Text

Amend 68 to License DPR 20,specifying New Limits for Radial Peaking Factors & Allowable Linear Heat Rates & Identifying Use of Excore Detectors for Core Power Distribution Monitoring
	ML20039D554
Person / Time
Site:	Palisades
Issue date:	12/08/1981
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	;
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ML20039D555	List: ML20039D554; ML20039D559; ML20039D560;
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	NUDOCS 8201050199
	Download: ML20039D554 (39)
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g UNITED STATES NUCLEAR REGULATORY COMMISSION y

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WASHINGTON. D. C. 20666 e,

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CONSUMERS POWER COMPANY DOCKET NO. 50-255 PALISADES PLANT AMENDMENT TO PROVISIONAL OPEP.ATING LICENSE Amendment No. 68 License No. DPR-20 1

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Consumers Power Company. (the licensee) dated July 21, 1981 as supplemented August 6,1981, October 22, 1981, November 9, 17, 20, 1981 and December 2,1981 complies with the standards and requirements 'of the Atomic' Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I;

.c B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering th'e health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance -with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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F82010501aa 811208 DF: ADOCK 05000

_g-Accordingly, the license is amended by changes to the Technical

- 2.

Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Provisional Operating License No. DPR-20 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendices A-and B (Environmental Protection Plan), as revised through Amendment No. 68, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR-THE NUCLEAR REGULATORY COMMISSION Dennis M. Crutchfield, Chief Operating Reactors Branch #5 Division of Licensing

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Attachment:

Changes to the Technical Specifications Date of Issuance:

December 8,1981 M'-

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9 ATTACHMENT TO LICENSE AMENDMENT NO. 6R PROVISIONAL OPERATING LICENSE N0. DPR-20 DOCKET NO. 50-255 Revise' Appendix A Technical Specifications by removing the following pages and by inserting the enclosed pages. The revised pages contain the captioned amendment. number and marginal lines indicating the area of change.

Remove Insert if 11 iii iii iv*

, 2 1-2 3-38 3-38**

3-58 3-58 3-59 3-61 3-61.

3-63 3-63 I

3-64 3-64 (Intentionally Blank) 3 3-66a 3 3-66d 3-81 a 3-81 a 3-87 3-87a 3-103 113 4 4-84 6-la 6-la**

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Included for pagination purposes only.

- These pages are included for the purpose of correcting errors which occurred -

durin the issuance of Amendment No. 62 (page 3-38) and Amendment No. 67 i

(6-la.

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TABLE OF CONTENTS (Cont'd)

Section Description Page 3.0 LIMITING CONDITIONS FOR OPERATION (Cont'd) 3.10 Control Rods 3-58 l

3.10.1 Shutdown Margin Requirements 3-58 3.10.2 Individual Rod Worth 3-58 I

3.10.3 Part-Length Control Rods 3-58 I

3.10.4 MisaligneJ or Inoperable Control Rod or Part-Length Rod 3-60 3.10.5 Regulatir Group Insertion Limits 3-60 3.10.6 Shutdown Rod Limits 3-61 3.10.7 Low Power Physics Testing 3-61 3.10.8 Center Control Rod Misalignment 3-61 3.11 Power Distribution Instruments 3-65 3.11.1 Incore Detectors 3-65 3.11.2 Excore Power Distribution Monitoring System 3-66a 3.12 Moderator Temperature Coefficient of Reactivity 3-67 3.13 Containment Building and Fuel Storage Building Cranes 3-69 3.14 Control Room Air Temperature 3-70 3.15 Reactor Primary Shield Cooling System 3-70 3.16 Engineered Safety Features System Initiation Instrumentation Settings 3-71 3.17 Instrumentation and Control Systems 3 3.18 Secondary Water Chemistry 3-82 3.19 Linear Heat Generation Rate Linits Associated With LOCA Considerations 3-84 3.20 Shock Suppressors (Snubbers) 3-88 3.21 Movement of Heavy Loads-Over the Spent Fuel Pool (To Be Submitted) 3-92 3.22 Fire Protection System

'3-96 3.22.1 Fire Detection Instrumentation 3-96 3.22.2 Fire Suppression Water System 3-98 3.22.3 Fire Sprinkler System 3-100 3.22.4 Fire Hose Stations 3-1 01 3.22.5 Penetration Fire Barriers 3-102 3

3.23 Power Distribution Limits 3-103 3.23.1 Linear Heat Rate (LHR) 3-103 3.23.2 Radial Peaking Factors 3-110 3-111 3.23.3 Quadrant Power T.H - Tq

