Text

Provides Addl Comments & Suggestions Re Proof & Review Version of Tech Specs.Marked-up Tech Spec Pages Encl
	ML20084E324
Person / Time
Site:	Byron
Issue date:	04/09/1984
From:	Tramm T; COMMONWEALTH EDISON CO.
To:	Harold Denton; Office of Nuclear Reactor Regulation
References
	8440N, NUDOCS 8405020232
	Download: ML20084E324 (51)
	v • d • e

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[D Commonwealth r

Edison

) One First Natiorrt Plan CNcago, ilhnois g ( O 7 Address Rep y to. Post Office Dox 767 (j Chicago, Illinois 60690 April 9, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Byron Generating Station Units 1 and 2 Technical Specifications NRC Docket Nos. 50-454 and 50-455 References (a): December 16, 1983 memorandum from Cecil 0.

Thomas.

(b): March 26, 1984 letter from T. R. Tramm to H. R. Denton.

(c): April 2, 1984 letter from T. R. Tramm to H. R. Denton.

Dear Mr. Denton:

This is to provide additional comments and suggestions regarding the proof and review version of the Byron 1 Technical Specifications that was distributed in reference (a). NRC review of specific changes proposed here is necessary before the Technical Specifications can be finalized.

Attachments A through J to this letter contain marked-up pages of various sections of the Technical Specifications. A summary explanation of the changes is provided for each attachment. Justifications are provided where appropriate.

A number of similar changes were submitted in references (b) and (c). We understand that the NRC will review each of these proposed changes and inform Commonwealth Edison of their acceptability.

Please direct any questions you may have regarding this matter to this office.

One signed original and fifteen copies of this letter and the attachments are provided for NRC review.

Very truly yours, l th .Adw _

T. R. Tramm Nuclear Licensing Administrator im f cc: Byron Resident Inspector \ g\

b0 O!000kS4 8440N A PDM

\.

ATTACHMENI A (Bases Section 2.0)

Circled items noted in this attachment have been previously submitted.

1. Section 2.1.1 (pg. 82-1) Reactor Core Delete paragraph beginning "The minimum value of the DNBR. . for all operating conditions." and replace with "The DNB design basis is as follows: There must be at least 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent confidence that the minimum DNBR for the limiting rods is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analysis using values of input parameters without uncertainties."

2. Section 2.2.1 (pg. 02-5) Reactor Trip System Instrumentation Setpoints In the first paragraph, delete "1.30" and replace with "the DNBR limit".

Per discussions between Westinghouse and CECO, the above changes ar e being made in order to better clarify the DNBR limit as it applies to Byron Station.

(0441M) a

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( ' '

, PROCF & REV!EW COPY

2.1 SAFETY LIMITS 1

BASES 2.1.1 REACTOR CORE <

The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DN8) and the resultant sharp reduction in heat transfer coefficient. DN8 is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the WRB-1 correlation. The WRB-1 DN8 correlation has ,

been developed to predict the DN8 flux and the location of DN8 for axially uniform and nonunifore heat flux distributions. The local DNS heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNS at a particular core location to the local heat flux, and is indicative of the margin to ONS.

of the Rjdur ng ateady- tat operation,, norma ~I~

operatibn, Jhe minimum valal Gandien j

add fpat pranfient 1 lin(tod toS.:Wf a "h typcal' celi aA1.32 fe.r a thfuble, cell. ,<TMs/v'alu,e crespondetoT,95Kg s probability at' a 95% confidenca level that DNKwi}hnot'9Ccdrfand As ghosen ^ '

as an'appropMate margin eto'ON8 fob al'1 operating' co'nditionsh curve fFig -1 6 shod e loci of points of THERMAL POWERrReactor lant tes-pressura_and - temperature for which the minimum DNB s no less than 1.34 for a typical call and 1.32 for a thimble cell, or the average enthalpy at the vessel exit is equal to the enthalpy of l saturated liquid.

N I.Y9 f These curves are based on an enthalpy hot channel factor, Fg, of p

! and a reference cosine with a peak of 1.55 for axial power shape. An allowance i

is included for an increase in F Ng at reduced power based on the expression:

N I* -

Fg = h 56 [1+ 0.3.(1-P)]

Where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the fg (M) function of the Overtemperature trip. When the axial power l l l

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l MiDOFTM can l( LIMITING SAFETY SYSTEM SETTINGS BASES Power Rance. Neutron Flux. Hiah Rates (Continued) p r The Power Range Negative Rata trip provides protaction far control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNOR to exi st. The Power Range Negative Rate tM p will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rata trip for those control red drop accidents for which DMSAs will be grestar than .

? I Dt h 924UI'..

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Intermediata and Source Ranos. Nuclear Flux The Intamediata and Sourca Range, Nuclear Flux tMps provide core protection during reactor START 1JP to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a suberitical 1C condition. These tM ps provide redundant protection to the Low setpoint tM p of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at aneut los counts per second unless manually blocked when P-6 becomes active. The Intermediata Range channels will initiata a Reactor tMp at a current level equivalent to approximately 2SX of RATED

- THOMAL POWER unless annually blocked when P-10 becomes active.

I Qvertamperaturs AT The Overtassersture AT trip provides core protection to prevent DNS for all combinations of pressure, power, coolant temperature, and axial power l '- distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

, and pressure is within the range between the Pressurizer High and Low Pressure I

trips. The Setpoint is automatically varied with: (1) coolant temperature to l correct for tamperature induced changes in density and heat capacity of water j and includes dynamic compensation for piping delays from the core to the loop i

temperature detectors, (2) pressuMzer pressure, and (3) axial power distribution.

With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1 '. If nial peaks are greater l

than design, as indicated by the difference between sop and bottom power range

' nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.1-1.

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ATTACHMENT B (Section 3/4.3)

Circled items noted in this attachment have been previously submitted.

1. Table 3.3.13 and Table 4.3-9 (pg. 3/4 3-67, 71) Radioactive Gaseous Effluent Monitoring Instrumentation Deletion of Ab Flow Rate Monitor, 4c. Sampler Flow Rate Monitor, 5d.

System Flow Rate Monitor and Se, Sampler Flow Rate Monitor are requested because these monitors are not needed to comply with the Technical Specifications. The only monitors required are vent stack monitors.

These instruments are not required to be operable for Noble Gas Activity Monitors.

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,,i TABLE 3.3-13 (Continued)

'E E RA010 ACTIVE GASEOUS EFFLUENT HONITORING INSTRUNENTATION r

x . *

MININUM CHANNELS APPLICABILITY ACTION

- OPERABLE E INSTRUMENT 7

r y. Gas Decay Tank System Delek

a. Noble Gas Activity k nitor - Providing 1 Alarm and Automatic Termination of

35 Release (ORE-PR002A and 28) 2 y l'

b. [10w' Rate,Nonitor' / /1 , /1 / ,/ ,/ ,/ * / / ./

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fn/_, I Sampler Flow Rate Monitor ' /

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, u y/. Containment Purge System 2 a. Noble Gas Activity Monitor - Providing

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[f Radioactivity h oitors Providing Alarm and Automatic Closure of Surge Tank Vent Component

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4 Cooling Wter Line (ORE-PR009 and IRE-PR009) 2

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5 TABLE 4.3-9 8

2 RAOIGACTIVE GASEOUS EFFLutNT MONITORING INSTRUNENTATION SURVEILLANCE REQUIRENENIS E DIGITAL CHANNEL MODES FOR IdilCH

_i CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE r

FUNCTIONAL UNIT CHECK CHECK CALIBRATION TEST IS REQUIRED Gas Decay Tank System

a. Noble Gas Activity Monitor - P)eleke P R(3) Q(1)

Providing Alarm and Automatic Termination of Release l ; (ORE-PR002A and 28)

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l ATTACHMENT C (Section 3/4.4)

Circled items noted in this attachment have been previously submitted.

