Text

Application for Amends to Licenses DPR-44 & DPR-56,proposing to Change Source Range Monitor to Range Neutron Monotors & Delete Intermediate Range Monitors
	ML20078A053
Person / Time
Site:	Peach Bottom
Issue date:	01/17/1995
From:	Hunger G; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
References
	NUDOCS 9501240265
	Download: ML20078A053 (92)
	v • d • e

-

cttion Support Department 10 CFR 50.90

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YECO ENER'3Y

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%5 Chesterbrook Boulevard Wayne. PA 19087-5691 i.

January 17,1995 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

SUBJECT:

Peach Bottom Atomic Power Station, Units 2 and 3 Faclity Operating License Change Request 93-18

Dear Sir:

PECO Energy Company hereby submits Ucense Change Request (LCR) 93-18, in accordance with 10 CFR 50.90, requesting changes to Appendbc A of the Peach Bottom Atomic Power Station (PBAPS) Facuity Operating Licenses.

The proposed changes are necessary to support PBAPL Modification P00271, which will replace the Source Range and Interrnediate Range Monitors with a new Wide Range Neutron Monitoring System (WRNMS).

Attachment I to this letter describes the proposed changes, Attachment 2 contains the revised l

1 Technical Specifications pages, and Attachment 3 contains NEDO-32368, ' Nuclear Measurement Analysis and Control Wde Range Neutron Monitoring System Licensing Report For Peach Bottom Atomic Power Station, Units 2 and 3," which was prepared by GE Nuclear Energy.

i if you have any questions conceming this submittal, please contact us.

Sincerely, 1

[.

G. A. Hunger, Jr. V Director - Licensing

Enclosures:

Affidavit, Attachments cc:

T. T. Martin, Administrator, Region I, USNRC W. L Schmidt, Senior Resident inspector, PBAPS, USNRC R. R. Janati, Commonwealth of Pennsylvania I

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9501240265 950117 PDR ADOCK 05000277 P

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i-COMMONWEALTH OF PENNSYLVANIA

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COUNTY OF CHESTER c

W. H. Smith, lit, being first duly sworn, deposes and says:

That he is Vice President of PECO Energy Company; the applicant herein; that he has read the attached License Change Request (LCR 93-18) for changes to the Peach Bottcm Facliity Operating Licenses DPR-44 and DPR-56, and knows the contents thereof; and that the stawments and matters set forth therein are true and correct to the best of his knowledge, information and belief.

4 Vice President

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Subscribed and swom to before me this [

ciay of CWA 1995.

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ATTACHMENT 1 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 t

Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 Facility Operating License Change Request f

93-18

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" Wide Range Neutron Monitoring instrumentation" Supporting information for C;mnges 0"

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PECO Energy Company (PECO Energy), Ucensee under Faculty Operating, Ucenses DPR-44 and DPR-56 for the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 respetively, requests that the Technical Specifications (TS) contained in Appendix A to the Operating Ucenses be amended contains the prng-1 handwritten Technical Specifications changes for Unit 2 only. The changes are identical for Unit 3. PECO Energy is providing the proposed changes in the improved Technical Specifications format as the requested approval date for these changes is beyond the I

implementation date of our previously submitted improved Technical Specifications application. contains the GE Nuclear Energy Licensing Topical Report, NEDO-32368, ' Nuclear Measurement Analysis and Control Wide Range Neutron Monitoring System thensing Report For Peach l

Bottom Atomic Power Station, Units 2 and 3." This report summarizes the analysis and evaluation performed which jus *!fy installing and operating a Nuclear Measurement Analysis and Control (NUMAC) l Wide Range Neutron Monitoring System (WRNMS) into PBAPS, Units 2 and 3. The WRNMS replaces the Source Range Monitor and intermediate Range Monitor. The PBAPS WRNMS is effectively identical to the NRC generically approved system documented in Ucensing Topical Report NEDO-31439-A, "The l

Nuclear Measurement Analysis and Control Wide Range Neutron Monitoring System (NUMAC-WRNMS)."

The NRC Safety Evaluation Report for NEDO-31439-A, was transmitted to GE Nuclear Energy by letter dated October 3,1990. The minor plant-specific differences between PBAPS and the generic system are evaluated in this report PECO Energy is requesting that the NRC appove Ucensing Report NEDO-32368 for PBAPS, Units 2 ard 3 by AprH 30,1995. PECO Energy is seeking NRC approval of this report prior to committing the substantial financial resources necessary to construct and assemble the WRNMS. PECO Energy further proposes that the actual Technical Specifications changes be approved by June 1,1996, to allow for installation of the modification during the Fall 1996 refueling outage. Implementation of the Technical Specifications changes wtl then occur upon completion of modification P00271.

Descriotion of Chanoes See Table 4-4, page 4-10, of Ucensing Report NEDO-32368, (Attachment 3) for a description of the proposed Technical Specifications changes.

Safety DiscustiQD l

The safety discussion for the proposed changes is provided in ettachment 3.

Ho Slanificant Hamds Consideration Ucensee proposes that this application does not involve significant hazards consideration for the following reasons:

1)

The orooosed chanoes do not involve a sionificant increase in the orchnhilitV or Consecuences of an accident previousiv evaluated.

The use of the WRNMS as discussed herein will not 8ncrease the probability or consequences of an accident previously evaluated.

The probability (frequency of occurrence) of design basis accidents (DBAs) occurring is not "

affected by the WRNMS. The only plant safety analysis affected by WRNMS is the Rod Withdrawal Error (RWE) at low power, and a reanalysis assuming use of WRNMS shows that the criteria of 170 cal /gm for fuel enthalpy increase under RWE is satisfied; thus, RWE is not a limiting event. Scram setpoints (equipment settings that initiate automatic plant shutdowns) wHl be established such that there is no increase in scram frequencydue to the WRNMS. No new challenges to safety-related equipment wHl result from WRNMS.,

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The oranomed changan rk not create the cosalbEkv of a new or different kind of accident from i

any oreviousiv svaluated.

As summarized below, this change wil not create the possibiky of a new or different kind of I

accident from any accident previously evaluated.

The components of the WRNMS wil be supplied to equivalent or better design and quellfication crkeria than is currently required for the plant. Equipment that could be affected by WRNMS has been evalmtart. No new operating mode, safety-related equipment lineup, accident scenario, system interaction, or equipment falure mode was identified. Therefore, the WRNMS '

wIl not adversely affect plant equipment.

111)

The oracomed changes do not involve a sinnificant reduction in a maroin of safety.

All the SRM/lRM functions required in the Technical Specifications are replaced with equivalent l

(more reliable) WRNMS functions. The accuracy and response times of the WRNMS are i

superior to those of the SRM/lRM subsystems. Implementation of the WRNMS does not affect any fuel or safety limit. The applicable Bases of the Technical Specifications have been rewritten, and the now Bases rnaintain the equivalent margin of safety as was provided by the SRM/lRM Bases.

j The WRNMS (a) does not decrease a channel trip occurmee beyond its acceptable limit, (b) i does not increase a channel response time beyond its acceptable limit. (c) increases indicated accuracles, and (d) does not cause any plant parameter for any analyzed event to fall outside of its acceptable limit (s).

The surveElance test frequency change of 7 to 31 days is based on the WRNMS having (1) fixed in-core detectors, (2) greater reliab81ty than the SRMs and IRMs, and (3) self test features. The 13 second allowable value for the WRNM Period-Short survellance, and the surve81ance test i

frequency change of 184 days to 24 months is based on trip setpoint calculations using GE's standard (NRC approved) setpoint methodology.

The MNMS will not involve a reduction in a margin of safety, as loac:s on plant equipment wel not increase, and reactions to or results of transients and postulated accidents wul not increase from those presently approved by the NRC.

Environmentalimoact Assessment I

An environmental impact assessment is not required for the changes proposed by this application because the changes conform to the criteria for " actions eligible for categorical exclusion" as specified in 10 CFR 50.22(c)(9).

The proposed change supports Modification P00271 which wRI replace the existing SRMs/lRMs with a i

new WRNMS. This Application involves no significant change in the types or significant incrsase in the l

amounts of any effluent that may be released offsite. As a result of the proposed changes, there wHi be j

no sigrdficant increase in individual or cumulative occupational radiation exposure, and no changes to power levels.

Conclusion The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed changes and have concluded that they do not involve an unreviewed safety question and that they are not a threat to the health and safety of the public.

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ATTACHMENT 2 PEACH BOTTOM ATOMIC POWER STAT'.ON UNITS 2 AND 3 l

Docket Nos. 50-277 50 278 Ucense Nos. DPR-44 DPR-56 Facility Operating Ucense Change Request 93-18

" Wide Range Neutron Monitoring I: strumentation" List of Revised Pages Unk 2 UnN2 Unh3 UnN3 (cont.)

(cont.)

i B 3.3-30 i

B 3.3-30 1.1-2 B 3.3-31 1.1-2 B 3.3-31 3.3-4 B 3.3-32 3.3-4 B 3.3-32 3.3-5 B 3.3-33 3.3-5 B 3.3-33 3.3 7 8 3.3-34 3.3-7 B 3.3-34 3.3-10 B 3.3-36 3.3-10 83.336 3.3-11 E 3.3 37 3.3-11 83.337 3.3-12 5 3.3-38 3.312 B 3.3-38 3.313 B 3.3-39 3.3 13 8 3.3-39 3.3-14 B 3.3-40 3.3-14 B 3.3 40 3.3-15 B 3.3-41 3.3-15 B 3.3-41 3.6-23 B 3.3-42 3.6 23 B 3.3-42 3.6-24 8 3.3-43 3.6 24 8 3.3-43 i (BASES)

B 3.3 44 I (BASES)

B 3.3-44 B 3.2-3 83.649 B 3.2-3 B 3.6-49 B 3.2-8 B 3.6-50 B 3.2-8 B 3.6-50 i

B 3.3-5 83.651 B 3.3-5 B 3.6-51 B 3.3-6 8 3.6 4 B 3.3-6 8 3.9-8 B 3.3-7 8 3.9-10 B 3.3-7 B 3.9-10 B 3.3-10 8 3.9-14 B 3.3-10 B 3.9-14 B 3.3-11 B 3.10 5 B 3.3-11 B 3.10-5 B 3.3-12 8 3.10-31 B 3.312 B 3.10-31 B 3.3-26 8 3.10-32 B 3.3-26 B 3.10-32 B 3.3-29 B 3.3-29 3

i

TABLE OF CONTENTS are:

1.0 USE AND APPLICATION 1.1-1 1.1 Definitions 1.1-1 1.2 Logical Connectors...................

1.2-1 1.3 Completion Times....................

1.3-1 1.4 Frequency 1.4-1 2.0 SAFETY LIMITS (SLs) 2.0-1 2.1 SLs 2.0-1 2.2 SL Violations 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY....

3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS...............

3.1-1 3.1.1 SHUTDOWN MARGIN (SDM) 3.1-1 3.1.2 Reactivity Anomalies................

3.1-5 3.1.3 Control Rod OPERABILITY 3.1-7 3.1.4 Control Rod Scram Times 3.1-12 3.1.5 Control Rod Scram Accumu'ators...........

3.1-15 3.1.6 Rod Pattern Control 3.1-18 3.1.7 Standby Liquid Control (SLC) System 3.1-20 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves.

3.1-26

.1 AV HkATdENERATIbNRATd

g (APLHGR) 3.2-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) 3.2-2 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) 3.2-4 i

WRWh 3.3 INSTRUMENTATION 3.3-1 3.3.1.1 Reactor Prote tj,on System (RPS) Instrumentation 3.3-1 3.3.1.2 hh'de O n e Ranfs" onitor (

) Instrumentation.....

3.3-10 3.3.2.1 Control Rod Block Instrumentation 3.3-16 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation..,..............

3.3-22 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation...

3.3-24 3.3.3.2 Remote Shutdown System...............

3.3-27 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation 3.3-29 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation 3.3-32 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation................. 3.3-44 3.3.6.1 Frimary Containment Isolation Instrumentation 3.3-48 3.3.6.2 Secondary Containment Isolation Instrumentation 3.3-55 3.3.7.1 Main Control Room Emergency Ventilation (MCREV)

System Instrumentation 3.3-59 3.3.8.1 Loss of Power (LOP) Instrumentation 3.3-61 i

3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring.....................

3.3-65 wear (continued)

PBAPS UNIT 2 i

Revision 0

Definitions 1.1

-1.1 Definitions (continued) i a

CHANNEL FUNCTIONAL-TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or ac'.ual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarm, interlock, i

display, and trip functions, and channel failure trips. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested.

CORE ALTERATION,

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel.

The following exceptions are not considered to be CORE ALTERATIONS:

d Hovement or. u de.

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range monitors, local power l

a.

range monitors,J"termediata r:r.;;. -itert, f

traversing incore probes, or special movable detectors (including undervessel replacement);

and i

b.

Control rod movement, provided there are no fuel assemblies in the associated core cell.

i Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

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CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5 Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) that alone woult produce the same thyroid dose as the quantit} and isotopic mixture of I-131, I-132, I-133, I-134, and 1-135 actually present.

The thyroid dose-conversion factors used for this calculation shal~.

be those listed in Table III of TID-14844, AEC,1962, " Calculation of Distance Factors for Power and Test Reactor Sites."

(continued) i PBAPS UNIT 2 1.1-2 Amendment s

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RPS Instrumentation 3.3.1.1

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SURVEILLANCE RE0VIREMENTS (continued)

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SURVEILLANCE FREQUENCY SR 3.3.1.1.3

NOTE-------------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

i Perform CHANNEL FUNCTIONAL TEST.

7 days SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST.

7 days

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SR 3.3.1.1.5 va"4fy the reurce

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4 "+ a "*a'4 4 ?+ a "?"ga '"criter (IRu) ;g;3gg)3 m4,ggy. j evarl2p.

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NOTE SR 3.3.1.1.6 Orly re';uired t: 5: m;t during entry,,,Lu "00E ? 'rer uCDE 1.

Va"ify the !"" and f?R". Channels everiep-7 d:yr-Perform CHMINEL fuflCMNAL TES7*

3) dewy $

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1 SR 3.3.1.1.7 Adjust the channel to conform to a 31 days calibrated flow signal.

SR 3.3.1.1.8 Calibrate the local power range monitors.

1000 MWD /T average core exposure (continued)

M.

W PBAPS UNIT 2 3.3-4 Amendment

RPS Instrumentation 3.3.1.1 SURVEILLANCE REOUIREMENTS (continued)

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SURVEILLANCE FREQUENCY SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.1.1.10

NOTE-------------------

Radiation detectors are excluded.

Perform CHANNEL CALIBRATION.

92 days 3.3.1.1.hh------------------NOTES------------------

SR 1.

Neutron detectors are excluded.

2.

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

n Perform CHANNEL CALIBRATION J Rd d'y: -

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SR 3.3.1.1.1,2'

NOTES------------------

i 1.

Neutron detectors are excluded.

i 2.

For Function 2.a, not required to be performed when entering MODE 2 from 1

MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering i

MODE 2.

l Perform CHANNEL CALIBRATION.

18 months i

SR 3.3.1.1.13 Verify Turbine Stop Valve-Closure and 24 months Turbine Control Valve Fast Closure, Trip 011 Pressure-Low Function; are not bypassed when THERMAL POWER is 2 30% RTP.

(continued)

ITF i

PBAPS UNIT 2 3.3-5 Amendment

RPS Instrumentation

.t 3.3.1.1 b

Table 3.3.1.1 1 (page 1 of 3)

Reactor Protection Systern Instrunentation APPLICARLE COWo!TIONS sans-M00ES OR REQUIRED REFERENCED r-OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWARLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

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Average Power Range Monitora StartLp High Flux 2

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SR 3.3.1.1.1 s 15.0% RTP SR 3,.3,.1.1.3, a.

scram SR 3.3.1.1.8 sa 3.3.1.1 g il SR 3.3.1.1.17 b.

Flow Blased High 1

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$R 3.3.1.1.1 5 0.66 W IDI st 3.3.1.1.2

+ 63.9% RTP scram SR 3.3.1.'1.7 SR 3.3.1.1.8

$R 3.3.1.1.9 SP.

3.3.1.1. X II SR 3.3.1.1.17 c.

Scram Clanp 1

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$R 3.3.1.1.1 s 118.0% RTP SR 3.3.1.1.2 SR 3.3.1.1.8 SR 3.3.1.1.9 SR 3.3.1.1.M il st 3.3.1.1.17 d.

Downscale 1

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C SR 3.3.1.1.8 NA SR 3.3.1.1.9 SR 3.3.1.1.17 (continued)

With any control red withdrawn f rom a core cell containing one or sere fuel asserelles.

