ML20041B126

Application for Amend of License DPR-44,requesting Various Changes for Fifth Refueling

ML20041B126
Person / Time
Site:	Peach Bottom
Issue date:	02/18/1982
From:	Bauer E, Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20041B124	List: ML20041B123 ML20041B126 ML20041B128 ML20041B130
References
NUDOCS 8202230298
Download: ML20041B126 (34)

Text

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BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of i

Docket No.

50-277 PHILADELPHIA ELECTRIC COMPANY APPLICATION FOR AMENDMENT N

OF FACILITY OPERATING LICENSE DPR-44 q

d Edward C.

Bauer, Jr.

Eugene J.

Bradley 2301 Market Street i

Philadelphia, Pennsylvania 19101 Attorneys for l

Philadelphia Electric Company l

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1 jjj2230298820219 p

ADCCK 05000g77 PDR

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BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION-In'the Matter of Docket No.

50-277 PHILADELPHIA ELECTRIC COMPANY APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSE DPR-44 Philadelphia Electric Company, Licensee under. Facility-Operating License DPR-44 for Peach Bottom Atomic Power Station Unit No.

2, hereby requests that the Technical. Specifications incorporated in Appendix A of the Operating License bc. amended lyr revising certain sections as indicated by a vertical har in the margin of attached pages iv, vi, 9,

13, 14, 103, 104, 133a, 133b, 133c, 140, 140a, 140b, 140c, 140d, 142, 142a, 142b, 142f, 142g,.

142h and by adding pages iv(a), 133d, 133e, 142i and 142j.. Page s 140e and 142c have been deleted.

The changes to the Technical Specifications are being requested to:

(1) accommodate the fifth refueling cf the Peach Bottom Unit ~2 reactor, (2) identify the operating limits for all

e ruel~ types;for cycle 6 operation, ( 3) permit continued operation of two previously irradiated fuel assemblies after each is re-constittted with six (6) new fuel rods, (4) accommodate continued use of developmental channels, (5) incorporate the 67B (designation for a BWR 4, 1967 product line) control rod drive t

scrum times,-(6) accommodate implementation of ODYN transient analysis code into reload licensing, (7) delete certain information in the Technical Speci fication Bases which is redundant to information contained in other previously submitted documents, (8) permit extended exposure of Lead Test Assemblies (LTA's) and 8x8R (Reload 3) fuel, (9) incorporate MAPLHGR limits for the Reload 5 fuel and extended exposure MAPLHCR 1imits fo r the Reload 1 (LTA's), Reload 3, and Reload 4 fuel, (10) add generic MAPLHGR curve for P8x8R fuel to reduce need for feiture cycle-dependent revisions, and (11) modify references given in the Basest regarding reload safety evaluations.

An analysis of the safety considerations involved in the reactor refueling and the cycle 6 operating limits for all fuel' types is set forth in a document entitled " Supplemental Reload Licensing Submittal for Peach Bottom Atomic Power Station Unit 2 Reload 5" (Y1003J01A34) December, 1981, which is filed herewith and incorporated herein by reference.

In addition, changes requested herein would permi t the re-constitution and continued operation of two previous.

Irradiated standard 8x8R fuel assemblies which were initially inserted 'into the Peach Bottom Unit-2 reactor as part of the Reload 4 new fuel loading.

The purpose of continued re-constituted fuel bundle.. operation is to obtain additional fission gas pressure measurements from the standard 8x8R fuel in conjunction with the General Electric extended fuel exposure test program.

These fuel assemblies will be re-constituted during the Peach Bottom Unit 2, Reload 5 refueling outage by replacing a total of twelve (12) irradiated fuel rods (six per fuel assembly) with new fuel rods prior to bundle re-insertion for Cycle 6 operation.

The twelve removed irradiated fuel rods will be subjected to fission gas pressure measurement puncture tests.

The estimated bundle average exposures at the Spring, 1982, refueling outage for the two bundles that will be reconsti tuted are approximately 20 GWD/MT and 12 GWD/MT respectively.

An additional cycle of operation will extend their fuel bundle average exposure to approximately 29 GWD/MT and 21 GWD/MT respectively, with a peak pellet exposure approaching 39 GWD/MT and 31 GWD/MT respectively.

An analysis of the safety considerations involved in continuing the use of the re-constituted fuel is set forth in Appendix B of the referenced document Y1003J01A34.

