ML20077S619

License Change Request 94-07 to Licenses DPR-44 & DPR-56, Revising Allowed Time Between Signal Calibrs of LPRM from Every 6 Wks (Approx 1,000 EFPH) to Every 2,000 Megawatt Days Per Std Ton

ML20077S619
Person / Time
Site:	Peach Bottom
Issue date:	01/13/1995
From:	Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20077S621	List: ML20077S619 ML20077S627
References
NUDOCS 9501240180
Download: ML20077S619 (9)

Text

-

Cttlen Support Department

.i - A 10 CFR'50.90

. =y PECO ENERGY rica,T,st;;L,,

965 Chesterbrook Boulevard Wayne, PA 19087-5691 January 13,1995 Docket Nos. 50-277 50-278 Ucense 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 Facility Operating Ucense Change Request 94-07

Dear Sir:

PECO Energy Company hereby submits Ucense Change Request (LCR) 94-07, in accordance with 10 CFR 50.90, requesting changes to Appendix A of the Peach Bottom Atomic Power Station (PBAPS) Facility Operating Licenses.

The proposed changes concern a revision to the frequency of calibration for the Local Power Range Monitor (LPRM) signals from every 6 weeks to every 2000 Megawatt Days per Standard Ton (MWD /ST). This change is similar to the request made by Plant Hatch as part of its improved Technical Specifications submittal.

Attachment 1 to this letter describes the proposed changes and Attachment 2  ?

contains the revised pages.

If you have any questions concerning this submittal, please contact us.

Sincerely,

. h. * .

G. A. Hunger, J .

Director - Ucensing

Enclosures:

Affidavit, Attachments cc: T. T. Martin, Administrator, Region I, USNRC a g

W. L Schmidt, Senior Resident inspector, PBAPS, USNRC R. R. Janati, Commonwealth of Pennsylvania

( iI 95o124o18o 95o113 PDR ADDCK oSooo277 P PDR

6

. t COMMONWEALTH OF PENNSYLVANIA  : i

SS.

COUNTY OF CHESTER  :

W. H. Smith,111, 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 94-07) for changes to the Peach Bottom Facility Operating Licenses DPR-44 and DPR-56, and knows the contents thereof: and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

1 J ,

Vice Presiden/t l

l Subscribed and sworn to before me this l3 day of gy 1995.

8J&a

" No}ary,Public N l

%kb % D

,.jstu @:Ch

s ATTACHMENT 1 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 Ucense Nos. DPR-44 DPR-56 Facility Operating Ucense Change Request 94-07 l

"LPRM Signal Calibration Frequency" l 1

1 Supporting Information for Changes I 4

I l

PECO Energy Company (PECO Energy), Licensee under Facility Operating Licenses .

DPR-44 and DPR-56 for the Peach Bottom Atomic Power Station (PBAPS), Units 2 and  :

.s 3 respectively, requests that the Technical Specifications (TS) contained in Appendix A  ;

to the Operating Licenses be amended. Handwritten proposed changes to pages 44  :

and 54 of the Technical Specifications are contained in Attachment 2.

The proposed changes revise the allowed time between signal calibrations of the Local  !

Power Range Monitor (LPRM) from every 6 weeks (approximately 1000 Effective Full l Power Hours (EFPH)] to every 2000 Megawatt Days per Standard Ton (MWD /ST).

Ucensee proposes that the changes be effective upon issuance of the amendment l approving the requested change. ,

b Descriotion of Chanoes Licensee proposes the following changes:

i

1) . Table 4.1.2, "ReacL' Protection System (SCRAM) instrument Calibration, [

Minimum Casibration Frequencies For Reactor Protection Instrument Channels,"  :

Page 44 i

Revise the " Minimum Frequency (2)" column for the LPRM Signal Instrument l Channel to read: t "Every 2000 MWD /ST."