1. Section 3/4.4.6 (pg. 3/4 4-18) Action Statement Delete the existing " Action" statement and replace with the following:

I a. With "a" of the above required Leakage Detection System inoperable, grab samples of the containment atmosphere must be obtained and analyzed for gaseous and particulate radioactivity at least onco per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With only two of the above required Leakage Detection Systems Operable, operation may continue for up to 30 days, otherwise, be in at least Hot Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ."

This change is for clarification of the existing statement.

2. Section 3/4.4.6 (pg. 3/4 4-18) Reactor Coolant System Leakage l In Surveillance 4.4.6.1.d delete " channel check" and "and Analog Channel l Operational Test". These two surveillances do not apply to the containment air pressure and reactor fan cooler outlet and inlet i temperature - performance.

3. Section 3/4.4.10 (pg. 3/4 4-38) Action a.

Delete the words "more than 50*F above the minimum temperature required by NDT considerations" and add "above 200*F."

The addition of this 200*F limit would prevent the plant from entering Mode 5 with a Class I component not conforming to ASME Section XI.

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(0441M) l

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REACTOR COOLANT SYSTEM

)

PRODF & i!MW COPY !

3/4.4.6 REACFo COOLANT SYSTEM LEAGGE 4 l

,1 s-

_ LEA GGE DETECTION SYSTEMS LIMITING CON 0! TION FOR OPERATION l

3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERA 8LE:

a. The Containment Atmosphere Particulate and Gaseous Radioactivity Nonitoring System,

b. The Containmaat Floor Orsin and Reactor Cavity Flow Monitoring System, and

c. The containment air pressure instrumentation and reactor containment fan cooler outlets and inlets Dewcall and dry bulb temperature . .

instrumentation.

APPLICA81GTY: MODES 1, 2, 3, and 4. , , , ,

.M* .

With on1h two'of the.above reg'uired Leakage,9etectief 5ystans OPE 8A8LE.

I4

- operation may continue for up to.30' days provided grab samples,ef the / g" 7 ' containment atmosphere are obtained and analyzed for_ gaseous and particulate-)

radioecitivity at least once per 24-hours phen the tequired 'Gaseousf or /

'.Particulata Radienctivity,Nonitoring System is operable;,etherwis'e, be in at least MOT STAN08Y within the Jtext 6 hess and(1'n COLQ,SHUTD0tef'within'the l'

~

. fol1owing 30 heves. 3 SURVEILLANCE REQUIREMerf5 .. -. . . .

j 4.4.6.1 The Leakage Detection Systems shall be demons,trated OPERA 8LE by: , ,

a. Containment Atmosphere Gaseous and Particulate Monitoring

{ Systam-performance of CHANNEL CHECK, CHANNEL CALIBRATION, '

CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3,

b. Containment Floor Orsin and Reacter Cavity Flow Monitoring Systarperformance of CHANNEL CAUSAATION at least once per I

i 18 months, and J. ) Containment air pressure and reactor containment fan coster outlet and inlet temperatures-performance of 4MANN56-4N64E, CHANNEL

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. 1 bO REACTOR COOLANT SYSTEM F'GCF & EEW COPY l 3/4.4.10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components 7'.

shall be maintained in accordance with Specification 4.4.10.

APPUCA81LITY: All M00ES.

1 ACTION

l D a. With the structural integrity of any ASME Code Class I component (s) '

I not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate i the affected component (s) prior to increasing the Reactor Coolant l System temperature acce-than-50*P1:evow-minfeum-tempeceh *

-regub'#try-NOT-consideratiore. *-- cxW 2.oo*F*.

O

b. With the structural integrity of any ASME Code Class 2 component (s) not co.1forving to the above recuirements, restore the structural integrity of the affected cogonefit(s) to within its limit or isolate the affected componect(s) prior to increasing the Reactor Coolant System temperature seove 200*F.

c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected cogonent(s) from service.

d. The provisions of Specification 3.0.4 are not gplicable.

l SURVEILLANCE REQUIREMENTS .

l 4.4.10 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1 August 1975.

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BYRON - UNIT 1 3/4 4*38

e.

ATTACHMENT D (Section 3/4.5)

Circled items noted in this attachment have been previ usly submitted.

1. Section 3/4.5.1 (pg. 3/4 5-1) Accumulators A contained borated water level of 34% has been changed to 64%. This agrees with the gallons quoted.

2. Section 4.5.3.2 (pg. 3/4 5-8) Surveillance Requirements Add an " asterisk" following the word " inoperable" on the second line in the paragraph. Also, add the following note and asterisk to the bottom of the page.

"N It may be desirable to operate the SI pumps for testing and also for filling the accumulator tanks while in mode 4. It is permissible to operate the SI pumps if the appropriate valves are locked closed to prevent pressurization of the RCS. To run the A-Pump, valves SI8802A and SI88210 must be locked closed. The 8-Pump may be used by locking closed valves SI8835, SI8802A and SI88028 with valves SI8821 A and SI8821B open.

When the accumulators are being filled block valves S188080, SI88088, SI6808C and SI88080 must also be locked closed."

Although this section requires that both safety injection pumps be demonstrated inoperable in Mode 4, it may be desirable to operate the SI pumps for testing and also for filling the accumulator tanks while in Mode 4. The above justifies the operation of the SI pumps.

(0441M)

9 PRODF & REVH COPY

.l 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator, shall be OPERABLE with:

a. The isolation valve open, '

lY*l*

b. A contained Forated water level of between 04% (6995 gallons) and 66% (7217 gallons),

c. A boron concentration of between 1900 21 m, and ,

d. A nitrogen cover pressure of between and psig.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 1 h ur or be in at least HOT STANOBY within tha next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 nours.

b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUT 00WN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS

4. 5.1.1 Each accumulator,shall be demonstrat,ed OPERABLE:

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

1) Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and

2) Verifying that each accumulator isolation valve is open.

" Pressurizer pressure above 1000 psig.

BYRON - UNIT 1 3/4 5-1

_ _ _ _ _ - _ - _ - _ _ _ _ - _ _ _ _ _ _ _ - - _ - _ _ - _ _ - - _ _ - - _ _ _ _ _ _ _ - _ - _ _ _ _ _ ~

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.-**4 EMERGENCY CORE COOLING SYSTEMS fgggg i SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERA 8LE per the applicable requirements of Specification 4.5.2.

I 4.5.3.2 All charging pumps and Safety Injection pumps, e.xcept the above required OPERA 8LE pumps, shall be demonstrated inoperable *by verifying that the motor circuit breakers are secured in the open position at least once per 12 houreddisisverwthe temperature of one or more of the RCS cold legs is less the ee-equel4aP'350*F

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f ATTACHMENT L (Section 3/4.8)

1. Section 4.8.1.1.2.d (pg. 3/4 8-3) A.C. Sources Surveillance Delete "obtained in accordance with ASTM-D2/0-1975" Reference to ASTM-D270-1975 has been deleted. The tank does not have sample taps located at the top, middle and bottom as specified in ASTM-D270-1975. Therefore, this part of the surveillance cannot be met with the present tank design.

2. Section 3/4.8.4 (pgs. 3/4 8-17 through 8-33) Electrical Equipment Protective Devices.

Justification The TRIP SETPOINT and RESPONSE 11ME columns are being deleted from the technical specifications Table 3.8-1 because the values shown are not representative of the data which should be conveyed to meet the intent of the technical specification Section 4.8.4.1. The data necessary to meet the intent of Section 4.8.4.1 consists of a definition of the acceptable functional criteria for all of the overcurrent protection devices such that they conform with the overcurrent protection design to maintain containment penetration integrity.