(a)

AW RTP when reset for single loop operstion per LCD 3.4.1, " Recirculation Loops 3

(b) 0.66 W + 63.9%

0.66 Operating."

p Amendment 3.3-7 PBAPS UNIT 2

i Lortam

,STW Instrumer.tation 3.3.1.2 INSTRUMENTATION 3.3

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uude Re,ge..deJ~s H Nov [UR Umb 3.3.1.2 S:"r: tr,ge ".or,itor (SPF.)- Instrumentation ts@ *1 LC0 3.3.1.2 The.SRff instrumentation in Table 3.3.1.2-1 shall be OPERABLE.

4 APPLICABILITY:

According to Table 3.3.1.2-1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME veam A.

One or more required A.1 Restore required SRMs 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months usum, SRMT inoperable in to OPERABLE status.

MODE 2r Mtr w

intermedi:t: r:r.;: --

.-- m;r,it;rs (IP" ) un Sng: 2 Or b:h.

vum B.

Three require JRMs' B.1 Suspend control rod Immediately inoperable in MODE 2.

withdrawal.

with IP": cr. tr.ge 2 e* beleu C.

Required Action and C.1 Be in MODE 3.

12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

associated Completion Time of Condition A or B not met.

(continued) 4 PBAPS UNIT 2 3.3-10 Amendment

UKWM SRM' Instrumentation 3.3.1.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

One or more required D.1 fully insert all I hour owng inoperable in insertable control MODE 3 or 4.

rods.

AND D.2 Place reactor mode 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months switch in the shutdown position.

E.

One or more required E.1 Suspend CORE Immedia.ely u y g inoperable in ALTERATIONS except MODE 5.

for control rod insertion.

AND E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies.

/

SURVEILLANCE REQUIREMENTS

NOTE--------------------------------------

Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified conditions.

SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued)

=/

PBAPS UNIT 2 3.3-11 Amendment

WAW" Instrumentation

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SRM 3.3.1.2

- SURVEILLANCE RE0VIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.2.2

NOTES------------------

1.

Only required to be met during CORE ALTERATIONS.

supMM 2.

One.SRM'may be used to satisfy more than one of the following.

wANM Verify an OPERABLE _SRfrdetector is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months located in:

a.

The fueled region; b.

The core quadrant where CORE ALTERATIONS are being performed, when is included in the the associated,SRM'hwng fueled region; and c.

A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated,SRM'is included seg py in the fueled region, kS9e4rfy.

t SR 3.3.1.2.3 Perform CHANNEL CHECK.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months '

(continued) f L

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PBAPS UNIT 2 3.3-12 Amendment n

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SRM' Instrumentation 3.3.1.2 s

SURVEILLANCE RE0VIREMENTS (continued)

SURVEILLANCE FREQUENCY SR. 3.3.1.2.4

NOTES------------------

1.

Not required to be met with less than or equal to four fuel assemblies adjacent ~ to the SRM'and no other fuel assemblies in the[ associated core quadrant.

WW 2.

Not required to be met during spiral unloading.

Verify count rate is:

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE a.

2 3.0 cps; or ALTERATIONS

b..Within the limits of ANQ Figure 3.3.1.2-1.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

-S P 3.2.1.2.:

P. Tv.m CHAHWEL runCliONAL TE5T and i days d:ter.T.inatien ei aiwnal Lu ne.ac

..L.e.

3.3.1.2.[>,------------------NOTE-------------------

SR Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after ICM: en Can e : er belew.-

Yf.em.L".d$L.s25 Mi~ E%

f 0 sup r e+ b e losJ Perform CHANNEL FUNCTIONAL TEST and 31 days determination of signal to noise ratio.

)

i SR 3.3.1.2.J'6 NOTES------------------

l 1.

Neutron detectors are excluded.

2.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after-IRM:

2

YJlucts-m2Een na-- Ls c. Leic&w -

C.K E.

m er

~ y n, v, Perform CHANNEL CALIBRATION.

4 84 d ays--

2 4 H o ff s PBAPS UNIT 2 3.3-13 Amendment w

1 wANM 4

sRf(Instrumentation 3.3.1.2 i

O Table 3.3.1.21 (page 1 of 1)

Y Range ronitor Instrunentatlon i

J~ e:

Widt Neuteea APPLICABLE MCCEs OR CTHER REQUIRED

$URVE!LLANCE.

I FUNCTION SPECIFIED CONDITIONS CHANNELS REQUIREMENTS 1

l kkN9 01eEtC

1. -Gewee Range Monitor 2(a) 3'fy) sa 3.!.1.2.1

^

SR 3.3.1.2.4 sa 3.3.1.2.#" I i

$R 3.3.1.2.f*6 3,4 2

st 3.3.1.2.3 st 3.3.1.2.4 i

st 3.3.1.2 # f st 3.3.1.2.t" 6 5

2(b)(c) st 3.3.1.2.1 SR 3.3.1.2.2 st 3.3.1.2.4

$R 3.3.1.2.5 SR 3.3.1.2 #'6 j

pt?NMS reddthlarE-5% power (a) With :'"- -."_,, 2 or below.

wxNh1 (b) Only one sam' channel Is required to be OPERABLE during spiral of flood or reload den the fueled region includes only that)JtM detector.

w

RWM wAlvMs wMWo+t

~

(c) special movable detectors may be used in place ofjaMd'If comected to nornistgeltcults.

(d) Chan= els mus't /> e, in 3 4 4 core gu a dra nts l

I

+

i l

1

.r.

J M

PBAPS UNIT 2 3.3-14 Amendment 4

knwm M instrumentation 3.3.1.2 c.

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6 10 14 18 22 26 30 Signal to Noise Ratio Figure 3.3.1.2-1 (page 1 of 1) tiinimum J.PfCount Rate Versus Signal to Noise Ratio unm a!.*

=

PBAFS UNIT 2 3.3-15 Amendment

Suppression Pool Average Temperature 3.6.2.1 C.

3.6 CONTAINMENT SYSTEMS J

3.6.2.1 Suppression Pool Average Temperature LC0 3.6.2.1 Suppression pool average temperature shall be:

wUe neatro n a.

s 95'F when any OPERABLE 4atcraedistr range 3 monitor (wavM) Mchannel is -en Sn:c];or above and no testing that af /.00E OFcpsuer ados heat to the suppression pool is being performed; wnum at t coso?., Po""

b.

s 105*F when any OPERABLEMchannel isjr tng: ? or above and testing that adds heat to the suppression pool is being performed; and wnNM c.

s 110*F when all 0PERABLEMchannels are below

-6 c 7.

3 i, co G og, you'er APPLICABIL;TY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 2

A.

Suppression pool A.1 Verify suppression Once per hour average temperature pool average

> 95'F but s 110*F.

temperature s 110*F.

'AND AND WANN Any OPERABLE X A.2 Restore suppression 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

,fgoog,g /'""ch m al er Pange 7 or pool average above.

temperature to e

s 95'F.

AND Not performing testing that adds heat to the suppression pool.

(continued)

PBAPS UNIT 2 3.6-23 Amendment

~~

("I Suppression Pool Average Temperature 3.6.2.1

'~

ACTIONS '(continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Required Action and B.1 Reduce THERMAL POWER 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion until all OPERABLE Time of Condition A w/fNM JMt'" channels are not met.

below Dan;; 7.-

h00E 0)g f. owe,.

l C.

Suppression pool C.1 Suspend all testing Immediately average temperature that adds heat to the

> 105'F.

suppression pool.

AND wtWM Any OPERABLE _IM-

,pg##8/*"'c h:nn:1 en hng: 7 or c

above.

+

E z

Performing testing m

that adds heat to the g

suppression pool.

i D.

Suppression pool D.1 Place the reactor Immediately

' average temperature mode switch in the

> 110*F but s 120*F.

shutdown position.

AND D.2 Verify suppression Once per pool average 30 minutes temperature s 120*F.

E i

D.3 Be in MODE 4 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 1

.l (continued) l m

e PBAPS UNIT 2 3.6-24 Amendment a.

TABLE OF CONTENTS B 2.0-1 B 2.0 SAFETY LIMITS (SLs)

-B 2.1.1 Reactor Core SLs.................

B 2.0-1 B 2.1.2 Reactor Coolant System (RCS) Pressure SL B 2.0-7 8 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY...

B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY B 3.0-10 B 3.1 REACTIVITY CONTROL SYSTEMS..............

B 3.1-1 B 3.1.1 SHUTDOWN MARGIN (SDM)

B 3.1-1 B 3.1.2 Reactivity Anomalies...............

B 3.1-8 8 3.1.3 control Rod OPERABILITY B 3.1-13 B 3.1.4 Control Rod Scram Times B 3.1-22 B 3.1.5 Control Rod Scram Accumulators..........

B 3.1-29 B 3.1.6 Rod Pattern Control B 3.1-34 8 3.1.7 Standby Liquid Control (SLC) System B 3.1-39 B 3.1.B Scram Discharge Volume (SDV) Vent and Drain Valves B 3.1-4B B 3.2 POWER DISTRIBUTION LIMITS B 3.2-1 B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE B 3.2-1 (APLHGR)

B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

B 3.2-6 B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

B 3.2-11 v6.%)

B 3.3-1 B 3.3 INSTRUMENTATION B 3.3.1.1 j Reactor Protection System RPS) Instrumentation B 3.3-1 B 3.3.1.2 4-f"-SeWG Raoye Monitor.pR$ Instrumentation....

B 3.3-36 B 3.3.2.1 e Control Rod Block Instrumentation B 3.3-45 B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation................

B 3.3-57 B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation..

B 3.3-64 B 3.3.3.2 Remote Shutdown System..............

B 3'.3-75 1

B 3.3.4J1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation B 3.3-82

)

B 3.3.5.1 Emergency Core Cooling System (ECCS)

Instrumentation................

B 3.3-91 B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System' Instrumentation................

B 3.3-129 B 3.3.6.1 Primary Containment Isolation Instrumentation B 3.3-140 B 3.3.6.2 Secondary Containment Isolation Instrumentation B 3.3-167 B 3.3.7.1 Main Control Room Emergency Ventilation (MCREV)

System Instrumentation B 3.3-17B B 3.3.B.1 Loss of Power (LOP) Instrumentation B 3.3-1B5 B 3.3.B.2 Reactor Protection System (RPS) Electric Power B 3.3-195 Monitoring (continued) i n

PBAFS UNIT 2 i

Revision 0 4

+

APLHGR B 3.2.1 7,

BASES LC0 multiplying the smaller of the MAPFAC, and MAPFAC, factors (continued) times the exposure dependent APLHGR limits.

With only one recirculation loop in operation, in conformance with the requirements of LCO 3.4.1, " Recirculation Loops Operating,"

the limit is determined by multiplying the exposure dependent APLHGR limit by the smaller of either MAPFAC, or MAPFAC.

APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels.

Design calculations (Ref. 6) and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases.

This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs.

When in MODE 2, the w </e ra9 e rif 7r inter-medie+e-r-ange -nn4 +"- scram function provides prompt

"'*"'# " / " '##

scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2.

Therefore, at THERMAL POWER levels s 25% RTP, the reactor is operating with substantial margin to the APLHGR limits; thus, this LC0 is not required.

W ACTIONS A.]

If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient analyses may not be met.

Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification.

1 fL.1 If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply.

To achieve this status, THERMAL PONER must be reduced to < 25% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The (continued)

W W

PBAPS UNIT 2 B 3.2-3 Revision 0

1 l

MCPR B 3.2.2 BASES APPLICABILITY flow conditions.

These studies encompass the range of key

)

(continued) actual plant parameter values important to typically limiting transients.

The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP.

This trend is expected to e/# ra e ne uN.

continue to the 5% to 15% power range when entry into MODE 2

'"8"* % p' /., y E f e,,,

occurs.

When in MODE 2, the 4* term diat; rang &Nt dA provides rapid scram initiation for any significant power 1

- " increase transient, which effectively eliminates any MCPR compliance concern.

Therefore, at THERMAL POWER levels

< 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LC0 is not required, 1

ACTIONS A.1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met.

Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is normally sufficient to restore the MCPR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of

)

specification.

M If the MCPR cannot be restored to within its required limits within the associated Compietion Time, the plant must be

)

brought to a MODE or other specified condition in which the LCO does not apply.

To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is 2 25% RTP and then every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

It is compared to the specified limits (continued)

PBAPS UNIT 2 B 3.2-8 Revision 0

k$3 '94114 11 FROM PECOlHUC GROUP E30-3 TO PEOPS TECHN! COL' PAGE.002/002-RPS Instrumentation B 3.3.1.1 BASES APPLICABLE.

The specific Applicable Safety Analyses, LCO, and SAFETY ANALYSES, Applicability discussions are listed below on a Function by LCO, and Function basis.

APPLICABILITY

~(continued)

Wlde h'ad'a WM 9.te-- dht - Ranae. Monitor (IAM)

Rance.Honitor '!=tre. r,el-M Nede Peru Wiale meg 1.a.

1-t:7..;d ut:

tm W%

3 fTheI monitor neutro ux levels om the up range of I#SerY A 9

the urce range mon' r (SRM) to e lower ra of the p rage power ran monitors (AP s). The I are capabl r f generating lp signals th can be us to prevent o

damage res ng from abno_

operatin transients the Ljnterme te power range. fin ggver range, the most signifir. ant snurre nt etivifyrMn'Wisduetocontrol-red withdrawal. The rovides diverse protection from the rod worth minimizer (RWM), which monitors and controls the movement of control rods at low power. The RWM prevents the withdrawal of an out of sequence control rod during startup that could result i[) an unacceptable. neutron flux excursion (Ref. 2). The M provides mitigation of the neutr flux excursion.

To demonstrate the capability of the System,to mitigate control rod withdrawal events,

=

& analysirs ha$e been performed (Ref. 3) to evaluate cui the consequences of control rod withdraw,,eventsduring/'7l%

startup that are mitigated only by the The withdrawal, **N l

g of a control rod out of sequence, duri;j.g startu, analysis (t f. 2 rd Ref. 3) assumes that on $e# channg each trip system is bypassed, demonstrates that the provide protection against local control rod withdrawal errors and results in peak fuel enthalpy below the 170 cal /gm fuel failure threshold criterion.

thens are also capable of limiting other reactivity-

~

excursions during startup, such as cold water injection event s.,. %n ". ' !:',' ", t ': ?": :

tr:f,it:f n =

.'.tr x t s...a........m._,,...

e.2...

...2..

.....4...

Q,.y g g

jj

,,,g N N, y

y (conti,nued)

The WRNMs provide signals to facilitate reactor scram in the event that core reactivity increase (shortening period) exceeds a predetermined reference rate. To determine the reactor period, the neutron flux signal is filtered. The period of this filtered neutron flux signal is used to generate trip signals when the respective trip satpoints are exceeded. The time to trip for a particular reactor period is dependent on the filter time constant, actual period of the signal and the trip setpoints. This period based signal is available over the entire operating range from initial control rod withdrawal to full power operation.

PBAPS UNIT 2 B 3.3-5 Revision 0

RPS Instrumentation B 3.3.1.1 c.

~"

BASES Wic) e h aTren feri oci-Sh er t APPLICABLE 1.a.

Inte,mediat RanceaMonitor '! cut-^

~' u" "ich SAFETY ANALYSES, (continued)

LCO, and wANbt WRNM APPLICABILITY The.1RWSystem is divided into two groups of.4RN channels, with four M channels inputting to each trip system. The analysis of References 2 :nd 3 assumes that one channel in each trip system is bypassed.

Therefore, six channels with three channels in each trip system are required for M W # N/+f OPERABILITY to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. -Th++-

t-ip ir v+4"e 4- 0:ch of th _10-y e nf t he _!pu, w;,ick

.urt be cele:tsd by the operator to maint:ir, thehn--

1"Y uithin trie munii.ured ievei v f a n IPS-range.

The analysis of Reference 3 has adequate conservatism to permit an.Rff Allowable Value of 120 divnien5 ef ;-/J.secc,,4, n r a4uieinn ech

~~

Peried-Shod wg lV pt The !*t:rm:siete Nng: ";nitor Neucron r Ur "ipcFunction must be OPERABLE during MODE 2 when control rods may be withdrawn and the potential for criticality exists.

In MODE 5, when a cell with fuel has its control rod withdrawn,

" ###5 W M Ms~ provide monitoring for and protection against unexpected reactivity excursions.

In MODE 1, the APRM System and the RWM provide protection against control rod withdrawal error events and the are not required. The R

wgvMs-.4Ms are automatically bypassed whe he mode switch >is in the Run position, wg,pN tuij tv'eatron s

1.b.

' t ^ ~eedist' Ranae, Monitor-Inoo ypt/Q This trip signal provides assurancejthat a minimum number of w AN/44Ms are OPERABLE. Anytime :n :PJt made switch is moved to any position other than " Operate,"-t"^ detect r valt;;; a h5s r/

gewer occurs, or He

-dr p; b 10'.-?

preset 10'! M or whan ' module n not plugged h/I-fed S/5f#"

4e, an inoper tive trip signal will be receiy4by the RPS dete<#s

/,//ure unless the T s bypassed.

