Text

Proposed Tech Spec 3.8/4.8 Re Radioactive Effluents & 3.10/4.10 Re Refueling
	ML20079F992
Person / Time
Site:	Quad Cities
Issue date:	09/30/1991
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20079F990	List: ML20079F987; ML20079F992;
References
	NUDOCS 9110080355
	Download: ML20079F992 (118)
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7 PROPOSED -TECH SPEC l

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' RADIOACTIVE EFFLUENTS' I

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QUAD CITIES UNITS 1 &-^2 DPR-29-&_DPR-30 3.8/4.8 RADIOACTIVE EFFLUENTS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

.A.

Explosive _ Gas Mixture-A.

Explosive Gas Mixture

-- 1 The concentration of hydrogen in At least once per shif t verifi-the Off-Gas. hold ~up system,.

cation will be made that the unit downstream-of-the recombiner is operating within the allowable shalli be limited by having a band of the base-line plot of re-recombiner ' OPERABLE ' within the combiner-outlet temperature vs.

tallowable band of the base-line reactor power.

_ plot of.

recombiner outlet-temperature vs. reactor pcwcr.

APPLICABILITY:

OPERATIONAL MODES 1

and 2,

whenever REACTOR VESSEL PRESSURE is greater than~900 psig.

ACTION:

1.

With the required recombiner inoperablei restore a recom-biner.to. OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

2.

With either the recombiners inoperable,.or all; charcoal

-beds bypassed for.more than 7' days in _a calendar quarter-

while operating above 30 percent of. RATED -THERMAL

POWER, prepare.and submit-to the Commission within 30 days, pursuant to Specifica-

-tion 6.6.B.3, a

Special zReport which-- includes the following information:

'a.

' Identification of the defective equipment.

b.

Cause of the-defective equipment.

-c.

Action (s) taken to re-3.8/4.8-1

= -. -

s

.o QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30

. store the equipment to Lan-operating status, d.

Length of time the above requirements were.not satisfied.

e.

Volume and curie content of the waste discharged which was not processed by the inoperable _

equipment but which required processing.

f.

Action'(s)-

'taken to

. prevent a

recurrence of equipment failures.

This. Special Report is in lieu-of a Licensee Event Report.

-3.-

The provisions of Specification 3.0.C are not applicable.

B.-

Main Condenser B.

Main Condensar The---release rate of the sum of 1..

The radioat t vity release the activities -from the noble cate of-noble-gases

-at'

gases'- measured ~'at the main (near) the: outlet _ of the condenser air ejector shall _ be main condenser air ejector LlimitedLto'less_thantor_ equal to shall be continuously moni-100 microcuries/sec per MWt tored -in accordance with the.

-(after 30 minutes decay),

provisions conteined in1the-ODCM.

' APPLICABILITY:

2.

The release rate of the sum Whenever the: main condenser air of the activities from noble ejector =is in-operation.

gases from the main _conden-ser air ejector : shall --be

, ACTION:

determined-to be within-the limits-of Specification

.With the release rate of the sum-3.8.B at the following fre-of the activities from noble quencies by performing an gases at the main condenser air isotope analysis of a repre--

_ ejector exceeding 100 microcuries sentative sample of gases

/sec per MWt1 (af ter ' 30 ' minutes taken at the recombiner out-3.8/4.8-2

.~. -.

=.-

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 decay), restoro the r acase rate let, or at the air ejector to within its limise within 72 outlet if the recombiner is

hours, or be in at least HOT bypassed.

STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a.

At least once per 31

days, b.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months follo-wing an

increase, as indicated by the SJAE Radiation Monitor, of greater than 50%,

after factoring out increases due to chan-ges in thermal power level and off-gas

flow, in the nominal steady-state fission gas release from the primary coolant.

C.

Mechanical Vacuum Pump C.

Mechanical Vacuum Pump The mechanical vacuum pump shall At least once during each be capable of being isolated and OPERATING

CYCLE, automatic secured on a

signal of main securing and isolation of tha steamline tunnel high radiation, mechanical vacuum pump shall be verified.

APPLICABILITY:

OPERATIONAL MODES 1,

2, and 3.

This Specification is applicable in OPERATIONAL MODES 1, 2,

and 3 only when the main steamline isolation valves are open.

ACTION:

1.

With the isolation and trip of the mechanical vacuum pump on a signal of main steamline tunnel high radiation inoperable, secure and isolate the mechanical vacuum pump within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

The provisions of Specifica-tion 3.0.C are not applicable, i

3.8/4.8-3

e QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 D.

Sealed Source Contamination D.

Sealed source contamination Each sealed source containing 1.

Test Requirements Each radioactive material either in sealed source shall be excess of 100 microcuries of beta tested for leakage and/or and/or gamma emitting material or contamination by:

5 microcuries of alpha emitting material shall be frec of 2 0.005 a.

The licensee, or microcuries of removable contamination, b.

Other persons speci-fically authorized by APPLICABILITY:

the Commission or an Agreement State.

At all times.

The test method shall have a ACTION:

detection sensitivity of at least 0.005 microcuries per 1.

With a sealed source having test sample, removable contamination in excess of the above limit, 2.

Test Frequencios Each withdraw the sealed source category of scaled sources, from use and either:

excluding startup sources and fission detectors pre-a.

Decontaminate and viously subjected to core repair the sealed flux, shall be tested at the source, or frequency described below; b.

Dispose of the sealed a.

Sources in Use At source in accordance least once per six with the Commission months for all sealed regulations.

sources containing radioactive material:

2.

The provisions of Specifica-tion 3.0.C are not 1)

With a half-life applicable.

greater than 30

days, excluding Hydrogen 3, and 2)

In any form other than gas, b.

Stored Sources not in Each sealed Use source shall be tested prior to the use or transfer to another licensee unless tested within the previous 6 months.

Sealed sour-l 3.8/4.8-4 l

s a

QUAD CITIES UNI'tS 1 & 2 DPR-29 & DPR-30 ces transferred with-out a certificate in-dicating the lhst test date shall be tested prior to being placed in use.

3.

Reports

-A Special Report shall be prepared and submitted to the Commission pursuant to Specification-6.6.B.3 if source leakage tests reveal the presence of 2

0.005 microcuries of removable contamination.

4.

A complete inventory of radioactive materials in the licensee's possession shall be maintained current.

E.

Control Room Emergency Filtration E.

Control Room Emergency Filtration System System The control room emergency 1.

At least once per month, filtration system, including at initiate 2000 cfm (i 10%)

least one booster fan shall be flow through the control OPERABLE.

room emergency filtration system for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months APPLICABILITY:

with the heaters OPERABLE.

When SECONDARY CONTAINMENT 2.

At least once per 1440 hours0.0167 days 0.4 hours 0.00238 weeks 5.4792e-4 months INTEGRITY is required by of system operation; or once Specification 3.7.I.

This speci-per OPERATING CYCLE but not fication is not applicable during to exceed 18

months, or post-maintenance

testing, or following painting, fire, or during removal of the charcoal toxic chemical release in test canister.

any "entilation zone communicating with the ACTION:

intake of the system while the system is operating that With the control room emergency could contaminate the HEPA filtration system inoperable:

filters or charcoal adsorbers, perform the 1.

In OPERATIONAL MODES 1,

2, following:

or 3, restore the inoperable system to OPERABLE status a.

In-place DOP test the within 14 days, or be in at I: EPA filter banks to least MOT SHUTDOWN within verify leaktight inte-3.8/4.8-5

t 0

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in at grity.

The results of least COLD SHUTDOWN within the in-place CDP tests the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

at 2000 cfm (i 10%) on llEPA f11ters shall 2.

When irradiated full or the show s 1% DOP penetra-fuel cask is being handled tion.

in-the reactor building or during CORE ALTERATIONS, or b.

In-place test the during operations with a po-charcoal adsorber tential for draining the banks with halogenated reactor vessel; suspend hydrocarbon tracer to

-these operations.

The pro-verify leak tight visions of Specification integrity.

The re-3.0.C are not applicable.

Sults of in-place halogenated hydrocar-bon tests at 2000 cfm (i 10%)

on the char-coal banks shall show s 11 penetration.

c.

Remove one carbon test canister from the charcoal adsorber.

The r'.sults of labora-

tor, carbon sample ar alysis shall show 2 SJ% methyl iodide re-moval ef ficiency when tested at 130'C and 95%

Relative ilumidity.

3.

At least onw ner OPERATING CYCLE, but not ts exceed 18

months, the

'ollowing conditions shaL1 be demonstrated a.

Pressure drop across the combined filters is less than 6 inches of water at 2000 cfm (x 10%) flow rate.

b.

OPERABILITY of inlet heater demonstrates heater delta T of 15'F.

4.

After any maintenance or testing that could affect 3.8/4.8-6

i e

l QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 testing that could affect the leak tight integrity of the IIEPA f11 tor or llEPA filter mounting

frame, perform in-place DOP tests on the llEPA filters.

The results of the in-placo DOP tests at 2000 cfm (1 10%) on llEPA filters shall shov 51%

DOP penetration.

5.

After any maintenance or testing that could affect the leak tight integrity of the charcoal adsorber banks, perform halogenated hydro-carbon tests on the charcoal adsorbors.

The results of in-place halogonated hydrocarbon tests at 2000 cfm (i 10%) shall show s 1%

penetration.

0 3.8/4.8-7

e e

i QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 3.8/4.8 RADIOACTIVE EFFLUENTS BASES A.

Explosive Gas Mixture This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the off gas system is minimized in conformanco with the requirements of General Design Criteria 60 of Appendix A to 10 CPR Part 50.

B.

Main Condenuer Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to a MEMBER OF THE PUBLIC at and beyond the SITE DOUNDARY will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment.

This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.

C.

Mechanical vacuum Pump The purpose of isolating the mechanical vacuum line is to limit the release of activity from the main condenser.

During an

accident, fission prodrets would be transported from the reactor through the main steamline to the main condenser.

The fission product radioactivity would be sensed by the main steamline radioactivity monitors which initiate isolation.

D.

Sealed Source Contamir.ation The objective of this specification is to assure that leakage from byproduct, source and special nuclear material sources does not exceed allowable limits.

The limitations on removable contamination for sources requiring Icak testing, including alpha

emitters, is based on 10 CPR 70.39(c) limits for plutonium.

B 3.8/4.8-1 l.

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QUAD CITIES Ul11TS 1 & 2 DPR-29 & DPR-30 E.

Control Room Emergency Filtration System The purpose of these specifications is to assure availability of the control room emergency filtration unit that has been i nst al.1 ed in response to NUREG-0737 Item III D.3.4.

G::aration of this unit is described in the Quad' Cities Updated FSAR Sections 10.10.4

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and 14.2.

B 3.8/4.8-2

EXISTING TECH SPEC TS 3.8/4.8

' RADIOACTIVE EFFLUENTS *

=

i QUAD ClllE5 DPR-29 3.8/4.8 RADI0 ACTIVE EFFLUENTS Limiting Conditions for Operation Surveillance Requirements Applicability:

Applicability:

Anplies t9 the radioactive effluents from Applies to the periodic reasurements of the plant, radioactive effluents.

Specifications A.

Ga aout Cff1':ents A.

Gaseous Effluents 1.

The Jos9 rate in unrestricted 1.

The dose rates due to ratio-ar.as (at or beyond the site active materials released in boundary, figure 4.8-1) due to gaseous offluents from the site radioactive materials released shall be determined to be within in gaseous effluents from the the prescribed limits by obtain-site shall be limited to the ing representative samples in fql'owing:

accordance with the sampling and analysis program specified in Table 4 8-1.

The dose rates are calculated using methods pre-scribed in the Off-Site Dose Calculation Manual (ODCM).

a.

For Noble Gases:

(1) Less than 500 mrem / year to the whole body.

(2) Less than 3000 mrem / year to the skin, b.

For iodine'131, for iodine-133, and for all

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radionuclides in particulate form with half-lives g eater than 8 days less than 1500 mrem / year.

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3.8/4.8-1 Amendment No. 114 i

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QUAD ClllES OPR-29 c.

If the dose rates exceed the above limits, without delay decrease the release rates to bring the dose rates within the limits, and provide prompt notification to the Commission (6.6.B.1) 2..

The air dose in unrestricted 2.

The air dose due to releases of areas (at or beyond the site radioactive noble gases in gas-eous ef fluents shall be deter-boundary) due to Noble Gases mined to be within the pre-released in gaseous effluents scribed limits by obtaining 'ep-r from the unit shall be limited resentative samples in accor-to the following:

dance with the sampling and analysis program specified in sections A and B of Table 4.8-1.

The allocation of ef-fluents between units having shared effluent control systems and the air doses are determined usthg methods prescribed in the ODCH at least once every 31 days, a.

For gamma radiation:

(1) Less than or equal to 5 mrad during any cal-endar quarter.

(2) Less than or equal to 10 mrad during any calendar

year, b.

For Beta radiation:

(1) Less than or equal to 10 mrad during any calendar quarter.

(2) Less than or equal to 20 mrad during any calendar year.

l 3.8/4.8-2 Amendment No. 114

1 i

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QUAD C111ES OPR 29 i

c.

With the calculated air dose i

from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 i

days,ifies the cause(s) fora Special Report which ident exceedingthelimit(s)and defines the corrective actions to be taken to ensure that future releases are in compliance with S.B.A.2.a & b..This is in lieu of a Licansee Event Report.-

- d.. With the calculated air dose i

from radioactive noble gases in gaseous effluents exceeding the limits of

' Specification 3.8.A.2.a. or-3.8.A.2.b., prepare and submit a Special Report to the Commission within 30 days and limit _the subse-quent releases such that the doses or dose commitment to a member of the public from all uranium fuel cycle sources is limited to less

.i than or equal to 25 arem to-the total body or any organ (except thyroid, which is -

limited to less than or equal to_75 mren) over 12 4

consecutive months. This Special Report shs11 include an analysis which demonstrates that radiation exposure to all members of the public;from all uranium fuel cycle sources'

-(including all effluent pathways and direct radiation) are i

3.8/4.8-3 Amendment No. 114

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_ ~ -. _ _ _, _ _... _ _. _ _. _. _..

e QUAD-CITIES DPR~29 lets than the 40 CFR Part 190 Standard. Otherwise, obtain a variance frcm the Commission to permit releases which exceed the 40 CFR Part 190 Standard.

The rediation exposure analysis contained in the Special Report shall use the me-thods prescribed in the ODCH. This report is in lieu of a Licensee Event Report.

3.

The dose to a member of the pub-3.

The dose to a member of the pub-lic in unrestricted areas (at or lic due to releases of iodine-beyond the site boLndary) from 131, iodine 133, tritium, and iodine-131, iodine-133 tritium, all radionuclides in particulate and all radionuclides In parti-form with half-lives greater culate form with half-lives than 8 days shall be determined greater than 8 days in gaseous to be within the prescribed lim-effluents released from the unit its by obtaining representative shall be limited to the fol-samples in accordance with the loving:

sampling and analysis program specified in Table 4.8 1.

Fer radionuclides not determined in each batch or weekly compo-site, the dose contribution to the current calendar quarter cu-mulative summation may be esti-mated by assuming an average monthly concentration based on the previous monthly or quar-terly composite analyses. How-ever, for reporting purposes, the calculated dose contri-butions shall be based on the actual composite analyses when possible.

The allocation of effluents between units having shared effluent control systems and the doses are determined using the methods prescribed in the ODCM at least once every 31 days.

3.8/4.8-4 Amendment No. 114

i e

e QUAD-CITIES i

DPR-29 l

a.

Less than or equal to 7.5 arem to any organ during any l

calendar quarter.

l b.

Less than or equal to 15 mrom to any organ during any-calendar year.

c.

With the calculated dose from the release of iodine-l 131.-iodine-133, tritium.

and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, a Special Report which identifies the cause(s) for.

exceeding the limit and defines the corrective actions taken and the proposed actions to be taken to ensure that future releases are in compliance with 3.8.A.3. a. & b.

This-is in lieu of a Licensee Event Repnrt, d.

With the calculated dose from the release of iodine-131, iodine-133, tritium, and all radionuclides in particulate form with t

half-lives greater than 8 days in gaseous effluents exceeding the limits of 7

Specification 3.8 A.3.a. or 3.8.A.3.b., prepare and-submit a Special' Report to the Commissibn within 30-days and limit subsequent l

releases such that the dose or dose commitment to a member of the public from L

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3.8/4.8-5 Amendment No. 114

e QUAD CITIES DPR-29 all uranium fuel cycle sources is limited to less ti. n or equal to 2$ aren to the total body or organ I

(except the thyroid, which is limited to less than or

-equal to 75 mres) over 12 consecutive months. This Special Report shall include an analysis which demonstrates that radiation exposures to all members of the public from all-urentum fuel cycle sources (incluJing all effluent i

pathways and direct radiation) are less than the 40 CFP Part 190 Standard.

Otherwise, obtain a variance from the Commission to permit releases which exceed the 40 CFR Part 190 Standard.

The radiation exposure analysis contained in the Special Report shall use the methods prescribed in the 00CM. This report is in lieu of a Licensee Event 1

Report.

4.

Off-Gas System _

4.

Off-Gas System a.

At all times during proces-Doses due to treated gases re-sing-for discharge to the leased to unrestricted areas at-environs, process and~ con-or beyond-the site boundary trol equipment provided to shall be projected at least once reduce the amount or con-per 31 days in accordance with centration of radioactive the ODCH.

materials shall be operated.

b.-

The above specification shall not apply for the Off Gas Charcoal Adsorber Beds-below 30 percent of rated thersal power.

l i-I i

3.8/4.8-6 6mendment No. 114 l

L,. -.

e QUAD C111ES OPR-29 5.

Explosive Gas Mixture b.

Explosive Gas Mixture a.

The concentraticn of hy-Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months verification drogen in the off gas hold will be made that the unit is up system, downstream of the operating within the allowable recombiner shall be limited band of the base line plot of by having a recombiner recombiner outlet temperature operable within the vs. reactor power.

allowable band of the base-line plot of recombiner outlet temperature vs. re-actor power, whenever the reactor is operating at a pressure greater than 900 psig.

b.

The recombiner may be inop-erable for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

6.

With either the recombiners in-operable, or all charcoal beds bypassed for more than 7 days in a calendar quarter while opera-ting above 30 percent of rated thermal power, prepare and sub-mit to the Commission within 30 days a special renort which in-cludes the follow lng information:

a.

Identification of the de-fective equipment.

b.

Cause of the defective equipment, c.