2) Bases Section 4.1, page 54 l Revise last paragraph to read:

"The sensitivity of LPRM detectors decreases with exposure to neutron  :

flux at a slow and approximately constant rate. This is compensated for ,

in the APRM system by calibrating twice a week using heat balance data  :

and by calibrating individual LPRM's every 2000 MWD /ST, using TIP  ;

traverse data. This calibration frequency is dependent on the use of 3D-  :

MONICORE as the core monitoring program." .

Justification of Chanoes l 1

BWR power oper? tion relies upon readings from fixod in-core neutron detectors known

- as Local Power Range Monitors (LPRM). LFRMs are small fission chambers with an  !

approximately linear response to the local neutron flux, and thus local thermal power.

The current surveillance Requirement to calibrate the LPRMs every 6 weeks employs a ' .

1-  !

i

, , - .r, - , , - . , , , s -,. ---_.-- ---- ----. - -- - - --- - - -

l second set of moveable detectors known as the Traversing in-core Probe (TIP) system. The required LPRM calibration relates the power distribution, measured by the TIP system, to the then existing LPRM readings. When the LPRMs are normalized -!

to one another, to the TIP readings, and to a plant heat balance calculation, these l LPRMs allow determination of the local power for each (approximately) 6 inch fueled i region of the core (node). j Outputs from the calibrated LPRMs are used in the Reactor Protection System i

- (Average Power Range Monitor (APRM)), the Rod Worth Minimizer, the Rod Block l Monitor (RBM), as well as daily surveillance of the Power Distribution Limits (reactor - j thermal limits monitoring). Accuracy requirements on the power distribution are j defined by GESTAR-il (NEDE-24011-P-A-10, Section 4.3.1.1.1) and GE Fuel Bundle  !

Designs (NEDE-31152P), which are part of the PBAPS licensing basis. In particular, i Table 3-3 o' NEDE-31152P requires calculated nodal powers to have a root mean square (rms) uncertainty of no more than 8.7% for reload cores. The attending Table l 3-3 comment states that this uncertainty also applies to the power distribution as  !

determined by the LPRM system. Thus, the accuracy in power distribution is  ;

determined by the LPRM system between TIP sets and must also meet the 8.7% rms  :

uncertainty.

4 Advances in process computer monitoring include the development of new mathematical techniques and algorithms combining reactor physics theory with on'-line j core data, (e.g., LPRM readings). One such methodology is 3D-MONICORE which is i currently in use at PBAPS. The 3D-MONICORE employs an adaptive learning i algorithm using on-line as well as historical core data inputs to improve power calculations within the reactor physics model by effectively modifying the neutron i leakage terms (adaptive coefficients) to force the calculated power distribution to match the measured power distribution as determined by the TIP system. An adequate number of LPRM calibrations is performed upon startup and initial operation to establish a base set of adaptive coefficients. Subsequent calculations use the adaptive coeffici; *s and LPRM readings during monitoring between LPRM i calibrations. Thi 49 of adaptive coefficients is updated at each LPRM calibration throughout the eph.

Corrections made within the monitoring process account for decay of LPRM sensitivity f due to depletion of the fissile coating within each LPRM. The 6 week calibration i interval at PBAPS was based upon older monitoring methodology and older LPRM ,

designs in use at the time, i Part of the study discussed below investigated the sensitivity decay characteristics of l the present generation of LPRM detectors in use at Plant Hatch. Calibration current  !

and expmure data demonstrate a significant reduction in the uncertainty associated  !

with LPRM sensitivity as a function of exposure. This reduced uncertainty, combined  !

with improved monitoring methods, allows lengthening of the LPRM calibration interval  ;

as discussed below.