Since ly one T W in each trip a

wA/c/. waald y,u// l. f4e system may be bypassed, only one in each RPS trip system l

"'Y U'

P

9 " ** # W ' 9"'

to n o f a s a h t;" relo U b IIda j This Function was not specifically credite-d'in the acci' dent analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

(continued) 4 PBAPS UNIT 2 B 3.3-6 Revision 0

RPS Instrumentation

~

B 3.3.1.1 BASES w.h. R - e Nw h*

APPLICABLE 1.b.

linWrkeddte Kabce Monitor-Inoo (continued) v' A R-p AA M" SAFETY ANALYSES, Six channels of the httdmehr.aMaii}e. Monitor-Inop LCO and Function, with three channels in each trip system, are APPtICABILITY required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this function on a valid Since this Function is not assumed in the safety signal.

analysis, there is no Allowable Value for this function.

This Function is required to be OPERABLE when the wide Rawp

./Jtde M~ihuqE Monitor function is Averace Power Ranoe Monitor 2.a. Averace Power Ranae Monitor Startuo Hiah Flux Scram The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core which provide an indication of the power distribution and local power The APRM channels average these LPRM signals to changes.

provide a continuous indication of average reactor power For from approximately 1% RTP to approximately 125% RTP.

operction at low power (i.e., MODE 2), the Average Power Range. Monitor Startup High Flux Scram Function is capable of generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range.

For most operation at low power levels, the Average Power Range Monitor Startup High Flux Scram Function will provide a de-secondary scram to the Raf s NrA %<b

"\\f1.dP411g5 Function because of the relative setpoints. JS-d+

th: I'""

t.~.e. s : cc 10, it is possible that the Average gg,,9 Power Range Monitor Startup High Flux Scram Function will h.p,o p,vg ]m /,3 provide the primary trip signal for a core wide increase in power.

No specific safety analyses take direct credit for the Average Power Range Monitor Startup High Flux Scram Function.

However, this Funct' ion indirectly ensures that before the reactor mode switch is placed,in the run position, reactor power does not exceed 25% RTP (SL 2.1.1.1)

(continuedl 8 3.3-7 Revision 0 PBAPS UNIT 2

c RPS Instrumentation e

B 3.3.1.1 BASES APPLICABLE 2.b.

Averace Power Rance Monitor Flow Biased Hich Scram SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY The Average Power Range Monitor Flow Biased High Scram Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL).

During MODES 2 and 5, other to w m M and APRM Functions provide protection for fuel cladding integrity.

2.c.

Averace Power Rance Monitor Scram Clamo The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases.

The Average Power Range Monitor i

Scram Clamp Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure.

For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Scram Clamp function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety / relief valves (S/RVs),

limit the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 5) takes credit for the Average Power Range Monitor Scram Clamp Function to terminate the CRDA.

The APRM System is divided into two groups of channels with three APRM channels inputting to each trip system.

The system is designed to allow one channel in each trip system to be bypassed. Any one APRM channel in a trip system can cause the associated trip system to trip.

Four channels of Average Power Range Monitor Scram Clamp with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

In addition, to provide adequate coverage of the entire core, at least 14 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located.

(continued) h'APS UNIT 2 B 3.3-10 Revision 0

^

RPS Instrumentation B 3.3.1.1 m

~

BASES APPLICABLE 2.c.

Averaae Power Ranoe Monitor Scram Clamo (continued)

SAFETY ANALYSES, LCO, and The Allowable Value is based on the Analytical Limit APPLICABILITY assumed in the CRDA and the loss of feedwater heater event analyses.

The Average Power Range Monitor Scram Clamp Function is i

required to be OPERABLE in MODE 1 where the potential j

consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Scram Clamp Function is i

assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Startup High Flux Scram Functicn conservatively bou the assumed trip and, together with the assumed rips, provides adequate protection.

Therefore, the Average Power Range Monitor Scram Clamp Function is not required in MODE 2.

2.d.

Averaoe Power Ranae Monitor-Downscale This signal ensures that there is adequate' Neutron Monitoring System protection if the reactor mode switch is ag, placed in the run position prior to the APRMs coming on scale.

With the reactor mode switch in run, an APRM w,W

=

y,y ~

downscale signal coincident with an associated -Int:rmedi;tc d

Range Monitor Meutrcr. FM-lligh or Inop signal generates a refjo)-Mort' trip signal.,This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The APRM System is divided into two groups of channel: with three inputs into each trip system.

The system is designed to allow one channel in each trip system to be bypassed.

wANAl g(y_gg. bypass switch.)However, the potential exists to bypass a s Four channels of Average Power Range Monitor-Downscale with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single failure will preclude a ggt,,

scram from this Function on a valid signal. The W,Je.

Interm;a.oi.aangeVMonitor Neutrer 4 = High and Inop ; p.

Functions are also part of the OPERABILITY of the Average, Mfg #

Power Range Monitor-Downscale Function.

If either of these 1

w AN/4 AftPr Functions cannot send a signal to the Average Power Range Monitor-Downscale Function either automatically when (continued) b' PBAPS UNIT 2 B 3.3-11 Revision 0

RPS Instrumentation E,. ;

S-3.3.1.1 BASES-APPLICABLE-2.d.

Averace Power Ranoe Mopitor--Downscale' (continued)

SAFETY ANALYSES, u nu w-LCO, and the trip conditions exist or ggnyally when the 1RN[is APPLICABILITY inoperable (e.g., 'when,lAMFfs taken out of operate), the associated Aver' age Power Range Monitor--Downscale channel is considered inoperable.

The Allowable Value is based upon ensuring that the APRMs are oa scale when transfers are made between APRMs and% W^'Ms This Function is required to be OPERABLE in MODE 1 since this is when the APRMs are the primary indicators of reactor 6

power.

This Function is automatically bypassed when the 1

mode switch is not in the Run position.

6 2.e.

Averace Power Ranoe Monitor--Inoo i

This signal provides assurance that a minimum number of APRMs are OPERABLE.

Anytime an APRM mode switch is moved to any position other than " Operate," an APRM module is unplugged, the electronic operating voltage is low, or the APRM has too few LPPM inputs (< 14), an inoperative trip signal will be received by the RPS, unless the APRM is bypassed.

Since only one APRM in each trip system may be bypassed, only one APRM in each trip system may be 1

inoperable without resulting in an RPS trip signal. This

- Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy. and diversity of the RPS as required by the NRC approved licensing basis.

Four channels of Average Power Range Monitor--Inop with two-channels in each trip system are required to be OPERABLE to ensure that no single failure will preclude a scram from e

this Function on a valid signal.

There is no Allowable Value for this Function.

This Function is required to be OPERABLE in the MODES where the APRM Functions are required.

(continued) l l

PBAPS UNIT 2 B 3.3-12 Revision 0

=.-

~~:--

RPS Instrumentation B 3.3.1.1 1

v f

? '} I c

'GAsiS I

ACTIONS C.1 (continued) w,wm on a valid signal.

For the typical / Function with one-out-of-two taken twice logic and the:lAMrand APRM Functions, this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip).

For Function 5 (Main Steam Isolation Valve--Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems)

OPERABLE or in trip (or the associated trip system in trip).

For Function 8 (Turbine stop Valve--Closure), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system.in trip).

For Functions 12 (Reactor Mode Switch--Shutdown Position) and 13 (Manual Scram), this would require both trip systems to have one channel, each OPERABLE or in trip (or the associated trip system in trip).

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is acceptable because it minimizes t

risk while allowing time for restoration or tripping of channels.

IL.1 Required Act*:on D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1.

The applicable-condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed.

Each time an inoperable channel has not met any Required Action 1

of Condition A, B, or C and the associated Completion Time has expired, Condition 0 will be entered for that channel and provides for transfer to the appropriate subsequent Condition.

(continued) i PBAPS UNIT 2 S 3.3-26 Revision 0

RPS Instrumsntation S 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.2 (continued)

REQUIREMENTS A restriction to satisfying this SR when < 25% RTP is provided that requires the SR to be met only at 2 25% RTP because it is difficult to accurately mainta.n APRM indication of core THERMAL POWER consistent with a heat balance when < 25% RTP.

At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR).

At 2 25% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with SR 3.0.2.

A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2.

In this event, the SR must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reaching or exceeaing 25% RTP.

Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

SR 3.3.1.1.3 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.

Any setpoint adjustment shall be made consistent with the assumptions of the ;urrent plant specific setpoint methodology.

As noted, SR 3.3.1.1.3 is not required to be perform'ed when entering MODE 2 from MODE 1, since testing of the MODE 2 required 47,." e.M-APRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links.

This allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2.

In this event, the'SR must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2 from MODE 1.

Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9).

(continued)

PBAPS UNIT 2 B 3.3-29 Revision 0

RPS Instrumentation B 3.3.1.1 ff BASES SURVEILLANCE SR 3.3.1.1.4 REQUIREMENTS (continued)

A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A Frequency of 7 days provides an acceptable level of system average availability over the Frequency and is based on the reliability analysis of References 9 and 10.

(The RPS Channel Test Switch Function's CHANNEL FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram Functions' Frequencies.)

SR 3.3.1.1.5 and SR 3.3.1.1.6 These Surveill ces are established to ernure that no gaps,\\

in neutron x indication exist from,su'bcritical to power operatio or monitoring core react,Wity status.

/

The erlap between SRMs and IRMs is required to be d onstrated to ensure thatp'eactor power will not be increasedintoaneutron/luxregionwithoutadequat.e indication. This is rpquired prior to withdrawing-SRMs from the fully inserted msition since indication isdeing transitioned from 1e SRMs to the IRMs.

/

yserp B +

The overlap tween IRMs and APRMs is o neern when reducing ower into the IRM range. Afi power increases, the system sign will prevent further' increases (by initiating a ro lock) if adequate overlap'is not maintained. Overlap between IRMs and APRMs exis,ts'when sufficient IRMs and APRMs/

oncurrently have onscale. readings such that the transitipn between MODE I and MODE'2 can be made without either APRM downscale rod block,pr' IRM upscale rod block. Overl ap '

between SRMs and JRMs similarly exists when, prior,to withdrawing the/SRMs from the fully inserted posjfion, IRMs are above mid/ scale on range 1 before SRMs have' reached the upscale ro lock.

As no d, SR 3.3.1.1.6 is only required o be met during ye,tyfintoMODE2fromMODE1.

en That -, after the overlap quirement has been met and indica ion has transitioned to

[,/the IRMs, maintaining overlap isen,ot required (AP reading downscale once in M0QV2).

l w_

(continued) misme PBAPS UNIT 2 B 3.3-30 Revision 0

l i

}

J Snsed B

\\

j A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the-intended function.

Any setpoint adjustment shall be made consistent with the assumptions of the current plant specific setpoint methodology.

As noted, SR 3.3.1.1.5 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required WANM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable i

links.

This allows entry into MODE 2 if the,3) day Frequency is not met per SR 3.0.2.

In this event, the SR must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering i400E 2 from MODE 1.

Twelve hours is based on operating experience and in consideration of providing a reasonable. time in which to complete the SR.

1 A Frequency of31 days provides an acceptable level of system average unavailability over the Frequency interval and is i

based on fixed awe detec%s, overall re lo'al,ilit '

I an d s e If-me., tt o<.'y featu res.

i I

r f

I

~*

RPS Instrumentation B 3.3.1.1

>b BASES SURVEILLANCE SR 3.3 M.5 and SR 3.3.1.1 I (continued)

REQUIREMENTS

/

y' If verlap for a group f channels is not demonstrated e.g.,

IRM/APRM ove ap),.the reason for thtf failure of the

/

'veillance shoul be determined and t)f appropriate inel(s) deenred inoperable. OnlyM hose appropriate anels thaf are required in the

'rrent MODE or condition /

anou declared inoperable.

Afrequency of 7 days is re_ onable based on engi ring

)

Judgment and the reliabiMfy of the IRMs and AP s.

/

j SR 3.3.1.1.7 The Average Power Range Monitor Flow Biased High Scram Function uses the recirculation loop drive flows to ' vary the trip setpoint. This SR ensures that the total loop drive flow signals from the flow units used to vary the setpoint is appropriately compared to an injection test flow signal to verify the flow signal trip setpoint and, therefore, the APRM Function accurately reflects the required setpoint as a M

function of flow.

If the flow unit signal is not within the appropriate limit, the affected APRMs that receive an input from the inoperable flow unit must be declared inoperable.

The Frequency of 31 days is based on engineering judgement, operating experience, and the reliability of this instrumentation.

SR 3.3.1.1.8 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)

System. This establishes the relative local flux profile for appropriate representative input to the APRM System.

The 1000 MWD /T Frequency is based on operating experience with LPRM sensitivity changes.

(continued)

S PBAPS UNIT 2 B 3.3-31 Revision 0

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.14 REQUIREMENTS (continued)

A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.

Any setpoint adjustment shall be consistent with the assumotions of the current plant specific setpoint zthodology.. The 92 day Frequency of SR 3.3.1.1.9 is based on the reliability analysis of Reference 9.

I The 24 month Frequency is based on the need to perform this Surveil. lance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown that these components will pass the Surveillance when performed at the 24 month Frequency.

SR 3.3.1.1.10. SR 3.3.1.1.11. SR 3.3.1.1.12.

SR 3.3.1.1.15. and SR 3.3.1.1.16 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.

This test verifies that the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the current plant specific setpoint methodology.

SR 3.3.1.1.16, however, is only a calibration of the radiation detectors using a standard radiation source.

As noted for SR 3.3.1.1.11 and SR 3.3.1.1.12,' neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.

Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 MWD /T LPRM calibration against the TIPS (SR 3.3.1.1.8).

A second note is provided for SRs 3.3.1.1.?1 and 3.3.1.1.12 that allows the APRM and,IMf' LdRNM i SRs to be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of entering MODE 2 from MODE 1.

Testing of the MODE 2 APRM and I

be performed in MODE 1 without utilizing)2CFunctions cannotjuin g%

leads or movable links.

This Note allows entry into MODE 2 from MODE 1, if the ""l8hy or 18' month Frequency is not met

$1 (continued)

PBAPS UNIT 2 B 3.3-32 Revision 0 4

~..

i

.1 RPS Instrumentation B 3.3.1.1 i-7 BASES l

?

SURVEILLANCE SR 3.3.1.1.10. SR 3.3.1.1.11. SR 3.3.1.1.12.

REQUIREMENTS SR 3.3.1.1.15. and SR 3.3.1.1.16 (continued)-

per SR 3.0.2.

Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. As noted for SR 3.3.1.1.10, radiation ~

detectors are excluded from CHANNEL CALIBRATION due to ALARA reasons (when the plant'is operating, the radiation detectors are generally in a high radiation area; the steam tunnel). This exclusion is acceptable because the radiation detectors are passive devices, with minimal drift. The i

radiation detectors are calibrated in accordance with SR 3.3.1.1.16 on a 24 month Frequency.-

The 92 day Frequency of SR 3.3.1.1.10 1s conservative with respect to the magnitude of equipment drift assumed.in the setpoint analysis. The Frequencies of SR 3.3.1.1:11'and

.3.1.1.12 are based upon the assumption of an14(.de/,o,,r,,

SR M af month calibration interval, respectively, in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequencies of SR 3.3.1~.1.15. and SR 3.3.1.1.16 are based upon the assumption of a 24 month

,E calibration interval in the determination of the magnitude of equipment drift in the applicable setpoint analysis.

SR 3.3.1.1.13 This SR ensures that scrams initiated from the Turbine Stop-o Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low functions will not be -inadvertently j

bypassed when THERMAL POWER is 2 30% RTP. This involves I

calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into i

the Allowable Value ($ 29.4% RTP which is equivalent to g

s 138.4 psig as measured from_ turbine first stage pressure).

and the actual setpoint.

Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THERMAL POWER 2 30% RTP to ensure that the calibration is valid.

(continued) i!9F PBAPS UNIT 2 B 3.3-33 Revisionj i

i n

RPS Instrumentation j

B 3.3.1.1 i

m-l BASES SURVEILLANCE SR 3.3.1.1.13 (continued)

REQUIREMENTS If any bypass channel's setpoint is nonconservative (i.e.,

the Functions are bypassed at 2 30% RTP, either due to open main turbine bypass valve (s) or other. reasons), then the affected Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass).

If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.

The Frequency of 24 months is based on engineering judgment and reliability of the components.

SR 3.3.1.1.17 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel.

The functional testing of control rods 3

(LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8),

overlaps this Surveillance to provide complete testing of f

the assumed safety function.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown that these components will pass the Surveillance when performed at the 24 month Frequency.

REFERENCES 1.

UFSAR, Section 7.2.

2.

UFSAR, Section Chapter 14.

3. - ZDC-23542, "Cuiiti.uoue C=trel Red "ithdcon.1 m L yr Startup R=ge," Spril 1 0, 1978.-

4.