Action (s) taken to restore the equipment to an opera-ting status, d.

Length of time the above requirements were not sat-isfied.

l t

3.8/4.8-7 Amendment No. 114 l

l

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QUAD-CITIES OPR-29 e.

Volume and curie content of the waste discharged which was not processed by the inoperable equipment but which required processing, f.

Action (s) taken to prevent a recurrence of equipment failures.

This is in lieu of a Licensee Event Report.

7.

The release rate of the sum of 7.

The radioactivity rate of vioble the activities from the noble gases at (near) the outlet of gases measured at the main con-the main condenser air ejector denserairejectorshallbelim-shall be continuously monitored ited to less than or equal to in accordance with Specification 100 microcuries/sec per HWt 3.2 H.

The release rate of the (after 30 minutes decay) at all sum of the activities from noble times. With the release rate of ga.es from the main condenser the sum of the activities from air ejector shall be determined noble gases at the main con-to be within the limits of denserairejectorexceeding100 Specification 3.8.A.7. at the microcuries/sec per HWt (after following f requencies by 30minutesdecay)Ithinitslim-performing an isotope analysis restore the release rate to w of a representative sample of itswithin72 hours lthinthe or be in at gases taken at the recombiner least HOT STANDBY w out?et, or at the air ejector next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , outlet if the recombiner is

bypassed, a.

At least once per 31 days, b.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase, as indicated by the main condenser air ejector noble gas activity monitor, of greater than 50%, af ter f actoring out increases due to changes in thermal power level and off gas flow, in the nominal l

steady-state fission gas i

release f rom the primary i

coolant, 3.8/4.8-8 Amendment No. 114

1 l

QUAD-C111ES DPR-29 B.

Liquid Effluents B.

Liquid Effluents 1.

The concentration of radioactive 1.

The concentration of radioactive material released from the site material in unrestricted areas to unrestricted areas (at or be-shall be determined to be within yond the site boundary, figure the prescribed limits by 4.8-1) shall be limited to the obtaining the representative concentrations specified in 10 samples in accordance with the CFR Part 20, Appendix D Table sampling and analysis program 11, Column 2 with the Table specified in Table 4.8-3.

The 4.8-2 values representing the sample analysis results will be HPC's for noble gases.

used with the calculational methods in the ODCM to determine With'the concentration of radio-that the concentrations are active material released from within the limits of the site to unrestricted areas Specification 3.8.B.1.

exceeding the above limits, without delay decrease the re-lease rate of radioactive mate-rials and/or increase the dilu-tion flow rate to restore the concentration to within the above limits.

2.

The dose or dose commitment 2.

a.

The dose contributions from above background to a member of measured quantities of the public from radioactive ma-radioactive material shall terials in liquid effluents re-be determined by calculation leased to unrestricted areas (at at least once per 31 days or beyond the site boundary) tnd a cumulative summation from the site shall be limited of these total body and to the following:

organ doses shall be maintained for (ich calendar a.

Durir.g any calendar quarter:

quarter.

(1) Less than or equal to 3 mrem to the whole body.

(2) Less tht.

,r equal to 10 mrem to n.

organ.

3.8/4.8-9 Amendment No. 114 i

\\

s QUAD-CITIES DPR-29 b.

During any calendar year:

o.

Doses computed at the near-est community water system will consider only the drinking water pathway and shall be projected using the methods prescribed in the ODCH at least once per 92 days.

(1) Less than or equal to 6 mrem to the whole body.

(2) Less than or equal to 20 mrem to any organ.

c.

With the calculated dose from the release of radio-active materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commis-sion within 30 days a Spe-cial Report which identi-fies the cause(s) for ex-ceeding the limit (s) and defines the corrective ac-tions taken and the pro-posed actions to be taken to ensure that future re-leases are in compliance with 3.8.B.2.a. & b.

This is in lieu of a Licensee Event Report, d.

With the calculated dose from the release of radio-active materials in liquid effluents exceeding the limits of Specification 3.8.B.2.a. or 3.8.B.2.b.,

prepare and submit a Spe-cial Report to the Commis-sion within 30 days and limit the subsequent re-leases such that the dose or dose commitment to a member of the public from all 3.8/4.8-10 Amendment No. 114

QUAD-ClllES DPR-29 uranium fuel cycle sources is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is limited to less thaa or equal to 75 mrem) over 12 consecutive months. This Special Report shall in-clude an analysis which de-monstrates that radiation exposures to all members of the public from all uranium fuel cycle sources (includ-ing all ef fluent pathways and direct radiation) are less than the 40 CFR Part 190 Standard. Otherwise obtain a variance from the Commission to permit re-leases which exceed the 40 CFR Part 190 Standard.

The radiation exposure analysis contained in the Special Report shall use methods prescribed in the ODCH.

This report is in lieu of a Licensee Event Report.

e.

Withtheprojectedannual whole body or any internal organ dose computed at the nearest downstream comu-nity water system is equal to or exceeds A mrem from all radioactive materials released in liquid ef-fluents from the Station, prepare and submit a Spe-cial Report within 30 days to the operator of the com-munity water system. The 3.8/4.8-11 Amendment No. 114

e QUAD-CITIES DPR-29 report is prepared to assist the operator in meeting the requirements of 40 CFR 141:

EPA Primary Drinking Water Standards. A copy of this i

report will be sent to the.

NRC.

This is in lieu of a Licensee Event Report.

3.

At all times during processing 3.

Liquid Waste Treatment-prior to discharge to the envi-rons, process and control equip-a.

Doses due to liquid re-ment provided to reduce the leases to unrestricted areas amount or concentration of ra-(at or beyond the site-dioactive materials shall be op-boundary) shall be projected eratedwhentheprojecteddose at least once per 31 days in-due to liquid effluent releases accordance with the ODCM.

to unrestricted areas (see Fig-ute 4.8-1), when averaged over 31 days, exceeds 0.13 mrem to the total body or 0.42 arem to any-organ.

4.

If liquid waste has to be or is i

being discharged without treat-ment as required above, prepare and submit to the Commission within 30 days, a-report which includes the following informa-tion:

a.

Identification of the de -

fective equipment.

b.

Cause of the defective-equipment.

c.

At: tion (s) taken to restore the' equipment to an operat-L ing status.

L d.

Length of time the above L

requirements were not sat-isfied.

l l:

3.8/4.8-12 Amendment No. 114 l-

~ _. _., _ -..,,.

t

\\

QUAD-C]T]ES DPR-29 e.

Volume and curie content of the waste discharged which was not processed by the appropriate equipment but which required processing.

f.

Action (s) taken to prevent a recurrence of equipment failures.

This is in lieu of a Licensee Event Report.

C.

Mechanical Vacuum Pump C.

Mechanical Vacuum Pump 1.

The mechanical vacuum pump shall At least once during each operating be capable of being isolated and cycle, automatic securing and isola-secured on a signal of main tion of the mechanical vacuum pump steam high radiation or shall be shall be verified, isolated and secured whenever the main steam isolation valves are open.

D.

Environmental Monitoring Program D.

Environmental Monitoring Program 1.

The environmental monitoring 1.

The radiological environmental program given in Table 4.8-4 monitoring.;amples shall be col-shall be conducted except as lected pursuant to Table 4.8-4 specified below, f om the locations specified in the ODCM, and shall be analyzed pursuant to the requirements of Table 4.8-6.

2.

With the radiological environ-2.

The results of analyses per-mental monitoring program not formed on radiological environ-being conducted as specified in mental monitoring samples shall Table 4.8-4 prepare and submit' be sum'aarized in the Annual Ra-to the Commission, in the Annual diological Environmental Operat-Radiological Operating Report, a ing Report.

3.8/4.8-13 Amendment No. 114

e QUAD CITIES DPR-29 description of the reasons for not conducting the program as required and the plans for pre-venting a recurrence. Devia-tions are permitted from the re-quired sampling schedule if spe-cimens are unobtainable due to hazardous conditions, seasonal unavailability, contractor omis-sion which is corrected as soon as discovered, malfunction of sampling equipment, or if a per-son who participates in the pro-gram goes out of business.

If the equipment malfunctions, cor-rective actions shall be com-pleted as soon as practical.

If a person supplying samples goes I

out of business, a replacement will be found as soon as pos-sible. All deviations from the sampling schedule shall be de-scribed in the annual report.

3.

With the level of radioactivity 3.

The land use census shall be in an environmental sampling me-conducted at least once per dium at one cr more of the loca-twelve months between the dates tions specified in the ODCH ex-of June 1 and October 1 by a ceeding the limits of Table door-to-door survey, aerial sur-4.8-5 when averaged over any vey, road survey, or by con-calendar quarter, prepare and sulting local agriculture au-submit to the Commission within thoritie:.

30 days from the end of the af-fected calendar quarter, a Spe-cial Report which includes an evaluation of any release condi-tions environmental factors or other aspects which caused the limits of Table 4.8-5 to be exceeded.

This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event the condition shall be reported and described in the Annual Radio-logical Environmental Operating Report.

3.8/4.0-14 Amendment No 114

e QUAD-C111ES DPR-29 4.

With milk samples unavailable 4.

The results of the land use cen-from one or more of the sample sus shall be included in the An-locations required by Table nual Radiological Environmental 4.8-4, identify locations for Operating Report.

obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The locations from which samples were unavail-able may then be deleted from the monitoring program, in lieu of a Licensee Event Report, identify the cause of the un-availability of samples and identify the new location (s) for obtaining repistement samples in the Annual Radiological Environ-mental Operating Report and also incide in the report a revised figure ( O a M table for the ODCH rr lecting t t new location (s).

5.

A censow of."edrest residences 5.

The results of the analyses per-and cf kr.his moducing milk formed as part of the required for human co%umption shall be crosscheck program shall be in-condscted annually (during the cluded in the Annual Radio-grazinc season for animals) to logical Environmental Operating determine their location and Report.

The analyses shall be number with respect to the done in accordance with the 00CM.

Site. The nearest residence in each of the 16 meteorological sectors shall also be determined within a distance of five miles.

The census shall be conducted under the following conditions:

a.

Within a 2-mile radius from the plant site, enumeration of animals and nearest res-idences by a door-to-door or equivalent counting technique.

3.8/4.8-15 Amendment No. 114

e QUAD-CITIES DPR-29 b.

Within a 5-mile radius, enumeration of animals by using referenced informa-tion from county agricul-tural agents or other reli-able sources.

A.

6 ith a land use census identi-

'ying location (s) of animals m'hich yield (s)an ODCH calculated dou or dose commitment greater than the values current!y being calculated in Specification 4.8. A.3, the new location (s) shall be added to the radiolo-gical environmental monitoring l

program within 30 days, if pos-sible.

The sampling location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program af-ter October 31 of the year in which this land use census was conducted.

7.

Radiological analyses shall be performed on samples represen-tative of those in Table 4.8-4, supplied as a part of the Inter-laboratory Comparison Program which has been approved by the NRC.

8.

With analyses not being per-formed as required, report the corrective actions taken to pre-vent a recurrence to the Commis-sion in the Annual Radiological Environmental Operating Report.

i l

I 3.8/4.8-16 Amendment No. 114 l

l

(

e QUAD-CITIES DPR-29 E.

Solid Radioactive Waste E.

Solid Radioactive Waste 1.

The solid radwaste system shall 1.

The PCP shall specify the method

-be used as applicable in and frequency to verify solidi-accordance with the PCP to fication of radioactive waste.

process wet radioactive wastes Actions to be taken if solidifi-to meet shipping and burial cation is not verified shall ground requirements.

also be specified in the PCP.

2.

With the provisions of the Process Control Program not saticfied, suspend shipments of defectively processed or defectively packaged solid radioactive waste from the site.

F.

Miscellaneous Radioactive Materials F.

Miscellaneous Radioactive Materials Sources Sources-Source Leakaga Test Each sealed source shall be tested for leakage and/or contamination by Specification the licensee or by other persons spe-cifically authorized by the Commis-Each sealed source containing radio-sion or an Agreement state, lhe test active material in excess of 100 mi-method shall have a detection sensi-crocuries of beta and/or gamma emit-tivity of at least 0.005 microcuries ting material or 5 microcuries of al-per test sample.

pha emitting material shall be free of > 0.005 microcuries of removablo Each category of sealed sources shall contamination.

be tested at the frequency described below:

Each sealed source with removable contamination in excess of the above 1.

Sources in use (excluding start-limit shall be immediately withdrawn up previously subjected to core from use and either decontaminated flux) - At least once per 6 and repaired or disposed of in accor-months for all sealed sources dance with Conmission Regulations, containing radioactive material:

A complete inventory of radioactive a.

With a half-life greater materials in the licensee's posses-than 30 days (excluding Hy-sion shall be maintained current at drogen 3), and all times.

b.

In any form other than gas.

3.8/4.8-17 Amendment No. 114

QUAD-CITIES DPR-29 2.

Stored sources not in use - lach sealed source shall be tested prior to the use or transfer to another licensee unless tested within the previous 6 nonths.

Sealed sources transferred with-out a certificate indicating the last test date shall be tested prior to being placed into use.

A Special Report shall be prepared and submitted to the Commission pur-suant to Specification 6.6.C.3 if source leakage tests reveal the pre-sence of > 0.005 microcuries of re-movable contamination.

G.

Qualification to Sections 3.8.A-E and 4.8.A-E in the event a limiting condition for operation and/or associated action

' requirements identified in Sections 3.8.A. through 3.8.E., and 4.8.A.

through 4.8.E. cannot be satisfied because of circumstances in excess of those addressed in the specifica-tions, no changes are required in the operational condition of the plant, and this does not prevent the plant from entry into an operational mode.

H.

Control Room Emergency Filtration H.

Control Room Emergency Filtration System System 1.

The control room emergency 1,

At least once per month, filtration system, including at initiate 2000 cfm (1 10%) flow least one booster fan shall be through the control room operable at all times when emergency filtration system for secondary containment integrity at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with the is required, except as specified heaters operable, in Sections 3.8.H.1.a. and b.

3.8/4.8-18 AmendmentNo.Ik4

1 4

s QUAD-CITIES OPR-29 a.

After the control room emergency filtration system is made or found to be inoperable for any reason, reactor operation and fuel handling are permissible only during the succeeding 14 days.- Within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months following the 14 days, the reactor shall be placed in a condition for which the control room emergency filtration system is not-required in accordance with Specification 3.7.C.1.a.

through d.

- b.

Specification 3.8 H.1.a..

above does not apply during performance or post-maintenance testing, or-during removal of the charcoal test canister._

2.

Periodic Performance Requirements 2.

Performance Requirement Tests a.

The results of the in-place a.

At least once per operating-DOP tests at 2000 cfm cycle but not to exceed 18 (1 10%)'on HEPA filters months, or following shall show < 1% DOP painting, fire, or toxic penetration 7 chemical release in any' ventilation zone communicating with the intake of the system while the system is operating that could contaminate the HEPA filters of charcoal adsorbers, perform the following:

1) In place DOP test the HEPA filter banks to verify leaktight integrity.

3.8/4.8-19

. Amendment No. 114

4 QUAD-ClllES DPR-29

2) In place test the charcoal adsorber banks with halogenated hydrocarbon tracer to verify leak tight integrity.

3) Remove one carbon test canister from the charcoal adsorber.

Subject this sample to a laboratcry analysis to verify methyl iodide removal efficiency.

b.

The results of in place b.

At least once per operat.ig halogenated hydrocarbon cycle, but not to exceed 18 tests at 2000 cfm (+ 10%) on months, the following the charcoal banks shall conditions shall be show < 1% penetration.

demonstrated:

1) Fressure drop across the combined filters is less than 6 inches of water at 2000 cfm (1 20%) flow rate.

2) Operability of inlet heater demonstrates heater AT of 15'f.

c.

The results of laboratory carbon sample analysis shall show > 90% methyl iodide removal efficiency when tested at 130'C and 95% R.H.

3.

Postmaintenance Requirements 3.

Postmaintenance Testing a.

Af ter any.naintenance or a.

Af ter any maintenance or

-heating that could affect testing tha.t could affect the HEPA filter or HEPA the leak tight integrity of filter mounting frame the HEPA filters, perform leaked tight integrity, the in place 00P tests on the results of the in place DOP HEPA filters in accordance tests at 2000 cfm (+ 10%) on with Specification 3.8.H.2.a.

HEPA filters shall ihow < 1%

~

00P penetration.

l i

I 3.8/4.8-20 Amendment No. 114 l

s QUAD-CITIES DPR-29 b.

After any maintenance or b.

After any maintenance or testing that could affect testing that could affect the charcoal adsorber the leak tight integrity of leak tight integrity, the the charcoal adsorber banks, results of in place perform halogenated halogenated hydrocarbon hyarocarbon tests on the tests at 2000 cfm (1 10%)

charcoal adsorbers in shall show < 1% penetration, accordance with

~

Specification 3.8.H.2.b.

l l

l 1

1 3.8/4.8-21 Amendment No. 114

l QUAD-CITIES DPR-29 BASES 3.8/4.8 Limiting Conditions for Operation and Surveillance Requirement Bases

}.0/4.O.A.1 GASECUS EriLUENi; DOSE-specification is provided to ensure that the dose at the unre:itricted ar boun from gaseous effluents from the units on the site will be withi e

annual do imits of 10 CFR Part 20 for unrestricted areas.

The a dose limits are the es associated with the concentrations of 10 art 20 Appendix B Table These limits provide reasonable a nce that radioactive material discharged in ous effluents will not te n the exposure of an individual in an unrestricto ea to annual av concentrations exceeding the limits specified in Appendix B, e II of,its restrict, at all tirres, the

,1 FR Part 20 (10 CFR Part 20.106(b)).

The specified release ra m

corresponding gamma and beta dose

- es a % background to an individual at or beyond the unrestricted area idary to less tt or equal to 500 mrem / year to the total body or to not s than or equal to 300 m/ year to the skin.

These release rate limit o restrict, at all times, the cortsqonding thyroid dose rate above bac nd to a child via the inhalation pathway tt. ?ess than or equal to 1500 a year.

For purposes of calculating doses resulting airborne rel the main chimney is considered to be an elevated release poin nd the re or vent stack is considered to be a mixed mode release point.

-3rB/4rer/.- DOSE, "0EM-GASES-specification is provided to implement the requirements of Sections III.

nd IV.A of Appendix I, 10 CFR Part 50.

The Limiting Conditio Operatio mplements the guides set forth in Section II.B of Appe I.