1 2-r,+ r m 2 , e

t rnis reactor physics mmhodology was used off-line to model four continuous mor*.s  ;

of rated operation at Plant Hatch during a recent cycle to study the effect of lengthening the LPRM calibration interval. This period included a control rod pattern j sequer:m exchange. Actual plant LPRM readings were modified using TIP set (OD1) ,

calibration currents to factor out the effects of the LPRM calibrations. The operational periods of interest were then re-depleted using the modified LPRM readings (i.e. i without calibrations). Comparisons of 2D bundle and 3D nodal power distributions ,

(with calibrations versus without) were made and percerd rms deviations were '

calculated for each exposure point of interest. Results of this analysis show that the licensing basis nodal power uncertainty of 8.7% was satisfied for up to approximately_ 3 3000 EFPH between LPRM calibrations. Additionally, peak power limits (core thermal limits) for the bum omitting LPRM calibrations were either in agreement with or were conservative with the power distribution which included all LPRM calibrations.

Therefore, an increase in the frequency for LPRM calibrations from 1000 EFPH to 2000 EFPH is justified.

Plant Hatch provided the above justification to the NRC as part their improved Technical Specifications submittal. Because of the similarities between the two  !

facilities, the information g!eaned from the testing performed at Plant Hatch is  ;

applicable to PBAPS. The preceding justification to extend the LPRM calibrations from 1000 to 2000 EFPH at Plant Hatch is consistent with PECO Energy's request to extend i the interval from 6 weeks to 2000 MWD /ST at PBAPS.

The following calculations were used to convert 2000 EFPH to 2000 MWD /ST. These l units are, for this application, effectively interchangeable based on the combination of  ;

two facts: j t

1) Straight Conversion: Using current core weights and rated powers, ,

converting 2000 EFPH to units of MWD /ST results in:

i 24 firs / Day, .1 97 Y .

and, 2411 8/ Day ,14 00 T.

Using projected core weights and re-rated power, a range of future MWD /ST values can be obtained. The projected core weight for an equilibrium full core of GE-11 is 142.9 ST; for GE-12 (the heaviest bundle currently projected for possible use), the core weight would be 151.8 ST. Therefore, the 2000 EFPH equivalences in MWD /ST are:

. GE-11 Equilibrium Full Core:

'2000EFPH y 3458 MW '

= 2016. 6 NWD/ST 24 Hrs / Day, ,14 2 . 9 ST and, GE-12 Equilibrium Full Core 2000 EFPH y ' 3458 MW '

= 1898. 3 NWD/ST 24 Hrs / Day, ,151. 8 ST .

s.

The GE-11 bundle is the current reload fuel design at PBAPS. Thus, for the 2000 EFPH equivalence, the expected range is bete;;;a 1854.2 MWD /ST and  ;

2016.6 MWD /ST. Values become larger as cores become lighter and re-rate power is implemented. This entir6 range is equivalent to only approximately 170 EFPH, or less than 10% of the original 2000 EFPH. .

2) Nodal Power Uncertainty Sensitivity to EFPH: The Hatch study results ,

indicate that, between approximately 2000 and 3000 EFPH, the nodal power '

uncertainty only increases by 0.1%. This increases the total TIP readings uncertainty by less than 0.1%, which will not violate the 8.7% rms uncertainty limits specified in NEDE-31152P. -

Thus, while it is recognized that the actual MWD /S slue corresponding to 2000 EFPH is a function of core weight and rated core power (and is therefore cycle  :

dependent), the selection of an even 2000 MWD /ST for implementation is reasonable and does not cause a violation of total TIP Readings uncertainty.

The proposed change in units from EFPH to MWD /ST is consir. tent with NUREG 1433, ,

" Standard Technical Specifications, General Electric Plants, BWR/4," and the proposed  :

PBAPS Improved Standard Technical Specifications, which was submitted to the NRC '

by letter dated September 29,1994. In this submittal, PECO Energy proposed an interval of 1000 MWD /T.  ;

' No Significant Harards Consideration Licensee proposes that this application does not involve significant hazards consideration for the following reasons i) The oronosed chanos does not involve a sionificant increase in the orobability  ;

or consecuences of an accident oreviously evaluated.

y

' This change does not affect tt'e operation of any equipment. The change does not affect the fundamental method by which the LPRMs are calibrated. The increased time between required LPRM calibrations does r ot affect either the 4

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

ll initiator of any accident previously evaluated or any equipment required to -

mitigete the consequences of an accident, or the isotopic inventory in the fuel.