The changen requested herein would permit the continued operation of 12 developmental fuel channels which were initially i -_

Installed during Roload 1.

The purpoco of continued operation is

' to.obtain channel-deflection data at higher exposures - (EOC exposure not to' exceed 40 GWD/MT) for use in conjunction with channel lifetime management.

An' analysis of the safety considerations involved in continuing-the_use of the developmental channels is set forth in Appendix E of the referenced document Y1003J01A34.-

The amendment request' proposes that the 67B control rod j

i r

drive (CRD) scram times be incorporated into the Te? nical h

e

(

. Speci fications.

The 67B CRD scram. times,- in replacang the 67A l

CRD scram times, require a 3.5 second average scram insertion j

i time, rather than 5.0 second average scram insertion time E

permitted for the 67A CRD, for the 90% inserted position from the fully withdrawn position.

The proposed change is in conformance

-with-the reload-unique transient analysis inputs utilized in the referenced document Y1003J01A34.

l i

i The amendment request provides for the implementation of the ODYN. transient analysis code for~ analysis of' rapid i

pressurization and over pressure protection events as set' forth in the referenced document Y1003J01A34.

The MCPR's calculated'

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for the pressurization events analyzed with ODYN have been n

adjusted in accordance with _ Option B of the ODYN implementation plan.' In the event that the Option B scram time specification is

- not-met, a linear interpolation between the Option A MCPR and the 4

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s Option B MCPR will be performed.

Both the Option A and Option B cycle-MCPR values and implementation procedure, including the ODYN Option B MCPR Scram Insertion Time Conformance-Procedure, are incorporated herein as described in the letter dated September 5,

1980, R. H[ Buchholz (General Electric) to P.

S.

Check (U.S. NRC), " Response to NRC Request for Information on ODYU Computer Model."

The proposed changes requested on pages 13, 14 140a, and deletion of page 140e would delete certain portions of the Bases of the Technical Specifications that contain information which is redundant to information contained in General Electric documents NEDE-24011-P-A, " General Electric Boiling Water Reactor Generic Reload Fuel Application" and NEDO 24081, " Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2."

The proposed changes to the Bases eliminate the necessity of revising these sections of the Technical Speci fications for each reload.

The changes requested herein would permit the extended operation of two Lead Test Assemblies (LTA's) and 210, Reload 3, 8x8R fuel assemblies.

The LTA extended exposure program is necessary to obtain information for a program which is directed toward improving uranium utilization.

An analysis of the safety considerations involved with LTA extended exposure is set forth 2n. Appendix C of the referenced document Y1003J01A34.

The 210, Reload 3, 8x8R fuel is being re-used as part of the Reload 5 design-Joading.

An analysis of the safety considerations involved with the Reload 3 fuel extended exposure is set forth in Appendix D of.the. referenced document Y1003J01A34.

The proposed changes incorporate MAPLHGR limits for the Reload 5 fuel and extended exposure MAPLHGR limits for the Reload 1'(LTA's), Reload 3 (8x8R), and Reload 4 (P8x8) fuel.. An analysis of the safety considerations involved in these MAPLHCR limits is set forth in a document entitled " Loss of Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2," Errata and Addenda Number 7 to NEDO-24081, which are filed herewith'and incorporated herein by reference.

The proposed generic P8x8R MAPLHGR curve addition on page 142j is added for the purpose of reducing-the need for.

future cycle-dependent revisions to.the Technical Specifications.

This curve has been constructed to bound the Reload 4 and Reload 5'P8x8R fuel.

Since'the proposed changes to the Technical Specifications do not involve a significant hazards consideration, pursuant to 10CFR Section 170.22, Philadelphia Electric Company, for fee purposes, proposes that the Application for Amendment be considered a Class III Amendment.

The Plant Operation Review Committee and the Operation and -Safety Review Committee have reviewed -these proposed changes to the Technical Specificiations, and have concluded'that they do t

not involve an.unreviewed safety question or a significant hazard consideration, and will.not endanger the health and safety of the public.

Respectfully submitted, PHILADELPHIA ELEC IC COMPANY

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By:

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COMMONWEALTH OF PENNSYLVANIA : -

ss.