The statements p ide the required operating flexibility anc. at t same time implenant the g s set forth in Section IV,A of Appendix o assure that the releaset, of radioat material in gaseous effluents w)M bc kept "as low as is reasonably achievable.'

e Surveillance Requireme f implement the requirements in Section III.A of Append that conformance h the guides of Appendix 1 is to be shown by calculational pr dures base n models and data such that the actual exposure of an individual th h

e appropriate pathways is unlikely to be substantially underestimated.

The calculations established in the ODCM for calculating the doses due to JA actua lease rates of radioactive noble gases in gaseous effluents wijMe consistent the methodology provided in Regulatery Guide 1.109, " 4tulation of Annual Dos to Man from Routine Releases of Reactor Effluents f the Purpose of Evaluating Com nce with 10 CFR Part 50, Appendix I," R sion 1, October 1977 and Regulatory J e 1.111, " Methods for Estimating ospheric Transport and Dispersion of Gaseous fluents in Routine Rei es from Light-Water Cooled Reactors", Revision 1, Ju 1977.

The ODCM e ons provide for determining the air doses at the unrestric boundary bas pon the historu.41 average atmospheric conditions.

NUREG-0133 pro s

ods for dost calculations consistent with Reguistory Guides 1.109 and 1.1 i

1 3.8/4.8-22 Amendment No.122

4 I

QUAD CITIES DPR-29 3.8/4.8.A.3 DDSE, RAD 1010 DINES, RADI0 ACTIVE MATERIAL IN PARilCULATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES 5 specification is provided to implement the requirements of Sections 11 111.

and IV.A of Appendix 1, 10 CFR Part 50.

The Limiting Conditions Operat are the guides set forth in Section 11.C of Appendix 1.

1 statement rovide the required operating flexibility and at the 5 time implement th utdel set forth in Section IV.A of Appendix 1 to sure that the releases of ra 1 ctive materials in gaseous effluents will b ept "as low as is reasonably achi ble." The ODCM calculational methods ecified in the surveillance requirem s implements the requirements i ection Ill.A of Appendix 1 that conforma with the guides of Appen I be sha.'n by calculational procedures ba on modelt, and data ch that the actual exposure of an individual through appro ate pathways i niikely to be substantially underestimated. The ODCM calcula nal msth approved by HRC for calculating the doses due to the actual release tes the subject materials are regnired to be consistent with the methodology p vided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man om tine Releases of Reactor Effluents for the Purpose of Evaluating Com ance with- 0 CFR Part P., Appendix 1",

Revision 1, October 1977 and R latory Guide 1.

1, "Heth m for Estimating f

Atmospheric Transport and Distersion of Gaseous Ef ents in Routine Releases f rom Light-Water-Conled R ctors," Revision 1, July 1 lhese equations also provide for determinin he actual doses based upon the torical average atmospheric conditi The release rate specifications fo adiciodines, radioactive mater in particulate form and radionuclides oth than noble

" gases are depe ent on the exitting radionuclide pathways to man, the unrestricte area.

The pathways which were examined in the developm of these specific ons were: 1) individual inhalation of airborne radionuclides,

)

depos on of radionuclides onto green leafy vegetation with subsequent co ption by man and 1) deposition onto grassy areas where milk animals gra '

with consumption of the milk by man.

3.S/4.G.A.4 CASC005 WAHHREMttE*T TtQRABILITY of the pseous waste treatment which reduces amounts or/

concentran f radioactive materials ensures that the syste e

. available for us r gaseous effluents requi pret Cent prior to release to the environment. Th tund.t ment that th pprDpriate portions of this system be operable when specifie reasonable assurance that the releases of radioactive n gaseou ents will be kept "as low as is reasonably ac This specification imp em he requirements of 10 CFR Par 50d Grneral Design Criterion 60 of Appendix A to Part 50, esign objective Section 11.0 of Appendix 1 to 10 CFR Part 50.

3.8/4.8-23 Amendment No. 114

w a

l QUAD-CITIES OPR 79 A. Esplosive Gas Hanture 3.0/4.0.A.5 EXPLOSIVE GA5 MtXTttRE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the off gas system is minimized in conformance with the requirements of General Design Criteria 60 of Appendix A to 10 CFR Part 50.

titU10 tiTLL'EUTS

. /4.8.B.1 CONCENTRATION This s cificati, is provided to ensure that the concentration of radioacti materials elease in liquid waste effluents from the site to unrestricte areas will be les han the concentration levels specified in 10 CFR Part 20 ppendix B, Table II, c umn 2.

The concentration limit for noble gases, MP n air (submersion), was onverted to an equivalent concentration in wa*rf using the International Commi on on Radiological Protection (ICRP) Pub)(cation

?.

3.8/4.C.B.2 DOSE

/

This specification is provi implement the re rements of Sections ll.A.

III.A and IV.A of Appendix I, 10 "R Part 50.

Limiting Condition for Operetion implements the guides set rth in ction II.A of Appendix 1.

The statements provide the required operat exibility and at the same time impbement the guides set forth in Secti V.A of Appendix 1 to assure that the releases of radioactive material in

uid luents will be kept "as low as is reasonably achievable". The dose lculations the ODCM implement the requirements in Section III.A Appendix I th*

nformance with the guides of Appendix 1 be shown by cale:

tional procedure bas on models and data such that the actual exposure an individual through appr - late pathways is unlikely to be substan ally underestimated.

The equatio specified in the ODCM for calculatin he doses due to the actual release ra of radioactive materials in li effluents will be consistent with the meth ology provided in Regulatory aide 1.109, "Calcule* bn of Annual Deses to Man f r Routine Releases of eactor Effluents for the Purpose of Evaluating Complia with 10 CFR Part

, Appendix I", Revision 1, October 1977 and Regulatory Gui t 1.113, "Esti ing Aquatic Dispersion of Effluents from Accidental and Routine ctor pTrovides methods for dose calculations consistent with Reg Guide 1.109 and 1(.1 Re) ases for the Purpose of Implementing Appendix 1", April 1977. NUREG-01 e main condenser-( Add per attac.hed) 3.8/4.8-24 Amendment No. 114 l

4 INCERT POk TECIINICAT. C'E0IPICATION PAGE 3.0/(.L4 4 B.

Main Condenser Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to a MEMBER OF THE PUBLIC at and beyond the SITE BOUNDARY will not exceed a small fraction of the limits of 10 CFR 50 Part 100 in the event this offluent is inadvertently discharged directly to the environment without treatment.

This specification implements the requirements 01 General Design criteria 60 and 64 of Appendix A to 10 CFR Part 50.

k 8

l

QUAD-CITIES DPR-29 8/4.8.B.3 LIQUID WASTE TREATMENT The erability of the liquid radwaste treatment system ensures that this syste will b available for use whenever liquid effluents require treatment prior tp

.elease o the environment. The requirement that the appropriate portions f t hi: syst be used when specified provides assurance that the releases o radwartive aterials in liquid effluents will be kept "as low as is re onably achievable".

his specification implements the requirements of 10 CF Part 50.35=, Genera esign Criterion 60 of Appendix A to 10 CFR Part 50 nd design objective Sectio 11.0 of Appendix I to 10 CFR Part 50.

3.8/4.8.0.1 MONITORI PROGRAM The radiological monitori program requireo by this sp ification provides measurements of radiation a of radioactive material in those exposure pathways and for those radion lides, which lead to e highest potential radiation exposures of individu s resulting from e station operation. This monitoring program thereby supple.nts the radi ogical etfluent monitoring program by verifying that the meas able conc.tr-ations of radioactive materials and levels of radiation are not high than.xpected on the basis of the effluent measurements and modeling of.. e nvironmental exposure pathways.

Program changes may be initiated based operational experience.

The detection capabilities required Tabl 4.8-6 are state-of-the-art f or routine environmental measurement in industr i laboratories.

The specified I-1 in water, m k and other food products lower limits of detection for /

correspond to approximately oJ quarter of the Ap ndix ! to 10 CFR Part 50 design objective dose-equiv rent of 15 mrem / year f o atmospheric releases and 10 mrem / year for liquid rele es to the most sensitive qan and individual. They ons C ven in Regulatory Guide

.109, " Calculation of i

are based on the assump Annual Doses to Man fr m Routine Releases e' Reactor Ef f ents for the Purpose of Evaluating Compi nce with 10 CFR Par' t.

Appendix I",

ctober 1977, except vear of drinki q water instead of the change for an 'nfant consuming 330 ti 510 liters / year.

3.8/4.8.0.

LAND USE CENSUS This s ecification is provided to ensure that changes in the use of un stricted are identified and that modifications to the monitoring program ar made are if/ required by the result" of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.

3.8/4.8-25 Amendment No. 114

e l

l QUAD-CITIES DPR-29 3.0/4.0.0.7 OR055 CHECK FROGRAM The rceiuirc; cat for partie4petion in the-4nterlabcretory campa-hc; crc = heck pregrc: is-provided-tc casere thet-independent checke en the preciticn end eccurecy cf -the meesttremente vi r adivadive malvriai h1 tmvtT'onmentai samptb matf4te: :re per4 e ed 2: p:rt--ef-the quality-usurance-prgrem hi-emironmental acniter4ng-in order te demonstrate-theO the rcMt+-are rc=mWy vel i d.-

c. mec.hanical Vacuum FLmp 3.0/4.S.C MECHANifAb-VAGUUM-PUMP-The purpose of isolating the mechanical vacuum line is to limit release of activity from the main condenser.

During an accident, fission products would be transported from the reactor through the main steamline to the main condenser.

The fission product radioactivity would be sensed by the main steamline radioactivity monitors which initiate isolation.

3.0/4.0.0 00!10 RA010ACTI": WA&TE The-operab!!!ty Of 14.c ;clid rcdionetive weetc system en3nres thet the eytta wi-li be eveilebic fer wie shenever 3elid redsestes rec wire Liecessing end packeg4ng pricr to bei-ng-shipped of f site.

This av ritientmn impl-mia ihr e

requ W ment: cf 10-CFR 50.36e-end Ocacreb 0c3ign Criteria CO-ef Appcadb A to 10 CTR Pert 50.

D. Scaled Source Con + amination 3.0/4.0.F MI6GEttANE005-RA&l0AttiVE-MATE RfAtt-50ttRtts The objective of this specification is to assure that leakage from byproduct, source and special nuclear material sources does not exceed allowable limits.

The limitations on removable contamination for sources requiring leal testing, including alpha emitters, is based on 10 CFR 70.39(t) limits for plutonium.

E. Control Roon1 Emergency Fe'Itratlen System 3.0/4.0.11 CONTR0t-R00tPAIRYltTRAT10M The purpose of these specifications is to assure availability of the control room emergency air filtration unit that has been installed in response to NUREG-0737 Item III D.3.4.

Operation of this unit is described in the *Eenteel-

-Room-Hab4 tabi44 ty-Study-fee-Quad-C4 t ies -Stat 4on-which-was-+ubmi tted--to-the-NRC in-December-1981. au.a cL Ci+o e-.s UpcLa)ed PSRR. 5cct50v6 lo.I0. 4 and. I4.2 h

i 3.8/4.8-26 Amendment No. 114 1

_. _ _ _ _ _. _ _. __ _ _ _ _ _ _ _, _ _ _ _ _ _ - - ~ _

y

-QUAD-CITIES DPR-29 TABLE 4.8-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANAL,YSIS PROGRAM MINIMUM LOWER LIMIT OF GASEOUS SAMPLING ANALYSIS,

TYPE OF DETECTION (LLD)

RELEASE TYPE FREQUENCY FREQUENCY ACTIVfTY ANALYSIS (pCl/ml)

Principal Ganna M.

MD Emitters' 1 x 10-8 A. Main Chimney

. Grab Sample

~ Reactor Bldg.

-Vent Stack M

Tritium 1x 10-6

8. All Release Continuousd We

_J.131 1

10-12 Types as-Charcoal

-Listed in Sample A Above-I-133 1 x 10 Continuousd Wc Principal Gamma 1x 10-33 Particulate Emitterse Sample (I-131. others).

Continuou$d Q

Sr-89 1 x 10-11 Composite Particulate Sample Sr-90 1x 10-11 Continboust M

Composite Gross Alpha 1x 10-II Particulate Sample-C. Main Chimney Continuousd Noble Gas Noble Gases 1x 10-6 Monitor l

10-4 l

0.~ Reactor Bldg Continuousd Noble Gas Noble Gases 1 x L

Vent Stack Monitor-p 3.8/4.8-27 Amendment No. 114

o e

QUAD-CITIES DPR-29 TABLE 4.8-1 (Continued)

TABLE N01AT10N a.

The lower limit of detection (LLD) is defined in table notation A. of lable 4.8-6.

b.

Sampling and analyses shall also be performed following shutdown, startup, or a thermal power change exceeding 20 percent of rated thermal power in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months unless (1) analysis shows that the DOSE EQUIVALENT l-131 concentration in the primary coolant has not increased more than a factor of 5, and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.

Samples shall be changed at least once per 7 days and the analyses completed c.

within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after removal from the sampler.

Sampling shall also be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following each shutdown, startup, or thermal power level change exceeding 20% of rated thermal power in one hour.

This requirement does not apply if (1) analysis shows that the DOSE EQUlVALENT l-131 concentration in the primary coolant has not increased more than a factor of 5, and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.

When samples col-lected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's may be increased by a factor of 10.

d.

The ratio of sample flow rata to the sampled stream flow rate shall be known.

The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr 88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions, and Mn-54, Fe-59, Co-60, 2n-65, Co-58, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions.

Other peaks which are measurable and identifiable by gamma ray spectrometry, together with the above nuclides, shall be also identified and reported when an actual analysis is performed on a sample.

Nuclides which are below the LLD f or the analyses shall not be reported as being present at the LLD level for that nuclide.

3.8/4.8-28 Amendment No. 114 l

+

s QUAD-CITIES DPR-29 TABLE 4.8-2 MAXIMUM PERMISSIBLE CONCENTRATION Of DISSOLVED OR ENTRAINED NOBLE GASES RELEASED FROM THE SITE TO UNRESTRICTE0 AREAS IN LlQUID WASTE NUCLIDE MPC(pCi/ml)*

Kr-5Bm 2x10'4

~4 Kr-85 5x10

-5 Kr-87 4x10

-5 Kr-88 9x10

-5 Ar-41 7x10

~4 Xe-131m 7x10

~4 Xe-133m 5x10 Xe-133 6x10'4

~4 Xe-135m 2x10 Xe-135 2x10'4

Computed from Equation 20 of ICRP Publication 2 (1959), adjusted for infinite cloud submersion in water, and R = 0.01 rem / week, density = 1.0 g/cc and Pw/Pt = 1.0.

3.8/4.8-29 Amendment No. 114 l

o e

QUAD-Cl?!ES DPR-29 TABLE 4.8-3 RADI0 ACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM MINIMUM LOWER LIMIT Of LIQUID

$AMPLING ANALYSIS TYPE OF DETECTION (LLD)

RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (pCi/ml)

Principal Gamma 5 = 10-7 A. Batch Haste Prior to Prior to Emitters' Release Tanks Each Batch Each Batch I-131 1 i 10-6 10 7 Gross Alpha 1 s Prior to M

Each Batch Compositeb 10-0 H-3 1 =

Fe-55 1x 10-6 Prior to 0

Each Batch Compositeb Sr-89. Sr-90 5x 10-8 10-5 Prior to M

Olssolved & Entrained I i One Batch /M Gasesf (Gamma Emitters) 10-6 B. Plant MC (Grab MC I-131 1 x Continuous tample)

Releasesd Principal Gamma 5x 10-7 Emitterse 10-5 Dissolved & Entrained 1 =

Gasesf (Gamma Emitters) i H-3 Ia 10-5 10 7 Gross Alpha 1 a i

Sr-89. Sr-90 5a 10-8 QC (Grab QC l

Samplo) 10-6 Fe-55 1 x 1

3.8/4.8-30 Amendment No 114 l

e s

QUAD-CITIES DPR-29 TABLE 4.8-3 (Continued)

TABLE NOTATION a.

The LLD is defined in Notation A of Table 4.8-6.

b.

A composite sample is one in which the quantity of liquid samples is proportional to the quantity of liquid waste discharged and in which the method of rampling employed results in a specimen which is representative of the liquids released.

c.

If the alarm setpoint of the service water effluent monitor as determined in the ODCM is exceeded, the frequency of analysis shall be increased to daily until the condition no longer exists, d.

A batch release is the discharge of liquid wastes of a discrete volume.

Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed to assure representative sampling.

A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., frun a volume or system that has an input flow during the release, The principal gamma emitters for which the LLD specification epplies e.

exclusively are the following radionuclides: Mn-54, f e-59, Co-60, 2n-65, Co-58, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. Other peaks which are measurable and identifiable by gamma ray spectrometry together with the above nuclides, shall be also' identified and reported when the actual analysis is performed on a sample. Nuclides which are below the LLD for the analyses shall not be reported as being present at the LLD level for that nuclide.

f.

The dissolved and entrained gases (gamma emitters) for which the LLD specification applies exclusively are the following radionuclides: K r-87,

Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138. Other dissolved and entrained gases (gamma emitters) which are measurable and identifiable by gamma-ray spectrometry, together with the above nuclides, shall also be identified and reported when an actual analysis is performed on a sample.

Nuclides which are below the LLD for the analyses shall not be reported as being present at the LLD leve. for that nuclide,

{

3.8/4.8-31 Amendment No. 114

e n

QUAD-CITIES DPR-29 TABLE 4.8-4 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Minimum Number of Samples Sampling and Col-Type and f requency and/or Sample and Sample Locations

1ection Frequency of Analysis

3. AIRBORNE

a. Particulates 16 locations Continuous opera-Gross beta and tion of sampler gamma isotopic as for a week specified in ODCM,

b. Radioiodine 16 lecations Continuous opera-I-131 as speci-tion of sampler fled in ODCM.

l for two weeks

2. DIRECT RADIATION Forty Locations Quarterly (Minimum of two TLDs per packet)

Sample locations are described in the ODCM.

l I

3.8/4.8-32 Amendment No. 114

e e

t QUAD-CITIES DPR-29 TABLE 4.8-4 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Minimum Number of Samples Sampling and Col-lype and frequency:

and/or Sample and Sample Locations

lection frequency of Analysis

3. WATERBORNE

a. Pub rr 2 Locations Monthly composite Gamma isotopic of weekly collec-analysis of_each ted samples composite sample

b. Sedimer wJstream location Annually Gamma isotopic

-iving body of analysis of each sample n

c. Plant Cooling In. e Discharge Weekly composite Gross Beta analy-Water sis of each sample :

Sample locations are described in the ODCM.