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4.0 SURVEILLANCE REQUIREMENTS 4-1 4.1 Instrumentation and Control 4-1 4.2 Equipment and Sampling Tests 4-13 4.3 Systems Surveillance 4 4.4 Primary Ccolant System Integrity Testing a-24 4.5 Containment Tests 4-25 l

11 AmendmentNo.JV,5/,68

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TABLE OF CONTENTS (Cont'd)

Section Description Page 4.6 Safety Injection and Containment Spray Systems Tests 4-39 4.7' Emergency Power System Periodic Tests 4-42

' 4.8 Main Steam Stop Valves 4-44

. 4.9 Auxiliary Feed-Water System 4-45 4.10 Reactivity Anomalies 4-46 4.11 Environmental Monitoring Program 4-47 4.12 Augmented Inservice Inspection Program for High Energy Lines Outside of Containment 4-60 4

4.13 Reactor Internals Vibration Monitoring 4-65 4.14 Augmented Inservice Inspection Program for Steam Generators 4 4.15 Primary System Flow Measurement 4-70 4.16 Inservice Inspection. Program for Shock Suppressors (Snubbers) 4-71 4.17 Fire Protection System 4-75 4.18 Power Distribution Instruments 4-81 4.18.1 Incore Detectors 4-81..

4.18.2 Excore Power Distribution Monitoring System 4-82 4.19 Power Distribution Limits 4-83 4.19.1 Linear Hea* Rate 4-83 4.19.2 Radial Peaking Factors 4-84 5.0-DESIGN FEATURES 5-1 5.1 Site 5-1 5.2 Containment Design Features 5-1 5.3 Nuclear Steam Supply System (NSSS) 5-2 5.4 Fuel Storage 5-3 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 -

Responsibility 6-1 6.2 Organization 6-1 6.3 Plant Staff Qualifications 6-1 6.4 Training 6-1.

6.5 Review and Audit 6-5 6.6 (Deleted) 6-9 i-6.7 Safety Limit Violation 6-9 6.8 Procedures

.6-10 6.9 Reporting Requirements 6-11 6.10 Record Retention 6-25 6.11 Radiation Protection Program 6-27 6.12 High Radiation Area 6-28 6.13

. Fire Protection Inspection 6-33 6.14 Environmental Qualification 6-33 6.15 Systems Integrity 6-33 6.16 Iodine Monitoring

.6-33 Amendment No. J, p, 63, 6[, 68 Ig fii l

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c-TABLE OF CONTENTS.(Cont'd)

Section Description Page S

SPECIAL TECHNICAL SPECIFICATIONS PURSUANT'TO AGREEMENT S-1 i

S-1 (Deleted)

S-5 S-2 S-2 Liquid Radwaste -System Modification i-S-3 through S-5 (Deleted) l b

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~ Amendment No. g, 44, 58,' N, 68 l-

l.1 REACTOR OPERATING CONDITIONS (Cont'd)

Low Power Physics Testing Testing. performed under approved written procedures to determine control rod worths and other core nuclear properties. Reactor power during these tests shall not exceed 2% of rated power, not including decay heat and primary system temperature and pressure shall be in the range of 260*F to 538"F and 415 psia to 2150 psia, respectively. Certain deviations from normal operating practice which are necessary to enable performing some of these tests are.

permitted in accordance with the specific provisions therefor in these Technical Specifications.

Shutdown Boron Concentrations Baron concentration sufficient to provide keff 4.0.98 with all control rods i

in the core and the highest worth control rod fully withdrawn.

Refueling Boron Concentration Boron concentration of coolant at least 1720 ppm (corresponding to a shutdown margin of at least 5% Ap with all control rods withdrawn).

Quadrant Power Tilt The difference between nuclear power in any core quadrant and the average in all quadrants.

Assembly Radial Peaking Factor - FrA The assembly radial peaking factor is the maximum ratio of individual fuel assembly power to core average assembly power integrated over the total core height, including tilt.

Total Radial Peaking Factor - FrT The total radial peaking factor is the maximum product of the ratio of individual assembly power to core average assembly power times the local peaking factor for that assembly integrated over. the total core height, including tilt. Local peaking factor is defined as the maximum ratio of the power in an individual fuel rod to assembly average rod power.

Interior Fuel Rod Any fuel rod of an assembly that is not on that assembly's periphery.

Total Interior Rod Radial Peaking Factor - FraH The maximum product of the ratio of individual assembly power to core average assembly power times the highest interior local peaking factor integrated over the total core height including tilt, i

Axial Offset c

The difference between the power in the lower half of the core and the upper i

half of the core divided by the sum of the powers in the lower half and upper half of the core.