COUNTY OF PHILADELPHIA s

S.'L.-Daltroff, _being first duly sworn, deposes and says:-

l That he is Vice President. of Philadelphia Electric Company, the Applicant herein; that he has~ read the foregoing Application for Amendment of Facility Operating License and knows the 1

contents.thereof; and that the statement's and matters set forth -

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therein'are true and correct to the best of-his knowledge, f

r information and belief.

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l Subscribed and sworn to i

this Y day before me of F N **V-b^~

, u bX:l 'j"

1. slAZ otary Public r

Ntary Puttle. Thi' del,nMa, Philadelphia Co.

Us D>mminoq Espires J y 28. MS3.

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3 CERTIFICATE OF SERVICE I certify that service of the. foregoing Application was made upon the Board of' Supervisors, Peach' Bottom Township, York County, Pennsylvania, by mailing a copy thereof, via first-class mail, to Albert R.

Steele, Chairman of the Board of Supervisors, R.D.

No.

l',

Delta, Pennsylvania 17314; upon the Board of Supervisors, Fulton Township, Lancaster County, Pennsylvania, by mailing a copy thereof, via first-class mail, to George K.

Brinton, Chairman of the Board of Supervisors, Peach Bottom, Pennsylvania 17563; and upon the Board of Supervisors, Drumore Township, Lancaster County, Pennsylvania, by mailing a copy thereof, via first-class mail, to Wilmer P. Bolton, Chairman of the Board of Supervisors, R.D.

No.

1, Holtwood, Pennsylvania 17532; all this 19th day of February 1982.

I r

ugege J.

Bradley Attorney for Philadelphia Electric Company

PBAPS

' Unit.2 LIST OF FIGURES Figure Title Page 1.1-1 APRM Flow Bias Scram Relationship To 16 Normal Operating Conditions 4.1.1 -

Instrument Test Interval Determination 55 Curves 4.2.2 Probability of? System Unavailability 98 Vs. Test Interval 3.4.1 Required' Volume and Concentration of 122 Standby Liquid Control System Solution 3.4.2 Required Temperature vs. Concentration 123 for Standby Liquid Control System Solution 3.5.K.1 MCPR Operating Limit vs. Tau 8x8R/LTA 142 3.5.K.2 MCPR Operating Limit vs. Tau,P8X8R fuel 142a 5

3.5.K.3 MCPR Operating Limit vs. Tau,P8DRB285 Fuel 142b 3.5.1.E Kf Factor Vs. Core Flow 142d 3.5.1.P MAPLHCR Vs.. Planar Average Exposure, 142e Unit 2, 8x8 LTA Fuel, 100 mil channels I

3.5.1.G.MAPLHCR Vs. Planar Average Exposure, 142f

-Unit 2, 8x8R Fuel, Type 8DRB284, 100 mil channels 3.5.1.H MAPLHGR Vs. Planar Average Exposure, 142g Unit 2, P 8X8R Fuel, Type P8DRB285, 100 mil channels 3.5.1.I MAPLHGR vs. Planar Average Exposure, 142h Unit 2, P 8x8R Fuel, Type P8DR8284 H, 80 mil & 100' mil channel & 120 mil channels 3.5.1.J.-MAPLHGR vs. Planar Average Exposure 142i Unit 2, P8X8R Fuel, Type P8DRB299, 100 mil channels 3.5.1.K. MAPLilGR vs. Planar Avorage Exposure 142j Unit 2, P8X8R Fuel (Generic)

-iv-1

PBAPS Unit 2 1

'a-LIST OF FIGURES H

Figure Title Page 3.6.1 Minimum Temperature for Pressure Tests 164

-such as required by Section XI 3.6.2 Minimum Temperature for Mechanical Heatup 164a or Cooldown following Nuclear Shutdown 3.6.3 Minimum Temperature for Core Operation 164b l

(Criticality) 3.6.4 Transition Temperature Shift vs. Fluence 164c 6.2-1 Management' Organization Chart 244 6.2-2 Organization for _ Conduct of Plant Operation 245' i

- iva -

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• s-PBAPS Unit 2 LIST OF TABLES it1e Page T