The information necessary to describe the acceptable functional criteria cannot be presented by a single trip setpoint and response time. Due to the nature of overcurrent protection devices, the operability limits can best be defined by a curve.

CECO therefore contends that the deletion of the TRIP SLIPOINT and RESPONSE TIME information will be less ambigious than tryinj to express either a single data point or a technical description for the designed response of the various overcurrent protectivo devices.

CECO will perform the testing of the overturrent protection devices in compliance with approved engineering procedures and vendor recommendations in order to prove that the devices remain within their operable limits as described in FSAR Q.40.13.

Also in Table 3.8-1 the information concerning the 125V D.C. feed through a penetration has been deleted. Upon investigation it was found that the overcurrent capacity of the D.C. feed cable to the penetration, is far less than the fault current capacity of the penetration. If a fault cordition appeared on this circuit ard the protectivo device failed to operate, the feed cable to the penetration for this circuit would rail before the integrity of the penetration was impaired. On this basis we propose the deletion of the 125VDC Pnl 114 circuit listed on page 3/4.8-33.

(0441M)

/ .

ELECTRICAL POWER SYSTEMS h hk SURVEILLANCE REQUIREMENTS (Continued)

a. In accordance with the frequency specified in Table 4.3-1 on a STAGGERED TEST BASIS by:

1) Verifying the fuel level in the day tank,

2) Verifying the fuel level in the fuel storage tank,

3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank,

4) Verifying the diesel starts from ambient condition and accelerates to at least 600 rps in less than or equal to 10 seconds. The 420 volts and generator voltage and frequency shall 60 3 1.2 Hz within 10 seconds after the start s be 4160 ignal. The diesel generator shall be started for this test by using one of the following signals:

a) Manual, or b) Simulated loss of ESF bus voltage by itself, or c) Simulated loss of ESF bus voltage in conjunction with an ESF actuation test signal, or d) An ESF actuation test signal by itself.

5) Verifying the generator is synchronized, loaded to greater than or ecual to 5500 kW in less than or eaual to 60 seconds. ooerates with a load greater than or equal to 5500 kW for at least 60 minutes, and

6) Verifying the diesel generator is aligned to provide standby power to the associated ESF busses.

b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by checking for and removing accumulated water from the day tanks;

c. At least once per 92 days by checking for and removing accumulated ,

water from the fuel oil storage tanks;

d. At least once per 92 days and from new fuel oil prior to its addi-tion to the storage tanks by vgrifying that a sample Cutd != -

r " r d e ^~ ~ 2:7; 1;7"*Esets the following minimum require-ments in accordance with the tests specified in ASTM-0975-1977:

1) A water and sediment content of less than or equal to 0.05 volume percent;

2) A kinematic viscosity of 40*C of greater than or equal to 1.3 centistakes, but less than or equal to 4.1 contistokes;

3) A specific gravity as specified by the manufacturer at 60/60*F of greater than or equal to 0.83 but less than or equal to 0.89 or an API gravity at 60*F of greater than or equal to 27 degrees but less than or equal to 39 degrees; BYRON - UNIT 1 3/4 8-3

). -

M UF & H EW COPY ELECTRICAL POWER SYSTEMS ,

) 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES _,

CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES _

LIMITING CONOITION FOR OPERATION 3.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be OPERA 8LE.

APPLICA81LITY: MODES 1, 2, 3, and 4.

)

ACTION: _,

With one or more of the above required containment penetration conductor overcurrent protective device (s) inoperable:

Restore the protective device (s) to OPERABLE status or de-energize

) a.

the circuit (s) by tripping the associated circuit breaker or racking out or removing the inoperable circuit breaker within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the circuit breaker to be tripped or the inoperable circuit breaker

racked out, or removed, at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent

)O' '-

devices in circuits which have their circuit breakers tripped, their inoperable circuit breakers racked out, or removed, or

b. Be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDChN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

) SURVEILLANCE REQUIREMENTS 4.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be demonstrated OPERA 8LE:

a. At least once per 18 months:

1) By verifying that the 7 kV circuit breakers are OPERA 8LE by selecting, on a rotating basis, at least 10% of the circuit ,

breakers, and performing the following: (

) a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated automatic actuation of the system-: d ...-ify' C i

thet-seeh-celey-end-eeeeefeted-c4 N i t :.. ..L., e e.%-e**te*L j -:i n;i t: f;;;d: :: C;i;;;d : d :: :;;:ifi.4 1., Te W 4.0 1. : d t's elemeHstra he C4af (Ae. vvef*// /tWetm 6,W1 pef tet/tn dest 3n teen,ws v/f Ah, sper=4/e /, wits,

l. 8YRON - UNIT 1 3/4 8-17 l

, l 1

+ '

O V

ELECTRICAL POWER SYSTEMS i SURVEILLANCE REQUIREMENTS (Continued) c) For each circuit breaker found inoperable during these ,

functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.

2) By selecting and functionally testing a representative sample of at least 10% of each type of 480-volt circuit breaker.

Circuit breakers selected for functional testing shall be selected on a rotating basis. The functional test shall consist of injecting a current input et t'r !??ct'f:d Setptid to each selected circuit breaker and verifying that each circuit breaker functions as designed : d 'l: rc:;; :: t b ' '---

t = = w eet te th: :;;;ifhd ;:?ue. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be func-O. tionally tested until no more failures are found or all circuit breakers of that type have been functionally tested; and

3) By selecting and functionally testing a representative sample of each type of fuse on a rotating basis. Each representative sample of fuses shall include at least 10% of all fuses of that type. The functional test shall consist of a nondestructive resistance measurement test which demonstrates that the fuse meets its manufacturer's design criteria. Fuses found inoper-able during these functional tests shall be replaced with OPERABLE fuses prior to resuming operation. . For each fuse found inoperable during these functional tests, an additional l representative sample of at least 10% of all fuses of that type shall be functionally tested until no more failures are found i

or all fuses of that type have been functionally tested,

b. At least once per 60 months by subjecting each 7 kV circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

O BYRON - UNIT 1 3/4 8-18

~

TABLE 3.8-1

]

o CONTAINNENT PENETRATION CONDUCTOR

[

E OVERCURRENT PROTECTIVE DEVICES T81P_ .

PROTECTIVE DEVICE SETPOINT '- ~~~ ~ ~~ RESPONSE IINE .

NUMBER AND LOCATION DEVICE (Amperes) (Sec/Cycie) ! /

1. 6.9 kV Switchgear IRC01PA-RCPA Primary Long time ; 1440x2.1 11.5 Bus 157 Cub 1 Inst. - 7686 N.A.

I \

l Bus 157 Nora. Feed Backup Long time - 4800x2 0:9 i w 1

ACB 1577 Bus 157 Emerg. Feed Backup Gr!. - 200 Long time - 4800

\ 0.3 0.9 l

cm AC8 1572 Inst. - 7680 N.A.

O I

1RC01P8-RCPS Primary Long time - 1400x2.1 11,6 Bus 156 Cub 2 Inst. - 7680 N/A.

I Bus 156 Nora. Feed Backup Long time - 4800x2 0.7 g AC8 1566 Gr. - 200 0.3 ::no Bus 156 Emerg. Feed Backup Long time - 4800x2 0.7 C3 l

l ACB 1562 Gr. - 200 0.3 M 1RC01PC RCPC Primary L time - 1440x2.1 11 5 N

Bus 158 Cub 5 Inst. - 7680 N. k.