3.8/4.8-33 Amendment No. 114

e e

QUAD-CITIES DPR-29 TABLE 4.8-4 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Minimum Number of Samples Sampling and Col-Type and f requency and/or Sample and Sample locations

1ection Frequency of Analysis

4. INJESTION

a. Milk 2 Locations At least once I-131 antlysis of weekly when ani-each sample animals are on pasture; at least once per month at other times,

b. Fish 1 location in receiv-Semi-annually Gamma isotopic ing body of water analysis on edible portions

Sample locations are described in the ODCM 3.8/4.8-34 Amendment No. 114

. -.... ~. _ -. -

y

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a QUAD-CITIES DPR-29 TABLE 4.8-5 REPORTING LEVELS FOR R4D10 ACTIVITY.

CONCENTRATIONS-IN ENVIRONMENTAL SAMPLES Reporting Levels tAnalysis

' Water AirborneParticylate Fish-.

Milk Food Products-or Gases (pCi/m )

(pCi/Kg, wet)

(pCi/1)

(pti/Kg, wet) '

4 l H '2 x 10 (a);

3 4

Mn-54:

l'x 10 3 x 10 2

4 Fe-59 4 x 10

-I x 10 t

3 4

Co-58

-lix 10 3 x 10 2

4

.Co-60 3-x.10 1 x 10 '

2 4

Zn 3'x 10 ~

2 x 10 2

Z r-Nb-95 4 x 10 I-131-

"2 0.9 3

1xId2 3

3 Cs-134-30-10 1 x 10 60 1 x 10 3

3 l2

Cs-137-

-50 20 1 x-10 70 2 x 10 l-

'Ba-La-140 2 x-10 3 x 10 2

2-

_; es-p L

a) for drinking water' samples. This.is 40 CFR Part 141 value.

m 3.8/4.8-35 Amendment No. 114 Y--=,:.:--....,-...,...._.---....~.-........-..-....__---_-_.-_-_-

e QUAD-C111ES DPR-29 TABLE 4.8-6 PRACTICAL LOWER LIMITS Of DETEC110N (LLD)

FOR STANDARD ENVIRONMENTAL RADIOLOGICAL MON 110 RING PROGRAM Sample Media Analysis LLD,B Units A

(4.66 o )

3 Airborne " Particulate" Gross Beta +

0.01 pCi/m 3

Gamma Isotopic 0.01 pCi/m 3

Airborne 1-131 lodine 131 0.10 pCi/m Milk /Public Water 1-131 5

pCi/1 Cs-134 10 pCi/1 Cs-137 10 a pCi/1 Tritium 200 pCi/1 Gross Beta +

5 pCi/1 Gamma Isotopic 20 pCi/1/nuclide Sediment Gross Beta +

2 pCi/g dry Gamma Isotopic 0.2 pCi/g dry Fish Tissue 1-131 - Thyroid 0.1 pCi/g wet Cs-134, 137 0.1 pCi/g wet Gross Beta +

1.0 pCi/g wet y Isotopic 0.2 pCi/g wet 0.5 pCi/1 on milk samples collected during the pasture season.

+

Referenced to Cs-137 a

5.0 pCi/1 on milk samples 3.8/4.8-36 Amendment No. 114

6 QUAD-ClllES DPR-29 TABLE 4.8-6 (Continued)

TABLE NOTA 110N A.

The LLD is the smallest concentration of radioactive material in the sample that will be detected with 95 percent probability with only 5% probability of falsely concluding that a blank observation represents a "real" sequal.

For a particular measurement system (which may include radiochemical separation) 4.55.s LLD = ----------------- D-----------------------

A ' E

V ' 2.22

Y ' exp

(- A at) ' t Where:

LLD is the "a priori" lower limit of detection for a blank sample or background analysis as defined above (as pCi per unit mass or volume),

s is the square root of the background count or of a blank sample count ; is gthe estimated standard error of a background count or a blank sample count as appropriate (in units of counts).

E is the counting efficiency (as counts per disintegration).

A is the number of gamma-rays emitted per disintegration for gamma-ray radionuclide analysis (A - 1.0 for gross alpha and tritium measurements).

V is the sample size (in units of mass or volume).

2.22 is the number of disintegrations per minute per picocurie.

Y is the fractional radio-chemical yield when applicable (otherwise Y = 1.0).

' is the radioactive decay constant for the particular radio.aclide (in units of reciprocal minutes).

At is the elapsed time between the midpoint of sample collection and the start time of counting.

(at = 0.0 for environmental samples and for gross alpha measurements).

t is the duration of the count (in units of minutes).

3.8/4.8-37 Amendment No. 114 l

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QUAD CITIES DPR TABLE 4.8-6 (Continued)

TABLE NOTATION The value of "s " used in the calculation of the LLD for a detection system shall-bebasedNnanactualobservedbackgroundcountorablanksamplecount (as appropriate) rather than on an unverified theoretically predicted value.

Typical values of "E", "V", "Y",

"t", and "ot" shall be used in the calculation.

For gamma-ray radionuclide analyses the background counts are determined f rom.

the total counts-in the channels which are within plus or minus one FWHM (Full Width at Half Maximum) of the gamma-ray photopeak energy normally used for the quantitrcive analysis for that radionuclide.

Typical values of the FWHM shall ha used in the calculation. -

The'LLD for all measurements is defined as an "A priori" (before-the fact)'

limit representing:the capability of a measurement system and not_as an "a posteriori" (after the fact) limit for a particular sample measurement.

'B.

Other radionuclides which are measureable and identifiable;by gamma-ray spectrometry, together with the nuclides indicated in Table 4.8-6, shall also be identified and reported when an actual analysis is-performed on a-sample.

Nuclides which are below the-LLD-for the analyses shall.not.be reported as being present at the LLD level for that nuclide.

1 1

3.8/4.8-38 Amendment No. 114

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Amencment No.II4

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SIGNIFICANT! HAZARDS CONSIDERATIONS-AND ENVIRONMENTAL ASSESSMENT EVALUATION t

PROPOSED TS 3.8/4.8 RADIOACTIVE EFFLUENTS" l

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3 EVALUATION-EQB SIGNIFICANT HAIARDA CONSIDEBATLQM Proposed Specification 3.8/4.8 Radioactive Effluents The proposed changes provided in this amendment request are made in order to provide a more user friendly document, incorporate desired technical improvements, provide changes consistent with the guidance of Generic Letter 89/01, and to incorporate some improvements from later operating BWRs.

These changes have been reviewed by Commonwealth Edison and we believe that they do not present a Significant Hazards Consideration.

The basis for our determination is documented as follows:

HASIS EQB HQ SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that-it involves no significant hazards consideration.

In accordance with the criteria of 10 CFR 50.92(c)ificant hazards a proposed amendment ~ to an operating. license involves no s:.gn 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,-because:

a.

The Generic Changes to the technical' specifications involve administrative changes to format and arrangement of the material.

As such, these changes cannot involve a significant increase in the probability or consequences of an-accident previously evaluated, b.

The_NRC issued Generic Letter 89/01 on January 31, 1989, in order to allow the technical specification provisions for the-Radiological Effluent Technical Specifications

_(RETS) to be relocated to the Administrative Controls section of the technical specifications (programmatic requirements) and to the ODCM or PCP (procedural details).

New programmatic controls-for radioactive effluents and radiological environmental monitoring are relocated to Section 6.0 of the technical specifications to conform to the regulatory requirements of 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a,.and Appendix I to 10 CFR Part 50.

The proposed changes to Section'3.8/4.8 of the Quad Cities technical specifications relocates to the ODCM, the provisions of present Specifications 3.8 A.1/4.8.A.1, 3.8.A.2/4.8 A.2, 3.8.A.3/4.8.A.3, 3.8.A.4/4.8.A.4, 3.8.B/4.8.B, 3.8.D/4.8.D, and Tables 4.8-1 through 4.8-6.

Present provisions of 3.8.E/4.8.E are relocated to the PCP.

The relocation of the present technical specification requirements is not intended to reduce the level of radiological effluent control.

The proposed change will-allow programmatic controls to remain in the

e technical specifications with procedural details controlled by the controls for changes to the ODCM and PCP.

Since the proposed changes to implement GL 89/01 provisions meet NRC requirements and retain adequate programmatic controls for RETS in the technical specifications, the proposed change does involve a significant increase in the probability or consequences of an accident previously evaluated.

c.

The remaining changes to Section 3.8/4.8 involve the rearrangement of present requirements, inclusion of some later operating plants' provisions or STS guidelines that are applicable at Quad Cities or the implementation of the intent of present provisions.

The proposed changes maintain operability of systems and equipment when needed to perform design functions.

No changes are made to existing setpoints or limits and, as such, the proposed changes do not 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 previously evaluated because:

a.

Since the Generic Changes proposed to the technical specifications are administrative in nature, they cannot create the possibility of a new or different kind of accident from any previously evaluated.

b.

The proposed changes to the technical specifications resulting from the implementation of Generic Letter 89/01 follow NRC guidelines for relocating RETS procedural requirements to the ODCM and PCP, and programmatic requirements to Section 6.0 of the technical specifications.

Since the present levn1 of radiological effluent monitoring is maintained by the proposed change, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

c.

The remaining changes proposed to the specifications in Section 3.8/4.8 are made to improve usability and incorporate some proven provisions from STS guidelines and later operating BWRs.

These changes maintain operability of the required systems and parameters when required '-

perform their design function.

No new modes of operat.on are introduced, considering present intent and later operating BWR provisions that are in use at these facilities and are applicable for use at Quad cities.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3)

Involve a significant reduction in the margin of safety 1

l

o because a.

The Generic Changes are administrative in nature and as such they cannot involve a significant reduction in the margin of safety.

b.

The proposed implementation of the guidelines of Generic Letter 89/01 retains required RETS programmatic controls in Section 6.0 of the technical specifications and specific procedural-controls in the ODCM and PCP.

As such, the present margins of safety are preserved and the changes do not involvo a significant reduction in the margin of safety.

The remaining changes to Section 3.8/4.8 are proposed in c.

order to retain the intent of present requirements for operability of systems and parameter limits or to implement proven requirements from STS guidelines or other operating BWRs with systems similar to those at Quad Cities.

Since the proposed changes do not decrease the present level of operability, do not change established sotpoints, and have been evaluated as acceptable for use at Quad Cities; they do not involve a significant reduction in the margin of safety.

4

MIROEMMTAL MRSHMI HAWATLQB PROPOSED SPECIFICATION SECTION 3.8/4.8 RADIOACTIVE EFFUUENTS Commonwealth Edison has evaluated the proposed amendment in accordance with the requirements of 10 CFR 51.21 and has determined that the amendment meets the requirements for categorical exclusion as specified by 10 CFR 51.22 (c) (9).

Commonwealth Edison has determined that the amendment involves no significant hazards consideration, there are no significant change in the types or significant increase in the amounts of any effluent that may be released offsite, and there is no significant increase in individual or cumulative occupational radiation exposure.

The proposed amendment does not modify the existing facility and does not involve any new operation-of the plant.

As such, the proposed amendment does not involve any change in the type or significant increases in effluents, or increase individual or cumulative occupational radiation exposure.

The proposed amendment to.Section 3.8/4.0,

" Radioactive Effluents" contains administrative changes such

-as including appropriate applicability statements within the specifications to define the applicability during operating mode and the required actions to be implemented in the event that spacification cannot be met.

The added requirements are based on Standard Technical Specifications and later operating plant requirements.

The proposed amendment implements the requirements of Generic Letter 89-01 which allows that the programmatic requirements be relocated to the Administrative Controls of the technical specifications and the procedural details are relocated to the Offsite Doce

' Calculation Manual or the Process Control _ Program, i

I

I QC-1/ QC-2 DIFFERENCES TS 3.8/4.8

' RADIOACTIVE EFFLUENTS'

e COMPhRISON OF UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS FOR THE IDENTIFICATION OF TECHNICAL DIFFERENCES SECTION 3.8/4.8 RADIOACTIVE EFFLUENTS Commonwealth Edison has conducted a comparison review of the Unit 1 and Unit 2 Technical Specifications to identify any technical differences in support of combining the The intent of Technical Specifications into one document.

the review was not to identify any differences in (e.g. table formats, use of capital presentation style letters, etc.) or punctuation but rather to identify areas which the Technical Specifications are technically different.

The review of Section 3.8/4.8, " Radioactive Effluents" revealed the following technical and administrative differences.

Page 3.8/4.8-1 Applicability Unit 1:

measuiements of radioactive Unit 2:

measurements radioactive Pace 3.8/4.8-3 3.8.A.2.d Unit 1:

radiation exposure to all members Unit 2:

radiation exposures to all members Page 3.8/4.8-4 4.8.A.3 Unit 1:

specified in Table 4.8-1 f

Unit 2:

specified Table 4.8-1 Egga 3.8/4.8-6 3.8.A.4 Unit 1:

Off Gas Charcoal Adsorber Unit 2:

Off Gas Charcoal Absorber

+

1 Eagg 3.8/4.8-8 4.8.A.7 Unit it. rate of noble gases at Unit 2:

rate of nobles gases at Unit 1 limits of Specification 3.8.A.7 at

=

-Unit 2:

limits-of Specification 3.8.H at The-Unit 1 Specification is correct in_that Specification.3.8.A.7 provides the limit of noble gas release at the main condenser air l

ejector.

Specification 3.8 H discusses the requirements for control Room Emergency Filtration System.

Pace 3,8/4.8-9 4.8.B.1 Unit 1:

obtaining the representative Unit 2:

obtaining representative 4.8.B.2.a Unit 12 total body and organ _ doses Unit 2:

total body and any organ doses Pace 3.8/4.8-10 3.8.B.2.c-Unit _1:

the ~cause(s) for exceeding the Unit-

~2:

the cause(s) for exceedint the y

Engg 3.8/4.8 3.8.C Unit 1:

The mechanical vacuum pump'shall Unit 2:

The mechanical-vacuum-shall 4.8.D Unit 1:

pursuant to Table 4.8-4

-Unit 2: _ pursuant _to Table 4.8-3 L

Table 4.8-4, Radiological Environmental Monitoring-' Program-is-the-correct-reference for this requirement.

Reference to Table 4.8-3, Radioactive Liquid Waste Sampling and Analysis Program in-this Specification is not consistent with the djscussion of the requirement.

Page 3.8/4.8-16 3.8.D.6.

Unit 1:

program-within'30 days if possible.

Unit 2:

program with 30 days, 1f possible.

Pace 3.8/4.8-11 3.8.G' Unit 1: ' Qualification to Sections 3.8.A-E and 4.8.A-E Unit 2:

No title provided

3-

)

i Page 3.8/4.8-19 4.8.H.2

- Unit la filter of charcoal adsorbers Unit 2:

filter =of charcoal absorbers Pace'3.8/4.8-20 4.8.H.2.a.2&3 Unit 1:

charcoal adsorber Unit.2:

charcoal absorber l

3.8.H.3 Unit 1:

frame leaked tight Unit 2:

frame leak tight Unit 1:- maintenance or heating that Unit 2:

maintenance or testing that Pace 3.8/4.8-21 3.8~H.3.b!

Unit'1:

charcoal adsorber 4.8.H.3.b Unit.2:

charcoal-absorber Eggg 3.8/4.8-22' Unit 1:- 3.8/4.8 Limiting Conditions for operations and surveillance-Requirements Bases

- Unit 2:

No titlo provided Eggg 3.8/4.0-25 4.8/4.8.D 1 Unit 1:

that the measurable concentrations Unit 2:

-that the measuremble concentrations Unit 1:

consuming 330 liters / year of

= Unit 2:

consuming 330 liter / year of Pace 3.8/4.8-26

- 3 ~. 8/ 4. 8. H Unit 1:--emergency air filtration Unit 2:

emergency air filtrations Page 3.8/4.8-28 Notes Unit 1 does not designate note "e" EAE2 3.8/4.8-37

. Note A

. Unit 1:

material in the sample Unit 2:

material in a sample D-M p

Mm-w-re d-E er--

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~

QUAD CITIES NUCLEAR POWER STATION TECHNICAL SPECIFICNFION UPGRADE PROGRAM PROPOSED AMENDMENT SECTION 3.10/4.10 " Refueling"

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EXECUTIVE

SUMMARY

Proposed Changes to TS 3.10/4.10

' REFUELING"

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EXRQUTIVE BURMARY QUAD CITIES TECHNICAL UPECIFICATION UPGRADE PROGRAM The Quad Cities Technical Specification Upgrade Program was conceptualized in response to lessons learned from the Dresden Diagnostic Evaluation Team inspection and the frequent need for Technical Opecification interpretations.

A comparison of the en'. sting Quad cities Technical Specification and, Standard Technical Specifications and later operating plants' Technical Specjfication provisions was cunducted to identify potential improver *nts in clarifying requirements and to identify requirements which are no longer consistent with current industry practices.

The comparison review identified approximately one-hundred and fifty suggested improvements.

The Technical Specification Upgrade Program was not intended to be a complete adoption of the Standard Technical specifications.

Overall, the Quad Cities custom Technical specifications provide for a safe oporetion of the plant and, therefore, only an upgrade was deemed aryropriate.

The comparison study revealtd a mix of recommended upgrades which included the relaxation of certain existing Technical Specification requirements, the addition of surveillances, the removal of allowances which would no longer be allowed under new plant.'.icensing, and better definition of appropriate action requirements in the event a Limiting Condition for operation cannot be met.

The Technical Specification Upgrade Program also implements NRC reconsended line item improvements to the Technics 1 specifications Nhich were issued under Generic Letters.

In response to an NRC recommendaticn, the Unit 1 and Unit 2 Technical Specifications are combined into one document.

To accomplish the combination of the Units' Technical Specifications, a comparison of the Unit 1 and 2 Technical Specifications was performed to identify any technical differences. The technical differences are identified in the proposed amendment package for each section.

The Technical Specification Upgrade Program was identified as a Station top priority during the development of Quad Cities Station's Performance Enhancement Program (PEP).

The Technical Specification Upgrade Program's goal is to provide a better t3ol to Station personnel to implement their responsibilities and co ensure Quad Cities Station is operated in accordance with current industry practices.

The upgraded specifications provide for more safe and reliable operation of the plant.

The program improves the operator's ability to use the Technical Specifications by more clearly defining Limiting Conditions for Operations and required actions.

The most significant improvement to the specifications is the aldition of equipment operability requirements during shutdown conditions.