' Table 4.2.B Minimum Test and Calibration Frequency 81

for CSCS 4.2.C Minimum Test and Calibration Frequency 83

'for Control Rod Blocks Actuation 4.2.D Minimum Test and Calibration Frequency 84 for Radiation Monitoring Systems 4.2.E Minimum Test and Calibration Frequency 85 for Drywell Leak Detection 4.2.F Minimum Test and Calibration Frequency 86 for Surveillance Instrumentation 4.2.G Minimum Test and Calibration Frequency 88 I

for Recirculation Pump Trip 3.5.K.2 Operating Limit MCPR Values for 133d lI Various Core Exposures 3.5.K.3 Operating Limit MCPR Values for 133e c

i Various Core Exposures 4.6.1 In-Service Inspection Program for Peach 150 Bottom Units 2 and 3 3.7.1 Primary Containment Isolation Valves 179 3.7.2 Testable Penetrations With Double 184 0-Ring Seal s 3.7.3 Testable Penetrations with Testable 184 Bellows 3.7.4

' Primary Containment Testable Isolation 185 Valves 4.8.1 Radioactive Liquid Waste Sampling 210 and Analysis

f 4.8.2 Radioactive Gaseous Waste Sampling 211 i

and Analysis I

3.11.D.1 Safety Related Shock'Suppressors 234d 3.14.C.1 Fire Detectors 240k

-vi-l

PBAPS Unit 2 SAFETY LIMIT LIMITING SAFETY SYSTEA SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability:

Applicability:

p The Safety Limits established The Limiting Safety System Settings to preserve the fuel cladding apply to trip settings of the instru-integrity apply to those ments and devices which are provided

[

variables which monitor the to prevent the fuel cladding integrity fuel thermal behavior.

Safety Limits from being exceeded.

Objectives:

Objectives:

The objective of the Safety The objective of the Limiting Safety Limits is to establish limits System Settings is to define the. level which assure the integrity of of the process variables at which auto- !

the fuel cladding.

matic protective action is initiated l

to prevent the* fuel cladding integrity i

Safety Limits from being exceeded.

I Speci fication :

Specification:

A.

Reactor pressure 2800 psia The limiting safety system settings l

and Core Flow 210% of Rated shall be as specified below:

A.

Neutron Flux Scram The existence of a minimum

1. APRM Flux Scram Trip Setting cri ti cal power ratio MCPR (Run Mode) less than 1.07 for two recirculation loop operation, When the Mode Switch is.in the a

or 1.08 for single loop RUN position, the APRM flux operation, sha)) constitute scram trip setting shall be :

violation of the fuel clad-u ding integrity safety limit.

S < 0.66W + 54%-0.66 A W To ensure that this safety where:

Jimit is not exceeded, neutron flux shall not be above the S = Setting in percent of rated scram setting established in thermal power (3293 MWt) speci fication 2.1. A for longer than 1.15 seconds as indicated W = Loop recirculating

.by the process computer. When flow rate in percent the process computer is out of of design W is 100 for core service this safety limit shall flow of 102.5 million 1b/hr be assumed to be exceeded if or greater.

the neutron flux exceeds its scram setting and a contrcl rod scram does not occur.,

PBAPS Unit 2

1.1 BASES

FUEL CLADDING INTEGRITY A. Fuel Cladding Integrity Limit at Reactor Pressure > 800 psia and Core Flow >10% of Rated The fuel cladding integrity safety limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Sir.ce the parameters which result in fuel damage are not directly observable during reactor operation the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur.

Although it is recognized that a departure from nucleate boiling would not i

necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been aLopted as a convenient limit.

However, the t

uncertainties in monitoring the core operating state and in I

the procedure used to calculate the critical power result in

(

an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncer tainties.

The Safety Limit MCPR is determined using the General Electric Therma] Analysis Basis described in references 1 and

3. -_

'o-PBAPS Unit 2 1.1.A BASES (Cont'd)

B.

Core Thermal Power Limit (Reactor Pressure < 800 psia on Core Flow < 10% of Hated)

The ur,e of the GEXL correlation is not valid for the critical power calculations at pressures below 800 psia or core flows less than 10% of rated.

Therefore, the fuel cladding integrity safety limit is established by other means.

This is done by establishing a limiting condition of core thermal power operation with the following basis.

Since the pressure drop in the bypass region is essentially all elevation head which is 4.56 psi the core pressure drop at low power and all flows will always be greater than 4.56 3

psi.

Analyses show that with a flow of 28 x 10 lbs/hr

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bundle flow, bundle pressure drop in nearly independent of

[

bundle power and has a value of 3.5 psi.