Bus 158 Norm. Feed Backup i

Long time , 4800x2 0.7 N

\

2 ACB 1582 Gr. - 2,0tr 0.3 \

Bus 158 Emerg. Feed Backup Long, time - 4800x2 0.7 \' n AC8 1587 Gr/- 200 0. 3 - C3

- N i

TABIE 3.8-1 (Continued)

E g CONTAIWtENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE PTOT N RFSPOW NtDEER AND LOCATICN DEVICE res) (Sec/ Cycl i

1. 6.9 kV Switchgear (Continued) 1RC01PD - RCPD Primary tong time - 1440x2.1 11,6 Bus 159 Cub 5 Ihst. - 76 /A.

Bus 159 Norm. Feed Backup tong time - 4880x2 0.7 ACS 1591 Gr. - 200 \ 0.3 ,

w Bus 159 Emerg. Feed Backup [ongtime-4800x 0.7

) ACB 1597 Gr. - 200 0.3 m -

4 2. 480V Switchgear o

1RYO3EA - Pzr. Primary MCC8 - 100 N.A.

f Htr. Backup Group A l Compt. Al-A6, 81 Backup MCC8 - 100 N.A.

I w 1RYO3E8 - Pzr. Primarv MCCB - 100 N.A. :23 C

Htr. Backup Group 8 l , o Compt. 81-86, A1 Backup MCCB - 100 N. . -ei i / o.

IRYO3EC - Pzr. Primary MCC8 - 100 N.A.

Htr. Backup Group C /

Compt. Al-A6, 81 Backup 4CC8 100 N.A.

IRYO3ED - Pzr. Primary 4CC8 - 100 N.A.

Htr. Sackup Group D ,/ c3 Compt. 81-86, A1 Backup MCEO - 100 F_A- I C

o o o - ' - - -o O ~~ o g g g ; g, O

~'

l l TA8tE 3.8-1 (Continued) ,

5 i CONTAll0ENT PENETRA110N CONDUCTOR g

OVEEURRENT PROTECTIVE DEVICES E -

Al W -s g

P901ECTIVE DEVICE POINT RESP 00tSE,TitIE IssGER ANO LOCATIGIt DEVICE _res) (Sec/Cylle)I

3. 490W A.C. Ckt. Skrs.

l IVPelCA - RCFC Fan ,

! 1A Low Speed Feed Skr Sugr 131X Primary Liwig time - 450 20-32

( C 4 4C Inst. - 4,500 N.A.

N1 Speed Feed Skr Primary Llongtime-900 20-32 Sugr 131X Cub SC Inst. - 7,500 N.A.

? Bus 1311 Blors.

p Feed 141 Swor., Backup Leag time - 960 3.4 i C d 14, Enst. - 3,960 N.A.

ACS 1415 IVrelCC - KFC Fan IC Low Speed , y Feed Skr Suge Primary thngt1 - 450 20-32 :::.o O

1311 C 4 4C Inst. 4.500 l N. A>\ M Nt Speed Feed Skr Primary time - 900 20-32 \I g.

Suge 131x C e SC I'n'st. - 7.500 ft. A.

_s E

o C3

v v v v v v v v v v o

\_>

o

i. .

TABLE 3.8-1 (Continued)

C006TAlleENT PENETRATIOg rnunwTOR

OVERCUBAENT PROTECTIVE DEVICES E

"t 1 RIP #

r 1EIPOINT RESP 000SE 0E P90iECTIVE KVICE IESSER AIS LOCATION DEVICE 4 Amperes)

(Sec/C e)!

. t

3. 400W A.C. Ckt. Skrs. (Continued)

IVPelCS - KFC Fan 18 *

- Low Speed Feed Skr Primary thag time - 450 20-32 Sugr 132X C 4 4C lhst. - 4,500 N.A. '

l I

I Ni Speed feed Skr Primary Long time - 900 20-32 Sugr 132X C d SC 3ast. - 7,500 N.A.

s

Bus 132X isors. Feed Backup ILong time - 960 3.4 142 Sugr. , Cd 14, inst. - 3,960 N.A.

? I l 53 ACS 1423 f

l

- IVPolCD - SCFC Fan 18 Low Speed feed Skr Primary Long tjee - 450 20-31

{ Inst./-4.500 N.A.

Smer 132X C e 2C NI Speed Feed Skr Primary eng time - 900 20-32 3 C:3 Sugr 132X C e 3C Test. - 7,500 N.A. s c.3

-T1 R

rcy 8

l

! . a h

5

.m . _- _ _ _ . ____ o

U U V U v v v v v ,v v ,

I ;

g' .

TAttE 3.8-1 (Continued)

E ,

B x

CONTAIIGENT PENEIRAT1000 rnumrTOR s

OVEEUARENT PROTECTIVE DEVICES 4 E

4

- Ni? W PA01ECTIVE MVICE %IPOINT KSP0p54 TifE INeeER AND t0 CAT 10N DEVICE [Asperes) (Sec/Cytle)

~

4. 480W 90elded Case Ckt. Skts. (NCCS) leCC 133x4 IKSIPA-A Prleary L5 N.A.

Ce 31 Sackup l5 N.A.

1E0lPA-B Primary le N.A.

4:* Cd 82 Backup '40 N.A.

T llIC22G Prleary 15 N.A.

O C4 83 Sackg 15 N.A.

IFBIA36 Prleary 15 N.A.

/ N.A.

Cd 84 Sacky 15 IVPSSCA Primary 30 N.A.

Ce C1 Sackup ,30 N.A. g 10F83P Prleary L

M N.A.

=

c.2 M

Ce C2 Sackup M N.A.

\ 83 IKSIPS-A Primary 15 If A. m C4 91 Sackup 15 N. .

IKSIPS-B C4 82 Primary Backup 40 N.A.

N.A.

Q 1AF02PS Primary 30 N.A.

b C4 94 Sackup 38 _

-N. AJ

O O O O O U Q O v N '

l, l TASTE 3.8-1 (Continued) I i

CONTAlteENT PENEIRATICIg rnasaurTOR e OVERCUAGENT P90TECTIVE DEVICES E

"4 P901ECTIVE DEVICE P618iT -BESPoetst TI9E IRSGER Als LOCATIcet DEVICE (Anseres) (Sec/ Cele) u 'l

4. 480V 90 sided Case Ckt. Skts. (9tCCS) (Continued) .

IICC 133m4 lifolP Primary Ib N.A. .

C4 95 Backup 1$ N.A.

lafelpA Prleary 40 N.A.

Backup 44 M.A.

y Ce 96

. = l

= Prleary 44 M.A.

IVPS2CA Backup M. A.

$ C4 El Primary 125 N.A IVP94CA Sackup 125 N.A

. C4 E2

. I 8

IvPO4CC Primary 125 N.4 y

! Cd F1 Backup 125 N.A. p t l C2 IEW11EA Primary . 125 125 N.J L N.A.

Q C4 F3 Back ,

~

IEW11ES Primary 12h 125 II. A.

IL A.

f C4 F3 SaCkuP IEW11EC Primary 1 18 A cd F3 Backup 125 N.A. n

\ o

- 11CS2fA ce F5 Primary Backup 26 -

ll.k W.A.

w I

U U U U U U . U U U U i C O O O.: ,

TASTE 3.5-1 (Continued) ;

CGIITAIISENT PfNE18ATIGN rmaryog #

OWACUBAENT P90TECTIVE DEVICES E

G -

r TRIP _

P901ECTIVE SEVICE ImeGER AIS 10 CAT 1011 SEWICE SEWIIII

( M res)

REh IIK (Sec/ Cycle)

I !

4. 480W Ignidad Case Ckt. Skts. (NCCS) (Continued)

E C 133m4 IICS2EB Primary 28, M.Al Cd El Back , 29 N.A.