Narrow Water Gap Fuel Rod A fuel rod adjacent to the narrow' inter-fuel assembly water gap (a gap not containing a control rod).

Narrow Water Gap Fuel Rod Peaking Factor - FN The maximum product of the ratio of individual fuel assembly power to core average fuel assembly power times the highest narrow water gap fuel rod local peaking factor integrated over the total core height including tilt.

Amendment No. M, FI, N,J7',68 1-2

i 35 SM AIFDAS E'ISTS A;plicability

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Applies to the operating status of the stes= and feedvater syste=s.

Obfective To define certain con:iitic=s of the stes= sud feedvater syste= neces-sary to assure scequate decay heat re=cval.

Specificatiens 0

be wated above 325 F unless the follow-351 The pri=ary coolant ska

=0 ing cc diticus are =et:

Ecth aux 111:ry feedvater pu=ps operable and one fire pu:.p operable.

l a.

b.

A =ini=u= of 100,000 ga'lons of water in the conde:: sate storage and pri=ary coolant syste: =akeup tanks ce=bined and a backup s0=ce of additic a1 vater frc= Lake Michigan by the operability of one of the fire protectics pu=ps.

All valves, interlocks a=d piping associated with the above c =-

c.

pone =ts required to functics dr ing accident conditicus, are cpert'le.

d.

The =ain stes= step valves are cperable and capable of clesing in five secc ds er less under no-flow conditit s.

o 3.5 2 With the pr P= / coolant syste= at a te=peratr e greater than 325 F, the require =ents of 3 51 =ay be = edified to pe '

"e following conditices to exist. If the syste= is ::t res:Ored t: =eet the require-

=ents of 3 5.1 vithin the ti=e pericd specified belov, the reactor s"a be_ placed in the coli shutdev :: ditic vithin 2!. hours.

Cne auxiliary feedvater pt=p =ay be incperable for a period of a.

I2 hours. Or b.

The firevater makeup to the auxiliary feedvater pu=p suction may be inoperable for a period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3.5.3 &If one auxiliary feedwater pump and the firewater makeup supply to

'the auxiliary feedwater pumps become inoperable, then the plant shall be placed in hot standby within I hour, in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.5.4 With both auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible and reduce power within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the lowest stable power level consistent with reliable main feedwater system operation.

3-38 Amendment No. 62 SS Jet:ba-20, W (Correctice,}

3.10 CONTROL RODS Apolicability Applies te _ operation of control rods and hot. channel factors during.

operation.

Objective To specify limits of control rod movement to assure an acceptable

-power distribution during power operation, limit' worth of individual rods to values analyzed for. accident conditions, maintain adequate shutdown margin after a reactor trip and to specify acceptable power limits for power tilt conditions.

-Soecifications 3.10.1 Shutdown Margin Requirements a.

With four primary coolant pumps in operation at hot shutdown and above, the shutdown margin shall be 2%.

b.

With less than four primary coolant pu.nps in operation at hot shutdown and above, the shutdown margin shall be.3.75%.

c.

At less than the hot shutdown condition, boron concentration shall be shutdown boron concentration, d.

If a control rod cannot be tripped, shutdown margin 'shall be in-creased by boration as necessary to compensate for the worth of the withdrawn inoperable rod.

e.

The drop time' of each control rod shall be no greater than 2.5 seconds from the beginning of rod motion to 90% insertion.

3.10.2 Individual Rod Worth l

a.

The maximum worth of any one rod in the core at rated power shall j

be equal to or less than 0.6%'in reactivity.

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7, b.

The maximum worth of any one rod in the core at zero power shall be equal to or less than 1.2*; in reactivity.

3.10.3 Part-Length Control Rods The part-length control rods will be completely withdrawn from the l

core (except for control rod exercises and physics tests).

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Amendment No.g, 4I, 5/, 68 3-58 l

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CCNTROL RCD A.D p0kB DISRIBUTICN LIMITS (Centd)

V 3.10 3.10.6 Shutdewn Rod Limits All shutdown rods shall be withdrawn before any regulating rods a.

are withdrawn.

b.

The shutdown rods shall not be withdrawn until normal water level is established is the pressuri:er.

c.

The shutdown rods shall not be inserted below their exercise limit until all regulat'ng rods are inserted.

3.10.7 Low Power physics Testing Sections 3.10.1.4, 3.10.1.b, 3.10.2.b, 3.10.3

, 3.10.4.b, 3.10.5 and

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3.10.6 may be deviated from during low power physics testing and CREM exercises if necessary to perform a test but only for the timq -

necessary to perform the test.