Thus, the bundle flow with a 4.56 psi driving head will be greater than 28 x

)

10 3 lbsihr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern.

Full scale atlas test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.

With the design peaking factors this corresponds to a core thermal power of more than 50%.

Therefore, a core thermal power limit of 25% for reactor pressures below 800 psia or core flow less than 10%

.is conservative.

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C.

Power Transient Plant t fety analyses have shown that the scrams caused by I

exceeding any safety setting will assure that the Safety l

Limi t of Specification 1.1. A or 1.1.B will not be exceeded.

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Scram times are checked periodically to assure the insertion E

times are adequate.

The thermal power transient resulting l

when a scram is accomplished other than by the expected scram signal (e.g.,

scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel

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damage, i

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i PBAPS

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Unit 2

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

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f 3.3.B Control Rods (Cont'd) 4.3.8 Control Rods (Cont'd)

4. Control rods shall not be with-4.

Prior to control rod with-drawn for startup or refueling drawal for startup or dur-unless at least two source ing refueling verify that range channels have an observed at least two sources range count rate equal to or greater channels have an observed j

than three counts per second.

count rate of at least i

three counts per second.

I

5. During operation with limiting

5. When a limiting control rod control rod patterns as deter-pattern exists, an instru-mined by the designated quali-ment functional test of the fled personnel, either:

RBM shall be performed prior j

to withdrawal of the desig-nated rod (s).

a. Both RMB chtnnels shall be operable, or

b. Control rod withdrawal shall be blocked, or

c. The operating power level sha)) be limited so that the MCPR will remain above the fuel cladding integrity safety j

limit assuming a single error that results in complete with-drawal of a single operable control rod.

l C. Scram Insertion Times C.

Scram Insertion Times

1. The average scram insertion time,

1. After each refueling outage, based on the deenergization of and prior to synchronizing the scram pilot valve solenoids the main turbine generator as time zero, of all operable initially following restart control rods in the reactor of the plant, all operable power operation condition fully withdrawn insequence i

shall be no greater than:

rods shall be scram time tested during startup from

% Inserted from Avg. Scram Inser-the fully withdrawn posi-Fully Withdrawn tion Times (sec) tion with the nuclear system pressure above 800 psig.

5 0.375 20 0.90 50 2.0 90 3.5

- 103 -

L

PDAhS Unit 2 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQbIREMENTS 3.3.C (Cont'd) 4.3.C (Cont'd)

.After-exceeding 30 percent power all previously untested operable control rods shall be tested - as described above prior to exceeding 40 percent power.

2. The average of the scram inser-

2. Whenever such scram time meas-tion times for the three fastest urements are made (such as when control rods of all groups of

a. scram occurs and the scram four control rods in a two-by-insertion time recorders are two array shall be no greater operable) an evaluation shall l

than:

be made to provide reasonable assurance that proper control E

rod drive performance is being

}

maintained.

t

% Inserted From Avg. Scram Inser-f Fully Withdrawn tion Times (Sec)

}

5 0.398 20 0.954 50 2.120 90 3.8

3. The maximum scram insertion time for 90% insertion of any operable control rod shall not exceed 7.00 ij seconds.

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PBAPS l

Unit 2 i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOCIREMENTS 3.5.I Average Planar LHGR 4.5.I Average Planar LHGR During power operation, the APLHGR The APLHGR for each type of fuel for each type of fuel as a function as a function of average planar of average planar exposure shall not exposure shall be checked daily exceed the limiting value shown in during reactor operation at >25%

the applicable figures during two rated thermal power.

recirculation loop operations.

4 During single loop operation, the APLHCR for each fuel type shall l

not exceed the above values mult-iplied by the following reduction factors:

0.71 for 7x7 fuel; 0.83 for 8x8 fuel; 0.81 for PTA, 8X8R and P8X8R fuel. If at any time during operation it is 4

determined by normal surveillance that the limiting value of APLHGR is being exceeded, action shall be initiated within one (1) hour to I

restore APLHGR to within prescribed limits.

If the APLHGR is not re-turned to wi thin prescribed limits within five (5) hours reactor power shall be decreased at a rate which 2

I would bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> unless APLHGR is returned to within limits during this period.

Sur-veillance and corresponding action 3

shall continue until reactor opera-tion is within the prescribed limits.