+

e i

I P.X3CUTIVE SCHMARY (continued)

Propocod Technical Specification Section 3.10/4.10, " Refueling" ine proposed change dcletes the present objective statement and provides Applicability statements within each specification similar to the STS.

The proposed Applicability statements include the Reactor Modes and other conditions for which the LCos must be satisfied.

An STS type of format is proposed while retaining the present two column layout.

STS guidelines allowing the reactor modo switch to be in either the Refuel or Shutdown posit!on during operational Mode 5 are adopted in the proposed changs7 to the Refueling Interlocks specifications.

Clarifications are added to the proposed Applicability.

proposed Actions from STS guidolines have been added.. Present Survej \\ lance Requirements are retained and STS provisions are added to verify that the reactor modo switch is locked in the Shutdown or Refuel position.

Present Core Monitoring requirements are enhanced by inclusion of STS guidelines as modified by General Electric recommendationc on the usage of " shorting links".

STS based remedial Action provisions are adopted and present Survoillanco Requirements are expanded by adoption of STS guidelines and General Electric recommendations.

STS based Action statements are incornorated in the proposed changes to the ruel Storage Pool Water Lovel specifications.

Present level limitations and daily level recording surveillance requirement are retained in the proposed change.

Prosent allowance for the removal of up to two control rods and/or control rod drive mechanisms for maintenance is replaced with more prescriptive STS guidelines allowing removal of only one control rod and/or control rod drivo mechanism during operational Modes 4 and 5.

Present Surveillanco Requirements are replaced with ones thst verify all the proposed LCo restrictions.

Present provisions for Extended Core Maintenance - Multiple Control Rod nemoval Specifications are replaced with STS guidelines containing more prescriptivo LCo, Action and Surveillance Requirements.

Present limitations on Spent Fuel Cask Handling are retained in the proposed rewrite of this specification.

An Action to suspend spent fuel cask handlint operations is added to the proposed specification.

Present Surveillance Requirements are ratained.

A new Control Rod Position specification, based on STS guidelines

l e

and not presently considered in the Quad Cities Technical Specifications, is added in the proposed specification.

A new Communications specification, based on STS guidelines and not presently considered in the Quad Cities Technical Specifications, is added in the proposed specification.

A new Reactor Vessel Water Level cpecification, based on STS guidelines and not presently considered in the Quad Citics Technical Specifications, is added in the proposed specification.

i.s i

SUMMARY

OF CHANGES i

Proposed TS 3.10/4.10 i

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' REFUELING'

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SUMMARY

OF CHANGES Proposed Specification 3.10/4.10 Refueling This amendment package is one in a series of proposals that will provide improvements to the present Quad cities Technical Specifications.

The summary of changes includes a general section to describe generic changes that are applicable to more than one section of the technical specifications and a section which provides the changes that are page by page specific.

GENERIC CHANGES Item it The present Applicability and objective statements at the beginning of each technical specification section are being deleted.

The Applicability statement is being included after the i

LCO statement in each individual specification.

Item 23 Each specification is rearranged to follow an STS type of format while retaining the present two column layout.

Each specification will contain an LCO, Applicability, Action and Surveillance Requirement section, as applicable.

SPECIFIC CHANGES Item 1 Pages 3.10/4.10-1 and 3.10/4.10-2, Specification 3.10.A/4.10.A, DPR-29 a.

Present requirements of Specification 3.10.A, Refueling Interlocks, are used to write the proposed LCO for new 3.10.A. Present requirements for the reactor mode switch to be in the Refuel position is changed to include the Shutdown position.

b.

Proposed Applicability for Refueling Interlocks is taken from the intent of present provisions and requires operability during Operational Mode 5 while performing Core Alterations with equipment associated with the Refuel position interlocks.

c.

Proposed Actions for Specification 3.10.A are taken from STS guidelines.

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e d.

Present SR 4.10.A is rewritten into proposed SRs 4.10.A.1, 4.10.A.2 and 4.10.A.3.

Proposed SR 4.10.A.4 is taken from STS guidelines to provide verification that the reactor mode switch is locked in the Shutdown or Refuel position.

Item 2:

Pages 3.10/4.10-2 and 3.10/4.10-3, Specification 3.10.B/4.10.B, DPR-29 a.

Present requirements of Specification 3.10.D, Core Monitoring, combined with STS guidelines on " shorting links",

modified to reflect the GE endorsed Quad Cities methodology, are used to write the proposed LCO for new 3.10.B.

b.

Proposed Applicability for Core Monitoring requirements of Operational Mode 5 during Core Alterations and control rod movement, is taken from present requirements.

c.

Proposed Actions for Specification 3.10.B are taken from STS guidelines.

d.

Utilizing STS guidelines and Quad Cities " shorting links" methodology, present SR 4.10.B in rewritten into proposed SRs 4.10.B.1, 4.10.D.2, 4.10.B.3 and 4.10.B.4.

Item 3:

Page 3.10/4.10-3, Specification 3.10.C/4.10.C, DPR-29 a.

Present requirements of Specification 3.10.C, Fuel Storage Pool Water Level, are used to write the LCO for new 3.10.C.

b.

For new specification 3.10.C, the proposed Applicability of whenever irradiated fuel is stored in the fuel storage pool is taken from present requirements.

c.

Proposed Actions for new Specification 3.10.C are taken from STS guidelines.

d.

Proposed SR 4.10.C is taken from present requirements of SR 4.10.C.

Item 4:

Pages 3.10/4.10-3 and 3.10/4.10-4, Specification 3.10.D/4.10.D, DPR-29 a.

Proposed Specification 3.10.D/4.10.D, control Rod and Control Rod Drive Maintenance - Single Control Rod Removal, is based on STS guidelines and replaces present Specification 3.10.D/4.10.D, Control Roa and Control Rod Drive Maintenance, i

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l b.

Proposed LCO 3.10.D/4.10.D allows only one control rod and/or control rod drive mechanism to be removed for maintenance provided several specified conditions are met, c.

Proposed Applicability is operational Modos 4 and 5.

d.

Proposed Actions for Specification 3.10.D ure taken from STS guidelines and require that with the provisions of the LCO not not, that control rod and/or control rod drive maintenance be suspended and action be initiated to satisfy the LCo requirements.

c.

Proposed Surveillance Requirements are based on STS guidelines and contain tests to verify that the conditions of the LCo are satisfied within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of removal of a control rod drive and/or control rod drive mechanism and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter until the control rod and/or control rod drive mechanism is reinstalled and the control rod in inserted in the core.

Item 5:

Pages 3.10/4.10-4 and 3.10/4.10-5, Specification 3.10.E/4.10.E, DPR-29 a.

Proposed Specification 3.10.E/4.10.E, Extended Core Maintenance - Multiple Control Rod Removal, is based on STS guidelines and replaces present Specification 3.10.E/4.10.E, Extended Core Maintenance.

b.

Proposed LCO 3.10.E allows any number of control rods and/or control rod drive mechanisms to be removed for maintenance provided the specified conditions are. net, c.

Proposed Applicability is linited to operational Mode 5.

d.

Proposed Actions for Specification 3.10.E are taken from STS guidelines and require suspension of control rod and/or control rod maintenance if the conditions of the LCo are not met.

The proposed Action further requires initiation of action to satisfy the conditions of the LCo.

e.

Proposed Surveillance Requirements in 4.10.E.1 are based on STS guidelines and contain tests to verify that the conditions of the LCO are satisfied within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of removal of a control rod drive and/or control rod drive mechanism and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter until the control rod and/or control rod drive mechanism is reinstalled and the control rod is inserted in the core.

Proposed SR 4.10.E.2 is based on ST3 guidelines and requires a functional test of the "one-rod-out" interlock, if this function had been bypassed, following replacement of all control rods and/or control rod drive mechanisms.

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Item 6:

Pages 3.10/4.10-5 and 3.10/4.10-6, Specification 3.10.F/4.10.F, DPR-29 a.

Present requirements of Specification 3.10.F, dpent Fuci Cask Handling, are-used to write the LCO for proposed Specification 3.10.F.

b.

Proposed Applicability for Specificaticn 3.10.F of during spent fuel cask handling operations is based on the intent of present requirements.

c.

Proposed Action 3.10.F.1 is written to add provisions which are not in the current technical specifications and that are needed to specify that when not operating in compliance with the LCO, steps are to be taken to suspend spent fuel cask handling operations and to initiate action to satisfy the LCO provisions.

Proposed Action 3.10.F.2 implements the provisions of present Specification 3.10.F.3.

d.

Proposed SRs 4.10.F.1, 4.10.F.2 and 4.10.F.3 are based on present provisions of section 4.10.F.

e.

New prcposed Specification 3.10.G/4.10.G, Control Rod Position, is added based on STS guidelines, f.

New proposed Specification 3.10.H/4.10.H, Commuaications, is added based on STS guidelines.

g.

New proposed Specification 3.10.I/4.10.I, Water Level -

Reactor Vessel, is added based on STS guidelines.

Item 7:

Pages 3.10/4.10-7 through 3.10/4.10-10, Bases 3.10/4.10, DPR-29 Rewrite present Bases 3.10/4.10 to allow implementation of the changes discussed above.

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DESCRIPTION OF CHANGES Proposed TS 3.10/4.10

'REFUELING

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DESCRIPTION OF PROPOSED AMENDMENT REQUEST Proposed Specification 3.10/4.10 Refueling The changes proposed in this amendment request are made to 1) improve the understanding and usability of the present technical specifications, 2) incorporate technical improvements, and 3) include some provisions from later operating BWR plants.

An item by item description of the proposed changes requested is provided below.

The Summary of Changes section can be referred to in order to reference back to a given change and its affected page.

GENERIC CllANGES Itom 1-Tbc present Quad Cities technical specifications contain Applicability and objective statements at the beginning of most sections.

These statements are generic in nature and do not provide any usoful information to the user of the technical specifications.

The proposed change will dolote tho objectivo statement and provide Applicability statements within each specification similar to the STS.

The proposed Applicability statement to be included in each specification will include the Reactor Hodos or other conditions for which the LCO must be satisfied.

Item,2 The changes proposed in t.iis item will provido an STS type of format while retaining the present two column layout.

The present format does not provide a separation of LCO, Applicability, and Action requiroments that are easily understood and identified.

The two column layout has been utilized at Quad cities sinco initial licensing of the plant and is preferred by the plant over the single column STS layout.

SPECIFIC CHANGES Item 1 The proposed changes described in Item 1 involve the rewrito of present Specification 3.10.A/4.10.A, Refueling Interlocks, using STS format and incorporating present provisions along with STS guidelines.

The Refueling Interlocks specification addresses the operability of the reactor modo switch and the refueling interlocks associated with the Refuel position of the reactor modo

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I switch.

Present provisions are used to writo proposed LCo 3.10. A with the addition of STS guidelines to allow the reactor modo switch to be in either the Shutdown or Rufuel position.

Proposed reactor modo switch in Refuel interlocks include control rod blocks, refueling platform reverso motion blocks, and refueling platform hoist blocks.

Proposed Applicability for Refueling Interlocks implements present intent of during Coro Alterations.

The proposed Applicability clarifles present requirements by requiring operability in operational Mode 5 during Coro Alterations with equipment associated with the Refuel position interlocks.

Proposed Actions for Specification 3.10.A are based on STS guidelines since present specifications do not contain remedial Action statements.

Proposed Action 3.10.A.1 requires that with the reactor modo switch not locked in the Shutdown or Refuel position, core alterations are suspended and the reactor modo switch is required to bc locked in the Shutdown or Refuel position.

Proposed Action 3.10.A.2 requires that with the one-rod-out interlock inoperable, the reactor mode switch be locked in the Shutdown position.

Proposed Action 3.10.A.3 requires that with any of the required Refuel position equipment interlocks inoperable, core alterations with equipment associated with the inoperable Refuel position equipment interlock be suspended.

Present Surveillance Requirements are retained and are rewritten into proposed SRs 4.10.A.1, 4.10.A.2, and 4.10.A.3.

Proposed SR 4.10.A.4 is added from STS guidelines in order to provide verification that the reactor modo switch is locked in the Shutdown or Refuel position during Core Alterations.

Item 2 This item describes the rewrite of present Specification 3.10.B/4.10.B, Coro Monitoring.

Present provisions are used for proposed LCO 3.10.B such that at least two SRMs are required operable and are required to be inserted to the normal operating level.

Present reatrictions on SRMs are retained so that one of the detectors is located in the quadrant where fuel or control rods are being moved and one is in an adjacent quadrant.

Also, the SRM is required to have a minimum count rate of 3 cps with all rods fully inserted except when no more than two fuel assemblies are present in the core quadrant associated with the SRM and those two fuel assemblics are in locations adjacent to the SRM.

Included in the proposed LCO is the present provision that allows the use of special movable detectors in place of the SRMs as long as they are connected to the normal SRM circuits.

Additional provisions regarding " shorting links" are added as proposed LCO 3.10.B.3.

These new provisions require the removal of shorting links from the RPS circuitry, in accordance with newly proposed Specification 4.10.B.4, prior to and during the time any control rod is withdrawn (except for control rods removed per proposed Specification 3.10.D or 3.10.E) and shutdown margin has

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not been cemonstrated.

Proposed Applicability of Operational Mode 5 during Core Alterations and control rod movement implemunts present intent in 3.10.B.

Present Specification 3.10.B does not contain remedial Action statements and STS guidelines are adopted.

Proposed Action 3.10.B requires that with the provisions of the LCO not met, that all o"erations involving Core Alterations and control rod movement be su.) ended and that all insertable control rods be inserted.

Present Surveillance Requirements in 4.10.B are augmented based on STS guidelines and rewritten as proposed SRs 4.30.B.1, 4.10.B.2, 4.10.B.3 and 4.10.B.4.

Under proposed SR 4.10.B.1, each SRM channel vill be demonstrated to be onorable once por 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

a) performance of a channel check; b)

erification that the detectors are inserted to the normal operating level; and c) during core alterations, verification that the detector of an aporablo SRM channel is located in the core quadrant where core alterations are being performed and another is located in an adjacent quadrant.

In addition, performance of a channel functional test will be required by SR 4.10.B.2, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of core alterations, and at least once por 7 days.

Under SR 4.10.B.3, verification that the channel count rate is at least 3 cps will be requiredt a) prior to control rod withdrawal; b) prior to and et least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during core alterations; and c) at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

To maintain assurance of adequate shutdown capability during refueling, new proposed SR 4.10.B.4 requires that with the control rod withdrawal function operable, the SRM circuitry " shorting links" be removed, or with four SRMs operable and monitoring the fueled region, two SRM " shorting links", placing the SRM scram logic in the coincident modo, be removed, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during the time any control rod is withdrawn (except for control rods removed per proposed Specification 3.10.D or 3.10.E) and shutdown margin has not been demonstrated.

Proposed SR 4.10.B.4 on the removal of " shorting links", was formulated using the General Electric guidelines outlined in Potentially Reportable Condition (PRC) 89-01 provided by GE pursuant to 10 CFR 21 on 3/1/89.

Item 3 Present Specification 3.10.C/4.10.C, Fuel Storage Pool Water Level, is rewritten into an STS format as proposed Specification 3.10.C/4.10.C, utilizing present provisions and STS action statements.

Proposed LCO 3.10.C implements present requirements to maintain at least 33 feet of water in the fuel storage pool.

The proposed Applicability implements present provisions of whenever irradiated fuel is stored in the fuel storage pool.

Since present provisions do not contain remedial Action statements, STS guidelines are adopted.

Proposed Action 3.10.C requires that with the spent fuel pool level not met, all operations involving handling of fuel assemblies and crane l

operations with loads in the spent fuel storage area be suspended, after the fuel assemblies and crane load is placed in a safe condition.

Present Surveillance Requirenent to record the fuel storage pool level at least once a day is retained as proposed SR 4.10.C.

Item 4 Item 4 describes the rewrite of present Specification 3.10.D/4.10.D, Control Rod and control Rod Drive Maintenance, using STS guidelines into proposed Specification 3.10.D/4.10.D, control Rod and control Rod Drive Maintenance - Single Control Rod Removal.

Present Specification 3.10.D/4.10.D allows two control rods and/or control rod drive mechanisms to be removed for maintenance provided the reactor mode switch is locked in Refuel, Shutdown Margin requirements are met, and required SRMs are operable.

Proposed Specification 3.10.D/4.10.D implements STS guidelines which are more restrictive than present provisions.

The proposed Specification will allow only one control rod and/or control rod drive mechanism to be removed for maintenance at a time.

Proposed LCO requirements also include requiring the reactor modo switch to be locked in the Shutdown or Refuel position, SRMs to be operable per Specification 3.10.D, the Shutdown Margin requirements to be mot, and all other control rods in a five-by-five array centered on the control rod being removed are inserted and disarmed or the fuel assemblics in the affected core cell are removed.

Proposed Applicability is operational Modes 4 and 5 in accordance with STS guidelines and clarifies present provisions to lock the reactor mode switch in the Refuel position.

The proposed restrictions on a single control rod removal are sufficient to allow this maintenance to be performed in the specified operational Modes.

Proposed Action 3.10.D is taken from STS guidelines since present specifications do not contain remedial action requirements.

Proposed Action 3.10.D requires that with the provisions of the LCO not met, removal of the control rod and/or associated control rod drive mechanism from the core and/or reactor vessel be suspended and that action be initiated to restore the LCO provisions.

Proposed Surveillance Requirement 4.10.D replaces present provisions which only address shutdown Margin requirements.

The proposed SRs require tests to be performed to demonstrate compliance with the conditions of the LCO within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of a control rod and/or control rod drive mechanism removal from the core and/or reactor pressure vessel, and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter until a control rod and associated control rod drive mechanism are reinstalled and the control rod is inserted in the core.

The SRs include verifying the reactor mode switch is operable and locked in the Shutdown or Refuel position with the "one-rod-out" interlock operable.

The SRs also verify

e that the required SRM channels are operable, Shutdown Margin requirements are met, rods in a five-by-five square array are inserted and disarmed or the affected core cell is defueled, and that all other control rods are inserted.

Item 5 Item 5 describes the rewrite of present Specification 3.10.E/4.10.E, Extended Ccro Maintenance, using STS guidelines into proposed Specification 3.10.E/4.10.E, Extended Core Maintenance - Multiple Control Rod Removal.

The proposed Specification contains provisions addressing the removal for maintenance, of more than one control rod and/or control rod drive mechanism.

The proposed use of STS guidelines for thia Specification will provide a more complete set of requirements for this maintenance task than are contained in present provisions.