3.10.8 center Control Rod Mialignment The requirements of Specifications 3.10.4.1, 3.10.4.a and 3.10.5 l

may be suspended during the performance of physics tests to deter-mine the isothermal temperature coefficient and power coefficient provided that only the center control rod is misaligned and the limits of Specification 3.23 are maintained.

i Basis

- Sufficient control rods shall be withdrawn at all times to assure that the reactivity decrease from a reactor trip provides adequate shutdown margis. The available worth of withdrawn rods must include the reac-tivity defect of power and the failure of the withdrawn rod of highest worth to insert. The requirement for a shutdown margin of 2.0% is're-activity with 4-pump operation, and of 3.75% in reactivity with less than 4 pump operation, is consistent with the assumptions used in the analysis of accident conditions (including steam line break) as reported is XN-NF-77-18 and additicnal analysis.(3) The change is insertion limit with reactor power shown on Figure 3-6 insures that the shutdown margis requirements for 4 pump operation is met at all power levels. The 2.5-second drop time specified for the control rods is the drop time used is the transient analysis.(3)

The maximum individual rod worth of. inserted control rods and associ-l ated peaking factors have been used to demonstrate reactor safety for l

the unlikely event of a red ejection accident as described in l

Reference 5.

The maximum worth of an inserted control rod will not l

exceed the values of the specification for the regulating group insertion limits of Figure 3-6.

i The insertion of part-length rods into the core, except for rod' exercises or physics tests, is not permitted since it has been demonstrated on other CI plants that design power distribution e=velopes can, under some circumstinces, he violated by using part-length rods. Further information may justify their use.

part-length rod insertion is permitted for physics tests, sisce resulting power distributions are closely monitored under test conditions. pa rt-length rod insertion for rod exercises (approximately 6 inches) is permitted since this amount of insertion has an insignificant effect on power distribution.

Amendment No. [)#, >[7, 68 3 I 4

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d 3.10 CONTROL ROD AND F0i4IR DISTRIBUTION LIMITS (Contd)

For a control rod misaligned up to 8 inches from the remainder of the i

banks, hot channel factors will be well within design limits. If a control rod is sisaligned by more than 8 inches, the maximum reactor power will be reduced so that hot channel factors, shutdown margin and ejected rod worth limits are met.

If in-core detectors are not available to measure power distribution and rod misalignments > 3 inches exist, then reactor power must not exceed 75*. of rated power to insure that hot chan M'i conditions are set.

Continued operation with that rod fully inserted vill only be permitted i'

if the hot channel factors,' shutdown margin and ejected rod verth limits are satisfied.

In the event a vithdrawn centrol rod cannot be tripped, shutdown margin requirements vill be maintained by increasing the boron concentration by an amount equivalent in reactivity to that control rod. The deviations

- per:nitted by Specification 3.10 7 are required in order that the control rod worth values used in *he reactor physics calculations, the plant safety analysis, and the Technical Specifications can be verified. These deviations vill only be in effect for the time period required for the test being performed.

The testing interval during which these deviations vill be in effect vill be kept to a minimum and special operating precau-tiens vill be in effect during these deviations'in accordance with approved written testing procedures.

Violation of the power dependent insertion limits, when it is necessary to rapidly reduce power to avoid or minimize a situation harmful to plant personnel or equipment, is acceptable due to the brief period of time that

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such a violation vould be expected to exist, and due to the fact that it is l

tmlikely that cere operating li=its such as ther=al margin and shutdown =ar-gin vould be violated as a result of the rapid rod insertion. Core ther=al i

=argin vill actually increase as a result of the rapid rod insertion. In addition, the required shutdevn margin vill mes.t likely not be violated as I

a result of the rapid rod insertion because present power dependent inser-tien li=its result in shutdevn nargin in excess of that required by the safety analysis.(5)

References (1) FSAR, Section ik.

(2) FSAR, Section 3 3 3

'(3) FSAR, Sectics T.b.2.2.

(h) FSAR, SEctica T.3 3.6.

(5) U-NF-TT-1S.

'kiendment No. }i, 33,J7, 68 -

3-63

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(INTENTIONALLY BLANK) 9 3-64 Amendment No. 68

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i 3.11 PCWER DISTRIBITTION INSTRUME.NTATION 3.11.1 INCORE DETECTORS

-vu LIMITING CONDITION FOR OPERATION 4

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- The incore detection system shall be operable:

a.

With at least 50* of the incore detectors - and 2 inceres per axial level per core quadrant.

b.