NEDC-32183P, " Power Rerate Safety Analysis Report for Peach Bottom 2 & 3," dated May 1993.

5.

UFSAR, Section 14.6.2.

NEpo-313 Q %'ael.,,, p1pasa,em ed frm lysi.s eOCs n bo / (continued)

Mon o'2'oeiy ;<.s re,L L t'cer, s>y depwr Fa, S

unja yt;)x fa p Weetion M c L B o t To ~

A to ic Powe,SrA<r.i I/n %s 2 +3, No vemb e@

PBAPS UNIT 2 B 3.3-34 Revision 0

wANM

_.SRWInstrumentation B 3.3.1.2 B 3.3 INSTRUMENTATION

~

B 3.3.1.2 Seur : Sng: "cai te. (SRM) Instrumentation i

Wide l{aye / Jeu ten pio,,, Tor [wRNM)

BASES wRwM.s nee capable ef-prcvidej-BACKGROUND The SPt pr:;ide the operator with information relative to the neutron fl x level at very low flux levels in the core.

As such, the -

indication is used by the operator to i

monitor the approach to criticality and determine when criticality is achieved.

na 5" 2r :: int:ia:d full,,

ineart oa ein+ 41 tha en"a+ "'+e ir greater th:n ; minimum

?11 ~ed coun+ *=te (a centrol red bl::h i: ::t at t!..a

)

enna4+4nn) arter eep +g <ntermednt: 7:ng: 7:3jt:7 r:ny;)

c'!:rh; ir da cart"ated (!! *e';ui ed by SP 2.2.1.1.5), the S": 2-a aer-211y #elly uithdra'r 'r = th: ::rt.

wRNM The.SRM'subsy(tgm of the Neutron Monitogi System (NMS) y gp3 l

consists of M channels.

Each of the

- channels can be t, N.

bypassed, but only one at any given time r by the operation r

of a bypass switch.

Each channel includes one detector thatis g eme.edly

n h phy;;cali., positioned in the core.

Each detector assembly consists of a miniature fission chamber with associated cabling, signal conditioning eAgipment, and Wp electronics associated with the various".SWfunctions. The signal conditioning equipment converts the current pulses from the fission chamber to analog DC currents that correspond to the count rate.

Each channel also includes indication, alarm, and control rod blocks.

However,.this LCO specifies OPERABILITY requireme.its only for the monitoring and indication functions of the SRMs'.waya,,

During refueling, shutdown, and low power operations, the i

primary indication of neutron flux levels is provided by. the WAN#f.SRMs~ or special mova)lg' detectors connected to the normal wym,.SRtf circuits. The"%4fs provide monitoring of reactivity changes during fuel or control rod movement and give the control room operator early indication of unexpected suberitical multiplication that could be indicative of an l

approach to criticality.

APPLICABLE Prevention and mitigation of prompt reactivity excursions SAFETY ANALYSES during refueling and low power operation is provided by LCO 3.9.1, " Refueling Equipment Interlocks"; LCO 3.1.1,

" SHUTDOWN MARGIN (SDM)"; LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation"; !" "cutrer % 'ligh and WANm Persed.Shert (continuedi PBAPS UNIT 2 B 3.3-36 Revision 0

(A W1 g instrumentation

..+

o 3..$.1.2 BASES APPLICABLE Average Power Range Monitor (APRM) Startup High Flux Scram SAFETY ANALYSES Functions; and LCO 3.3.2.1, " Control Rod Block (continued)

Instrumentation."

g q,th *,n, w,,. n e. ave 4 um

,,.cr. v w y h v / k /s TheSRMlfhavenosafetyfunction[andarenotassumedto function during any UFSAR desion bajis accident or transient anal ysi sy, However, the,SJtMf %rMile the only on-scale (As h ve* ',"/ '"[,monitoring of neutron flux levels during startuo and af "'l "f

refueling.

Therefore, they are being retained in Technical CM '" A Specifications.

During startup in MODE 2, three of the,;A4-U x.

ccT-channels are LC0 required to be OPERABLE to monitor the reactor flux level a

prior to and during control rod withdrawal, subcritical

" /. -

64 '

multiplication and reactor criticalityy,=d wu'

ry

,a

- u

---+ ~ s - _

_,g de/c 8tv 7a'*[

-t;...oni m a tne ans w. Ro..g : cr d:;;.

11 L vno v.

W d-

. '.; ;r:

pir:d in order to provide a 2

barvls %f / h /"""

representation of tne overall core response during those g A (.f w ( c a v <_

periods when reactivity changes are occurring throughout the yuakyah$$

C0YU*

In MODES 3 and 4, with the reactor shut down, twoJRfC OANM channels provide redundant monitoring of flux levels in the core.

Larw m in MODE 5, during a spiral offload or reload, an SRK outside l

the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the fueled region.of the core.

Thus, CORE ALTEg g y,S are allowed in a quadrant with no OPERABLE,S? Kin an adjacent quadrant provided the Table 3.3.1.2-1, footnote (b),

requirement that the bundles being spiral reloaded or spiral offloaded are all in a gjg),e, fueled region containing at i

least one OPERABLE SRN'is met.

Spiral reloading and offloading encompass reloading or offloading a cell on the edge of a continuou:; fueled region (the cell can be reloaded or offloaded in any sequence).

U6fNim in nonspiral routine operations, two,SRMf are required to be i

OPERABLE to provide redundant monitoring of reactivity changes occurring in the reactor core.

Because of the local nature of reactivity changes during refuel (n y.;iequate coverage is provided by requiring one,SRM'to oe OPERABLE in j

the quadrant of the reactor core where CORE ALTERATIONS are (continued)

I PBAPS UNIT 2 B 3.3 37 Revision 0

wAWM SAH-Instrumentation B 3.3.1.2 BASES m

wnNM LCO being performed, and the other-GRM to bc OPERABLE in an (continued) adjacent quadrant containing fuel.

These requirements ensure that the reactivity of the core will be continuously monitored during CORE ALTERATIONS.

Special movable detectors, according to footnote (c) of Table 3.3.1.2-1, may be used in place of the normal SRM" WANM nuclear detectors.

These special detectors must be connected to the normalMMPcircuits in the HMS, such that the applicable neutron flux indication can be generated.

These special detectors provide more flexibility in monitoring reactivity changes during fuel loading, since they can be positioned anywhere within the core during refueling.

They must still meet the location requirements of SR 3.3.1.2.2 and all other required SRs for SRMr' W g ypf

- nserY 3

T For an SRM channel to be considered OPERABLE, it must.be providing neutron flux monitoring indication.

WANMs APPLICABILITY The SRMs are q " quired to be OPERABLE in MODES 2, 3, 4, and 5

p,,,dA, # p./Trior to the-

bcir.; "- v'la D$nga 4 to provide for j

na

~.

Igg.yg 'j r neutron monitoring.

In MODE 1, the APRMs provide adequate j

c.;.

m nitoring of reactivity changes in the core; therefore, the 17"[hy.fy, f,N* " beve, N TDb provides adequate m 4

are not required.

In MODE 2, with-IRM: en Range : enW Nr d

"6 s

am typff not required.

"d,Qs

% /s*.9 ACTIONS A.1 and B.1 tk u/ANM ek an e ls In MODE 2, "ith-the IRM: cn Range--? cr belew, SRMs provide the means of monitoring core reac ivity and criticality.

With any number of the required

' inoperable, the ability to monitor neutron flux is degraded.

Therefore, a limited time is allowed to restore the inoperable channels to OPERABLE status.

won Provided at least one -SRWremains OPERABLE, Required unuMr Action A.1 allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore the required.SRMf to

{

OPERABLE status. This time is reasonable because there is adequate capability remaining to monitor the core, there is limited risk of an event during this time, and there is sufficient

'ma to take corrective actions to restore the required". f~o OPERABLE status.

During this time, control rod withdrawal and power increase is not precluded by this it il Y~

N7

~

ued)

$ Al e f Yc 5 5. l. J ~ l, Cv$mc $e 60, w t * *" $ """'l'd N*

(#&* #

t 1

meay Spa +ie I cw e

<.c w n) 0 v

A i

B 3.3-38 evision 0

7 4

g.Arv'rbt o'

_.4Rf'rinstrumentation B 3.3.1.2 1

BASES

.g pecceda te wnm indaahk y

ACTIONS A.1 and B.1 (continued) p g,fg. A afs-rg n-er Required Action. Havin the ability to monitor the core with at least one-SPM, cree::d hg to IR.". nr.g; 2 cc greete, bdt' everl:p required by in 2.2.1.1.0), and thereby exiting the Applicability of this LCO, is acceptable for ensuring adequate core monitoring and allowing continued operation.

wANMs With three required SRHr inoperable, Required Action B.1 allows no positive changes in reactivity (control rod withdrawal must be immediately suspended) due to inability to monitor the changes.

Required Action A.1 still applies and allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore monitoring capability prior to requiring control rod insertion. This allowance is based on the limited risk of an event during this time, provided that no control rod withdrawals are allowed, and the desire to concentrate efforts on repair, rather than to immediately shut down, with no SAMs OPERABLE.

' &/rs w

C.1 WANN.s In MODE 2, if the required number of.4RMT is not restored to w

OPERABLE status within the allowed Completion Time, the reactor shall be placed in MODE 3.

With all control rods fully inserted, the core is in its least reactive state with the most margin to criticality. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly mar.ner and without challenging plant systems.

D.1 and D.2 wpNPs With one or more required.SAHr inoperable in MODE 3 or 4, the neutron flux monitoring capability is degraded or nonexistent.

The requirement to fully insert all insertable control rods ensures that the reactor will be at its minimum reactivity level while no neutron monitoring capability is available.

Placing the reactor mode switch in the shutdown position prevents subsequent control rod withdrawal by maintaining a control rod block. The allowed Completion Time of I hour is sufficient to accomplish the Required Action, and takes into account the low probability of an event requiring the SRM' occurring during this interval.

WPNM (continued) u2 PBAPS UNIT 2 B 3.3-39 Revision 0

~

RiWM

, SRM'Instruinentation B 3.3.1.2 BASES ACTIONS E.1 and E.2 (continued)

L>dA/ **s With one or more required SRM' inoperable in MODE 5, the ability to detect local reactivity changes in the core during refueling is degraded.

CORE ALTERATIONS must be immediately suspended and action must be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Suspending CORE ALTERATIONS prevents the two most probable causes of reactivity changes, fuel loading and control rod withdrawal, from occurring.

Inserting all insertable control rods ensures that the reactor will be at its minimum reactivity given that fuel is present in the core.

Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative position.

Action (once required to be initiated) to insert control rods must continue until all insertable rods in core cells containing one or more fuel assemblies are inserted.

vom SURVEILLANCE As noted at the beginning cf the SRs, the SRs for each JWdt REQUIREMENTS Applicable MODE or other specified conditions are found in the SRs column of Table 3.3.1.2-1.

SR 3.3.1.2.1 and SR 3.3.1.2.3 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on another channel.

It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.

Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious.

A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CAllBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability.

If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.

(continued)

PBAPS UNIT 2 B 3.3-40 Revision 0

WAWNI SR rInstrumentation B 3.3.1.2

~

BASES SURVEILLANCE SR 3.3.1.2.1 and SR 3.3.1.2.3 (continued)

REQUIREMENTS The Frequency of once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for SR 3.3.1.2.1 is based on operating experience that demonstrates channel failure is rare. While in MODES 3 and 4, reactivity changes are not expected; therefore, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is relaxed to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for SR 3.3.1.2.3.

The CHANEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

SR 3.3.1.2.2 To provide adequat coverage of potential reactivity changes in the core, one

. is required to be OPERABLE in the quadrant where C0 ALTERATIONS are being performed, and the other OPERABLE

.ust be in an adjacent quadrant containing fuel. Note 1 states that the SR is required to be rr.et only during CORE ALTERATIONS.

It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring.

This Surydllance consists of a review of plant logs to ensure that%Ki' required to be a

OPERABLE for given CORE ALTE IONS are, in fact, OPERABLE.

m.,

In the event that only one is required to be OPERABLE, per Table 3.3.1.2-1, footnote (b), only the a. portion of this SR is required.

Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRK."'u,

n The 12 hcur Frequency is based upon operating experience and supplements operational controls over refueling activities that include steps to ensure that the SR& required by the LC0 are in the proper quadrant.

"' MW i

SR 3.3.1.2.4 w@r1 This Surveillance consists of a verific ion of the MP instrument readout to ensure that the reading is greater than a specified minimum count rate, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. The nal-to-noise ratio shown in Figure 3.3.1.2-1 is the nt rate at which there is a 95Y. probability that the signal indicates the presence of neutrons and only a 5Y. probability that the-SRM -m.;

(continued) a PBAPS UNIT 2 B 3.3-41 Revision 0 i

wAWM 4 **-

,.f '

.-GRM~ Instrumentation B 3.3.1.2 e

^

BASES SURVEILLANCE SR 3.3.1.2.4 (continued)

REQUIREMENTS signal is the result of se (Ref. 1).

With few fuel assemblies loaded, the will not have a high enough count rate to satisfy the SR.

Theref made for loading sufficient " source"fre, allowances are material, in the form of irradiated fuel assemblies, to establish the minimum count rate.

To accomplish this, the SR is modified by Note 1 that tates that the count rate is not required to be met on that has le than or equal to four fuel assemblies adjacent to the and no other fuel assemblies are in the associated core quadrant With four or less fuel assemblies loaded around each* Mand no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical.

In addition, Note 2 states that this requirement does not have to be met during spiral unloading.

If the core is being unloaded in this manner, the various core configurations encountered will not be critical.

The Frequency is based upon channel redundancy and other information available in the control room, and ensures that the required channels are frequently monitored while core reactivity changes are occurring. When no reactivity changes are in progress, the Frequency is relaxed from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SR 3.3.1.2.5 ?"d !"

2.2.1.2.C Performance of a CHANNEL FUNCTI0t!AL TEST demonstrates the associated channel will function properly.

SR 3.3.1.2.5 is j

9 JM *"#

required in MODESP and the @ day Frequency ensures that the r

channels are OPERABLE while core reactivity changes could be

(.p,;,/,;,,,,, j, f,,f,g' in progress. This Frequency is reasonable, based on o veral/ r,/,a fI/' /

operating experienceVand on other Surveillances (such as a 7/

,s e/f. %,7* */ re b es CHANNEL CHECK), that ensure proper functioning between f

CHANNEL FUNCTIONAL TESTS.

to 1 2 1 9 A4e eman4,.ma in unnr 9 v4+h tou, --

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e --

hal nu snA 4n Mont*U 7 snd a C4nen enq E.f+

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de net normally take p12 e i-MODES 3 r.d ' :nd ::r:

""'c+ 4 "4 +y ch ?";"" 2ra "ly t^

^"+r0! -rd :;;;;;,t t;;

d""

(continued) c.

i 1

PBAPS UNIT 2 B 3.3-42 Revision 0 i

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u/ l FN

/

SR+t Instrumentation 5 3.3.1.2 BASES

+7 SURVEILLANCE SR 3.3.1.2.5 :nd P'

'.2.1.:.

(continued)

REQUIREMENTS "00E 2, the caequeacy her beea e.vtended 4 rem ' day: t:

?! day.

The M day r*aa"anny 4 s bseed na epeetting evnn.4nnen sna nn n+har c,,mm 411,.,7 y 3 roy;q',

CHECY.) th:t on:ure preper functiering bett : r C""!E' FU"CTIC"^.L TESTS.

dorre&/ '"#"'D5

!erification of,the signal to noise ratio also ensures that a

p he detectors are daterted te

ptdl; p; rating lovci.

th 3NN f/n,

in 3 s,.11y u4thda?"a conditica, the deteater: 2r tuf#iciaat'y ae'e"ed #ae'" the #"e' ed veg4 ea of the care t:

er ent4 211" a'4minata neutrans + ram *=? china +he datentar

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ano,. niin t rs+e nh+s4ned wh41a +he An+nr+nec see f,11 u 2 ir5Y iE'5:i~"-Ed +5 be ReiEE"'Ed. [ ~

th The Note to the Surveillance allows the Surveillance to be

,v delayed until entry into the specified condition of the f

Applicability (THERMAL POWEF. decreased to

!"" ".or ; x ** *

,.,dg d arE-5%g e r --* below).

The SR must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after MMr W eg 7 /Ars -rgy a 9r.g: 2 r below.

Tr e allowance to enter the readi

per Applicability with the 31 c'ay Frequency not met is reasonable, based on the inmited time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed after entering the Applicability,2nd t'c i nibi'ity to

~

narrnrm +ho enroe<11 3nce ok41e at tigh r p:;;;7 1;;;13, Although the Surveillance could be performed while er Inu at

/,iyAv-pow <-

".r.g : 2, the plant would not be expected to maintain steady state operation at this pcwer level.

In this v nt, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable, based on the"

^%eing I

otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.