The proposed LCO allows any number of control rods and/or control rod drive mechanisms to be removed from the core and/or reactor vessel provided certain conditions are met.

These conditions include having an operable reactor modo switch locked in the Shutdown or Refuel position, SRMu operable per Specification 3.10.B, Shutdown Margin requirements met, all other control rods inserted or their core cells defueled, and the core cell being worked on, being defueled.

Proposed Applicability of Operational Mode 5 follows STS guidelines and present intent of locking the reactor mode switch in Refuel for these operations.

Proposed Action 3.10.E is added from STS guidelines since present specifications do not contain remedial action requirements.

Proposed Acticn 3.10.E requires that with the provisions of the LCO not met, removal of the control rods and/or control rod drive mechanisms from the core and/or reactor vessel be suspended and that action be initiated to satisfy the above requirements.

Proposed Surveillance Requirements in 4.10.E aro based on STS guidelines and replace present provisions.

Present provisions only require certification that a control rod's control cell contains no fuel assemblies prior to control rod withdrawal for extended core maintenance.

The proposed SRs will verify all conditions specified in the LCO within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core.

The conditions verified include that the reactor mode switch is operable and locked in the Shutdown or Refuel position, the SRM channels are operable per Specification 3.10.B, Shutdown Margin requirements are met, all other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell, and the core cell on which maintenance is being performed is defueled.

Proposed SR 4.10.E.2 implements STd i

1 a

guidelines and requires the performance of a functional test of the "one-rod-out" interlock following replacement of all control rods and/or control rod drive mechanisms, if this function had been bypassed, l

r Item 6 item 6 of this amendment request details the rewrite of present Specification 3.10.F/4.10.F, Spent Fuel Cask Handling into proposed Specification 3.10.F/4.10.F.

The proposed changes implement present provisions and some 3TS guidelines for similar equipment.

Proposed Spucification 3.10.F/4.10.F contains the restrictions on spent fuel cask handling operations in the reactor building.

Proposed LCO 3.10.F contains present provisions that restrict spent fuel cask handling above the 623 foot level of the reactor building to the restricted mode of operation except in emergency or equipment failure situations and then only to get the cask to the closest acceptable stable location.

Proposed Applicability implements present intent of during spent fuel cask handling operations.

Present Specification 3.10.F.3 is rewritten as proposed Action 3.10.F.2 and requires that with a failed controlled area limit switch, operation is permissible for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provided an operator verifies operation of the crane within the restricted zone painted on the floor.

proposed Action 3.10.F.1 is added based on STS guidelines for similar situations, to address conditions where the 4

LCO is not met, other than the one addressed in proposed Action 3.10.F.2 for a failed controlled area limit switch.

Proposed Action 3.10.F.1 requires that with the provisions of the LCO not mot, that spent fuel cask handling operations be suspended and action be initiated to comply with the LCO provisions.

Present Surveillance Requirementu in 4.10.F are retained in proposed SR 4.10.F.

Present Quad Cities Technical Specifications do-not contain explicit provisions requiring that all control rods be inserted (except control rods removed per proposed Specification 3.10.D or 3.10.E) while in Operational Mode 5 during core alterations.

Proposed Specification 3.10.G/4.10.G, based on STS guidelines, is added in order to address the necessary requirements for these conditions.

Proposed LCO 3.10.G provides the explicit requirement that all control rods be inserted (except control rods removed per proposed Specification 3.10.D or 3.10 E) while in-Operational Mode 5 during core alterations.

With all control rods not inserted, proposed Action 3.10.G requires suspension of all core alterations, except that one control rod may be withdrawn under control of the reactor modo switch Refuel position one-rod-out interlock.

In accordance with STS guidelines, proposed SR 4.10.G requirns that all control rods (except control rods removed per Specification 3.10.D or 3.10.E) be verified to be inserted within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to: a) the start of core alterations; ana b) the i

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withdrawal of one control rod under the control of the reactor mode switch Refuel position one-rod-out interlock.

Proposed SR 4.10.G further requires that this verification be re-performed at 1 cast once overy 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Present Quad Cities Technical Specifications do not contain a requirement that direct communication be maintained between the control room and refueling floor personnel while in Operational Mode 5 during core alterations.

Proposed Specification 3.10.H/4.10.H, based on STS guidelines, is added in order to address the necessary requirements for those conditions.

Proposed LCO 3.10.H provides the requirement that direct communication bc maintained between the control room and refueling floor personnel i

while in operational Mode 5 during core alterations.

When direct communication cannot be maintained between the control room and

)

refueling floor personnel, proposed Action 3.10.H requires immediate suspensior, of core alterations.

In accordance with STS guidelines, proposed SR 4.10.H requires that direct communications between the control room and refueling floor personnel be demonstrated within one hour prior to the start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during core alterations.

Present Quad Cities Technical Specifications do not contain provisions for reactor vessel water level during handling of fuel assemblies or control rods within the reactor pressure vessel while in operational Mode 5 when the fuel assemblies being handled are irradiated or the fuel assemblies seated within the reactor vessel are irradiated.

Proposed Specification 3.10.I/4.10.I, based on STS guidelines, is added in order to address the necessary requirements for these conditions.

Proposed LCO 3.10.1 requires maintenance of 23 feet of water over the top of the reactor pressure vessel flange.

The proposed LCO provides the minimum water level-required during handling of fuel assemblics or control rods witnin the reactor pressere vessel while in Operational Mode 5 when the fuel assemblies being handled are irradiated or the fuel assemblics seated within the reactor vessel are irradiated.

When this minimum reactor vessel water level cannot be satisfied, proposed Action 3.10.1 requires suspension of all operations involving handling of fuel assemblies or control rods within the reactor pressure vessel, after all fuel assemblies and control rods have been placed in a safe condition.

In accordance with STS guidelines, proposed SR 4.10.I requires that the reactor vessel-water level be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during handling of fuel assemblics or control rods within the reactor pressure vessel.

Item 7 The rewrite of the Bases-for Section 3.10/4.19 is limited to j

changes that are necessary to implement the changes proposed to l

the individual specifications.

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PROPOSED TECH SPEC TS 3.10/4.10

' REFUELING'

4 0

QUAD CITIES UllITS 1& 2 DPR-29 & DPR-30 3.10/4.10 REFUELIliG LIMITING CollDITIONS FOR OPERATION SURVEILLANCE REQUIREMEllTS A.

Refueling Interlocks A.

Refueling Interlocks The reactor modo switch shall be 1.

The refueling interlocks OPERABLE and locked in the SHUT-shall be demonstrated DOWN or REFUEL position.

When OPERABLE by performance of a the reactor mode switch is locked CHAN!!EL FUNCTIOllAL TEST in the REFUEL position the fol-prior to CORE ALTERATIO!(S lowing interlocks shall be associated with the REFUEL OPERABLE:

position interlocks.

1.

Control Rod Blocks 2.

A CHANNEL FUNCTIONAL TEST of the refueling interlocks a.

Mode switch in shall also be performed at STARTUP/ HOT STAllDBY least once per week after and refueling platform the initial test in 4.10. A.1 over the reactor,

above, until no longer required.

b.

Fuel on any refueling holet and refueling 3.

The refueling interlocks platform over the shall be demonstrated reactor.

OPERABLE by performance of a CilAllNEL FUNCTIONAL TEST c.

Mode switch in REFUEL following any repair work with one control rod associated with the withdrawal permit, interlocks.

2.

Refueling Platform Reverse 4.

The reactor mode switch Motion (toward reactor shall be verified to be vessel) Blocks locked in the SilUTDOWN or REFUEL position as a.

Mode switch in specified:

STARTUP/ HOT STANDBY.

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior b.

Any control rod out to:

and fuel on any refueling hoist.

1)

Beginning CORE ALTERATIONS, and 3.

Refueling Platform Hoist Blocks 2)

Resuming CORE ALTERATIOllS when a.

Any control rod out the reactor modo and fuel on any switch has been refueling hoist over

unlocked, the vessel.

3.10/4.10-1

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 b.

Iloist overload.

b.

At least once por 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

ti l a h position limitation.

APPLICAIllLLIy1, OPERATIOliAL MODE 5 during CORE ALTERATIONS with equipment associated with the REFUEL position interlocks.

ACTIONt 1.

With the renctor modo switch not locked in the SilUTDOW!1 or REFUEL position as speci-fiod, suspend CORE ALTERA-TIO!1S and lock the reactor modo switch in the SliUTDOWii or REFUEL position.

2.

With the oneurod-out intor-lock inoperable, lock the reactor modo switch in the SHUTDOWii position.

3.

With any of the abovo required REFUEL position equipment interlocks inoperable, suspend CORE ALTERATIONS with equipment associated with the inoperablo REFUEL position equipment interlock.

B.

Core Monitoring B.

Core Monitoring At least two SRM's shall be Each of the required SRM channels OPERABLE and inserted to the shall be demonstrated OPERABLF.

norral operating lovel.

The use by:

of special movable detectors during fuel loading and CORE 1.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months !

ALTERATIONS in place of the normal SRM nuclear detectors is a.

Performance of a

permissible as long as these CHANNEL CllECK, special detectors are connected to the normal SRM circuits.

Core b.

Verifying the detec-3.10/4.10-2

e QUAD CITIES UllITS 1 & 2 DPR-29 & DPR-30 monitoring shall moet the tors are inserted to following the normal operating level, and 1.

One of the requirod SRM detectorn shall be located c.

During CORE ALTERA-in the quadrant where fuel TIollS, verifying that or control rods are being the detector of an noved and one in an adjacent OPERADLE SRM channel

quadrant, is located in the core quadrant where CORE 2.

The SRM or special movable ALTERATIONS are being detector shall have a

performed and another minimum of 3 cps with all is located in an rods fully. inserted in the adjacent quadrant.

coro except when both of the following conditions are 2.

Performanco of a

CitA!111EL mott FUNCTIONAL TEST:

a.

No more than two fuoi a.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior assemblics are present to the start of CORE in the coro quadrant ALTERATIONS, and associated with the SRM.

b.

At least once por 7 days.

b.

While in core, those fuoi assemblies are in 3.

Verifying that the channel locations adjacent to count rato is at least a the SRM.

cps:

3.

The " shorting links" shall a.

Prior to control rod be removed from the RPS withdrawal, circuitry per 4.10. B. 4 prior to and during the time any b.

Prior to and at least control rod is withdr Wn once por 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (except for control rods during CORE ALTERA-removed por Specification TIONS, and 3.10.D or 3.10.E) and SHUTDOWN MARGIN has not boon c.

At least onco por 24 demonstrated, hours.

APPLICABILITY:

4.

With the control rod vith-draws 1 function

OPERADLE, OPERATIONAL MODE 5 during CORE verifying, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ALTERATIONS a r.d control rod prior to and at least once

movement, por 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during, that:

ACTION!

a.

The SRM circuitry

" shorting links" have With the requirements of the been removed, or 3.10/4.10-3 r

-iw

.ce.--

=,-e

.m.

e.e-w

---v

,-w--+

.+

e

-+----1,

= - -,

wwe e

e QUAD CITIES UllITS 1 fr 2 DPR-29 & DPR-30 above specification not b.

With four SRMs OPERA-satisfied, suspend all operations BLE and monitoring the involving CORE ALTERATI0lls and fueled region, two SRM control rod movement and insert

" shorting links",

all insortablo control rods, placing the SRM scram logic in the coinci-dont modo, have boon

removed, during the timo any control rod is withdrawn (except for control rods removed por Specification 3.10.D or 3.10.E) and SilUTDOWil MARGIll has not boon demonstrated.

C.

Puol Storage Pool Water Lovel C.

ruel Storage Pool Water Lovel At least 33 foot of water shall At least onco por day, record the be maintained in the fuel storage fuel storago pool lovel.

pool.

APPLICABILITYt Whenover irradiated fuel is stored in the fuel storago pool.

ACTIOf41 With the requiromonta of the above specification not satisfied, suspend all operations involving handling of fuel assemblios and crano operations with loads in the spent fuel storage pool area after placing the fuel assemblies and crano load in a safo condition.

D.

Control Rod and Control Rod Drive D.

Control Rod and Control Rod Drive Maintenance - Single Control Rod Maintenance - Single Control Rod Romoval Romoval one control rod and/or the Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start associated control rod drive of removal of a

centrol rod mechanit.m may be removed f rom the and/or the associated control rod core and/or reactor proscuro drive mechanism from the core vossal provided that at least the and/or reactor pressuro vossol following requirements are and at least onco por 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.10/4.10-4

e e

j QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 satisfied until a control rod and thereaf ter until a control rod associated control rod drive and associated control rod drive mechanism are reinstalled and the mechanism are reinstalled and the control rod is fully inserted in control rod is inserted in the the cores core, verify that:

1.

The reactor mode switch is 1.

The reactor mode switch is OPERABLE and locked in the OPERABLE per Surveillance SHUTDOWN position or in the Requirement 4.1.A.1 or REFUEL position per Table 4.10. A.4, as applicable, and 1-2 and Specification locked in the SHUTDOWN 3.10.A.

position or in the REFUEL position wii.h the "one-rod-2.

The source range monitors out" REFUEL position inter-(SRM) are OPERABLE per lock OPERABLE per Specifica-Specification 3.10.B.

tion 3.10 A.

3.

The SilUTDOWN MARGIN 2.

The SRM CHANNELS are requirements of Specifi-OPERABLE por Specification cation 3.3.A are satisfied, 3.10.B.

except that the control rod selected to be removed; 3.

The-SilDTDOWN MARGIN requirements of Specifica-a.

May be assumed to be tion 3.3. A are satisfied per the highest worth specification 3.10.D.3.

control rod required to be assumed to be 4.

All other control rods in a fully withdrawn by the five-by-five array centered SHUTDOWN MARGIN test, on the control rod being and removed are inserted and electrically or hydrauli-b.

Need not be assumed to cally disarmed or the four be immovable or fuel assemblies surrounding untrippable.

the control rod or control rod drive mechanism to be 4.

All other control rods in a removed from the core and/or-five-by-five array centered reactor vessel are removed on the control. rod being from the core cell, removed are inserted and electrically or hydrauli-5.

All other control rods are cally disarmed or the four inserted.

fuel assemblies surrounding the control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.

3.10/4.10-5

e QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 5.

All other control rods are inserted.

APPLICABILITYt I

OPERATIONAL MODES 4 and 5.

ACTIONt j

With the requirements of the above specification not satisfied, suspend removal of the control rod and/or associated control rod drive mechanism from the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.

E.

Extended Core Maintenance E.

Extended Core Maintenance Multiplo Control Rod Removal Multiple Control Rod Removal Any nuinber of control rods and/or 1.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the control rod drive mechanisms may start of removal of control be removed from the core and/or rods and/or control rod reactor pressure vessel provided drive mechanisms from the that at least the following core and/or reactor pressure requirements are satisfied until vessel and at least once per all. control rods and control rod 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter until drive mechanisms are reinstalled all control rods and control and all control rods are inserted rod drive mechanisms are in the core.

reinstalled and all contrcl rods are inserted in the 1.

The reactor mode switch is core, verify that:

OPERABLE and locked in the SHUTDOWN position or in the a.

The reactor mode REFUEL position per Specifi-switch is OPERABLE per cation 3.10.A.

The REFUEL Surveillance Require-position "one-rod-out" ment 4.1.A.1 or interlock may be bypassed, 4.10.A.4, as applica-as

rsquired, for those ble, and locked in the control rods and/or control SHUTDOWN position - or rod drive mechanisms to be in the REFUEL position

removed, after the fuel per Specification assemblies have been removed 3.10.A.

as specified below.

E b.

The SRM CHANNELS are 2.

The source range monitors OPERABLE per Specifi-(SRM) are OPERABLE per cation 3.10 B.

Specification 3.10.B.

3.10/4.10-6

f QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 3.

The SHUTDOWN MARGIN c.

The SHUTDOWN MARGIN i

requirements of Specifica-requirements of Speci-tion 3.3.A are satisfied, fication 3.3.A are satisfied.

4.

All other control rods are i

either inserted or have the d.

All other control rods 1

surrounding four fuel are either inserted or assemblies removed from the have the surrounding core cell.

four fuel assemblies 1

removed from the core 5.

The four fuel assemblies

cell, surrounding each control rod or control rod drivo e.

The four fuel assem-mechanism to be removed from blies surrounding each the core and/or reactor control rod and/or vessel are removed from the control rod drive me-core cell.

chanism to be removed from the core and/or APPLICABILITY r reactor vessel are removed from the core OPERATIONAL HODE 5.

cell.

ACTION!

2.

Following replacement of all control rods and/or control With the requirements of the rod drive mechanisms removed above specification not in accordance with this satisfied, suspend removal of specification, perform a

control rods and/or dontrol rod functional test of the drive mechanisms trom the core "one-rod-out" REFUEL inter-and/or reactor pressure vessel lock, if this function had and initiate action to satisfy been bypassed, the above requirements.

F.

Spent Fuel Cask Handling F.

Spent Fuel Cask Handling Fuel cask handling shall ' be re-1.

Prior to fuel cask handling stricted as follows:

operations, the redundant crane including the rope,

1. -

Fuel cask handling above the hooks, slings, shackles and 623-foot level of the other operating mechanisms reactor building shall be shall be inspected.

-The I

done with the reactor rope will be replaced if any building crane in the of the following conditions restricted mode only, except exist:

as speciflod below, a.

Twelve randomly dis-2.

Fuel cask handling in other tributed broken wires-than the restricted mode in one lay or four will be permitted in broken wires in one 3.10/4.10-7 I

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 emergency or equipment strand of one rope failure situations only to lay.

the extent necessary to get the cask to the closest b.

Wear of one-third the acceptable stable location.

original diameter of outside individual APPLICABILITYt wire.

During spent fuel cash handling c.

Kinking, crushing, or operations.

any other damage re-sulting in distortion ACTION:

of the rope.

1.

With the above requirements d.

Evidence of any type not satisfied, except as of heat damage.

specified below in ACTION 3.10.F. 2, suspend spent fuel e.

Reductions from cask handling operations and nominal diameter of initiate action to satisfy more than 1/16 inch the above requirements.

for a

rope diameter from % inch to l\\ inch 2.

With a

failed controlled inclusive, area limit switch, operation is permissible for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 2.

Prior to operations in the providing an operator is on restricted

mode, the the refueling floor to following shall be tested assure the crane is operated within the restricted zone a.

The controlled area painted on the floor, limit switches.

b.

The "two-block" limit switches.

c.

The

" inching hoist" controls.