3.5.J Local LifGR 4.5.J Local LHCR i

During power operation, the linear The LHGR as a function of core heat generation rate (LHGR) of any height shall be checked daily rod in any fuel assembly at any axial during reactor operation at location shall not exceed design LHGR. >25% rated thermal power.

LHGR < LHGRd l

LHCR I Design LHCR 13.4 kW/ft for all 8x8 fuel l

-133a-

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PBAPS Unit 2-LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

3. 5.J Local LHGR (Cont'd)

If at any time during operation it is determined by normal surveillance that limiting value for LHGR is be-ing exceeded, action shall be-initi -

ated within one (1) hour to restore LHGR to within prescribed limits.

If the LHGR is not returned to with-in five (5) hours, reactor power shall be decreased'at a rate which would bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> unless LHGR is. returned to within limits during this period.

Sur-veillance and corresponding action l

shall continue until reactor oper-ation is within the precribed limits.

[

9 3.5.K.1 Minimum Critical Power 4.5.K Minimum Critical Power 3

Ratio (MCPR)

Ratio 2

1.

During power operation the MCPR

1. MCPR shall be checked daily for the applicable incremental during reactor. power operation cycle core average exposure and at >25% rated thermal power.

for each type of fuel shall be 2.

Except as provided in Specifi-equal to or greater than the value cation 3.5.K.3, the verifica-given in Speci fication 3.5.K.2 or tion of the applicability of 3.5.K.3 times Kf, where Kf is as 3.5.K.2.a Operating Limit MCPR shown in Figure 3.5.1.E.

If at Values shall be performed every j

any time during operation it 120 operating days by scram time p

is determined by normal surveil-testing 19 or more control rods 1ance that the limi ti ng on a rotation basis and-per-value for MCPR is being exceeded, forming the following:

action shall be initiated within one (1) hour to restore MCPR to a.

The average scram time to f

within prescribed limits.

If the 20% insertion position the MCPR is not returned shall be:

i to within prescribed limits

]Fave <] B within five (5) hours, reactor 1

power shall be decreased at a

b. The average scram time to j

rate which would bring the the 20% insertion position j

reactor to the cold shutdown is determined as follows:

condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> n

unless MCPR is returned to 3~ ave = 1 Ni71 within limits during this period.

i=1 I

Surveillance and corresponding n

action shall continue until re-

)[ Ni actor operation is within the i=1 prescribed limits.

s where: n = number of surveillance

]

tests performed to date in the cycle

- 133b -

PBAPS Unit 2 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS i

3.5.K.

Minimum Critical Power 4.5.K. Minimum Critical Power Ratio MCPR (Cont'd)

Ratio-MCPR (Cont'd) j 2.

Except as speci fied in 3.5.K.3, Ni = number of active control the Operating Limit MCPR Values rods measured in the'ith are as follows:

surveillance test, a.

If requirement 4.5.K.2.a is met:

The Operating Limit MCPR values are as given in Table 3.5.K.2 7'1 = average scram time to the 20% insertion position of all rods measured in l

b.

If requirement 4.5.K.2.a is not the ith surveillance test.

}

met:

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The Operating Limit MCPR

c. The adjusted analysis mean values as a function of 7" scram time ( 7"B) is calculated are as given in Figures as follows:

I 3.5.K.1, 3.5.K.2, and 3.5.K.3.

t

( Ni 7"B =j# +1.651 i

n

( 1 Ni i=1 i

Where:

Where:

l I

7~ = Tave - 7 B l

+

// = mean of the distribution for 0.90 - 7B average scram insertion time ty the 20% position

0.710 sec.

3.

The Operating Limit MCPR_ values Ni = total number of active control I shall be as given in Table 3.5.K.3 rods measured in specification [

if the Surveillance Requirement 4.3.C.1 of Section 4.5.K.2 to scram time test control rods is not performed o' = standard deviation of the distribution for average scram insertion time to the 20% position = 0.053.

P

-133c-

PBAPS Unit 2 Table 3.5.K.2

. OPERATING LIMIT MCPR VALUES FOR VARIOUS CORE EXPOSURES

• MCPR Operating Limit **

Fuel Type For Incremental Cycle Core Average Exposure

' BOC to 2000 MWD /t 2000 MWD /t'before EOC Before EOC To EOC 8x8R/LTA-1.23 1.26 P 8x8R 1.25 1.29 P8DRB285 1.29 1.29 If requirement 4.5.K.2.a is met.