I /

IICS2fC Primary 20 4.A.

Cd G2 Back, 20 N.A.

t.

T 11CS2EF Primary 3dl N.A.

3 Cd Al Seckg 34 N.A.

1 '

IICS2fE Primary 34 N.A.

C d A2 Back, 30 N.A.

IICS2EB Primary 30 N. A. 7 __

Cd A3 Cack, 30 N.A. y

! M Iftm2J Primary 15 N.A. C3 Cd G1 Sack, 15 N.A. %

IFIIB3J Primary gM.A.

Backup .A. N Cd G2 15 I '

IStalPS-B Prleary 44 , A.

Cd 81 Sackup 44 N.'A.

I \ c3 18ESIPS Prleary 70 N.A C::3 Cd E3 Backg 72 -

NTA.

m Q

1

v v -v v- v v ,v v 3 v- v ,

~

i _ ,

- l

.= ; ; ,

^'z's TABLE 3.8-1 (Continued) l Y l? f

' g CONTAlletENT PENETRATION CONDUCTOR t x ,

j s 8 .' OVERCURRENT PROTECTIVE DEVICES

, E e

a: -4 ,

" ' ~.6Effetttf-PROTECTIVE DEVICE RESPONSE-HME.

', I' NUfeER A V LOCATION DEVICE (Amparad- -(c.c /cyc %}

\

4. 480V MoWuc J - = Ckt. 8kts. (NCCS) (Continued)

MCC 134x5

! r

) 1RColPC-A ,

Primary :L5 N;A.

Cub C1 ~/ Backup :L5 N,A.

l 1Rr0lPC-8 Primary 40 N. A.

l Backup 40 N. A.

j ,.. y Cub C2

,

i j g IVPJ5C8 '

Prle.ary 30 N. A.

g ' Cub 'J1 . Sackup 30 N. A.

l.l IRC31PS-A, .- .

Primary :l5 N. A.

Cub C3 i .Ba::kup :l5 N. A.

j Ep Prleary N.A. N 1HC656-A d}0 Cub D3 , i '

Backup 40 N.A. c.3

l r .. I M IVP02C8 Primary 4(l N.f. gge

-( +f Backup 40 N. 4.

Cub F1 m

! 1RColR-A Prleary 1) N.$. b Cub,F2 Backup 1h N. 4. g

'F' Primary 30 C2 1Rf02FA - N.f. O j ';

, Cub G3 Backup 3( 8.A.

i j IEW12EA Primary 145 N.A.

4 Cub F3 Backup- It5 N.A. ,

v

U U U U U U U U U. ; U U TABLE 3.8-1 (Continued) 5 -

8 z

CONTAINMENT PENETRATION CONDUCTOR

' OVERCURRENT PROTECTIVE DEVICES E

4 g 7R!W PROTECTIVE DEVICE SHPetMT- RESPONS N Ml#EER AND LOCATION DEVICE (^ r er) - (.S c/cyd e}_

4. 480V Molded Case Ckt. 8kts. (MCCB) (Continued) .

MCC 134x5 IEW12E8 Primary 1 !S N A.

Cub F3 Backup 135 N A.

1EW12EC Primary 125 N, f

A.

Cub F3 Backup ITS N,A.

1:a

I T IVP04C8 Primary 12l5 N A.

tj Cub F4 Backup 125 NfA.

IVPO400 Prlaary 1 N.A.

Cub f5 Backup 1 N.A.

ISI8808C Primary 7 N. A. N m

Cub A2 Backup 7i N. 4. c::3 c3 T

IS188088 Primary 73 N.A.

Cub A3 Backup 70 N. 4. Q*

az 1RH87028 Primary 15 H. A.

Cub 81 Backup 15 N.A. Q-

\

IRH87018 Primary 15 N.$.

Cub 83 Backup l$ N.$. @

i.

C C 0 0 W V U O V 0 0 O O >

I O.. .

1 9

TABLE 3.8-1 (Continued) .

CONTAlletENT PENETRATION CONDUCTOR ,

2 . !

e OVERCURRENT PROTECTIVE DEVICES .

C 5

4 T;'" .

PROTECTIVE DEVICE SEffetiti F45e-.st IIm -

(Ampumes)

NLNSER Ate LOCATION DEVICE f-/Cid:) -

480V Molded Case Ckt. 8kts. (MCC8) (Continued)

MCC 132x2 1CV8112 Primary N A.

Cub 84 Backup 5 N A. l 10G079 Primary 15 N. A.

Cub C1 Sackup 1 i N. A. j R IWD056A Primary ! , N. A.

Cub C2 Backup ! i N. A.

T

$$ Inc. nan Primary 1h N. A.

Cub C3 Backup 16 N. A.

1RY80008 Primary l i N A. !

i Cub C4 Backup 1 s . N A.

1RY8003C Primary 15 N A.

Cub C5 Backup 15 N A. @

11P06E Primary 23 N A. g -

Cub El Backup 23 N. A.

IRC80038 Cub 04 Primary Backup 15 15 N. 4.

N. 4.

b ,

ILL43J Prleary 73 N..4. ca Cub E2 Backup 73 N.iL. O i

. IRC8002A Primary 40 N. d h.

' Backup Cub G1 4p N.).

c

v

~' ~

O O O O U l

T" O ~O i t

!O. .

TABLE 3.8-1 (Continued) j E '

E CONTAINMENT PENETRATION CONDUCTOR z

' OVERCURRENT PROTECTIVE DEVICES E

-s r 1 RIP--- .

PROTECTIVE DEVICE 4ETrotNT- 1ESPONSETIE DEVICE .(Ampare 4 4k / Cycle)-

NUMBER AND LOCATION

4. 480V Molded Case Ckt. Bkts. (MCCB) (Continued)

MCC 132x2 i 1RC80028 Primary [0 N A.

Backup .10 N A.

Cub G2 Prinary 40 N .A.

IRC8002C Backup lla N A.

t' Cub G3 J

Primary ' 0 N. A.

? 1RC80020 40 N A.

Di Cub G4 Backup MCC 131x2A i

i0 N. 4. up 15188080 Primary Cub A2 e Backup 105 N. 4. c3 1AP25E-A ~'1 NCC 131x2 Cub 82 ,

ISI8808A Primary 70 N.$. m ,

I Cub A3 f Backup 12 N.A.

1AP2SE-A -

MCC 131x2 Cub 82 I Q

i n C3 i

O O O O O 4 O Q V

.' P f -

! n I j TABLE 3.8-1 (Continued}

= CONTAINNENT PENETRATION CON 00CTOR i

3 8 OVERCURRENT PROTECTIVE DEVICES E

"4 TRIP g RESPONSt-TIME -

PROTECTIVE DEVICE S H P &fftf' (Ampace4- (4ec/fyt.4eh NUMBER AND LOCATION DEVICE

4. 480V Molded Case Ckt. Bkts. (NCC8) (Continued) i MCC 131x2 Primary 0 N: A.

1RC8001A Backup N. A.

Cub Primary 4) N. A.

IRC80018 8ackup 4 H.A.

4:* Cub

+

Primary 4 N.A.

? 1RC8001C Backup 40 H.A.

g Cub Prleary 40 N. A.

e IRC80010 Backup 40 N.A.

I' Cub N.A. 7 Primary 1 IRH8701A Cub Backup 1 N.A. $

C. 3 N.A. M 1RH8702A Prinary 15 Backup :.5 N.A. Gro Cub

3 N.A. r ILL42J Prleary /0 N.A. -

Cub Backup 70 Primary : 0 N.A.

IVQ001A Cub Backup N.'A. Q "t7 Primary N.f.

IVQ002A -

Cub Backup lh N.p.