With the incore alarming function of the datalogger 2

operable and alarm setpoints entered into the -datalogger.

1 5

APPLICABILITY (1) Item a. above is applicable when the incore' detection system is ij' used, for:

f Measuring quadrant power tilt, Measuring radial peaking factors, i

Mea,suring linear heat rate (LHR), or Determining target Axial Offset.(AO) and excore monitoring allowable power level.

(2) Items a. and b. above are applicable when the incore detectien-system is used for monitoring LHR with automatic alarms.

(Incore Alarm System.)

t-ACTION 1:

l L

With less than the required number of incore detectors, do not use the l

system for the measuring and calibration functions under -(1) above, i-l ACTION 2: With the alarming function of the datalogger inoperable, do l

not use the system for autcmatic monitoring of LHR (Incperable Incore Alarm System).

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f Amendment No.,JI,;E6, 68

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POWER DISTRIBUTION INSTRLM NTATIO!!

3.11.1 INCORE DETECTORS LIMITING CONDITION FOR OPERATION ACTION 2:

(Contd)

Operation may continue using the excore monitoring system as specified in 3.11.2 c. - by meeting the requirements of 3.23.1.

Basis The operability of the incere detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The operability of the incere alarm system depends on the availability of the datalogger as well as the operability of a minimum number of incore detectors. Incore alarm setpoints must be updated periodically based on measured power I

distributions. The incore detector Channel Check is normally performed by an off-line computer program that correlates readings with one another and with computed power shapes in order to identify inoperable detectors.

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Amendment No. )f, H, 56, 57, 58, 68

3-66

POiER DISTRIBLTION INSTRUMENTATION 3.11.2 EXCORE POWER DISTRIBLTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION The excore monitoring system shall be operable with:

a.

The target Axial Offset (AO) and the Excore Monitoring Allowable Power Level (APL) determined within the previous 31 days using the incere detectors, and the measured A0 not deviated from the target A0 by more than 0.05 in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

The A0 measured by the excore detectors calibrated with the A0 measured -

by the incore detectors.

c.

The quadrant tilt measured by the excore detectors calibrated with the quadrant tilt measured by the incore detectors.

APPLICABILITY:

(1) Items a., b. and c. above are applicable when the excore detectors are used-for monitoring LHR.

(2) Item c. above is applicable when the excore detectors are used for monitoring quadrant tilt.

ACTION 1:

With the excore monitoring system inoperable, do not use the system for monitoring I

LHR.

ACTION 2:

i If the measured quadrant tilt has not been calibrated with the incores, do not use the system for monitoring quadrant tilt.

l Basis l

The excore power distribution monitoring system consists of Power Range Detector Channels 5 through 8.

l The operability of the excore monitoring system ensures that the assumptions; employed in the PDC-II analysis ( ) for determining A0 limits that ensure operation l

within allowable LHR limits are valid.

3-66a A mendment No. jpg,,)d, 68

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PCWER DISTRIBUTION INSTRC"NTgIj.y 3.11.2 EXCORE PCWER DISTRIBLTION MONITORING SYSTEM LIMITING CCNDITION FOR OPERATION i

Basis (Contd) i Surveillance requirements ensure that the instruments are calibrated to agree with the incore measurements and that the target A0 is based on the current operating conditions. Updating the Excere Monitoring APL ensures thst.the core LHR limits are protected within the ! 0.05 band on AO. The APL considers both LOCA and DNB based LHR limits, and factors are included to account for changes in rtiial power shape and LHR limits over the calibratien interval.

The APL is determined from the following:

LHR(Z)TS x Rated Power x 1.02]gg LHR(!)3, x V(Z) x E (Z) p i

l Where:

(1) LHR(Z)TS I" * *

8

8 * ('# # * * ' "

i (2) LHR(Z) is the measured peak LER including uncertainties vs Core Height, i

(3) V(Z) is the function (shown in Figure 3.11-1),

(4) E (Z) is a factor to account for the reduction of allowed LHR in the peak rod p

with increased exposure (Figure 3.23.2) such that:

For fuel rod burnups less than 27.0 GWd/ E - E = 1.0 p

For fuel rod burnups greater than 27.0 GWd/MT but less than 33.0 GVd/ E -

E = 1.0 + 0.0064 x LHR

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P For fuel rod burnups greater than 33.0 GVd/MT - E = 1.0 + 0.0012 x LHR Where LHR is the measured fuel rod average LHR in kW/ft, Amendment No. 68 3-6%

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a POWER DISTRIBLTION INSTRUMENTATION 3.11.2 EXCORE POVEF. DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION l

Basis (Contd)

(5) The factor of 1.02 is an allowance for the effects of upburn, (6) The quantity in brackets is the minimum value for the entire core at any elevation (excluding the top and bottom 10% of core) considering limits for peak rods, interior fuel rods and narrow water gap fuel rods. E (Z) p is only applied if the minimum value is based on limits for the peak rod.