SR 3.3.1.2.7 6 Performance of a CHANNEL CALIBRATION at a requency of M matas 181 d:y: verifies the performance of the detectors and associated circuitry.

The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the 4

system or component status.

Note 1 excludes the-neutron detectors from the CHANNEL CALIBRATION because they cannot i

readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant sensitivity 1

over the range and with an accuracy specified for a fixed useful life.

j (continued)

M F' PBAPS UNIT 2 B 3.3-43 Revision 0

WANIN SRM' Instrumentation B 3.3.1.2 i

BASES SURVEILLANCE SR 3.3.1.2."P (continued)

REQUIREMENTS Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the (c/, Applicability. The SR must be performed in MODE 2 within wgunf, re./J, /ASE 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or enter 3ng MODE 2 witH-I""; :n R:ngc 2 or be hw.

er gdlo The allowance to enter the Applicability with the dei d:y41anX4 w

Frequency not met is reasonable, based on the limited time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed after entering the Applicability,: d th:-

an,h414+y +n norsnr. +wo en uep,,,ce ma q : :t 'igher p: :r 1 /1:.

Although the Surveillance could be performed while or Ajp/e, pauser,enIoMo.:ng: 2, the plant would not be expected to maintain steady state operation at this power level.

In thi

vent, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable, based on the being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillance.

REFERENCES 1.

NRC Safety Evaluation Report for Amendment Numbers 147 and 149 to Facility Operating License Numbers DPR-44 and DPR-56, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, August 28, 1989.

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PBAPS UNIT 2 B 3.3-44 Revision 0

h' * *

Suppression' Pool Average Temperature B 3.6.2.1 4

E BASES APPLICABLE Reference' 1 and Reference 2 analyses.

Reactor shutdown at a SAFETY ANALYSES pool temperature of 110'F and vessel depressurization at a (continued) pool temperature of 120*F-are# assumed for the Reference 2

-analyses. The limit of 105'F, at which testing is terminated, is not used in the safety analyses because DBAs are assumed to not initiate during unit testing.

Suppression pool average temperature satisfies Criteria 2 and 3 of the NRC Policy Statement.

LCO A limitation on the suppression pool average temperature is required to provide assurance that the containment-conditions assumed for the safety analyses are met. This limitation subsequently ensures that peak primary containment pressures and temperatures do not exceed maximum allowable values during a postulated DBA or any transient resulting in heato of the suppression pool. The LCO requirements are-Average temperdure s 95'F when any OPERABLE wI/e r8^78 a.

r,edtn.,

c.,ib &aN)O.t;mdi:t; cenv. munii.wi UTe>T channel is aa " _:;;; P of f,fa cqg f,w er y or above and no testing that adds heat to the amp suppression pool is being performed. This requirement ensures that licensing bases initial conditions are met.

b.

Average temperature s 105'F when any OPERABLE M W 4#M af/.04E08 F*e- --

cnannel is#: ":..;; 7 or above and testing th't. adds a

heat to the suppression pool is being performed. This required value ensures that the unit has testing flexibility, and was selected to provide margin below the 110'F limit at which reactor shutdown is required.

When testing ends, temperature must be restored to s 95'F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months according to Required Action A.2.

Therefore, the time period that the temperature is > 95'F is short enough not to cause a significant increase in unit risk.

c.

Average temperature s 110*F when all OPERABLE 4Mt-WMA1 j,00 E 07, /

P r. Mannels are below'"...v. 7.

This requirement ensures

~

that the unit will be shut down at > 110'F. The pool is designed to' absorb decay heat and sensible heat but could be heated beyond design limits by the steam generated if the reactor is not shut down.

(continued)

PBAPS UNIT 2 B 3.6-49 Revision 0

,i' Suppression Pool Average Temperature l

B 3.6.2.1 t

T,

BASES uRNr1 an I,coEog swer LCO Note thataindication = "'" hag; 7 is a convenient measure g

(continued).

of when the reactor is producing power essentially equivalent to 1% RTP. At this power level, heat input is approximately equal to normal system heat losses.

j APPLICABILITY In MODES 1, 2, and 3, a DBA could cause significant heatup of the suppression pool.

In MODES 4 and 5, the probability and consequences of these events are reduced due to..the i

pressure and temperature limitations in these MODES.

Therefore, maintaining suppression pool average temperature within limits is not required in MODE 4 or 5.

i ACTIONS A.1 and A.2

~

With the suppression pool average temperature above the specified limit when not performing testing that adds heat to the suppression pool and when above the specified power indication, the initial conditions exceed the conditions assumed for. the Reference 1, 2, and 3 analyses. However, primary containment cooling capability still. exists, and the r$

primary containment pressure suppression function will occur i

at temperatures well above those assumed for safety.

analyses. Therefore, continued operation is allowed for a limited time. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is adequate to i

allow the suppression pool average temperature to be restored below the limit. Additionally, when suppression pool temperature is > 95'F, increased monitoring of' the suppression pool. temperature is required to ensure that it i

remains s 110'F. The once per hour Completion Time is adequate based on past experience, which has shown that pool temperature increases relatively slowly except when testing

]

that adds heat to the suppression pool is being performed.

Furthermore, the once per hour Completion Time is considered adequate in view of other indications in the control room, i

including alarms, to alert the operator to an abnormal suppression pool average temperature condition.

B.d If the suppression pool average temperature cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the power must be reduced to below hag; 7-for all OPERABLE 4W&within h002ci;powe,r WAWPts PBAPS UNIT 2 B 3.6-50 Revision 0

~

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Suppression Pool Average Temperature B 3.6.2.1 BASES.

. ACTIONS M (continued) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is reasonable, based on operating experience, to reduce power from full power conditions in an orderly manner and without challenging plant systems.

t M

wANbtchonell$'all'0

W Suppression pool averag temperature is allowed to be > 95'F when any OPERABLE-I"" :..::::1 h e-H aar 7-or above, and when testing that adds heat to the suppression pool is being performed. However, if temperature is > 105'F, all testing must be immediately suspended to preserve the heat absorption capability of the suppression pool. With the testing suspended, Condition A is entered and the Required Actions and associated Completion Times' are applicable.

0.1. D.2. and 0.3 T.

Suppression pool average temperature > 110*F requires that "r'

the reactor be shut down immediately. This is accomplished t

by placin'g the reactor mode switch in the shutdown position.

l Further cooldown to M00E.4 is required at normal cooldown l

rates (provided pool temperature remains s 120*F).

Additionally, when suppression pool temperature isi> 110*F, increased monitoring of pool temperature is required to i

ensure that it remains s 120'F. The once per 30 minute i

completion Time is adequate, based on operating experience.

Given the high suppression pool average temperature in this Condition, the monitoring Frequency is increased to twice i

that of Condition A.

Furthermore, the 30 minute Completion Time is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition.

E.1 and E.2 If suppression pool average temperature cannot be maintained l

at s 120*F, the plant must be brought to a MODE in which the

.l LCO does not apply. To achieve this status, the reactor i

pressure must be reduced to < 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and the plant must be brought to at least MODE 4 within

. u,

~

(continued) l PBAPS UNIT 2 B 3.6-51 Revision 0

t Control Rod Position B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Control Rod Position BASES BACKGROUND Control rods provide the capability to maintain the reactor suberitical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction _in the Reactor Manual Control System. During refueling, movement of control rods is limited by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2) or the control rod block with the reactor mode switch in the shutdown position (LCO 3.3.2.1).

UFSAR design criteria require that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions :(Ref.1).

The control rods serve as the system capable of maintaining the reactor subcritical in cold conditions.

The refueling interlocks allow a single control rod to be withdrawn at any time unless fuel is being loaded into the core. To preclude loading fuel assemblies into the core

7J with a control rod withdrawn, all control rods must be fully inserted. This prevents the reactor from achieving criticality during refueling operations.

APPLICABLE Prevention and mitigation of prompt reactivity excursions SAFETY ANALY3ES during refueling are provided by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the SDM (LCO 3.1.1), the **Ja my e n o b e

,,,eatter peried-san 1 Sterediete range rMter neutren flux-scram (LCO 3.3.1.1),

and the control rod block instrumentation (LCO 3.3.2.1).

The safety analysis fcr the control rod withdrawal error during refueling in the UFSAR (Ref. 2) assumes the functioning of the refueling interlocks and adequate SDM.

The analysis for the fuel assembly insertion error (Ref. 3) assumes all control rods are fully inserted. Thus, prior to fuel reload, all control rods must be fully inserted to minimize the probability of an inadvertent criticality.

Control rod position satisfies Criterion 3 of the NRC Policy Statement.

(continued) 2k u

PBAPS UNIT 2 B 3.9-8 Revision 0

i Control Rod Position Indication B 3.9.4 sem B 3.9 REFUELING OPERATIONS B 3.9.4 Control Rod Position Indication BASES BACKGROUND The full-in position indication for each control rod provides necessary information to the refueling interlocks to prevent inadvertent criticalities during refueling operations. During refueling, the refueling interlocks (LCO 3.9.1 and LCO 3.9.2) use the full-in position indication to limit the operation of the refueling equipment and the movement of the control rods. The absence of the full-in position indication signal for any control rod removes the all-rods-in permissive for the refueling equipment interlocks and prevents fuel loading. Also, this condition causes the refuel position one-rod-out interlock to not allow the withdrawal of any other control _ rod.

UFSAR design criteria require that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions (Ref.1).

The control rods serve as the system capable of maintaining 3

the reactor suberitical in cold conditions.

APPLICABLE Prevention and mitigation of prompt reactivity excursions SAFETY ANALYSES during refueling are provided by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the SDri (LCO 3.1.1), the, wise r-*Je

,,,d,...,,,,,, L, g e rie d-sl e r t % t e rme d i : t : r:ns: m: nit:r ~stren fis scram (LCO 3.3.1.1),

and the control rod block instrumentation (LCO 3.3.2.1).

The safety analysis for the control rod withdrawal error during refueling (Ref. 2) assumes the functioning of the refueling interlocks and adequate SDM. The analysis for the fuel assembly insertion error (Ref. 3) assumes all control rods are fully inserted. The full-in position indication is required to be OPERABLE so that the refueling interlocks can ensure that fuel cannot be loaded with any control rod withdrawn and that no more than one control rod can be withdrawn at a time.

Control rod position indication satisfies Criterion 3 of the NRC Policy Statement.

(continued) w PBAPS UllIT 2 B 3.9-10 Revision 0

i i

Control Rod OPERABILITY-Refualing

~"

B 3.9.5 B 3.9 REFUELING OPERATIONS B 3.9.5 Control Rod OPERABILITY-Refueling BASES BACKGROUND Control rods are components of the Control Rod Drive (CRD)

System, the primary reactivity control system for the reactor.

In conjunction vith the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes during refueling operation.

In addition, the control rods provide the capability to maintain the reactor subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System.

UFSAR design criteria require that one of the two required independent reactivity control systems be capable of holding the reactor core suberitical under cold conditions (Ref.1).

The CRD System is the system capable of maintaining the reactor suberitical in cold conditions.

APPLICABLE Prevention and mitigation of prompt reactivity excursions during refueling are provided by refueling interlocks /* m/e SAFETY ANALYSES (LCO 3.9.1 and LC0 3.9.2), the SDM (LCO 3.1.1), the wi nedfre mm.T" Pe" j,f,,.1 "ter-edhte rze writer rutron flux scram (LCO 3.3.1.1),

and the control rod block instrumentation (LCO 3.3.2.1).

The safety analyses for the control rod withdrawal' error during refueling (Ref. 2) and the fuel assembly insertion error (Ref. 3) evaluate the consequences of control red withdrawal during refueling and also fuel assembly insertion with a control rod withdrawn.

A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment. Control rod scram provides protection should a prompt reactivity excursion occur.

Control rod OPERABILITY during refueling satisfies Criterion 3 of the NRC Policy Statement.

LCO Each withdrawn control rod must be OPERABLE. The withdrawn control rod is considered OPERABLE if the scram accumulator pressure is 2 955 psig and the control rod is capable of (continued) aN PBAPS UNIT 2 B 3.9-14 Revision 0

p

~

Reactor Mede Switch Interlock Testing B 3.10.2

+

j!

B 3.10 SPECIAL OPERATIONS B 3.10.2 Reactor Mode Switch Interlock Testing BASES BACKGROUND The purpose of this Special Operations LCO is to permit operation of the reactor mode switch from one position to another to confirm certain aspects of associated interlocks during periodic tests and calibrations in MODES 3, 4, and 5.

The reactor mode switch is a conveniently located, multiposition, keylock switch provided to select the necessary scram functions for various plant conditions (Ref. 1). The reactor mode switch selects the appropriate trip relays for scram functions and provides appropriate bypasses. The mode switch positions and related scram interlock functions are summarized as follows:

Shutdown-Initiates a reactor scram; bypasses main a.

steam line isolation and main condenser low vacuum scrams; Refuel-Seiects Neutron Monitoring System (NMS) scram -

E b.

function for low neutron flux level operation (btrt ud r~je n e a, ~~.b5 ddec: not di::bic ee-average power range monitor.5e+de scram); bypasses main steam line isolation and main condenser low vacuum scrams; Startup/ Hot Standby-Selects HMS scram function for low c.

neutron flux level operation (46termediat; r:r.ge vule. mp re& n monitors and average power range monitors); bypasses main steam line isolation and main condenser low vacuum scrams; and d.

Run-Selects HMS scram function for power range operation.

The reactor mode switch also provides interlocks for such functions as control rod blocks, scram discharge volume trip bypass, refueling interlocks, and main steam isolation valve isol ations.

APPLICABLE The acceptance criterion for reactor mode switch interlock SAFETY ANALYSES testing is to prevent fuel failure by precluding reactivity excursions or core criticality. The interlock functions of (continuedl PBAPS UNIT 2 B 3.10-5 Revision 0 am

e

SDM Test-Refueling B 3.10.8

,m B 3.10 SPECIAL OPERATIONS B 3.10.8 SHUTDOWN MARGIN (SDM) Test-Refueling BASES BACKGROUND The purpose of this MODE 5 Special Operations LCO is to permit SDM testing to be performed for those plant configurations in which the reactor pressure vessel (RPV) head is either not in place or the head bolts are not fully tensioned.

LCO 3.1.1, " SHUTDOWN MARGIN (SDM)," requires that adequate SDM be demonstrated following fuel movements or control rod replacement within the RPV. The demonstration must be performed prior to or within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after criticality is reached. This SDM test may be performed prior to or during the first startup following the refueling.

Performing the SDM test prior to startup requires the test to be performed while in MODE 5, with the vessel head bolts less than fully tensioned (and possibly with the vessel head removed).

While in MODE 5, the reactor mode switch is required.to be in the shutdown or refuel position, where the applicable i

control rod blocks ensure that the reactor will not become critical. The SDM test requires the reactor mode switch to be in the startup/ hot standby position, since more than one control rod will be withdrawn for the purpose of demonstrating adequate SDM. This Special Operations LCO

)

provides the appropriate additional controls to allow withdrawing more than one control rod from a core cell containing one or more fuel assemblies when the reactor vessel head bolts are less than fully tensioned.

APPLICABLE Prevention and mitigation of unacceptable reactivity SAFETY ANALYSES excursions during control rod withdrawal, with the reactor mode switch in the startup/ hot standby position while in MODE 5, is provided by the int:rm: dict: r:ngo acnitor (I"")wth rup asuf,..

..,tr,r[wahr';a.dA ae"trer fh:x scram (LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation"), and control rod block instrumentation (LCO 3.3.2.1, " Control Rod Block Instrumentation"). The limiting reactivity excursion during startup conditions while in MODE 5 is the control rod drop accident (CRDA).

(continued)

=m 5#

PBAPS UNIT 2 B 3.10-31 Revision 0

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SDM Test--Refueling B 3.10.8 BASES APPLICABLE CRDA analyses assume that the reactor operator follows SAFETY ANALYSES prescribed withdrawal sequences.

For SDM tests performed (continued) within these defined sequences, the analyses of References 1 and 2 are applicable.

However, for some sequences developed fer the SDM testing, the control rod patterns assumed in the safety analyses of References 1 and 2 may not be met.

Therefore, special CRDA analyses, performed in accordance with an NRC approved methodology, are required to demonstrate the SDM test sequence will not result in l

unacceptable consequences should a CRDA occur during the testing.

For the purpose of this test, the protection provided by the normally required MODE 5 applicable LCOs, in addition to the requirements of this LCO, will maintain normal test operations as well as postulated accidents within the bounds of the appropriate safety analyses (Refs. I and 2).

In addition to the added requirements for the RWM =APRM, and control rod coupling, the notch out mode g

is specified for out of sequence withdrawals.

Requiring the notch out mode limits withdrawal steps to a single notch, v4Any which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur l

during the test.

As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of i

the NRC Policy Statement apply.

Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LC0 As described in LC0 3.0.7, compliance with this Special Operations LC0 is optional.