O O O O O O O O O M O- 0 O o !

O..

I f

TABLE 3.8-1 (Continued) !

5 CONTAll8 TENT PENETRATION CONDUCTOR i

g

OVERCURRENT PROTECTIVE DEVICES 1

C

-t :

INIr e N l SiiFGiWi-i PROJECTIVE DEVICE NulSER Als LOCATION DEVICE ( ^-m;)- W e)

4. 480V Molded Case Ckt. 8kts. (MCC8) (Continued)

MCC 131x2 Prleary 1!5 N' A.

! 1RC80030 Backup 15 N. A.

Cub

' Primary 1 5 N. A. '

1RC8003A Backup 15 N. A.

4:* Cub Primary 15 N. A. .

? 10G057A Backup ]5 N.A.

M Cub Primary 'l5 N. 4.

ICC9416 Cub Backup L5 N. 4. $

C3 1CC9438 Primary 15 l5 N. L.

N. 4.

Q ,

Cub Backup g N. A. N Primary 45 10G081 Cub Backup 15 N. A. Q

$ NCC 133x6

, o Primary 135 N. L. . c:3 IHC016 - Cub B2 l

Cub 81 Backup 175 N. 4. ]

Primary 2f5 N.A.

, ILLO4E - Cub C3 N.A.

Cub C1 Backup 225 ,

j j .

W W v- v v v v v v

'v v 3 O O '

O.-i .

I .

I ..

TABLE 3.8-1 (Continued) i' ,!

g .

CONTAINNENT PENETRATION CON 00CTOR I g ,

OVERCURRENT PROTECTIVE DEVICES E ""

4 _..

.m

~

p t

y PROTECTIVE DEVICE MfretstT- RESP 9NGE-TIME- !

Nup6ER AND LOCATION DEVICE (? ;: ::)

Cec / Cide) 480V Molded Case Ckt. 8kts. (NCC8) (Continued)

1. ,EC 155=6 - j l I !

IVP03CA Prleary 1M5 N .A.

Cub A3 Backup 135 N A.

IVP0300 Prleary 1.25 N .A. ,

Cub C4 Backup 1.t5 N .A. !

R as MCC 132x5

, ICC9414 Primary 5 N A.

Cub 84 Backup i .A. N. . ,

i MCC 134x7 ILLO5E Prleary 225 N.4. m Cub Backup 22 s N. it. zu

! C3 IVP03C8 Primary Backup 12 i 12i N.aL.

N. , L.

M Cub g Prleary 12 i N. 4. "

I IVP03CC i Cub Backup 125 N.

f. h E

o .

c:a i

v v v v U U U ! LT V' D

J s ! -

. i l

TABLE 3.8-1 (Continued) g CONTAINNENT PENETRATION CONDUCTOR ,'

2 OVERCURRENT PROTECTIVE DEVICES E

H P80TECTIVE DEVICE SEffeHtf- RESPONSE-HME-NUMBER AND LOCATION DEVICE ( b res) - (O /Cyc k)

4. 480V Molded Case Ckt. 8kts. (MCC8) (Contlnued) i MCC 131x28 IWD0568 Primary i N.' A.

Cub A1 Backup i N.A.

IRY8000A Primary 1p N. A'. .

4:

8 Cub A5 Backup 10 N A.

y -;5 'OC F..i 114 Oc. E U

id52a 4.;-ary- 1D N .A.

C4 l'2 " -9? ? ) N.A.

N

5. 260 VAC RCO Power (53 rods, 5 panels) g i C.3 Stationary Gripper Primary 10 - fhse M.A. v1 -

Coils (all panels) Backup 10 - Fose N .A. De Lift Colls Primary S i - Fo se N A.

(all panels) Backup 5) - Fose N. A. --

Movable Gripper Primary lil - FLse N.A.

Coils (all panels) Backup 18 - fdse N. . C"3 C3 N

M a

f

-_L-_- - - _ - - _ - - - _ _ - _ _ _ _ _ _ _ . -

af ATTACHMENT F 4

\

(Section 3/4.9) I l

1. Section 3/4.9.6 (pg. 3/4 9-6) Refueling Machine In step 3.9.6.a.1; " minimum" capacity is changed to " rated" capacity and "3250" pounds to "2850" pounds.

In step 3.9.6.b.1; " minimum" capacity is changed to " rated" capacity and "3000" pounds to "2500" pounds.

Typically, a crane has a " rated" capacity and not a " minimum" capacity, thus the necessary change. The rated capacity of the manipulator crane was changed per Westinghouse letter CAW 6825. The rated capacity of the auxiliary hoist was changed to prevent the handling of heavy loads.

2. Section 4.9.6.1 (pg. 3/4 9-6) Surveillance Requirements Replace "3250" pounds with "3563" pounds. The manipulator crane should be tested at 125% of the rated capacity of 2850 pounds.

3. Section 4.9.6.2 (pg. 3/4 9-6) Surveillance Requirement Replace "3000" pounds with "2500" pounds. This was changed to reflect change in rated capacity in 3.9.6.b.1.

4 i (0441M) t

I l

, ~ . _ . _ r i ~~ REFUELING OPERATIONS M o 3/4.9.6 REFUELING MACHINE Ph00F& R&TW copy

[ ___ _ _ ._

l l ,, __

LINITING CONDITION FOR 0.'ERATION 1

3.9.6 The refueling machine shall be used for movement of drive rods or fuel assemblies and shall be OPERA 8LE with:

a. The refueling machine used for movement of fuel assemblies hav'ing:

l'ated 2850 A e6n4anse capacity of 4466 pounds, and 1)

2) An overload cutoff limit less than or equal M 2850 pounds,

b. The auxiliary hoist used for latching and unlatching drive rods having:

r~afed 2[0

1) A e4e4anse capacity of -3000* pounds, and

2) A load indicator which shall be used to prevent lifting loads

in excess of 1000 pounds.

! APPLICA8ILITY: Ouring movement of drive rods or fuel assemblies within the reactor vessel.

4 ACTION: ,

With the requirements for crane and/or hoist OPERA 81LITY not satisified, suspend use of any inoperable manipulator crane and/or auxiliary hoist from operations ,

l involving the movement of drive rods and fuel assemblies within the reactor l vessel.

I

SURVEILLANCE REQUIREMENTS J

4.9.6.1 Each manipulator crane used for movement of fuel assemblies within 1 the reactor vessel shall be demonstrated OPERA 8LE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior l to the start of such operations by performing a load test of at least 1990- 3N3 pounds and demonstrating an automatic load cutoff when the crane load exceeds 2850 pounds.

l 4.9.6.2 Each auxiliary hoist and associated load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERA 8LE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of such operations by performing a load test of at least 46dt pounds.

2fo l-ll 8YRON - UNIT 1 3/4 9-6

ATTACHMENT G (Section 3/4.11)

1. Section 3.11.1.4 (pg. 3/4 11-8) Liquid Holdup Tanks The curie limit for the Primary Water Storage Tank (PWST) is based upon the maximum concentration in the PWST that, if released from the tank, will result in concentations not greater than the 10CFR20 MPC limits for unrestricted areas. For Byron, two cases are evaluated with the most conservative value chosen as the curie limit. Case 1 is the nearest potable water supply (a well 1960 feet East-South-East of the Aux.

Bldg.). From the FSAR, Section 2.4.13.3 a dilution factor of 2200 is given with a transient time of 30.5 years.

Case 2 is the nearest surface water body (a spring 3630 feet Northwest of the Aux. Bldg.). From the FSAR, Section 2.4.13.1.1 a dilution factor of 655 is given with a transient time of 64.12 years.

