).-

2

r

~

I

{,

i REFUELING OPERATIONS

}

g SURVEILLANCE REQUIREMENTS (Continued) 9 b.

Performance of a CHANNEL FUNCTIONAL TEST:

1.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of CORE ALTERATIONS, and 2.

At least once per 7 days.

c.

Verifying that the channel count rate is at least 0.7* cps:

1.

Prior to control rod withdrawal,-

2.

Prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS, and 3.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l-d.

Verifying, within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during,.that the RPS circuitry " shorting links" have been removed during the time any control rod is withdrawn ** unless adequate shutdown margin has demonstrated per Specification 3.1.1.

)

• ' THE cocer SLATE mAy ee_ REbocEb To O.3 coomT5 PEJL SEcouD

'P oon. To kAne viitis a Bu""uP or 2000 nisuoV on 7"'

P'UT c o c e. Pee vi o E o r u a.

S tGNAz -To-ocesE_ RAT /o E 2.. AFTEX A B00/00P

. of: F_cco Mu)D/T T masr ee AT L. EAST O,7 CPS PPWIDED

,,, TH4,HWA4SP-b'he_ croor PATE.

SE car /o is 2 2. oruecutsq 3ces, 14 s iW15d/A1NidikbflMt'tWiM14/A%IlltlVMtWWIM6l,( tlhVf(/

M

• • Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

g FERMI - UNIT 2 3/4 9-4 3

' TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS' g

-TRIP FUNCTION-

~ TRIP'SETPOINT-

' ALLOWABLE VALUE e

1.

ROD BLOCK MONITOR E-a.

Upscale

< 0.66 W + 40%*

< 0.66'W + 43%*

Z b.

Inoperative NA NA',,

m c.

Downscale'

> 5% of RATED THERMAL POWER

. _.3% of RATED: THERMAL POWER 2.

APRM a.

Flow Biased Neutron Flux - High

< 0.66 W + 42%*

'< 0.66 W.+ 45%*

b.

Inoperative WA NA c.

Downscale

> 5% of RATED THERMAL POWER

> 3% of RATED THERMAL POWER d.

Neutron Flux - Upscale, Setdown 514%ofRATEDTHERMALPOWER 5 12% of RATED THERMAL POWER 3.

SOURCE RANGE MONITORS a.

Detector not full in NA NA.

b.

Upscale 5 1.0 x 105 cps 5 1.6 x.10s cps w

c.

Inoperative NA NA 1

d.

Downscale

> 3 cps **

> 2 cps **

T 4.

INTERMEDIATE RANGE MONITORS a.

Detector not full in NA NA b.

Upscale

< 108/125 divisions of

< 110/125 divisions of Tull scale Tull scale c.

Inoperative NA NA~

d.

Downscale

> 5/125 divisions of

> 3/125' divisions of 5.

SCRAM DISCHARGE VOLUME uH scale YuH scale a.

Water Level-High

< 589'11 "

< 591'0" b.

Scram Trip Bypass NA NA-t 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale

< 108/125% of rated flow

< 111/125% of rated flow b.

Inoperative NA '

NA c.

Comparator 5 10% flow deviation

. < 11% flow deviation 7.

REACTOR MODE SWITCH SHUTDOWN POSITION NA NA

• The rod block function is varied as a function of recirculation loop drive flow (W).

The trip setting of this function.aust be maintained in accordance with Specification 3.2.2.

• • Ydf lW4 /Wddddd /fd /s /7 /dg/ #Ayyded /tyd /dfdANWW/Al#dd/fWf6/ i#/N A.

M E. D ot00 5cALE. Rph SE.TibtA/T CcO/0T8 ATE. /rMV BE. ~REbocED Tb O.3 c.ps PR.rA Tu M/EutMGy A b u oP of: 2cco Mv/D on THE F:sRS7 core. PROVIDED THE. SIGNAL.-Tb-AjoiSE. ratio l$ > 2., MTElt i

egesuis oF 2 cc MWD T' oM ThE FIR 5r ccWE., TtW. Cccor ftATE. tyMY OEF. Rebec.EO 7t> O,7 c.PS PRcVdf.D M S\\ M -To~CotSt.GbTio R 2.,

r e' i -

p INSTRUMENTATION SOURCE RANGE MONITORS

}

l

[

LIMITING CONDITION FOR OPERATION 3.3.7.6 At least the following source range monitor channels shall be OPERABLE:

a.

In OPERATIONAL CONDITION 2^, three.

b.

In OPERATIONAL CONDITIONS 3 and 4, two.

APPLICABILITY: OPERATIONAL CONDITIONS 2*, 3, and 4.

ACTION:

a.

In OPERATIONAL CONDITION 2* with one of the above required source

' range monitor channels inoperable, restore at least 3 source range monitor channels to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hoars.

b.

In OPERATIONAL CONDITION 3 or 4 with one or more of the above required source range monitor channels inoperable, verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within I hour.

SURVEILLANCE REQUIREMENTS

)

4.3.7.6 Each of the above required source range monitor channels shall be demonstrated OPERABLE by:

a.

Performance of a:

1.

CHANNEL CHECK at least once per:

a) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in CONDITION 2*, and b) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in CONDITION 3 or 4.

2.

CHANNEL CALIBRATION ** at least once per 18 months.

b.

Performance of a CHANNEL FUNCTIONAL TEST:

- 1.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to moving the reactor mode switch from the Shutdown position, if not performed within the previous 7 days, and 2.

At least once per 31 days.

c.

Verifying, prior to withdrawal of control rods, that the SRM count rate is at least 0.7*** cps with the detector fully inserted.

"With IRM's on range 2 or below.

• • Neutron detectors may be excluded from CHANNEL CALIBRATION.

• • • FMWf 66W M#%3 ttl4tM4Mt( tM% (iM(E f4( l D%hqfM1W,t 4 tfpy TRt_ c.ovmr ram may em. POOCEE) To Ol3 c_ps PRcR. To g

Ac.M10hsV6 A' bcRJocP eF 2Cc0 ND/T etc THE_ f:iRST Coke.

PRcVetn Tkt. GUAR-76-No65E. FArto > 2.s AFrER A eeReueP es: Ecco t%AD/T, Tk NYkno I.

ne E 3 c 94, r

O

.