. Thus, the change does not increase either the probability or the radiological  !

consequences of an accident previously evaluated.

The orooosed chanons do not creda the oossibliity of a new or different kind of  ;

i) t accident from any oreviousiv evaluated.

l The proposed change does not introduce a new mode of plant operation and 4 does not involve the installation of any new equipment or modifications to the l plant. Therefore, it does not create the possibility of a new or different kind of ~

accident from any accident previously evaluated.

iii) _Tbg orooosed chances do not involve a significant reduction in a margin of safety.

The GETAB determination of the Maximum Critical Power Ratio (MCPR) Safety Limit allows a maximum total nodal uncertainty of the TIP readings (of which the  !

LPRM Update uncertainty is a part) of 8.7%. The change in LPRM calibration frequency results in an LPRM Update unwrtainty of 4.2% nodal power. This, ,

combined with the other uncertainties which comprise the total TIP readings {'

uncertainty, yields a total TIP readings uncertainty of less than the allowed ,

8.7%. Thus the change in LPRM calibration frequency will not affect the MCPR Safety Limit. i l

The LPRMs are utilized as input to the APRM and RBM systems. The primary safety function of the APRM system is to initiate a scram during core-wide neutron flux transients before the actual core-wide neutron flux level exceeds the safety analysis design basis. This prevents fuel damage from single )

operator errors or equipment malfunctions. The APRMs are calibrated at least ,

twice per week to the plant heat balance, utilize a radially and axially diverse l group of LPRMs as input and are utilized to detect changes in average, not  !

local, power changes. Therefore, the effects of decreasing the LPRM calibration frequency on the APRM system responses will be minimal due to any individual LPRM drift being practically canceled out (due to diversity of input) and/or due to the frequent recalibration of the APRMs to an independent power ,

calculation (the heat balance). Thus, decreasing the LPRM calibration  !

frequency will not significantly impact the performance of the APRM system's scram function, and there is no impact on transient delta-CPRs. j 1

The RBM system is utilized in the mitigation of a Rod Wdhdrawal Error (RWE). l The RBM system is designed to prevent the operator from increasing the local I power significantly when withdrawing a control rod. On each selection of a i 1

F i -4, w w ---- > - - . --, _ -- , , - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . ___ _ _ _ _ _ _ _ _ _ _ _ _ _

.- . . . 4

~

control rod, the average of the assigned, unbypassed LPRMs is adjusted to equal a 'iOO% re*arence signal for each of the two RBM channels. Each RBM .

channel automatically hmits the loca! thermal margin changes by limiting the - '

allowable change in local average neutron flux to the RBM setpoint. If the local average neutron flux change is greater than that allowed by the RBM setpoint, within either RBM channel, the rod withdrawa! permissive is removed preventing further movement. Since the change in local neutron flux is calculated from the change in the average of the LPRM readings, and calibrated on every rod selection to the reference signal, offsets in individual LPRM readings due to l cenbration differences are effectively eliminated for a given RBM setpoint.

Therefore, the constraints on the withdrawal of any given rod are unchanged and there will not be any increase in RWE delta-CPR. '

i Since the MCPR Safety Limit is unaffected and the delta-CPR values are i unchanged, the cycle CPR limits are unchanged. Therefore, the change in the frequency of LPRM calibration does not result in a reduction in a margin of safety. ,

Environmentalimoact Assessment  ;

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 the extension of LPRM calibrations from every 6 weeks to every 2000 MWD /ST. The Application involves no significant change in the types or significant increase in the amounts of any effluent that may be released .

offsite, and there will be no significant increase in individual or cumulative occupational ,

radiation exposure.

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.

1 I

1

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