The major source of activity input to the PWST is from the Recycle Evaporator Monitor Tank. To obtain the estimated activity in this tank, the Recycle Holdup Tank inventory (FSAR, Table 2.4-20) was used. Each isotopes concentration was divided by its evaporator decon factor (FSAR, Table 12.2-34) to obtain the Recycle Monitor Tank activity. This activity is assumed to be equivalent to the PWST activity.

Because of the transient times involved (as specified above) Cs-137 is the only isotope that has a long enough half-life to still be present in appreciable amounts after transient; thus, Cs-137 is the limiting isotope.

Isotope Recycle Holdup Tank uCi/ml Decon Factor Recycle Monitor Tank uCi/ml Rb-88 3.7E-2 10000 3.7E-6 RB-89 2.1E-3 10000 2.1E-7 Mo-99 5.3E-2 10000 5.3E-6 I-131 2.5E-2 1000 2.5E-5 l I-132 2.8E-2 1000 2.8E-5 I-133 4.0E-2 1000 4.0E-5 I-134 5.6E-3 1000 5.6E-6 I-135 2.2E-2 1000 2.2E-5 Cs-134 2.3E-2 10000 2.3E-6 Cs-136 2.8E-2 10000 2.8E-14 Cs-137 1.5E-2 10000 1.5E-6 Cs-138 9.8E-3 10000 9.8E-7 Ba-137M 1.4E-2 10000 1.4E-6 1.36E-4 uCi/ml (0441M)

, ATTACHMENT G (Continued)

(Section 3/4.11) 6

% Cs137 1;5 _4 = 1.1%

CASE 1 - Potable Water Dilution factor - 2200 MPC CS137 - 2E-5 uCi/ml PWST volume = 1.89E9 ml (2200)(2E-5uCi/ml)(1.89E9ml)j10 6 uCi/Ci = 83 Ci Cs137 in PWST CASE 2 - Surface Water Dilution factor = 655 (655)(2E-5 uCi/ml)(1.89E9 ml)

106 uCi/Ci = 25 Ci Cs137 in PWST Case 2 is limiting 25 Ci Cs-137/.011 = 2270 Ci limit in PWST T 2,000 Ci i

1 l

(0441M)

)

i e ,

RADI0 ACTIVE EFFLUENTS PROOF & REY!EW COPY g

LIQUID HOLDUP TANKS ,

LINITING CONDITION FOR OPERATION l

( 3.11.1.4 The quantity of radioactive material, excluding tritium and dissolved or entrained noble gases contained in any outside tanks shall be limited to the following:

a. Primary Water 5,torage Tank 1 1000 Curies, and

< b. Outside Temporary Tank i 10 Curies.

APPLICA8ILITY: At all times.

ACTION:

, a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Sesiannual Radioactive Effluent Release Report, pursuant to Specification 6.7.1.7.

~

' _ b. The provisions of Specifications 3.0.3 and.3.0.4 are not applicable.

SURVEILLANCE PEQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when g

radioactive materials are being added to the tank.

THIS PAGE OPEN PENDING RECEIPT 0? '

(

INFORMATION FROM THE APPLICANT i I

(

1 i BYRON - UNIT 1 3/4 11-8 l l

. ... . . _ _ . . .. .m - - . - y wr v*;t. ar w

o ATTACHMENT H t

(Bases Section 3/4)

1. Section 3/4.2 (pg. B 3/4 2-1) Power Distribution Limits In the first paragraph, third line, insert " design" between the words

" minimum" and "DNBR".

2. Section 3/4.2 (pg. B 3/4 2-4)

Delete the paragraph beginning " Fuel Rod Bowing reduces. . . partially offset rod bow penalties." and insert " Fuel rod bowing reduces the value of DNB ratio. The Safety Analysis for Byron /Braidwood cores maintained sufficient margin between the safety analysis limit DNBR's and the design limit DNBR's to accommodate full flow and low flow penalties identified in WCAP-8691, Revision 1, " Fuel Rod Bow Evaluation" which is applicable to 17 x 17 Optimized Fuel Assembly analysis utilizing the WRB-1 DNB Correlation.

3. Section 3/4.2.5 (pg. B 3/4 2-6) DNB Parameters First paragraph, fif th line, delete thewords "a minimum" and replace with "d e s ig n" . Also on that same line delete the words "of 1.30".

4. Section 3/4.4.1 (pg. B 3/4 4-1) RC Loop and Coolant Circulation First paragraph, second line, delete the words "above 1.30" and replace with "above the applicable Safety Analysis DNBR".

Per discussions between Westinghouse and CECO, the above 4 changes are being made in order to better clarify the DNBR limit as it applies to Byron Station.

5. Section 3/4.4.9 (pg. B 3/4 4-10) Cooldown This has been added to be consistent with the change that was made on page 3/4 4-32.

Insert the following paragraph as the last paragraph on Pg 3/4 4-10:

"The notch in the cooldown curve of Figure 3.4-3 is due to the added constraint on the vessel closure flange given in Appendix G of 10 CFR

50. This constraint requires that, at pressures greater than 20% of the preservice system hydrostatic test pressure, the flange regions that are highly stressed by the bolt preload must exceed the RTNDT of the material by at least 120"F. In the case of Byron 1, the flange RTNDT +

120*F impinges on the cooldown curves and therefore the notch is required.

(0441M)

. _ _ _ . ,m . _ _ _ _ - - - _ _ __ _ _

t PR00F & REMW COPY f 3/4.2 POWER DISTRIBUTION LIMITS ..

BASES --

cG5th H The specifications of this section vide assurance of fuel integrity during Condition I (Normal Operation) a II (Incidents of Moderate Frequency) events by: (1) maintaining the minimus DN8R in the core greater than or equal to 1.34 for a typical cell and 1.32 for a thimble cell during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

Fq (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of

( Fh the integral of linear power along the rod with the highest integrated power to the average rod power; and F,y(Z) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

3/4.2.1 AXIAL FLUX OIFFERENCE The limits on AXIAL FLUX OIFFERENCE (AFO) assure that the Fq (Z) upper bound envelope of 2.32 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions. The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux ,

difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the l associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER 1evel. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations, g

f 8YRON - UNIT 1 8 3'/4 2-1

9 l

-_ _ s POWER DISTRIBUTION LIMITS PM0E & HM MPY .

u- .

BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

, c. The control rod insertion limits of Specification 3.1.3.6 are l maintained, and

d. The axial power distribution, expressed in terms of AXIAL FLUX OIFFERENCE, is maintained within the limits.

N F

g will be maintained within its limits provided Conditions a. through

d. above are maintained. As noted on Figure 3.2-3, RCS flow rate and Fg N may be " traded off" against one another (i.e. , a low measured RCS flow rateisacceptableifthemeasuredFhisalsolow)toensurethatthecalcu-lated ON8R will not be below the design DN8R value. The relaxation of Fg as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

R as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts N

for F g less than or equal to 1.49. This value is used in the various accident analyses where Fh influences parameters other than ON8R, e.g. , peak clad A

temperature, and thus is the maximum "as measured" value allowed.

4 Fuel rod bowing reduces the value of DN8 ratio. Credit is available to frtSEA1' partially offset this reduction. This credit comes from a generic design

~~

"8 ,,

margin which totals 9.1% when the analysis is performed with the approved interim methods. The margin used to partially offset rod bow penalties is 9.1%.