These falues shall be increased by 0.01 for single loop operation, i

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PBAPS Unit 2 Table 3.5.K.3 OPERATING LIMIT MCPR VALUES FOR VARIOUS CORE EXPOSURES

• j MCPR. Operating Limit **

Fuel Type For Incremental' Cycle Core Average Exposure BOC to 2000 MWD /t 2000 MWD /t before EOC Before EOC To EOC 8x8R/LTA 1.33 1.38 P 828R 1.35 1.41 f

P8DRB285 1.35 1.41 4

i If surveillance requirement. 4.5.K.2 is not performed.

These values shall be increased by 0.01 for single loop operation.

i 4

i

- 133e -

Unit 2 3.5 BASES (Cont 'd. )

H.

Engineering Safeguards Compartments Cooling and Ventilation l

One unit cooler in each pump compartment is capable of providing j

adequate ventilation flow and cooling.

Engineering analyses indicated that the temperature rise in safegurards compartments j

without adequate ventilation flow or cooling is such that t

continued operation of the safeguards equipment or associated auxiliary equipment cannot be assured.

Ventilation associated with the High Pressure Service Water Pumps is also associated i

with the Emergency Service Water pumps, and is specified in i

Speci fication 3.9.

I.

Average Planar LHGR This specification assures that the peak cladding temperature j

following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10CFR Part 50, i

l Appendix K.

l The peak cladding temperature (PCT) following a postulated loss-f of-coolant accident is primarily a function of the average heat t

generation rate of all the rods of a fuel assembly at any axial location and is only dependent, secondarily on the rod to rod j

power distribution within an assembly.

The peak clad temperature is calculated assuming a LHGR for the highest powered rod which I

is equal to or less than the design LHGR. 'This LHGR times 1.02 is used in the heat-up code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking j

factors.

The Technical Specification APLHGR is the LHGR of.the j

highest powered rod divided by its local peaking factor.

The j

limiting value for APLHGR are shown in the applicable figure for each fuel type.

i The calculational procedure used to establish the APLHGR is based

[

on a loss-of-coolant accident analysis.

The analysis was performed using General Electric (G.E.) calculational models f

which are consistent with the requirements of Appendix K to 10CFR Part 50.

A complete discussion of each code employed in the

{

analysis is presented in Reference 4.

Input and model changes in l

the Peach Bottom loss-of-coolant analysis which are different r

from the previous analyses performed with Reference 4 are described in detail in Reference 8.

These changes to the

. analysis include:

(1) consideration of the counter current flow limiting (CCFL) effect, (2) corrected code inputs, and (3) the effect of drilling alternate flow paths in the bundle-lower tie plato.

- 140 -

I

PBAPS Unit 2 3.5.I BASES (Cont'd)

J.

Local LHGR This specification assures that the linear heat generation rate in any 8X8 fuel rod is less than the design linear heat generation.

The maximum LHGR shall be checked daily during reactor operation at 125% power to determine if fuel burnup, or control rod movement has caused changes in power distribution.

For LHGR to be at the design LHCR below 25% rated thermal power, the peak local LHGR must be a factor of approximately ten (10) greater than the average LHCR which is precluded by a considerable margin when employing any permissible control rod pattern.

K.

Minimum Critical Power Ratio (MCPR)

Operating Limit MCPR The required operating limit MCPR's at steady state operating conditions are derived from the established fuel cladding integrity Safety Limit MCPR and analyses of the abnormal operational transients presented in Supplemental Reload Licensing Analysis and Reference 7.

For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting

-given in Specification 2.1 To assure that the fuel cladding integrity Safety Limit is not violated during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR).

The transients evaluated are as described in reference 7.

- 140a -

PBAPS Unit 2 3.5.K.

BASES (Cont'd)

The largest reduction in critical power ratio is then added to the fuel cladding integrity safety limit MCPR to establish the MCPR Operating Limit for each fuel type.

Two codes are used to analyze the rod withdrawal error transient.

The first code simulates the three dimensional BWR core nuclear and thermal hydraulic characteristics.

Using this code a limiting control rod pattern is determined; the following assumptions are included in this determination:

(1) The core is operating at full power in the xenon-free condition.

(2) The highest worth control rod is assumed to be full i nserted.

(3) The analysis is performed for the most reactive point in the cycle.