If the quantity in brackets is greater than one, the AFL shall be the I

I rated power level.

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Reference I.

(1) XN-ST-80-47 Amendment No. 68 3-66c

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Amendment flo. 68

r Table 3.17.4 (Cont'd) t Minimum Minimum Permissible Operable Degree of Bypass No.

Functional Unit Channels Redundancy Conditions 8.

Pressuri:er Water 2

1 Not Required in Level (LI-0102)

Cold or Refueling Shutdown 9.

Pressurizer Code Safety 1 per None Not Required below Relief Valves Position valve 325'F Indication (Acoustic Monitor or Temperature Indication) l'O.

Power Operated Relief I per None Not required when PORY isolation valve Valves (Acoustic ~

valve' is closed and its Monitor or Temperature Indication) indication system is operable u

i 11.

PORY Isolation Valves 1 per None-Not required when Position Indication valve reactor is dEpressurized and vented through a vent gl.3 sq. in.

12.

Subcooling Margin 1

None Not Required j

Monitor Below 515'F i

None NotRequihed 13.

Auxiliary Feed Flow 1 per Rate Indication Steam Below 325*F Generator

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j 14 Auxiliary Feed Pump l per (e)

None Not Required Auto Initiation Pump Below 325'F l

Circuitry

~15.

Execre Detecicr L

Nc=e

'Ncue

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(e) With one auxiliary feed pump automatic initiation circuit inoperable, in lieu of the requirement of 3.17.2, provide a second licensed operator in the control room within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . With both inoperable, in lieu of following the requirements of 3.17.2, start and maintain in' operation the turbine driven auxiliary feed pump.

l

.( f) Calculste the Quadrant ?cver '."ilt using the execre readings at least c ce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the excere detectors deviatic: ala:=s are L:cperable.

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3-81 a A endment No.,67, 68

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3.23 PCWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE (LHR)

LIMITING CONDITION FOR OPERATION The LHR in the peak power fuel rod at the peak power elevation Z shall not exceed the value in Table 3.23-1 times F ( ) times F ( ) I*

I"""*' "

A B

F (Z) is shown in Figure 3.23-1 and the function F (E) where E is the fuel A

B rod burnup is shown in Figure 3.23-2].

The LHR at the peak power elevation in any interior fuel rod or narrow water gap fuel rod shall not exceed the value in Taele 3.23-1 times F (Z) [the function F (Z) is shown C

C I

in Figure 3.23-3].

APPLICABILITY: Power operation above 50% of rated power.

ACTION 1:

When using the incore alarm system to monitor LHR, and with four or more coincident incore alarms, initiate within 15 minutes corrective action to reduce the LHR to within the limits and restore the incore readings to less than the alarm setpoints within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or failing this, be at less than 50% of rated power within the following 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACTION 2:

When using the excore monitoring system to monitor LHR and with the A0 deviating from the target A0 by more than 0.05, discontinue using the excore monitoring system for monitoring LHR.

If the incore alarm system is inoperable, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be at 85% (or less) of rated thermal power and follow the procedure in ACTION 3 below.

t 3-103 Amendment No. 68 I

P0kTR DISTRIBLTION LIMITS 3.23.1 LINEAR HEAT RATE (LHR) l LIMITING CONDITION FOR OPERATION ACTION 3:

If the incore alarm system is inoperable and the excore monitoring system is not being used, operation at less than or equal to 85% of rated power may continue provided that incere readings are recorded manually.

Readings shall be taken on a minimum of 10 individual detectors per quadrant (to include 50% of the total number of detectors in a 10-hour period) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and at least every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thereafter. If readings indicate a local power level equal to or greater than the alarm setpoints, the action specified in ACTION 1 above shall be taken.

Basis The limitation on LHR ensures that, in the event of a LOCA, the peak temperature of the cladding will not exceed 2200*F.II) In addition, the limitatien on LHR for the highest power fusi rod, narrow water gap fuel rod and interior fuel rod ensures that the minimum DNER will be maintained above 1.30 during anticipated transients; and, that fuel damage during Condition IV events such as locked rotor will not exceed acceptable i

limits.( )( ) The inclusien of the axial power distribution term ensures that the operating power distribution is enveloped by the design power distributions.