This margin breaks down as follows:

1) Design limit DN8R (1.6)%

2) Grid spacing Ks (2.9)%

3) Thermal Diffusion Coefficient (1.2)%

4) DN8R multiplies (1.7)%

5) Pitch Reduction (1.7)%

The margin used to partially offset rod bow penalities is (5.9)% with the remain-ing (3.2)% used to trade off against measured flow which may be as much as (2))

lower than thermal design flow plus uncertainties. The penalties applied to Fg to account for rod bow as a function of burnup are consistent with those des-cribed in Mr. John F. Stolz's (NRC) letter to T. M. Anderson (Westinghouse) dated April 5,1979 with the difference being due to the amount of margin each unit uses to partially offset rod bow penalities. j When an qF measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate i for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.

h THIS PAGE OPEN PENDING RECEIPT OF I

1 8vRON - UNIT 1 INFORIQUQyf, ROM THE APPUCANT 1

1 i

- . . ., . . . 8

0* el F-JL g h"~~d *k #" % D N 6 A u S% Wy 6p/ SM ww e mahm M@ DMGR.'sW A h pEatp Dese.'s

%Ah ua. A 4 e M .' a W a w c.s e - a m , m i ,

FA M % %onra" M sa app g n <i, O & M % } Q & j A4. Med-/

DN8 Gew.bZk .

'4 . - . . . . - - . . . _ . _ . _ . . . . . _ _ - . . ~ . . .. - . - - -

I

]

{

t,.,

g ., ..

1 i' E PROOF & REVIEW COPY l REACTCR COOLANT SYSTEM 3ASES D

( DRESSL'RE/TEWPERATURE LIMITS (Continued)

% Kgg = constant provided by the code as a function of temperature relative p to the RT NOT of the material,

+

j C= 2.0 for level A and B service limits, and

[ C= 1.5 for inservice hydrostatic and leak test operations.

k' At any time during the heatup or cooldown transient, K gg is determined by

~

, t.*e netal temperature at the tip of the postulated flaw, the appropriate value "or ATNOT, and the reference fracture toughness curve. The thermal stresses esulting free temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, KIT, f r the eference flaw is computed. From Equation (2) the pressure stress intensity f actors are obtained and, from these, the allowable pressures are calculated.

'TLX%N For the calculation of the allowable pressure versus coolant temperature curing cooldown, the Code reference flaw is assumed to exist at the inside of M t.7e vessel wall. During cooldown, the controlling location of the flaw is V always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable

ressure-temperature relations are generated for both steady-state and finite

'c coldown rate situations. From these relations, composite limit curves are

=nstructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because

. =ntrol of the cooldown procedure is based on measurement of reactor coolant

?- temperature, whereas the limiting pressure is actually dependent on the material

tancerature at the tip of the assumed flaw. During cooldown, the V4T vessel g iccation is at a higher temperature than the fluid adjacent to the vessel 10.

t 3fs condition, of course, is not true for the steady-state situation. It e fo11cws that at any given reactor coolant temperature, the AT developed during

, (. =olcown results in a higher value of K IR at the V4T location for finite l[ =oldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in K gg exceeds K gg,thecalculatedhiowaDN9 h l, ressure during cooldown will be greater than the steady-state vaTU C '

l' l . The above procedures are needed because there is no direct control on i; tancerature at the 1/4T location; therefore, allowable pressures say unknowingly l

M violated if the rate of cooling is decreased at various intervals along a s

cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

~') Add (%y<.,} )9 ~

3Y9CN - UNIT 1 . 8 3/4 4-10 4

4 S

9

( 3/4.4 REACTOR COOLANT SYSTEM f% h@@ M -

l BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION .

The plant is designed to operate with all reactor coolant loops in operation and maintain DN8R d r: 1.3^*during all normal operations and anticipated 1 transients. In MODES 1 a 2 with one reactor coclant loop not in operation this specification requ s that the plant be in at least HOT STAN08Y within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . A&ws 78; AAoucastr

.rsrery ANaYS/$ DNBA In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERA 8LE.

In MODE 4, and in M00E 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RhR or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERA 8LE.

I The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a reactor coolant pump with one or more RCS cold legs less than or equal to 380*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed tre limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.

((.

SYRON - UNIT 1 8 3/4 4-1 l

l

+*

, , , , , , , . . a I

-- -.--o= o-efv rnen

- 3 POWER DISTRIBUTION LIMITS

( .

BASES _ . . _.

, QUADRANT POWER TILT RATIO (Continued) _

l not correct the tilt, the margin for uncertainty on Fq is reinstated by reducing i

the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistant with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

3/4.2.5 DN8 PARAMETERS .

The limits on the DNS-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistant with the initial FSAR ass tions and have been analytically demonstrated adequate to maintain a efef-- - - N8R ;f 1.1*. f.hroughout each analyzed transient.

DE5 m er The 12-hour periodic surveillance of these parameters through instrument

(. readout is sufficiant to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

l l

8YROM - UNIT 1 B 3/4 2-6 O e, 4 .e

e o

I fr.rayec.jsb k Tl1 a nsfc b In bbc cooNown C u. r ve ef T 'f" 2 /f dut bo bbt Oclcled C o n J f rein & as $b < VerJe / C la J u r t i in A.G o{ /d C /*/t Co, 7~b is Co nc lrsin b hlctnfe f ven ff Ye g u.i e t t bb ^ b, & h press <<rer- 9 r ca fe r (ban 20 2 ch bke p r e s e evie e s, < /e ,n hya r. < M . 6,, ty e < << c e ed 64e 2rivp, ef 64e ma fee; / 1, a / /ea s &

I20 % I; de caz. .f By .,,f, dAc f/a , e g g,,t12o v i mp ; ny << o. 61. cos Ue w. cur u r, aJ di er . fa r e &L, n. /a /

ic rep ;r</.

ATTACHMENT I 0

(Section 6.0)

1. Section 6.3.2.c (pg. 6-12) Authority The word "pases" has been changed to " phases" for typographical error.

(0441M)

r i hk ADMINISTRATIVE CONTROLS ONSITE (Continued)

3) Review of all proposed changes to the Technical Specifications;

4) Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety;

5) Investigation of all violations of the Technical Specifications including the preparation and foWarding of reports covering evaluation and recommendations to prevent recurrence to the Division Vice President and General Manager - Nuclear Stations and to the Supervisor of the Offsite Nuclear and Investigative

- Function; .

6) Review of all REPORTABLE EVENTS;

7) Performance of special reviews and investigations and reports thereon as requested by the Supervisor of the Offsite Review and Investigative Function;

8) Review of the Station Security Plan and implementing procedures and submittal of recommended changes to the Division Vice President ano General Manager - Nuclear Stations;

9) Review of the Emergency Plan and station implementing procedures and shall submit recommended changes to the Division Vice President - Nuclear Stations;

10) Review of Unit operations to detect potential hazards to nuclear safety;

11) Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Division Vice President and General Manager - Nuclear Stations and the Supervisor of the Offsite Nuclear Review and Investi-gative Function; and

12) Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE 00SE CALCULATION MANUAL, and the Radwaste Treatment Systems.

c. Authority A I

The Technical Staff Supervisor is responsible to the Stati an Superintendent and shall make recommendations in a timely panner in all areas of review, investigation, and quality control peces of plant maintenance, operation, and administrative procedures relating to facility operations and shall have the authority to request the action necessary to ensure compliance with rules, regulations, and procedures when in his opinion such action is necessary. The Station Superintendent shall follow such recommendations or select a course BYRON - UNIT 1 6-12

c ATTACHMENT J (Section 3/4.2)

1) Figure 3.2-3 (pg. 3/4 2-9) RCS Total Flow Rate Versus R.

The new graph has been supplied by Westinghouse l

l l

a I

p'# .

-- -. - m-- == =_ _ ,

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tI :- -M 1 ----f 1

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9 Fy 3.1-5 N kop Ogh vs. F,( 6df

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ML20084E324

Text

Subject:

Dear Mr. Denton:

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