SDM tests may be' performed while in MODE 2, in accordance with Table 1.1-1, without meeting this Special Operations LCO or its ACTIONS.

For SDM tests performed while in MODE 5, additional requirements must be met to ensure that adequate protection against potential reactivity excursions is available.

To provide additional scram protection beyond the normally required VuN.JRE, the APRMs are also required to be OPERABLE (LC0 3.3.1.1, Functions 2a and 2e) as though the reactor were in MODE 2.

Because multiple control rods will be withdrawn and the reactor will potentially become critical, the approved control rod withdrawal sequence must'be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2), or.must be verified by a (continued)

PBAPS UNIT 2 B 3.10-32 Revision 0

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j ATTACHMENT 3 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 Facility Operating License Change Request 93-18

" Wide Range Neutron Monitoring Instrumentation' NEDO-32368 Nuclear Measurement Analysis and Control Wide Range Neutron Monitoring System Ucensing Report For Peach Bottom Atomic Power Station, Units 2 and 3 fr--

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GE Nuclear Energy y;

115 Curtner Avenue San Jose. CA 95125

['li NEDO-32368 DRF C51-00129/1 Class I November 1994 1

Nuclear Measurement Analysis and Control a.

Wide Range Neutron Monitoring System

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Licensing Report For Peach Bottom Atomic Power Station, Units 2 & 3

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e Prepared by:

K.T. Schaefer GE Nuclear Energy Reviewed by:

S. Chakraborty GE Nuclear Energys l

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Approved by:

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D. Demick, Projdtt Manager GE Nuclear Energy e

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NEDO-32368 c',

CLASS I

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t Id IMPORTANT NOTICE REGARDING 3

CONTENTS OF TIIIS REPORT

+-

Please Read Carefully The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between PECO Energy Company (PECO Energy) and GE, Plant Services Contrat effective August 1,1977, and the Contract for the NUMAC-WRNMS Program, effective May 4,1994, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract.

The use of this information by anyone other than PECO Energy, or for any purpose other than e

that for which it is intended, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the y

completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

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NEDO-32368 CLASSI IA TABLE OF CONTFETS Title P_ ate Executive Summary lii l.

General Introduction and Summary 1-1

~'

1.1 Summary 11 i

1.,2 Conclusion 1-2 j

===2.

System Description===

2-1 2.1 Introduction 2-1 2.2 Hardware Configuration 2-3 2.2.1 Generic vs. PBAPS Design Differences 2.3 System Operation 2-3 3

2.3.1 System Outputs Table 2-1 WRNMS Indication Ranges Table 2-2 WRNMS Outputs

~

Figure 2-1 Core Locations of Neutron Monitoring System Detectors Figure 2-2 Configuration of the WRNMS (One Channel) 3.

Design Considerations

, 3-1 3.1 Safety Classification 3-1 3.2 Instrument Power 3-1 3.2.1 WRNMS Control Room Chassis Power 3.3 Qualification 3-2 3.3.1 General Requirements 3.3.2 Sensor, Cable, and Connector Qualification 3.3.3 NUMAC Electronics Dynamic Qualification 3.3.4 Cable Qualification 3.4 Compliance With IEEE 279 3-6 3.5 System Application Considerations 3-8

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NEDO-32363 4

'CLASSI k;

1.

TABLE OF CONTENTS (Continued)

W Illk East a

4.-

Safety Concerns 4-1 4.1 RWE At Low Power 4-1

)

4.1.1 Identification of Causes and Frequency Classification l

4.1.2 Sequence of Events and Systems Operations j

4.1.2.1 Sequence of Events l

4.1.2.2 Identification of Operator Actions l

4.1.2.3 Effects of Single Failures and Operator Errors i

4.1.3 Core and System Performance 4.1.3.1 Analysis Method and Assumptions u

4.1.3.2 Analysis Conditions and Results 4.1.3.3 Evaluation Based on Criteria

-H 4.1.4 Barrier Performance 4.1.5 Radiological Consequences

[

4.2 Failure Evaluation 4-4 4.2.1 Single Failures

~

4.2.2 Software Common Cause Failures 4.2.3 EMI and RFI Concerns 4.3 Changes to Technical Specifications 4-6 4.4 Conclusion 4-6 l

^

Table 4-1 Comparison of Outputs Between SRM/IRM and WRNMS l

Table 4-2 Causes of Control Rod Withdrawal Error Table 4-3 Sequence of Events i

Table 4-4 Proposed Technical Specification Changes 5.

References 5-1 l

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NEDO-32368 C1. ASS 1

+

l1 EXECUTIVE

SUMMARY

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This report summarizes the analyses and evaluation performed that justify installing and a

operating a Nuclear Measurement Analysis and Control (NUMAC) Wide Range Neutron

=

Monitoring System (WRNMS)into Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3.

The WRNMS replaces the Source Range Monitor and Intermediate Range Monitor. The PBAPS WRNMS is effectively identical to the NRC generically approved system documented in licensing topical report (LTR) NEDO-31439-A (Reference 1).

The minor plant-specinc differences between the PBAPS and the generic system are evaluated in this report.

a This evaluation for the WRNMS has reviewed the predominant plant licensing challenges, and demonstrates that implementing the WRNMS at PBAPS can be accomplished without a significant increase in the probaMlity or consequences of an accident previously evaluated, without creating the possibility of a new or different kind of accident from any accident previously evaluated, and without exceeding any presently existing regulatory limits applicable to the plant, which might cause a reduction in a margin of safety. This modification involves no signi0 cant hazards consideration.

M M

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NEDO-32368

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Cl. ASS I ba 5

1.

GENERAL INTRODUCTION AND

SUMMARY

u The Wide Range Neutron Monitoring System (WRNMS) combines the long life capabilities of the regenerative in-core fission detector with the reliable characteristics of the Nuclear Measurement Analysis and Control (NUMAC) nuclear instrumentation to provide a major improvement in wide range neutron monitoring. The system uses fixed location in-core breeder detectors, locally mounted preamplifiers, and control room mounted signal conditioning electronics to continuously monitor neutron fluxes over the range of startup through full power, and provide an operator interface.

It replaces the heretofore used Source Range and Intermediate Range Monitor (SRM/IRM) starup and intermediate range neutron sensors, the detector retraction mechanisms and controls, and the accompanying monitoring electronics (see Appendix A of Reference 1).

1.1 Summary The WRNMS monitors approximately 11 decades of neutron flux over the startup, intermediate, and power ranges. It operates in the counting and Mean Square Voltage (MSV) modes using y

a in-core ion chamber having a fissionable and regenerative emitter to cover all of the ranges, and a microprocessor-based NUMAC-WRNMS control room monitor to calculate reactor parameteis. The NUMAC-WRNMS is capable of performing the following functions:

(1) Using the measurements from a breeder type neutron sensitive ion chamber to calculate reactor power and period over 11 decades.

Providing continuous monitoring of neutron flux from 10% to full power.

(2)

(3)

Providing power and period output data to remote meters and recorders, in analog form and over digital data link.

(4) Providing alarm, trip and control rod block signals.

(5) Semi-autoniatic self-calibration of control room monitor, including the preamplifier.

(6) Providing security against tampering (by password and keylock control).

1-1

NEDO-32368

[

CLASS I p

(7) Reducing human error through the use of a digital display, and a simple front panel design incorporating human factor guidelines.

re (8) Reducing possibility ofinoperable or inadvertent trip outputs by improved calibration and testing techniques.

(9) Supply both analog and digital outputs for remote meters and recorders.

(10) Providing operator interfaces and remote readouts.

Section 2 of this report contains a summary description of the PBAPS NUMAC-WRNMS, Section 3 discusses the design considerations and performance standards, and Section 4 discusses the safety concerns for the system.

1.2 Conclusion m

The PBAPS NUMAC-WRNMS meets all of the applicable regulatory requirements, and offers improved reliability and performance over prior wide range monitoring instrumentation. The

~

^

WRNMS provides (Class 1E qualified) neutron monitoring backup for post-accident applications.

Digital circuitry and microprocessor capabilities, together with the self-test system, improve the overall accuracy and availability of the control room instrumentation, which enhances nuclear power plant availability at:d safety.

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w tt 1-2

NEDO-32308 CLASSI t-W A

2.

SYSTEM DESCRIPTION

[

The Wide Range Neutron Monitoring System (WRNMS) uses a fixed in-core regenerative fission chamber with associated NUMAC electronics to provide neutron monitoring over the entire range of normal BWR operation and post-accident neutron monitoring from 10*% to 100% (11

~

decades) of full power.

2.1 Introduction Currently, PBAPS uses four source range monitors (SRMs) and eight intermediate range monitors (IRMs) to provide approximately 10 decades of startup range coverage. The detectors have to be inserted during shutdown and withdrawn after startup to preserve their limited life using non-Class IE retract mechanisms. In addition, range switches are used for the linear IRM output to switch between decades during power ascension (see Appendices A.1 and A.3 of 2

Reference 1). The present SRM/IRM system will be replaced with 8 NUMAC-WRNMS channels (2 trip systems with 4 channels / system), using fixed location in-core regenerative breeder detectors. The number and locations of the WRNM sensors have been analytically and

]

experimentally determined to provide suf6cient flux level information under the most limiting bypass and sensor failure conditions. Each quadrant of the reactor has provision for three sensor locations (previously the SRM and IRM locations). Any two locations out of the three in each of the quadrants for a total of eight sensors (refer to Figure 2-1) provide sufficient coverage.

The vertical portion will be the same as the IRMs in their fully inserted condition. The four SRM sensors would be eliminated from the system.

Use of regenerative sensors permits permanent in-core locations corresponding to present " full in" detector positions.

.m The WRNMS indication ranges versus what are normally termed as source and intermediate ranges are shown in Table 2-1.

The current SRMs are required to have 3 operable channels, and the current IRMs are required to have 3 operable channels per trip system. To provide adequate coverage, equivalent to the SRMs, the WRNMS in Range 0 needs to have 3 operable channels with each channel located in a separate quadrant of the core. To provide adequate coverage, equivalent to the IRMs, the WRNMS in Ranges 1-5 needs to have 3 operable channels per trip system. However, as the

~

WRNMS operates continuously over both source and intermediate ranges (providing monitoring and trip functions), WRNMS source range monitoring will usually have at least 6 operable 2-1

s'*

NEDO-32368 CLASSI k

channels. Therefore, the WRNMS sensor arrangement provides equal or better coverage than 12 the current SRMs and IRMs.

A single NUMAC-WRNMS channel consists of a fixed in-core (no retract mechanism required) regenerative sensor, dry tube, environmentally qualified cable, preamplifier, NUMAC control room monitor electronics which utilize counting, mean square voltage and constant current techniques to process the signal and provide controls and readouts for the operator. Protective trip functions are provided based on stable reactor period, which eliminates the need for range switches. By using a regenerative uranium mixture of U-234 and U-235, the wide range sensor can remain in-core and provide a useful life of approximately 7 full power years. Since the sensor is fixed in-core, the detector drive equipment is eliminated. Bottom entry type sensor a

assemblies will be used for ease of detector replacement.

A graphic representation of the neutron flux monitoring function is shown in Figure 2-2 of Reference 1. There is a near one-to-one relationship between true and indicated neutron flux d

from startup to full power. Figure 2-2 shows a block diagram of a single channel NUMAC-WRNMS.

a The sensor is housed in a dry tube in the vessel and connected by coaxial cable to a preamplifier which is located outside the primary containment. The preamplifier provides a signal output to the control room monitor which processes the flux signals.

~

Wide range neutron monitoring by the NUMAC-WRNMS consists of a combination of a source range and an intermediate range. The source range uses a count rate function that covers the same low neutron flux range and supplies the low count rate block as the present SRM subsystem (see Table 4-1). Since the signalis continuous between the SRM and IRM functions, the trips associated with SRM upscale and IRM downscale are not needed, and thus, are eliminated. All range switch based trips are eliminated, and instead, a period based trip is

^

introduced over the entire range. The intermediate range uses a Mean Square Voltage (MSV) technique that is consistent with the present IRM subsystem in providing the middle range of neutron flux monitoring. Outputs from the source range and intermediate range functions of the WRNMS interface with control room and remote equipment. The combined count rate /MSV 3

wide range function covers the source and intermediate neutron flux range of 1.0 x 10 nv to 33 l3 7.5 x 10 nv (Above 1.5 x 10 ny, constant de current mode operation is employed instead of the constant voltage method). In the constant de current mode the flux calculated by the MSV method at constant voltage is compensated for high voltage reduction caused by the current limit.) and provides continuous logarithmic and digital signal outputs with no discontinuity 2-2

NEDO-12.1.28 m.

CLASS I M

between the two functions over the complete range. Trips are combined in the new WRNMS

~

Trip Auxiliary Units to interface with the. Reactor Protection System (RPS) and the Reactor Manual Control System (RMCS).

a As shown in Reference 1, the WRNMS originally provided upscale (high) neutron flux rod block and scram functions. A generic Rod Withdrawal Error (RWE) analysis justified these functions.

However, a PBAPS specific RWE (see Section 4) demonstates that the short period rod block and scram functions provide adequate protection against flux excursions. Plus, the APRM Setdown scram (in non-Run mode) at PBAPS provides additional backup protection against flux

~

excursions. Therefore, the upscale neutron flux rod block and scram functions are not required, and thus are not included in the PBAPS WRNMS.

2.2 Hardware Configuration The components for the NUMAC-WRNMS include the sensor / integral cable assembly, dry tube, preamplifier and control room monitor with operator interface (s). Figure 2-3 of Reference 1 shows the mechanical configuration for the system. The general details of the system are described in Section 2.2 of Reference 1. Specific design differences between the generic design in Reference 1 and the PBAPS design are provided below.

~

2.2.1 Generic vs. PBAPS Design Differences Appendix C of Reference 1 describe; the configuration variation of the WRNMS. The eight channel replacement of the SRM and IRM subsystems is applicable to the WRNMS at PBAPS.

The range switches are eliminated, as lincar power indication is performed via auto-ranging.

Stable period trip is used instead of manual range switch related trips.

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2.3 System Operation Operation of the WRNMS at PBAPS will be consistent with the description provided in Section 2.4 of Reference 1.

2.3.1 System Outputs j

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The system outputs are ' :ed in Table 2-2.

2-3

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- NEDO-32360 CLASSI m

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Table 2-1 j

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[s WRNMS Indication Ranges w

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.ra Range WRNMS Range WRNMS Indication Range i

i Source 0

0.1 - 1E6 counts per second

,a 1

Intermediate 1

0 - 125E-5 % power 2

0 - 125E-4 % power 3

0 - 125E-3 % power i

ca 4

0 - 125E-2 % power

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5 0 - 125E-1 % power i,r)

I' Power 6

0 - 125E0 % power L

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NEDO-32368 CLASSI m

Table 2-2 WRNMS Outputs n

Description Signal Level Range Type a

Percent Power (Recorder) 0 - 1 Volt 104 to 100%

Analog log-scale Percent Power (Recorder) 0 - 1 Volt 0 - 12.5 %

Analog auto-ranged linear scale

-J Percent Power (Computer) 0 - 160 mv 0 to 12.5%

Analog Auto-ranged linear scale Reactor Period (Recorder) 0 - 1 Volt

-100 to inf to Analog non-linear s

+3 secs.

Reactor Period (Meter) 0 - 1 ma

-100 to inf to Analog non-linear

+3 secs.

e Reactor Period (Computer) 0 - 160 mv

-100 to inf to Analog non-linear

+3 secs.

Log Count Rate Hi Trip Variable Set point Trip Logic Log Count Rate Hi Hi Trip Variable Set point Trip Logic HI Reactor Period Trip Variable Set point Trip Logic Hi HI Reactor Period Trip Variable Set point Trip Logic Downscale Power Trip Variable Set point Trip Logic INOP Alarm (non-fatal) Trip Trip Logic INOP Trip (fatal) Trip Trip Logic Percent Power and Period Digital Serial RS 232 Range Indication BCD 0 - 10 Discrete a

Percent of Full Power Internal 104 to 100%

Graphic Bar Percent to Full Power Internal 10* to 150%

Alphanumeric Display Period Internal

-100 to inf to Graphie Bar

+3 secs Period Internal

-100 to inf to Alphanumeric Display

+3 secs.