(4) The control rods are assumed to be the worst possible pattern without exceeding thermal limits.

(5) A bundle in the vicinity of the highest worth control rod is assumed to le operating at the maximum allowable linear heat generation rate.

(6) A bundle in the vicinity of the highest worth control rod is assumed to be operating at the minimum allowable critical power ratio.

The three-dimensional BWR code then simulates the core response to the control rod withdrawal error. -Tne second code calculates the Rod Block Monitor response to the rod withdrawal error.

This code simulates the Rod Block Monitor under selected failure conditions (LPRM) for the core responso (calculated by the 3-dimensional BWR simulation code) for the control rod withdrawal.

The analysis of thd rod withdraw 31 error for Peach Bottom' Unit 3 considers the continuous withdrawal of the maximum worth control rod at i ts maximum drive speedJ from the reactor which is operating with the limiting control rod pattern as discussed above.

- 140b -

i

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~

PBAPS Unit 2 3.5.K.

BASES (Cont'd)

A brief summary of the analytical method used to determine the nuclear characteristics is given in Section 3 of Reference 7.

Analysis of the abnormal operational transients is presented in Section 5.2 of Reference 7.

Input data and operating conditions used in this analysis are shown in Table 5-8 of Reference 7 and in the Supplemental Reload Licensing Analysis.

L.

Average Planar LHGR (APLHGR), Local LHGR and Minimum Critical Power Ratio (MCPR)

In the event that the calculated value of APLHGR, LHCR or MCPR exceeds its limting value, a determination is made to ascertain j

the cause and initiate corrective action to restore the value to within prescribed limits.

The status of all indicated limiting l

fuel bundles is reviewed as well as input data associated with the limiting values such as power distribution, instrumentation data (Traversing In-Core Probe TIP, Local Power Range Monitor -

LPRM, and reactor heat balance instrumentation), control rod configuration, etc., in order to determine whether the calculated values are valid.

4 In the event that the review indicates that the calculated value exceeding limits is valid, corrective action is immediately undertaken to restore the value to within prescribed limits.

Followinc corrective action, which may involve alternations to the control rod configuration and consequently changes to the core power distribution, revised instrumentation data, including changes to the relative neutron flux distribution,for up to 43 incore locations is obtained and the power distribution, APLHGR, LHGR and MCPR calculated.

Corrective action is initiated within one hour of an indicated value exceeding limits and verification that the indicated value is within prescribed limits is obtained within five hours of the initial indication.

In the event that the calculated value of APLHGR, LHGR or MCPR exceeding its limiting value is not valid, i.e.,

due to an erroneous instrumentation indication, etc., corrective action is initiated within one hour of an indicated value exceeding limit.

Verification that the indicated value is within prescribed limits is obtained within five hours of the initial indication.

Such an invalid indication would not be a violation of the limiting condition for operation and therefore would not constitute a reportable occurrence.

- 140c -

s

~'

PBAPS Unit 2 3.5.L BASES (Cont'd)

Operating experience has demonstrated that a calculated value of APLHGR, LHGR or MCPR exceeding its limiting value predominately occurs due to this latter cause.

This experience coupled with the extremely unlikely occurrence of concurrent operation exceeding APLHGR, LHCR or MCPR and a Loss of Coolant Accident or applicable Abnormal Operational Transients demonstrates that the times required to initiate corrective action (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) and restore the calculated value of APLHCR, LHGR or MCPR to within prescribed limits (5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) are adequate.

3.5.M. References 1.

" Fuel Densification Effects on General Electric Boiling Water Reactor Fueld, Supplements 6,7, and 8 NEDM-10735, August 1973.

2.

Supplement 1 to Technical Report on Densifications of General Electric Reactor Fuels, December 14, 1974 (Regulatory Staff).

3.

Communication:

V.

A.

Moore to I.

S. Mitchell, " Modified GE Model for Fuel Densi fication", Docket 50-321, March 27, 1974.

4.

General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE 20566 i

(Draft), August 1974.

5.

General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by latter, G.

L.

Gyorey to Victor Stello, Jr.,

dated December 20, 1974.

6.

DELETED 7.

General Electric hoiling Water Reactor Generic Reload Fuel Application.

NEDO-24011-P-A.

8.

Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2, NEDO-24082, December 1977, and for Uni t 3, NEDO-24082, December, 1977.

i

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