Either of the two core power distribution monitoring systems (the incore l

1 alarm system or the excore monitoring system) provides adequate monitoring of the core pcwer distribution and is capable of verifying that the LHR I

does not exceed its limits. The incere alarm system performs this 3-104 Amendment No. 68

- - - - -. - _ _ = _

l l

POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE (LHR) i LIMITING CONDITION TOR OPERATION Basis (Contd) function by continuously monitoring the local power at many points throughout the core and comparing the measurements to predetermined setpoints above which the limit on LHR could be exceeded. The excore nonitoring system performs this function by providing comparison of the measured core A0 with predetermined A0 limits based on incere I

measurements. An Excore Monitoring Allowable Power Level (APL), which may be less than rated power, is applied when using the excore monitoring a

system to ensure that the A0 ilmits adequately restrict the LKR to less 4

than the Itaiting values.I')

~

i If the incore alarm system and the excore monitoring system are both inoperable, power will be reduced to provide margin between the actual peak LHR and the* LHR limits and the incors readings will be manually collected at the terminal blocks in the control room. utilizing suitable signal detector. If this is not feasible with the manpower available, the l

i l

reactor power will be reduced to a point below which it is improbable that l

l tho'LHR limits could be exceeded. The time interval of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the core power distribution to detect significant changes until the monitoring systems are returned to service.

{

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To ensure that the design margin of safety is maintained, the

(

determination of both the incere alarm setpoints and the APL takes into account a measurement uncertainty-factor of 1.10, an engineering I

_ 3-105~

Amendment No. 68 t

P0bTR DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATES (IRR)

LIMITING CONDITIONS OT OPERATION Basis (Contd) uncertainty factor of 1.03, a thermal power measurement uncertainty factor i

i of 1.02 and allowance for quadrant tilt.

I References (1) XN-ST-77-24 i

j (2) XN-NT-77-18 (3) XN-ST-78-16 (4) XN-hT-80 -47 i

i i

i I

3-106 Amendment No. 68

TABLE 3.23-1 LINEAR HEAT RATE LIMITS Fuel Rod Type No of Fuel Rods in Assembly 208 216 Peak Rod 15.28 14.72 Narrow k*ater Gap Rod 15.12 14.47 Inter.or Rod 14.17 13.89 TABLE 3.23-2 RADIAL PEAKING FACTOR LIMITS, Fg Peaking Factor No of Fuel Rods in Assembly 208 216 Assembly

[

1.43 1.45 r

Ff 1.77 1.77 Peak Rod Narrow Gap Rod F 1.75 1.74 A

Interior Rod F

1.64 1.67 r

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3-110 Amendment No. 68 e

o POWER DISTRIBLTION LIMITS g

3.23.2. RADIAL PEAKING' FACTORS

+

LIMITING CONDITION FOR OFERATION TheradialpeakingfactorsF,Ff'[r andFfshallbelessthanorequalto j-A r

the value in Table 3.23-2 times the quantity [1.0 + 0.5(1-P)]'where P is'the core thermal power in fraction of rated power.

APPLICABILITY: Power operation above 50*. of rated power.

i ACTION:

With any radial peaking factor exceeding its' limit within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , reduce f

l' thermal power to-less than the lowest value of:

F 1

[1-2(f-1)]xRated' Power l.

L

-j A

T N

Where F is the measured value of.either F, p, p or F and F is the r

r r

L corresponding limit from Table 3.23-2.

j Basis i

A T

The limitations on F, F, F and [ are provided to. ensure that assumptions r

r r

r used in the analysis for establishing DNB margin, LHR and the thermal margin /

low-pressure and high-power trip setpoints remain valid during operatieru 1.

Data from the incere detectors are used for determining the measured radial l

peaking factors. The periodic surveillance requirements for determining the f

measured radial peaking factors provide assurance that they remain withir.

prescribed limits. Determining the measured radial peaking factors after each j

fuel loading prior to exceeding 50*. of rated power provides additional assurance that the core is properly loaded.

I 3-111 Amendment No. 68

POWER DIS'mIBLTION LIMITS 3.23.3 QUADRANT POWER TILT - T_

LIMITING CONDITION FOR Ol'ERATION The quadrant power tilt (r ) shall not exceed 5%.

APPLICABILITY: Power operation above 50% of rated power.

ACTION:

.1.

With the quadrant power tilt determined to exceed 5% but less than or equal to 10%, correct the power tilt within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or determine within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and at least once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, that the radial peaking factors are within the limits of Section 3.23.2, or reduce power at the normal shutdown rate to less than 85% of rated power.