Log Count Rate Internal 0.1 to 10' cps Alphanumeric Display Setpoints Internal As set Alphanumeric Display

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i Polarizing Voltage Internal 0 to 400V Alphanumeric Display i

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NEDO-32368 CLASS I W

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DETECTOR & CONTROL ELEMENT ARRANGEMENT (TOP VIEW OF CORE)

- CONTROL RODS (185)

A WRNM SPARE DRY TUBE LOCATIONS (4) OR ALT.WRNM DET. LOCATIONS FOR THAT GUADRANT

? LPRM DETECTOR ASM IN THE PWR. RANGE (43)

_ SPARE EMITTING SOURCE POSITIONS (7) d EMITTING SOURCE POSITIONS (7) FOR INITIAL STARTUP 2 WRNM DETECTOR LOCATIONS (8)

O LPRM DETsIN TRIP SYS *B'

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LPRM DETs IN TRIP SYS *A" Figure 2-1. Core Location Of Neutron Monitoring System Detectors

a'i NEDO-32368 CLASSI WC M

um RPS RMCS PROCESS OPERATOR SYSTEM SYSTEM COMPUTER PANEL S4A a;

- ANNUNCIATORS METERS

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-RECDRCERS RANCE trolCATORS REGENERATNE g

SENSOfL (8)TO y

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PANEL S44 i

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DRY TUDE (12) NEW i

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AUX TRIP y

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PEDESTALCONNECTION SCX ll MOOF'ED 2JAATCHING TRANSFORMERS Q

(IF REQUIRED)

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TEST CABLE (75 OHM)

N HV CABLE DEDICATED CONDUlf (PEDESTAL TO PRE

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l SIGNAL CABLE F5 OHM) l TE OHM Mt SloMAL CARLE (UNDER VESSEL TO PEDESTAL)

NOTE

1) TESTM V /L.V/ SIGNAL CABLES BETWEEN CONTROL ROOM AND PREAMPLIFIER MAY BE REUSED FROM SRM/lRM SYSTEMS

2) SIGNAL CABLE F5 OHM) TO BE INSTALLED

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M?N FROM UNDER VESSEL TO PREAMPLIFIER a

Figure 2-2. Configuration of the WRNMS (One Channel)

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NEDO 32368 CLASS I C.

3.

DESIGN CONSIDERATIONS -

The PBAPS WRNMS design criteria, requirements and qualification are consistent with those described in Section 3 of Reference 1. The one exception is that at PBAPS the WRNMS is not j

required to meet Regulatory Guide 1.97, since this requirement is met by other equipment.

However, the PBAPS WRNMS does provide an alternate (redundant) method to perform the Regulatory Guide 1.97 neutron monitoring function.

3.1 Safety Classification The WRNMS is designed to perform the same functions as the SRM/IRM subsystems, while eliminating the insert-retract hardware and drive control electronics. No SRM function is safety-related. However, the IRM high neutron flux scram is a safety-related function. Therefore, the WRNMS (in part) is safety-related, and is qualified as a Class IE system. To ensure maximum d

system availability, a Self-Test System (located in the control room monitor) periodically tests the system electronics, and if a failure of a safety-related function is detected (fatal inop.), the e

system trip.

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3.2 Instrument Power.

a The WRNMS is designed to operate off of existing plant power supplies used for the SRM/IRM 24 Vdc power. This power is backed up by standby power sources.

3.2.1 WRNMS Control Room Chassis Power Two types ofinternal power supplies are provided within the NUMAC-WRNMS control room monitor; (1) a pair of low voltage power supplies for preamp'ifier and chassis modules; and (2) a detector high voltage power supply. Power to operate the control room chassis is as follows:

i u

Voltage 19.5 to i29 1

Current 1 Amp s

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3-1 i

k

m NEDO-32368 CLASS 1 3.3 Qualification j

u The qualification of the WRNMS has been reviewed against the environmental and seismic qualification requirements for PBAPS, and the qualification of the WRNMS bounds the PBAPS requirements. WRNMS qualification is described below.

a 3.3.1 General Requirements The WRNMS has been qualified to the applicable requirements of:

L NRC Regulatory Guide 1.89 (IEEE 323-1974)

NRC Regulatory Guide 1.100 (IEEE 344-1975)

NRC Regulatory Guide 1.97 m

The qualification was established using qualification methods set forth in General Electric Environmental Qualification Program, Document NEDE 24326-1-P (Reference 2). The generic

x WRNMS qualification program is provided in Section 4 of Reference 1.

1 m

Control Room eauioment l

NUMAC equipment is qualified to the environment shown below:

i i

a.

Temperature:

5 to 50 C b.

Humidity:

10 to 90% RH c.

Atmospheric Pressure:

Standard Atmospheric Pressure, +/- 1.0 Inch of Water d.

Radiation:

0.0005 Rad /Hr,1000 Rad (TID) l e.

Seismic:

Response spectra at instrument location, both horizontal and vertical,3% damping Fault (SSE)

Upset (OBE) 1 Hz 0.1 g 0.1 g 4 Hz 10.8 g 7.2 g 19 Hz 10.8 g 7.2 g 30 Hz 6.0 g 4.0 g j

100 Hz 6.0.g 4.0 g

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emus 3-2

NEDO-32368 CLASS I m

This envelops the Peach Bottom specific requirements as follows:

Temperature 114*F max Humidity 27 to 50 % RH ai Pressure None Radiation None a

Seismic 1.92 g,0.5% damping max C#J Preamotifier

~

Detector Preamplifier is located in the reactor building in harsher environment and will be

.a qualified by type testing to perform its safety-related functions in the environment shown below:

Temperature Normal 65 to 103*F, Transient temperature 149*F steady state,207

F peak for 10 minutes) a Humidity Normal 20 to 90 % RH, Transient 100%

~

Pressure

-0.25 IN WG, Transient 14.7/16.7 psia Radiation 6.49 E4 Rad TID Dry Tube. Inteeral cable and Incore detector These items is are located in the reactor vessel in harsher environment and the type test qualification envelops the Peach Bottom environment shown below:

Temperature:

Normal 50 to 545 F, Transient temperature 575*F steady state Pressure 0 - 1050 psig, Transient 1250 psig Neutron Flux 1.8 E14 ny, Transient 3.4 E14 nv max Gamma Flux 1.68 E8 R/hr max Under Vessel cables and Connectors The under vessel cables and connectors are located in the containment in harsh environment.

The standard type tested qualification will envelop the Peach Bottom environment shown below:

Normal 65 to 150 F, Transient temperature 340 F for a short period Temperature:

Humidity Normal 20 to 90 % RH, Transient 100%

3-3

1 NEDO-32363 CLASS I

.~

Pressure

-0.5 - 2.0 psig, Transient 58.8 psig Radiation 4.69 E7 Rad TID k,

J Individual component qualification outlines follow.

Sensor, Cable, and Connector Qualification 3.3.2 Qualification to meet the requirements of NUREG 0588 and IEEE 323-1974 was done by a combination of type testing and analysis.

a.

Type Testing 1)

Sensor Seismic 2)

Environmental Testing of Connector Thermal Age

~

Radiation Age Seismic Post-LOCA Environmental (100 days) b.

Analysis 1)

Sensor and Integral Cable 2)

Materials 3)

Construction 4)

Similarity to Other Sensors 3.3.3 NUMAC Electronics Dynamic Qualification Operational Test Under Dynamic Coi,J.tions Type Testing a.

3-4

NEDO-32368 CLASSI m

1)

Fixture sa The interface fixture was subject to a sine sweep survey along each of the three orthogonal axes to the following levels:

4 1-100 Hz 0.2 Peak Sweep Rate One Octave / Minute a

The fixture was found to be free of resonance at any given frequency of 100 Hz or below. A resonance is defined as an amplification of response acceleration over input acceleration by a factor of four.

n 2)

Fixture / Hardware w

The hardware under test, mounted rigidly to its interface fixture, was subject to a sine sweep survey along each of its three orthogonal axes to the levels above. Identify and record product resonances exceeding an amplificadon factor of four to one.

b.

Seismic Exposure (Upset Condition) a Five random vibration bi-axial seismic events at the upset levels were performed.

The test was performed twice, one in each horizontal axis, with the interface fixture

~

rotated 90 degrees in the horizontal plane between tests.

Seismic Exposure (Faulted Condition) c.

Perform one random vibration bi-axial seismic event at the faulted levels in

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accordance with the requirements ofIEEE 344-1975, Section 2.2.3.b, except that the test duration is twenty seconds minimum. The test is performed twice, once in each horizontal axis, with the interface fixture rotated 90 degrees in the horizontal plane between tests.

3.3.4 Cable Qualification

~

Cable provided as part of this project for use inside the primary containment or reactor building has been conditioned and tested to demonstrate that it will function during a Loss of Coolant 3-5

~

NEDO-32368 CLASSI v

u Accident (LOCA) postulated to occur at any time during 40 years of operation under conditions as prescribed by IEEE-383-1974 The cables were subjected to the LOCA pro 61e ofIEEE-323-1974 Appendix A for combined PWR/BWR while energized with rated voltage.

a The cable has also been tested in accordance with IEEE-383-1974, Section 2.5.6. Samples were subjected to and passed the vertical flame test as described in Part 6 of ICEA S-19-81.

3.4 Compliance With IEEE 279 IEEE 279-1968 is the licensing basis for PBAPS. However, for conservatism the following

~

assessment is based on the 1971 version of IEEE 279.

The WRNMS replaces the IRM subsystem. All functions performed by the IRM subsystem are performed by the WRNMS.

1.

The WRNMS meets the same IEEE-279 bypass and inoperable status indication requirements as the IRM subsystem.

2 2.

The WRNMS meets Paragraph 4.1 by providing automatic initiation of scram and rod block trips.

/

med 3.

The neutron monitoring system WRNM trips comply with the (Paragraph 4.2) single failure criterion through the use of physics! panel barriers and electrical isoiation provisions to provide independence among the : aur redundant WRNM channels in each trip

' ~ '

system.

4.

The quali6 cation and design of the WRNMS trip meet the quality and reliability requirements of Paragraph 4.3.

5.

Consistent with Paragraph 4.4, all required equipment is either type tested or analyzed, in conjunction with Geld experience of the components, to demonstrate that the equipment will meet performance requirements. "In-situ" operational testing of the sensor channels and protection system will be performed during preoperational testing.

6.

The components are qualified to operate (as needed) under abnormal conditions of environment, energy supply and accidents, and thus, maintain channelintegrity and meet the requirements of Paragraph 4.5.

3-6

t NEDO-32368 CLASSI 7.

Paragraph 4.6 requires redundant and physically separated RPS trip channels. The eight WRNM channels are electrically isolated and physically separated from one another so as to comply with this requirement, except immediately under the reactor vessel where complete physical separation is not practical.

+-

3 8.

The channels for the WRNMS-RPS trip variables are electrically isolated from plant control systems in compliance with Paragraph 4.7. Within the WRNM modules, (i.e.,

prior to their output trip unit driving the RPS), analog outputs are derived for use with control room meters, recorders and the process computer. Electrical isolation has been incorporated into the design at this interface to prevent any single failure from influencing the protective output from the trip unit. The trip unit outputs are physically separated and M5 electrically isolated from other plant equipment in their routing to the RPS panels.

9.

The WRNMS directly measures neutron flux to determine and protect against (short period z

RPS trip) a reactor over-power condition b compliance with Paragraph 4.8.

10. The WRNMS is not required to supply a RPS trip when the reactor is in the "run" mode.

~

Therefore, the requirements of Paragraph 4.9 are not applicable. However, the WRNMS would normally be in operation, and would respond to local neutron flux and provide the operator with indication.

11. The WRNMS-RPS trip has provisions for sensor test and calibration during reactor operation in compliance with Paragraph 4.10.

12. With respect to Paragraph 4.11, there are sufficient WRNM channels to permit any one channel in a given trip system to be manually bypassed and still ensure that the remaining operable channels comply with the IEEE 279 single failure design requirements. One WRNM manual bypass switch is provided for each RPS trip system. The characteristics of the switch permit only one of the four WRNM channels of the trip system to be bypassed at any time. With any channel bypassed in a given trip system, three channels remain operable to satisfy the protection system requirements.

13. With respect to Paragraph 4.12. an operating bypass is provided for the WRNM system trip, when the reactor is in the "run" mode.

14. With respect to Paragraph 4.13, when any WRNM instrument channel output to the RPS is bypassed, it is indicated by a light in the main control room.

t..

3-7

n NEDO-32368

~..

Cl. ASS 1 c

15. With respect to Paragraph 4.14, manual bypassing of any WRNM channel is accomplished

~

with control room selector switches.

16. With respect to Paragraph 4.15, the analyticallimits for the WRNMS setpoint are provided

~

in Table 4-1.

C

17. With respect to Paragraph 4.16, the WRNMS output is connected to the RPS trip relays

~

which drive the scram actuators.

l

18. Paragraph 4.17 is not directly applicable to the WRNMS.

19. With respect to Paragraph 4.18, access to setpoint adjustments, calibration controls and test

~

points for the WRNMS-RPS variables is under administrative control.

20. With respect to Paragraph 4.19, when any one of the redundant sensors exceeds its setpoint value for a WRNMS-RPS trip variable, a control room annunciator is initiated to identify the particular variable.

21. The WRNMS-RPS system trip data presented to the operator complies with Paragraph 4.20.

a

22. With respect to Paragraph 4.21, during reactor operation the control room operator will be able to determine failed WRNMS sensors, but subsequent repair can orily be accomplished during reactor shutdown.

3.5 System Application Considerations Appendix G of Reference 1 lists a number of (SRM/IRM and Regulatory Guide 1.97 related) plant-specific systern application considerations to be confirmed. Some of the considerations are applicable for using the WRNMS to meet Regulatory Guide 1.97. As the WRNMS is not required to meet Regulatory Guide 1.97 at PBAPS, these considerations need not be addressed.

Each of the SRM/IRM considerations is addressed below.

l.

The Rod Withdrawal Error event was reanalyzed (see Section 4.1), and specific period based trip setpoints are based on the analytical limit shown in Table 4-1.

3-8

e-NEDO-32368 CLASSI 2.

The PBAPS specific location of detectors and cables are based on Reference 1, and summary descriptions are provided in Section 2.

m r?

3., 4., 6., & 7.

Not applicable to PBAPS,

^

The WRNMS system meets the existing IRM subsystem separation requirements. The 5.

separation between safety-related and nonsafety-related circuits is via coil :o contact

~

separation in relays. The WRNMS is not designed or required to meet Regulatory Guide 1.75.

8.

The plant specific Technical Specification changes are provided in NUREG-1433 format and content, which are different from the original BWR/5 type standard Technical Specification changes provided in Reference 1.

However, the technical content of the PBAPS Technical Specification changes meet the intent of the proposed Technical Specification changes from Reference 1.

1, 9.

Plant procedures including emergency procedures will be modified, as needed, during the implementation of the WRNMS.

4

-9 1

3-9

/

NEDO-32368 Ct ASS I t

y 4.

SAFETY CONCERNS M

Of all of the plant safety analyses only a rod withdrawal error (RWE) event at low power is potentially affected by the implementation of a WRNMS at PBAPS, No other safety analysis takes credit for a source or intermediate range neutron scram function. For example, in the w

feedwater controller failure (FWCF) event (re-analyzed as part of the Peach Bottom Power Rerate program) the scram function assumed in the analysis resulted from a high water level In the trip, and no credit was taken for possible trips from the neutron monitoring system.

original design the IRM scram function would mitigate a RWE. Table 4-1 shows that the WRNMS will continue to perform required scram functions and all other (as needed) SRM and

-'i IRM type functions. A plant specific low power RWE event was analyzed and is summarized below.

y As the plant-specific RWE analysis only takes credit for the short period trips, no PBAPS safety analysis takes credit for the WRNMS upscale level functions, and these functions are not needed

^

to meet any design or regulatory criterion. Therefore, the WRNMS upscale level functions are not included in the PBAPS design (see Table 4-1).

a 4.1 RWE At Low Power (During Reactor Startup) 3 4.1.1 Identification of Causes and Frequency Classification

~~

lt is postulated that during a reactor startup, a single control rod is inadvertently withdrawn

~

continuously due to a procedural error by the operator and operator failure to acknowledge continuous alarm annunciations prior to safety system actuation. The probability of in'itial causes or errors of this event alone is considered low enough to be categorized as an infrequent incident. The probability of further development of this event is low, because it is contingent upon the failure of the Rod Worth Minimizer (RWM) system, together with a high worth out-of-

~

scquence rod selection contrary to procedures.

2 The WRNMS has period-based functions that will stop a continuous rod withdrawal by initiating a rod block if the flux excursion, caused by rod withdrawal, generates a period shorter than 20 seconds. The period-based functions also initiate a scram trip if the flux excursion generates a period shorter than 10 seconds. Any single WRNMS rod block trip initiates a rod block. One out or iwo taken twice logic is used to initiate a scram. A detailed description of the period-based trip function is presented in Reference 1.

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NEDO-32368

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Ct. ASS I For this transient to happen, a large reactivity addition must be introduced. The reactor must be critical, with control rod density greater than 50%. The causes of the event are summarized in Table 4-2. The probability for this event to occur is considered low enough to warrant its les being categorized as an infrequent incident, w.

4.1.2 Sequence of Events and Systems Operation 4.1.2.1 Sequence of Events Control rod withdrawal errors are not considered credible in the startup aad low power ranges.

~

The RWM system prevents the operator from selecting and withdrawing an out-of-sequence' control rod. However, the sequence of events of a postulated continuous control rod withdrawal error during reactor startup is shown in Table 4-3.