3.

Prior to any series of fuel cask handling operation, the empty spent fuel cask will be lifted free of all support by a maximum crf 1 foot and left hanging for 5 minutes.

G.

Control Rod Position G.

Control Rod Position All control rods shall be All control rods shall be inserted, except for control rods verified to be inserted, except 3.10/4.10-8 l

l

e 5

QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 removed per Specification 3.10.D control rods removed per or 3.lO.E.

Specification 3.10.D or 3.10.El APPLICABILITYr 1.

Within.t2 hours prior to OPERATIONAL MODE 5, during CORE a.

The start of CORE ALTERATIONS.

ALTERATIONS.

ACTIONt b.

The withdrawal of one control rod under the With all control rods not control of the reactor inserted, suspend all other CORE mode switch Refuel ALTERATIONS, except that one position one-rod-out control rod may be withdrawn interlock.

under control of the reactor modo switch Refuel position one-rod-2.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

out interlock.

H.

Communications H.

Communications Direct communication shall be Direct communications between the maintained between the control control room and refueling floor room and refueling floor personnel shall be demonstrated personnel, within one hour prior to the start of and at least once per 12 APPLICABILITY hours during CORE ALTERATIONS.

OPERATIONAL MODE 5, during CORE ALTERATIONS.

ACTION:

When direct communication between the control room and refueling floor

. personnel cannot be maintained, immediately suspend CORE ALTERATIONS.

I.

Water Level - Reactor Vessel I.

Water Level - Reactor Vessel At least 23 fact of water shall The reactor vessel water level be maintained over the top of the shall be determined to be at reactor pressure vessel flange.

least its minimum required depth-within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start APPLICABILITY:

of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during handling of fuel assem-During handling of fuel assem-blies or control rods within the blies or control rods within the reactor pressure vessel, reactor pressure vessel while in 3.10/4.10-9 l

. ~

~

e QUAD CITIES UllITS 1 & 2 DPR-29 & DPR-30 OPERATIONAL MODE 5 whan the fuel assemblies being handled are irradiated or the fuel assemblics scated within the reactor vessel are irradiated.

ACTIONL With the requirements of the above specification not satisfied, suspend all operations involving handling of fuel assemblics or control rods within the reactor pressure vessel af ter placing all fuel assemblies and control rods in a safe condition.

3.10/4.10-10

.e QUAD CITIES UNITS 1 & 2 DPR-29 & DPR-30 3.10/4.10 RETUELING BASES A.

During refueling operations, the reactivity potential of the core is being a* ered.

It is necessary to require u

certain interlocks and restrict certain refueling i

procedures such that there is asourance that inadvertent criticality does not occur.

I To minimize the possibility of loading-fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the1 reactor core.

This requirement assures that during refueling, the refueling interlocks will prevent inadvertent criticality as designed. The core reactivity-limitation of Specification 3.3 limits the CORE ALTERATIONS to assure that the resulting core loading can be aontrolled with the reactivity control system and interlocks at any-time during shutdown or the following OPERATING CYCLE.

The addition of large amounts of reactivity to the core is prevented by operating procedures, which are in turn backed up by refueling interlocks on rod withdrawal and movement of the refueling platform. When the mode switch is in the REFUEL position, interlocks prevent the refueling platform from being moved over the core if a-control-rod is withdrawn and fuel is on a hoist.

Likewise, if the refueling platform is over the _ core with fuel on a hoist, control rod motion is blocked by the

' interlocks. With the mode switch in the REFUEL position, only one control rod can be withdrawn.

B.

The SRMs are provided to monitor the core during periods of station shutdown and to guide the operator during refueling operations and station startup.

Requiring two-operable SRMs in or adjacent to_any core quadrant where f uel or control-rods are being moved assures adequate monitoring _of that quadrant during such alterations.

Requiring a minimum-of 3 counts per second whenever criticality is possible provides assurance that neutron flux is being monitored. Criticality is considered to be L

impossible if there are no more than two assembliss in a quadrant and if these are in locations adjacent to the SRM.- In this case -only, the SRM or dunking type detector count rate is permitted to be less than 3 counts per second.

B 3.10/4.10-1 l1

-.--.-. -.w..

c.

l QUAD CITIES UNITS 1 &2 DPR-29 & DPR-30 C.

To assure that there is adequate water to shield and cool the irradiated fuel assemblies stored in the pool, a minimum pool water level is established.

The minimum water level of 33 fest is established because it would be a significant change from the normal level (37 feet 9 3

inches), well above a level to assure adequate cooling (just above active fuel), and above the level at which GSEP action is initiated (5 feet uncontrolled loss of level with lovel decreasing).

D.

The intent of this specification is to permit the removal of one control rod and/or control rod drive mechanism from the core and/or reactor pressure vessel for maintenance purposes.

This operation is performed with the mode switch in the REFUEL or SHUTP/.RN position during OPERATIONAL MODES 4 or 5 to ensure th3t maintenance or repair of control rod drives will be performed under conditions that limit the probability of inadvertent criticality.

E.

The intent of this specification is to permit the unloading of a significant portion of the reactor core for such purpt.ses as inservice inspection requirements, examination of the core support plate, etc.

This specification provides assurance that inadvertent criticality does not occur during such operation.

This operation is performed w!.th the mode switch in the REFUEL or SHUTDOWN position during OPERATIONAL MODE 5 in order to limit the probahtlity ef inadvertent criticality.

T.

-The operati ) of the redundant crane in the Restricted Mode during fuel cask handling operations acsures that the cask remsins within the controlled area once it has been removed from its trnisport vehicle (i.e., once it is above the 623 foot elevation).

Handling of the cask on the Refueling Floor in the Unrestricted Mode is allowed only in the case of equipment failures or emergency ccnditions when the cask is already suspended.

The Unrestricted Mode of operation is allowed only to the extent necessary to get the cask to a suitable stationary position co that required repairs can be made. Operation with a failed position so that required repairs can be made.

Operation with a

failed controlled area microswitch will be allowed for a 48-hour period provided an Operator is on the floor in addition to the crane operator to assure that the cask handling is limited to the controlled area as marked on the floor.

This will allow adequate time to make repairs but still wil? not restrict cask handling operations unduly.

B 3.10/4.10-2

2 e

i I

QUAD CITIES UNITS 1 d 2 DPR-29 & DPR-30 The Surveillance Requirements specified assure that the redundant crane is adequately inspected in accordance with the accepted ANSI Standard (B.30.2.0) and manufacturer's recommendatiuns to determine that the equipment is in satisfactory condition.

The testing of the controlled area limit switches assures that the crane operation will be limited to the designated area in the Restricted Mode of operation.

The test of the "two-block" limit switch assures the power to the hoisting motor will be interrupted befora an actual "two-blocking" incident can occur.

The test of the inching haist assures that this mode of load control is available when required.

Requiring the lifting ar.d holding of the c?,sk for 5 minutes during the initial lift of each series of cask handling operations puts a load test on the entire crane lifting mechanism as well as the braking system.

Performing this test when the cask is being lifted initially from the cask car assures that the system is operable prior to lifting the load to an excessive height.

G.

The requirement that all ccatrol rods be inserted during other CORE ALTERATIONS minimizes the possibility that fuel will be loaded into a cell without a control rod, although one rod may be withdrawn under control of the reactor mode switch refuel position one-rod-out interlock.

H.

The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during movement of fuel within the reactor pressure vessel.

I.

The restrictions on minimum water level for the reactor vessel ensure that sufficient water depth is available to remove the assumed iodine activity released from the rupture of an irradiated fuel assembly.

This minimum water depth is consistent with the assumptions of the accident analysis.

B 3.10/4.1's-3

I EXISTING TECH SPEC TS 3.10/4.10

' REFUELING"

-... - ~ -. -.. _ -.. -...

ce QUAD-C111ES OPR 29-i 3.10/4.10 -REFUELING' t

I LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

Applicability:

Applicability:'

Applies to fuel handling and core Applits to the periodic testing of those-i reactivity limitations.~

interlocks and instruments used during refueling.

Objectiv' e:

Objective:

To assure core reactivity is within To verify the operability of instrumenta-capabilit) of the_ control rods and to tion and interlocks used in refueling, prevent criticality during refueling.

SPECIFICA110NS-i L

A.

Refueling Interlocks A.

Refueling Interlocks I

The reactor modo switch shall be-Prior to any fuel handling, with the J

locked in the Refuel position during head off the reactor vessel, the re-core alterations, and the refueling fueling interlocks shall be: function-l interlocks listed below shall be ally tested. _They shall also be operable except as specified in' tested at weekly intervals-thereafter

. Specifications 3.10.0 and 3.10.E.

until no longer required and fol-lowing any repair work associated with the interlocks.

1.

Control Rod Blocks a.

Mode switch in Startup/ Hot-

-Standby and refueling platform over the reactor.

-b.

Fuel on any refueling hoist and refueling platform over.-

the reactor, c.

Mode switch in Refuel with' one control rod withdrawal L

_ permit.

~2.

Refueling Platform Reverse l.

Motion (toward recctor vessel)

Block f-l-

L l-

~

L 3.10/4.10-1 Amendment No. 114

QUAD-ClllES DPR-29 V

a.

Mode switch in Startup/

Hot Standby.

b.

Any control rod out and fuel on any refueling hoist.

3.

Refueling Platform Holst Blocks a.

Any control rod out and fuel on any refueling hoist over the vessel, b.

Hoist overload, c.

High position limitation.

B.

Core Monitoring B.

Core Monitoring During core alterations, two SRM's Prior to any alterations to the core, shall be operable, one in the core the SRM's shall be functionally test-quadraat where fuel or control rods ed and checked for neutron response, are being moved and one in an Thereaf ter, the SRM's shall be adjacentquadrant. For an SRM to be checked daily for response, except considered operable, the following when the conditions of 3.10.B.2.a and conditions shall be satisfied:

3.10.B.2.b are met.

1.

The SRM shall be inserted to the normal operating level (use of special movable, dunking type detectors du .g initial fuel loading and major core al-terations in place of normal detectors is permissible as long as the detector is connected into the proper circuitry which contains the required rod blocks).

3.10/4.10-2 Amendment No. 114

e l

QUAD-CITIES DPR-29 2.

The SRM or dunking type de-tector shall have a minimum of 3 cps with all rods fully inserted in the core except when both of the following conditions are fulfilled:

a.

No more than two fuel assemblies are present in the core quadrant associated with the SRM.

b.

While in core, these fuel assemblies are in locations adjacent to the SRM.

C.

Fuel Storage Pool Water Level C.

Fuel Storage Pool Water Level Whenever irradiated fuel is stored in Whenever irradiated fuel is stored in the fuel storage pool, the pool water the fuel storage pool, the pool level level shall be maintained at a level shall be recorded daily.

of at least 33 feet.

D.

Control Rod and Control Drive D.

Control Rod and Control Rod Drive Maintenance Maintenance.

A maximum of two nonadjacent control rods separated by more than two con-trol cells in any direction may be withdrawn from the core for the pur-pose of performing control rod and/or control rod drive maintenance pro-vided the following conditions are satisfied:

1.

The reactor mode mitch s' hall be 1.

Sufficient control rods shall be locked in th"

...ael po'sition.

withdrawn prior to performing The refueling interlock which this maintenance to demonstrate prevents more than one control with a margin of 0.25% ak that rod from being withdrawn may be the core can be made subcriti-bypassed for one of the control cal at any time during the rods on which m intenance is maintenance with the strongest being performed. All other re-operable control rod fully fueling interlocks shall be withdrawn and all other operable operable.

rods fully inserted.

3.10/4.10-3 Amendment No. 114

e QUAD-ClilES OPR-29 Alteraately, if a minimum of eight control rods surrounding each control rod out of service for maintenance are to be fully inserted and have their di-rectional control valves electrically disarmed, the 0.25%

Ak margin will be met with the gg strongest control rod remaining in service duriag the maintenance period fully withdrrwn.

2.

Specification 3.3. A.1 shall be met, or the control rod directional control valves for a minimum of eight control rods surrounding each drive out of service for maintenance will be disarmed electrically and suf-ficient margin to criticality demonstrated.

3.

SRM's shall be operable (a) in each core quadrant containing a control rod on which maintenance is being performed, and (b) in a quadrant adjacent to one of the quadrants specified in Specification 3.10.D.3.(a) above. Requirements for an SRM to be considered operable are given in Specification 3.10.B.

E.

Extended Core Maintenance E.

Extended Core Maintenance More than two control rods may be Prior to control rod withdrawal for withdrawn from the reactor core extended core maintenance, that con-provided the following conditions are trol rod's control cell shall be satisfied:

certified to contain no fuel assemblies.

r 3.10/4.10-4 Amendment No. 114

e o

QUAD-CITIES DPR-29 1.

The reactor mode switch shall be locked in the Refuel position.

The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed o*i a withdrawn control rod after the fuel assemblies in the cell containing (controlled by) that control rod have been removed from the reactor core.

All other refueling interlocks shall be operable.

2.

SRM's shall b3 operable in the core quadrant where fuel or control rods are being moved and in an adjacent quadrant. The requirements for an SRM to be considered operable are given in Specification 3.10.B.

F:

Spent Fuel Cask Handling F.

Spent f uel Cask Handling 1.

Fuel Cask handling above the 623' level of the Reactor 1.

Prior to fuel cask handling Building will be done with the operations, the redundant crane reactor building crane in the including the rope, hooks, RESTRICTED H0DE only, except as slings, shackles and other specified in 3.10.F.2.

operating mechanisms will be inspected.

The rope will be replaced if any of the following conditiuns exist:

a.

Twelve (12) randomly distributed broken wires in one lay or four (4) broken wires in one strand of one rope lay.

b.

Wear of one-third the original diameter of outside individual wire.

3.10/4.10-5 Amendment No. 114 1

no:

QUAD CITIES DPR 29

- c;

-Kinking, crushing, or any-other damage resulting in-distortion of the rope, d.

Evidence of any type of heat damage.:

e.

Peductions from nominal di'ameter of more than:l/16 inch for a rope diameter from 7/8" to 1 1/4" inclusive.

2.

Fuel cask handling in other than 2.

Prior to operations in the the RESTRICTED MODE will be-RESTRICTED MODE permitted in' emergency or-equipment failure situations-

- only to the extent necessary to

- get the cask to the closest acceptable stable location, a.

The controlled area limit switches will be tested;=

b.

the "two-block" limit' switches'will be tested; c.

the _" inching-hoist" controls will be tested.-

3.

- Operation with a failed 3,

The empty spent fuel' cask will controlled area-limit switch is be lifted-free of all-support-by-permissible for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months a maximum of 1 foot and left providing an operator is on the hanging for'5 minutes prior to

- refueling floor to assure.the any_ series of_ fuel ~ cask handling

' crane is operated within the-operations, restricted zor.e painted on t_he floor.

1 3.10/4.10-6 Amendment No. 114 l

m

-..,-e

=

C e

L QUAD-C11105 DPR-29 3.10 UMillNG CCNDlHONS TCR OPERMiON 0 ALES S.10 /4.10.hroc uN Cr SMSES A.

Durirg refueling operations, the reactivity potential of the core is being altered.

It is necessary to require certain interlocks and restrict certain refueling procedures such that there is assurance that inadvertent criticality does not occur.

To minimize the possibility of loading fuel into a c' ell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the reactor core. This requirement assures that during refueling, the refueling interlocks will prevent of Specification 3"/ limits the //N //////////>7 activity limitation inadvertent criticality as designed. The core r 03 to assure that the resulting core loading can be controlled with the reactivity control s stem and interlocks at any time during shutdown or the following

/M The addition of large amounts of reactivity to the core is prevented by operating procedures, which are in turn backed up by refueling interlocks on rod withdrawal and movement of the refueling platform.

When the mode switch is in the R4W./ position, interlocks prevent the refueling platform from being moved over the core if a control rod is withdrawn and fuel is on a hoist.

Likewise, if the refueling platform is over the core with fuel on a hoist, control rod motion is blocked by the interlocks. With the mode switch in the R//Mf position, only one control rod can be withdrawn.

B.

TheSR$ireprovidedtomonitorthecoreduringperiodsofstation shutdown and to guide the operator during refueling operations and station startup, RequiringtwooperableSRginoradjacenttoany core quadrant where fuel or control rods are being moved assures adequate monitoring of that quadrant during such alterations.

Requiring a minin.um of 3 counts per second whenever criticality is possible provides assurance that neutron flux is being monitored.

Criticality is considered to be impossible if there are no more than two assemblies in a quadrant and if these are in locations adjacent to the SRM. In this case only, the SRM or dunking type detector count rate is permitted to be less than 3 counts per second.

C.

To assure thet there is adequate water to shield and cool the irradiated fuel assemblies stored in the pool, a minimum pool water level is established. The minimum water level of 33 feet is established because it would be a significant change from the nore-1 level (37 feet 9 inches), well above a level to assure adequate cooling (just above active fuel), and above the level at which G7f P action is initiated (5 feet uncontrolled loss of level with level decreasing).

3.10/4.10-7 Amendment No. 114 I.

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E.

The intent of this specification is to permit the unloading of a significant portion of the reactor core for such purposes as inservice inspection requirements, examination of the core sLpport plate, etc.

This specification provides assurance that inadvertent criticality does not occur during such operation.

Thi; ernretica is perfor;;d with itc med: ;;it:F :- th " fuel p%4t,ie,r t:__ ___, _i_ a pt:';!d: the,r-fueli

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F.

The operation of the redundant crane in the Restricted Mode during fuel cask handling operations assures that the cask remains within the controlled area once it has been removed from its transport vehicle (i.e., once it is above the 623 foot elevation). Handling of the cask on the Refueling Floor in the Unrestricted Mode is allowed only in the case of equipment f ailures or emergency conditions when the cask is already suspended. The Unrestricted Mode of operation is allowed only 3.10/4.10-8 Amendment No. 114 1

e INSERT A FOR TECHNICAL SPECIFICATION BASES PAGE 3.10/4.10-8 The intent of this specification is to permit the removal of one control rod and/or control rod drive mechanism from the core This/or reactor pressure vessel for maintenance purposes.

and operation is performed with the mode switch in the REFUEL or SHUTDOWN position during OPERATIONAL MODES 4 or 5 to ensure that maintenance or repair of control rod drives will be performed under conditions that limit the probability of inadvertent criticality.

INSERT B FOR TECHNICAL SPECIFICATION BASES PAGE 3.10/4.10-8 This operation is performed with the mode switch in the REFUEL or SHUTDOWN position during OPERATIONAL MODE 5 in order to limit the probability of inadvertent criticality.

l 1

4-g QUAD-C111ES DPR-29 to the extent necessary to get the cask to a suitable stationary position so the required repairs can be made.

Operation with a failed controlled area microswitch will be allowed for a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period eJ providMg an Operator is on the floor in addition to the crane operator to assure that the cask handling is limited to the controlled area as marked on the floor. This will allow adequate time to make repairs but still will not restrict cask handling operations unduly.

The Surveillance Requirements specified assure that the redundant crane is adequately inspected in accordance with the accepted ANSI Standard (B.30.2.0) and manufacturer's recommendations to determine that the equipment is in satisfactory condition.

The testing of the controlled area limit switches assures that the crane operation will be limited to the de*;ignated area in the Restricted Mode of operation. The test of the "two-block" limit switch assures the power to the hoisting motor will be interrupted before an actual "two-blocking" incident can occur.

The test of the inching hoist assures that this mode of load control is available when required.

Requiring the lifting and holding of the cask for 5 minutes during the initial lif t of each series of cask handling operations puts a locd test on the entire crane lifting mechanism as well as the braking system.

Performing this test when the cask is being lifted initially from the cask car assures that the system is operable prior to lifting the load to an excessive height.

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3.10/4.10-9 Amendment No. 114 l

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3.10/4.10-10 Amendment No. 114 i

4 0

INSERT C FOR TECHNICAL SPECIFICATION BASES PAGE 3.10/4.10-9 The requirement'that all control rods be inserted during other CORE ALTERATIONS minimizes the possibility that fuel will' be loaded into a cell without a control rod, _ although one rod.may be withdrawn

- under control of the-reactor mode switch refuel position one-rod-

-out interlock.

INSERT D FOR TECHNICAL SPECIFICATIONS BASES '# AGE 3.10/4.10-9 The requirement for communications capability ensures that refueling station personnel can be' promptly informed of significant changes in the f acility status or core reactivity condition during

' movement of: fuel within-the reactor pressure vessel.

INSERT E FOR TECHNICAL SPECIFICATION BASES PAGE 3.10/4.10- 9 The restrictions.- on minimum water level for the reactor vessel ensure -that. sufficient water depth is available to remove the assumed iodine activity released from the-rupture of an irradiated fuel-assembly.

This-minimum water-depth is consistent-with.the assumptions of'the-accident analysis.

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SIGNIFICANT HAZARDS-CONSIDERATIONS AND-ENVIRONMENTAL ASSESSMENT EVALUATION l

PROPOSED TS 3.10/4.10 l

' REFUELING' l-

n EVALUATION FOR SIGNIFICANT HAZARDS CONSIDERATION Proposed Specification 3.10/4.10 Refaeling The proposed changes provided in this amendment request are made in order to provide a more user friendly document, incorporate desired technical improvements, and to incorporate some improvements from later operating BWRs.

These changes have been reviewed by Commonwealth Edison and we believe that they do not present a Significant Hazards Consideration.

The basis for our determination is documented as follows:

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards consideration.

In accordance with the criteria of 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, because:

a.

The Generic Changes to the technical specifications involve administrative changes to format and arrangement of the material.

As such, these changes cannot involve a significant increase in the probability or consequences of an accident previously evaluated.

b.

Proposed Changes to the Refueling Interlocks Specifications The proposed changes to the Refueling Interlocks specifications include the retention of present Refueling Interlock requirements in Operational Mode 5, during core alterations, and the addition of STS guidelines to allow the reactor mode switch to be in either the Refuel or Shutdown position.

The proposed addition of the Shutdown reactor mode switch position provides consistency with the addition of the STS definition for Operational Mode 5 in Table 1-2.

Present Applicability is during core alterations which is retained in the proposed changes.

Clarifications are added to the proposed Applicability such that Refueling Interlocks are required in Operational Mode 5 during core alterations with equipment associated with the Refuel position interlocks.

Proposed Actions are added where none presently exist and are taken from STS guidelines.

The proposed Actions ensure that with Refueling Interlock LCO requirements not met, safe conditions are established so that Core Alterations are suspended and/or the teactor mode switch is locked in

n the Shutdown or Refuel position.

Present Surveillance Requirements are retained in the proposed specification and STS provisions are added to verify that the reactor mode switch is locked in the Shutdown or Refuel position.

The proposed changes maintain necessary operability of the Refueling Interlocks when required to perform their design function.

No new Operational Modes are introduced by the proposed changes and the intent of present provisions is maintained.

Therefore, the proposed changes do not involve a s!gnificent increase in the probability or consequences of an accident previously evaluated.

c.

Proposed Changes to the Core Monitoring Specifications Present Core Monitoring requirements are retained in the proposed rewrite of these specifications.

The present requirement for two SRM's to be operaolo during core alterations is retained along with the requirements for the SRM's to be inserted to the normal operating level, for one to be located in the quadrant where core alterations are taking place and one in an adjacent quadrant, and the present limitations associated with requiring a minimum of 3 cps.

Thu present allowance for use of a special movable detector in place of an SRM is retained.

To maintain assurance of adequate shutdown capability during refueling, an enhancement to the present LCO is provided by inclusion of a " shorting link" removal requirement based on STS guidelines, as modified by General Electric recommendations and Quad Cities station practices.

Present Surveillance Requirements are expanded by incorporation of STS surveillance provisions coupled with guidelines provided by General Electric regarding the removal of " shorting links" from the RPS circuitry prior to and during the time any control rod is withdrawn and shutdown margin has not been demonstrated.

The present Core Monitoring specifications do not contain remedial action provjsions.

Proposed Actions are taken from proven STS guidelines and require that with the requirements of the LCO not met, i Core Alterations and control rod movement is st pended and all insertable control rods are inserted.

Since at least the present level of Core Monitoring is retained by the proposed changes, they do not involve a significant increase in the probability or consequences of an accident previously evaluateG.

d.

Proposed Changes to the Fuel Storage Pool Water Level Specifications Present provisions are used to develop the proposed changes to the Fuel Storage Pool Water Level specifications with the addition of STS Action guidelines.

Present water level requirements of 33 feet I

n l-in une fuel storage pool are maintained when irradiated fuel is stored in the fuel storage pool.

Present daily Surveillance Requirement to record the fuel storage pool level is retained.

The proposed Actions added from the STS are needed since no remedial action provisions currently exist.

The proposed Actions require with the water level requirement not met, that all operations involving handling of fuel assemblies and crane operations with loads in the spent fuel storage pool area be suspended after placing the fuel assemblies and crane load in a safe condition.

The proposed changes help to ensure that safe conditions are establiched in the fuel storage pool when the pool water level is less than the lower limit of 33 feet.

Other present provisions, that are in current use, are retained.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

e.

Proposed Changes to the Control Rod and Control Rod Drive Maintenance - Single Control Rod Removal Specifications Present provisions currently in use at Quad Cities for Control Rod and Control Rod Drive maintenance are replaced with more prescriptive STS guidelines.

Present limitations of a maximum of two nonadjacent control rods that are allowed to be withdrawn for maintenance are replaced with requirements that only allow one control rod and/or control rod drive mechanism to be removed.

Restrictions in place in order to allow this maintenance include requiring the reactor mode switch to be operable and locked in the Shutdown or Refuel position, the required SRMs to be operable, Shutdown Margin requirements to be met, and all other control rods in a five-by-five square array centered on the control rod to be removed, to be inserted and disarmed or the affected core cell being worked on to be defueled.

The proposed LCO is more restrictive than present provisions and provides necessary controls on this maintenance activity.

The proposed Applicability of Operational Modes 4 and 5 reflects STS guidelines and provides acceptable plant conditions to perform the required maintenance considering the restrictions on these activities which are contained in the LCO.

Present Surveillance Requirements are replaced with SRs that verify all the restrictions on control rod and/or control rod drive removal that are contained in the LCO.

The addition of STS Action provisions require suspension of maintenance activities if the provisions of the LCO are not met, and require that action be initiated to satisfy the LCO requirements.

The addition of the STS allowance for the reactor mode switch to be in the Shutdown or Refuel position for this maintenance adds

6 n

operational flexibility without reducing any protective features.

In order to withdraw control rods, the reactor mode switch must still be placed in the Refuel position.

Control Rod Drive Mechanisms can be removed with the reactor modo switch in either the Shutdown or Refuel position.

The proposed changes add more restrictive STS guidelines that are applicable for use at Quad Cities.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated, f.

Proposed Changes to the Extended Core Maintenance -

Multiple control Rod Removal Specifications The proposed changes for the Extended Core Maintenance -

Multiple control Rod Removal specifications are similar to those for Control Rod and Control Rod Drive Maintenance - Single Control Rod Removal.

Present provisions currently in use at Quad Cities for Extended Core Maintenance are replaced with STS guidelines that contain more prescriptive LCO, Action and Surveillance Requirements.

The LCO restrictions on removal of any number of control rods and/or control rod drive

-mechanisms include requiring the reactor mode switch to be operable and locked in the Shutdown or Refuel position, the SRMs to be operable per Specification 3.10.B, the Shutdown Margin requirements to be met, all other control rods to be either inserted or their core cells defueled, and the core cell being worked on is defueled.

Proposed Applicability is restricted to l

Operational Mode 5 in accordance with STS guidelines.

l Present specifications do not contain Action provisions and STS guidelines are followed for this addition.

The proposed Action requires suspension of the removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel if the conditions of the LCO are not satisfied.

Present Surveillance Requirements only verify that a core cell is defueled prior to maintenance activities.

This present SR is replaced with STS SRs that verify all the conditions of the LCO within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to start of i

the removal of control rods and/or control rod drive i

mechanisms and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter I

until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the l

core.

The addition of the STS allowance for the reactor mode switch to be in the Shutdown or Refuel position for this i

maintenance adds operational flexibility without reducing any protective features.

In order to withdraw control rods, the reactor mode switch must be in the Refuel position.

Control Rod Drive Mechanisms can be removed with the reactor modo switch in either the i

O n

Shutdown or Refucl position.

The proposed changes add more restrictive STS provisions on control rod and/or control rod drive mechanism removal for maintenance that-are applicable for use at Quad Cities.- Therefore, the propoqed changes do not involve.a significant increase-in the probability or consequences of an accident previously evaluated.

g.

Proposed Changes to the Spent Fuel Cask Handling Specifications Present limitations on Spent Fuel Cask Handling are retained in the proposed rewrite of this specifiaation.

The proposed addition to this specification is an Action to specify that when operations are not in compliance with the LCO, that steps are to be taken to suspend spent fuel cask handling operations and to initiate action to satisfy the LCo provisions.

Present limitations that are retained include limiting the fuel cask handling above the 623 foot level of the reactor building to the rest'icted mode and allowing handling in other than the restricted mode at this level, only for emergency or equipment failure situations.

The present Action that is retained addresses a failed controlled area limit switch and allows operations to continue for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months as long as an operator verifies the crane is operated within the restricted zone painted on the floor.

Present Surveillance Requirements are retained and have been shown through use to provide assurance of safe Spent Fuel Cask Handling.

Since present, proven specifications are retained and provisions added help to ensure continued safe handling of the Spent Fuel Cask, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

h.

Proposed Addition of Control Rod Position Specifications The proposed addition of Specification 3.10.G/4.10.G, Control Rod Position, represents incorporation of a requirement that does not presently exist in the Quad Cities Technical Specifications.

These new provisions are based on STS guidelines and require that all control rods be inserted (except control rods removed per proposed Specification 3.10.D or 3.10.E) while in operational Mode 5 during core alterations.

With all control rods not inserted, proposed Action 3.10.G requires suspension of all core alterations, except that one-control rod may be withdrawn under control of the reactor mode switch Refuel position one-rod-out interlock.

In accordance with STS guidelines, proposed SR 4.10.G requires that all control rods (except control rods removed per Specification 3.10.D or 3.10.E) be verified to be inserted within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to: a) the

O' r3 I

start of core alterations; and b) the withdrawal of one control rod under the control of the reactor mode switch Refuel position one-rod-out-interlock.

Proposed SR 4.10.G further requires that this verification be re-performed at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The proposed addition of Specification 3.10.G/4.10.G, adds restrictions not presently in the Quad Cities Technical Specifications and helps to ensure that all control rods are inserted (except control rods removed per proposed Specification 3.10.D or 3.10.E) while in Operational 1

Mode 5 during core alterations.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

i.

Proposed Addition of Communications Specifications The proposed addition of Specification 3.10.H/4.10.H, communications, represents incorporation of a requirement that does not presently exist in the Quad Cities Technical Specifications.

These new provisions 1

are based on STS guidelines and require that direct communication be maintained between the control room and refueling floor personnel while in operational Mode 5 during core alterations.

When direct communication cannot be maintained between the control room and refueling floor personnel, proposed Action 3.10.H requires immediate suspension of core alterations.

In accordance with STS guidelines, proposed SR 4.10.H requires that direct communications between the control room and refueling floor personnel be demonstrated within one hour prior to the start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during core alterations.

The proposed addition of Specification 3.10.H/4.10.H, adds restrictions not presently in the Quad cities Technical Specifications and helps to ensure maintenance of direct communication between the' control room and refueling floor personnel while in Operational Mode 5 during core alterations. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

j.

Proposed. Addition of Reactor Vessel Water Level Specifications The proposed addition of Specification 3.10.I/4.10.I, Water Level - Reactor Vessel, represents incorporation of a requirement that does not presently exist in the Quad Cities Technical Specifications.

These new provisions are based on STS guidelines and require maintenance of 23 feet of water over the top of the reactor pressure vessel flange during handling of fuel assemblies or control rods within the reactor pressure vessel while in Operational Mode 5 when the fuel

o o

assemblies being handled are irradiated or the fuel assemblies seated within the reactor vessel are irradiated.

When this minimum reactor vessel water level cannot be satisfied, proposed Action 3.10.1 requires suspension of all operations involving handling of fuel assemblies or control rods within the reactor pressure vessel, after all fuel assemblies and control rods have been placed in a safe condition.

In accordance with STS guidelines, proposed SR 4.10.I requires that the reactor vessel water level be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during handling of fuel assemblies or control rods within the reactor pressure vessel.

The proposed addition of Specification 3.10.I/4.10.I, adds restrictions not presently in the Quad cities Technical Specifications and helps to ensure maintenance of a minimum amount of water over the top of the reactor pressure vessel flange during specific Operational Mode

5. activities.

Therefore, the proposed changes do not 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 i

from any previously evaluated because:

1 a.

Since the Generic Changes proposed to the technical specifications are administrative in nature, they cannot create the possibility of a new or different kind of accident from any previously evaluated, b.

The proposed changes for Quad Cities Technical Specification Section 3.10/4.10 are based on present provisions and STS guidelines or later operating BWR plants' NRC accepted changes.

These proposed changes have been reviewed for' acceptability'at the Quad cities Nuclear Station considering similarity of system or component design versus the STS or later operating BWRs.

No new modes of operation are introduced by the proposed changes, considering the acceptable operational Modes in present specifications, the STS, or later operating BWRs.

The proposed changes do not modify existing setpoints or design assumpt).ons for system or component operation.

Present, proven Surveillance Requirements are retained.

Proposed changes to Action statements in many places add requirements that are not in the present technical specifications or adopt requirements that have been used successfully at other operating BWRs with designs similar to Quad Cities.

The proposed changes either maintain at least the 1: resent level of operability or adopt relaxations to present requirements which still provide a proven acceptable level of operability.

Therefore, the proposed changes do not l

create the possibility of a new or different kind of

C n

accident from'any previously evaluated.

3)

InvolveLa significant reduction in the margin of safety because:

a.

Due to the administrative nature of the Generic Changes, they do_not involve a significant reduction in the margin of safety.

b.

The proposed changes to Technical Specification Section 3.10/4.10 implement present requirements, the intent of present requirements, or provisions that have been found acceptable for use on other operating BWRs with system designs similar to that at Quad Cities.

The proposed changes are intended to improve readability, usability, and the understanding of technical specification requirements while maintaining acceptable levels of safe operation.

The proposed changes have been evaluated and l

found to be acceptable for use at Quad Cities based on system design, safety analysis requirements and operational performance.

Since the proposed changes are based on NRC accepted provisions at other operating plants that are applicable at Quad Cities and maintain necessary levels of system, component or parameter operability, the proposed changes do not involve a significant reduction in the margin _of safety.

n ENVIRONMENTAL ASSESSMENT EVALUATION PROPOSED SPECIFICATION SECTION 3.10/4.10 REFUELING Commonwealth Edison has evaluated the proposed amendment in accordar.co with the requirements of 10 CFR 51.21 and has determined that the amendment meets the requirements for categorical exclusion as specified by 10 CFR 51.22(c)(9).

Commonwealth Edison has determined that the amendment involves no significant hazards consideration, there is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite, and there is no significant increase in individual or cumulative occupational radiation exposure.

The proposed amendment does not modify the existing facility and does not involve any new operation of the plant.

As such, the proposed amendment does not involve any change in the type or significant increases in effluents, or increase individual or cumulative occupational radiation exposure.

The proposed amendment to Section 3.10/4.10, " Refueling", contains administrative changes such as including appropriate applicability statements within the specifications to define the applicability during operating mode and the required actions to be implementei in the event tha* specification cannot be met.

The information is consistent with.ae Standard Technical Specifications or later operating plants.

In addition, some existing requirements have been updated and new requirements added to reflect the Standard Technical Specifications oi later operating plant requirements.

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COMPARISON-OF UNIT-1.AND-UNIT 2 TECHNICAL SPECIFICATIONS

- 1 FOR THE IDENTIFICATION OF TECKNICAL' DIFFERENCES SECTION 3.10/4.10 REFUELING Commonwealth Edison has conducted a comparison review of i

the Unit 1 and Unit 2 Technical Specifications to identify any technical differences in support of combining the

- Technical Specifications into one document.

The intent of-

- the review was.not to identify any differences in 1 presentation ~ style (e.g. table formats, use.of capital j

letters, etc.) or punctuation but rather to identify areas-

- which the Technical Specifications are technically or H

administratively different.

)

The-review of Section-3.10/4.10, Refueling" did not i

reveal any technical differences.

The following administrative. changes were identified:

1

-Pact.3.10/4.10-5 4.10.F.1.b-Unit 1:

wires in one strand of one rope lay Unit 2:

wires in one strand of rope' lay Unit 1:

original diameter of outside-Unit 2:

original diamater or outside Eggg 3.10/4.10-8 e

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ML20079F992

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ML20079F992

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