4.1.2.2 Identification of Operator Actions No operator actions are required to preclude or terminate this event, since the plant design, as discussed above prevents the event's occurrence, and WRNMS period-based trip functions will initiate and terminate this event.

4.1.2.3 Effect of Single Failures and Operator Errors Any additional single failure or operator error will not prevent the event mitigation fun'ctions (e.g., rod blocks and scrams) from automatically occurring prior to any design or safety limit being exceeded.

4.1.3 Core and System Performance 4.1.3.1 Analysis Method and Assumptions

~

The analysis uses the reactivity insertion analysis code described in Reference 3. It is a two-dimensional adiabatic code assuming no heat transfer to the coolant. The analysis consists of four steps. In Step 1, with the error rod being continuously withdrawn from full in, the model is used to calculate the average power and period change as a function of time with a continuous reactivity insertion simulating the RWE. In Step 2, the power versus time data are used as input to a calculation of the WRNMS rod block and scram trip times. Both the rod block trip and scram trip times are then determined. In Step 3, the reactivity insertion input to the adiabatic 42

- -. ~.

NEDO-32368 -

CLASS I d

model is adjusted such that after the period reaches the rod block setpoint (20 sec), there is no

.further reactivity insertion. The RWE transient is then recalculated by the model with the a

adjusted reactivity input. The reactor scram time is also adjusted based on the time determined in Step 2. The calculated fuel enthalpy does not consider local peaking effect. In Step 4, the 2

peak fuel enthalpy that includes the local peaking effect is calculated.

Other assumptions used in the analysis are:

(1) The standard BWR data of the adiabatic modelis used.

3 (2) The scram reactivity shape is derived from the design core, assuming no failing rods and same scram speed for all rods.

w (3) Six delayed neutron groups are assumed.

4.1.3.2 Analysis Conditions and Results y

(1)

Analysis Conditions (a) The reactor is assumed to be in the critical condition before the control rod with<i.wal, with an initial power of 0.1 % rated, and a core average temperature of

~

286*F.

(b) The worth of the withdrawn rod is 2.5% Ak from full-in to full-out.

(c) The control rod withdrawal speed is 3.11 in/s, the nominal withdrawal speed.

(2)

Analysis Result With this 2.5%6k reactivity insertion, the flux excursion generates a period of approximately 0.5 second. The rod block trip is initiated at 7 seconds after the start of the transient. The scram is initiated at about 8 seconds. The event is terminated by the scram.

The peak fuel enthalpy reached is approximately 19.9 cal /g, which is 2.9 cal /g higher than the initial fuel enthalpy.

-m e

j, NEDO-32358 CLASSI C.

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~4.1.3.3 Evaluation Based On Criteria Due to the effective protection function of the period-based trip function, the fuel enthalpy increase is small. The criterion of 170 cal /gm for fuel enthalpy increase under RWE event is y

satisfied, it is concluded that a RWE at low power is not a limiting event, and thus, need not

~

be reanalyzed for future fuel reloads.

Q 4.1.4 Barrier Performance

9d An evaluation of the barrier performance is not made for this event, because there is no fuel damage in this event and only with mild change in gross core characteristics.

4.1.5 Radiological Consequences u

An evaluation of the radiological consequences is not required for this event, because no i

.j radioactive material is released from the fuel.

4.2 Failure Evaluation

-i The following failure evaluation examines (1) potential single failures and determine their effects f

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on plant safety, and (2) design and process control features to prevent software common cause failures.

4.2.1 Single Failures The WRNh1S meets all the redundancy requirements, and can accommodate potential single

~

failures as the SRhi and IRh1 subsystems that the WRNh1S is replacing. Therefore, compliance to single failure criteria at PBAPS does not change with the implementation of the WRNMS, and plant safety will not be adversely affected.

'~

4.2.2 Software Common Cause Failures The design and process control features that prevent software common cause failures are described in Section E.2 of Reference 1.

~

Potential errors caused by human-machine interface (Hhil) are prevented by the design features described in Section 3 of Reference 1, The system performs reactor trip functions only during 4-4

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NEDO-32368 CLASSI m

startup operation. When the reactor is in the Run mode, the trips are bypassed. There are several channels which are compared against each other for consistency of data. The system a

itself protects against a RWE, by providing rod blocks. This function backs up the (nonsafety-related) RWM system.

i 4.2.3 EMI and RFI Concerns a

Numerous tests, in accordance with electro-magnetic interference (EMI) and radio frequency interference (RFI) related industry standards, have been performed to demonstrate protection against EMI/RFI related disturbances. Complete immunity to EMI/RFI is dependent upon the equipment installation methods (conduits and grounding, etc.). The NUMAC-WRNMS has been

~

installed in a number of plants. No adverse affect due to EMI events has been encountered.

An EMI/RFI mapping data were collected for the Peach Bottom site in compliance with MIL-Std 461D and NUREG/CR-5941. A comparison was made between the EMI susceptibility characteristics of the NUMAC WRNMS and the EMI emission environment of the plant, as reported in the site survey data.

Conducted Susceptibility When the conducted susceptibility and worst case emission spectra for the plant are compared, the susceptibility spectra envelope the emission spectra with a margin of at least 18 dB. Therefore, the WRNMS will not be affected by conducted noise present at the Peach Bottom site.

~

Rndiated Susceptibility (Magnetic Fields)

When the radiated magnetic field susceptibility and worst case emission spectra for the plant are compared, the susceptibility spectra envelope the emission spectra with a margin of at least 52 dB. Therefore, the WRNMS will not be affected by radiated magnetic fields present at the Peach Bottom site.

Rndiated Susceptibility (Electric Fields)

When the radiated electric field susceptibility and worst case emission spectra for the plant are compared with the WRNM's susceptibility of at least 10 V/m over the entire frequency range, then a margin of 12 dB exists at the emis, ion peak located at 450 MHz, and a

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NEDO 32368 CLASS I margin of at least 52 dB elsewhere. Therefore, the WRNMS will not be affected by radiated electric fields present at the Peach Bottom site.

In addition, the system employs fail safe logic for RPS trips. The EMI/PFI disturbances generally cause' signal spikes (transients), which in turn would produce a reactor period trip, I

causing a reactor scram. Thus, the EMI related disturbances do not prevent the sytem from performing its safety-related function.

M 4.3 Changes To Technical Specifications p

The changes to the Technical Specifications (TS) are provided in the NUREG-1433 format. The TS changes are applicable to both Units 2 and 3. Except for the plant-specific licensing basis l

differences, the proposed PBAPS TS are technically the same as those provided in Appendix F of Reference 1. Per the criteria for what items should be incorporated into NUREG-1433 based l

~~

TS, Table 4-4 presents the proposed TS changes needed to implement the WRNMS at PBAPS.

4.4 Conclusion

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The WRNMS performs all the necessary functions of the original SRMs and IRMs, and the RWE at low power remains non-limiting. Therefore, the WRNMS will continue to ensure that PBAPS remains within its licensed safety basis.

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Table 4-1 td Comparison of Outputs Between SRM/lRM and WRNMS (The following table shows the outputs provided by the original SRM l

7

',lJ and IRM systems compared to those provided by the WRNMS.)

SRM/IRM Outputs (nominal values)

. WRNMS Outputs SRM IRM (analytical limits)

Downscale

Alarm Downscale

Rod Block Downscale

Alarm and

~

(3 cps)

Rod Block (2.5/125 of scale) *

(1 cps)

Rod Block INOP

Scram and INOP #1

Scram and Rod Block (fatal)

Rod Block INOP #2

Alarm (Non fatal)

'~~

9 Retract

Rod Not needed

~

Permissive Block l

(100 cps) f High Level

Rod Flux High

Rod Block 7

(1 05 cps)

Block Level for Fuel l

5 (3x10 cps)-

Loading j

High Level

Rod Block Not needed for Startup m

(108/125 (Replaced by period trip) of scale)

High High

Red Flux High

Scram for Level Light High Fuel l

(5 x 10' cps)

(1x10' cps)

Loading

.n High High

Scram Not needed for Startup Level (Replaced by period trip) t (120/125 of scale)

Period High

Alarm Selected

Rod Block Period Based : Alarm and (10 sec)

Range Low Trip High Rod Block (20 sec)

Selected

Scram Period Based : Scram Range High Trip High High (10 see) e i

These output:; eliminated RC Time Constant = 40 4-7 p_

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NEDO-32368 CLASSI ra Table 4-2 95 Causes of Control Rod Withdrawal Error G

Operator Selects Out-of-sequence Control Rod and initiates Rod Withdrawal m

V RWM Rod Block Fails V

Operator Procedural Errors --

"5 Alarms Are ignored and Control Rod is Continuously Withdrawn m

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Hpe Table 4-3 l

Sequence of Events 1

Time (sec)

Events The reactor is critical and operating in the startup range.

e

> 0 The operator selects and withdraws an out-of sequence control rod at the maximum normal drive speed of 3.11 in/sec.

The RWM fail to block the selection (selection error) and continuous withdrawal (withdraw error) of the out-of-sequence rod.

l Neutron flux increases rapidly (due to the continuous reactivity addition) with i

a very short period.

a 7

The WRNMS Period-Based Rod Block Trip initiates rod block due to short j

period (less than the 20 seconds).

i 8

The WRNMS Period-Based Scram Trip initiates reactor scram due to short period (less than the 10 seconds).

l 9

Reactor is scrammed and event is terminated.

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Table 4-4

[s Proposed Technical Specification Changes (Based on NUREG-1433)

Z

.Eagg Section General Descriotion of Chance i

1.1-2 1.1 Change " source range monitor" to " range neutron monitors,"

and delete " intermediate range raonitors."

3.3-4 SR 3.3.1..l.5 &

Change text and test frequency to 31 days to reflect WRNMS, SR 3.3.1.1.6 overuil reliability from fixed incore detectors, high re!! ability components, and self-monitoring (self test feature) feature.

o 3.35 SR 3.3.1.1.11 & 12 Reverse the order of these SRs to be consistent with improved Tech Specs format. Change channel calibration to reflect the

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24 months frequency applicable to the WRNM.

[

3.3-7 Table 3.3.1.1-1 Change Function I reference from IRM to WRNM.

Item 1.a.

Change to period-short trip, delete SR's that are no longer applicable due to the m,plementation of the WRNM, re-number SR's as applicable to the implementation of the f

WRNM, and supply allowable values.

l

\\n item 1.b.

Re-number SR's as applicable to the implementation of the WRNM.

/'m 2.a., 2.b. & 2.e.

Delete SR that is no longer applicable due to the j

implementation of the WRNM, and re-number SR's as applicable to the implementation of the WRNM 3.3-10 & 11 Spec. 3.3.1.2 Change title and text from SRM to WRNM, including LCO.

1 Modify ACTIONS A, A.1, B, D & E to reflect WRNM l

instead of IRM and SRM.

3.3-12 SR 3.3.1.2.2 Change title and text from SRM to WRNM.

3.3-13 SR 3.3.1.2.4 Change title and text from SRM to WRNM.

i 4

SR 3.3.1.2.5 Delete this SR, as it is no longer relevant. Re-number i

following SRs accordingly here and on page 3.3-14.

^

1 (now) SR 3.3.1.2.5 Change note to indicate the WRNMS value.

(now) SR 3.3.1.2.6 Change note to indicate the WRNMS value.

Change SR frequency to reflect WRNMS.

-6 4-10

NEDO 32368 r

CLASS I Table 4-4 (continued)

~

bre Sntion General Description of Chance 3.3-14 Table 3.3.1.2-1 h1odify the titles, function and notes applicable to the WRNh1 instead of the SRh1 instrumentation.

L 3.3-15 Figure 3.3.1.21 Change SRh1 to WRNht.

~

3.6-23 Spec. 3.6.2.1 Change all 4 (3 in LCO and I in ACTION A.) it.dicated

~

values to reflect WRNht.

3.6-24 CONDITIONS, Change indicated values in B.1 and C. to reflect WRNht.

ACTIONS Change intermediate range monitor to WRNht.

B3.2-3 B3.2.1 APPLICABILITY B3.2-8 P 3.2.2 Change reference from IRh1 to wide range neutron

~

APPLICABILITY monitor period-short function.

J B3.3-5 & 6 B 3.3.1.1 in item 1.a, change the titles, and add new text to reflect Neutron Period-Short design change from IRh1 to the WRNh1S, describe the new period measurement and trip functions, and replace Ref. 2 and j

3 with just Ref. 3. Replace paragraph addressing " limiting other reactivity excursions" with the equivalent p'tragraph based on NUREG-1433, and changing IRh1 to WRNht.

B3.3-6 & 7 B 3.3.1.1 Item 1.b. change title and description to refleet design change WRNht-Inop.

from IRh1 to the WRNh1S.

e B3.3-7 B 3.3.1.1 Item 2.a. description changed to retleet design change from APRh1 Flux-High IRh1 to the WRNhtS.

B3.3-10 &

B 3.3.1.1 Change IRN1 to WRNh1 in items 2.b and 2.e.

~

11 B3.311 &

B 3.3.1.1 Description change in item 2.d. to retlect design change rom

~.

12 APRh1 Downscale IRh1 to the WRNh1S.

B3.3-26 B 3.3.1.1 Change IRh1 to WRNht.

ACTION C.!

B3.3-29 B 3.3.1.1 Delete reference to IRN1 functions.

SR 3.3.1.1.3 B3.3-30 &

B 3.3.1.1 Replace descriptive text to retleet WRNN1.

31 SR 1.3.1.1.5 &

Add SRs basis for WRNh1 functional tests.

SR 3.3.1.1.6 4-11 1

r NEDO 32360 CLASS I

~

Table 4-4 (continued) l' axe Section Genern! Description of Chan.;g B3.3-32 &

B 3.3.1 1 Change IRM to WRNM Modify text to refleet WRNMs 24

~

33 SR 3.3. '.l.11 &l2 month SR frequency, based on setpoint analysis.

a 83.3 34 B 3.3.1 1 Replace Reference 3 with the licensing report on the REFER'!NCES WRNMS.

B3.3-36 B 3.3.1.2 Titles and description change to retleet design change BACKGROUND from SRM to the WRNM.

~~

B3.3-36 &

B 3.3.1.2 Description change to reflect design change from 37 APPLICABLL SRM/IRM to the WRNMS.

.~.

SAFETY ANALYSIS B3.3-37 &

B3.3.1.2 Description change to reflect design change from y

38 LCO SRMllRM to the WRNMS, and change text to address location requirements for detectors.

B3.3-38.

B 3.3.1.2 Description change to retleet design change from 39 & 40 APPLICABILITY, SRM/lRM to the WRNMS, and update associated ACTIONS.

ACTIONS A.1, B.I.

C.1, D.1, D.2, E.1 & E.2 9

B3.3-40, B3.3.1.2 Description change to reflect design change from SRM 41.42,43 SURVEILLANCE to the WRNMS, provide the WRNM indicated value for

& 44 REQUIREMENTS the bottom of intermediate range, delete reference to SR 3.3.1.2.6 and re-number following SR's accordingly, and' revise channel calibration frequency to 24 months based on WRNMS setpoint analysis.

B3.6-49.

B3.6.2.1 Modify LCO and ACTIONS B.1 and C.1 to refleet the 50 & 51 applicable power range that is indicated by the WRNMS.

B3.9-8 B 3.9.3 Description change to reflect design change from IRM APPLICABLE to the WRNMS.

SAFETY ANALYSES B3.9-10 B 3.9.4 Description change to refleet design change from IRM APPLICABLE to the WRNMS.

SAFETY ANALYSES B3.9-14 B 3.9.5 Description change to refleet design change trom IRM APPLICABLE to the WRNMS.

SAFETY ANALYSES 4 12

e NEDO-32368 r

CLASS I 4+

Table 4-4 (continued) m hwx Section General Description of Chance B3.10-5 B 3.10.2 In items b. and c. change text to reflect the WRNMS BACKGROUND design, and clarify APRM and WRNMS functions.

w B3.10-31 & B 3.10.8 Description change to reflect design change from IRM

~

32 APPLICABLE to the WRNMS.

SAFETY ANALYSES 4-.

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5.

REFERENCES l.

GE Nuclear Energy, Licensing Topic Report, "The Nuclear Measurement Analysis and Control Wide Range Neutron Monitoring System (NUMAC-WRNMS)," NEDO-31439-

. A, Class I (non-proprietary), October 1990.-

x w

2.

General Electric Co., " General Electric Environmental Qualification Program," NEDE-24362-1-P, Revision 1, Class III (proprietary), January 1983.

3.

C. J. Paone and J. A. Woolley, " Rod Drop Accident Analysis for Large Boiling Water Reactors," GE Licensing Topical Report, NEDO-10527, and Supplements 1 and 2, March 1972, July 1972 and January 1973.

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ML20078A053

Contents

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ML20078A053

Text

SUBJECT:

Dear Sir:

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SUMMARY

SUMMARY

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