2.

With the quadrant power tilt determined to exceed 10%, correct the quadrant power tilt within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce power to less than 50% of rated power within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

3.

With the quadrant power tilt determined to exceed 15%, be in at least hot standby within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

lp Basis i

Limitations on quadrant power tilt are provided to ensure that design safety

{i margins are maintained. Quadrant power tilt is determined from excore

{

detector readings which are calibrated using incere detector measurements.(1)

Calibration factors are determined from incore measurements by performing a -

j two-dimensional, full-core surface fit of deviations between measured and i

theoretical incere readings and integrating the fitting function over each core quadrant. Values of LER and radial peaking factors are increased by the value of quadrant tilt.

3-112.

Amendment No. 68

FCWER DISTRIBUTION LIMITS i

3.23.3 QUADRANT PCWER TILT - Tq LIMITING CONDITION FOR OPERATION i

References 1

(1) FSAR, Section 7.4.2.2

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~~ 3-113 Amendment No. 68

.~

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

c 4.18 PCVER DISTRIBLTION INSTRUMENTATION 4.18.1 INCORE DETECTORS SURVEII. LANCE REQUIREMENTS i

4.18.1.1 The incore detection system shall be demonstrated operable:

i a.

By performance of a Channel Check prior to its use following 1

a core alteration and at least once per 7 days during power j

4 operation when required for the functions-listed in 1

Section 3.11.1.

t i

b.

At least once per refueling by performance of a Channel

'-l Calibration which exeepts the ne'utron detectors but includes i

electronic components.

{

f 4.18.1.2 The incere alarm system is demonstrated operable through use of the datalogger program out-of-sequence alarm. The out-of-sequence alarm is demonstrated operable once per refueling by per,formance of a Channel Check.

I 4

l

^ 4-81 Amendment No. 68 I

4 POWER DISTRIELTICN INSTRL?!ESTATION 4.18.2 EXCORE MONITORING SYSTEM SURVEImNCE REQUIREMENTS 4.18.2.1 At least every 31 days of power operation:

A target A0 and excore monitoring allowable power level a.

shall be determined using excore and incere detector readings at steady state near equilibrium conditions.

b.

The excore measured A0 shall be compared to the incore measured AO.

If the difference is greater than 0.02, the excore monitoring system shall be recalibrated.

The excore measured Quadrant Power Tilt shall be compared to c.

the incore measured Quadrant Power Tilt. If the difference is greater than 2%, the excore monitoring system shall be recalibrated.

l i

i l

4-82 Amendment No. 68

w -- -

4.19 POWER DISTRIBIRION LIMITS 4.19.1 LINEAR HEAT RATES SURVEILLANCE REQUIRE.TSTS 4.19.1.1 When using the incore alarm system to monitor LHR, prior to operation above 50% of rated power and every 7 days of power operation thereafter, incere alarms shall be set based on a measured power distribution.

4.19.1.2 When using the excere monit ring system to monitor LHR:

a.

Prior to use, verify that the measured A0 has not deviated i

from the target A0 by more than 0.05 in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

Once per day, verify that the measured Quadrant Power Tilt is less than or-equal to 3%.

c.

Once per hour, verify that the power is less than er equal to the APL and not more than 10% of rated power greater than the power level used in determining the APL.

d.

Once per hour, verify that the measured A0 is'within 0.05 of the established target AO.

r i

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I

. 4-83 Amendment No. 68

6 4.19 POWER DISTRIBLTION LIMITS 4.19.2 RADIAL PEAKING FACTORS SURVEILLANCE REQUIREMENTS The measured radial peaking factors (d. F, [ and [)

4.19.2.1 r

r r

r obtained by using the incere detect 2cn system, shall be deter:nined to be less than or equal to the values stated in the i

LCO at the following intervals:

a.

After each fuel loading prior to operation above 50*. of rated power, and b.

At least once per week of power operation.

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l 4-84 Amendment No'. 68

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I 6.3.3 The Shift Technical Advisor (STA) shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients.and accidents.

6.4 TRAINING

6. 4.1 A retraining and replacement training program for the plant staff shall be maintained under the direction of the Nuclear Training.

Administrator and shall meet or exceed the requirements and recom-mendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR, Part 55.

6.4.2 A training program for the fire brigade shall be maintained under the direction of the Plant Training Coordinator and shall, as practical, meet or exceed the requirements of Section 27 of the NFPA Code.

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i 6

e (Correction) 6-la' Acendment No. If, J7,,jH7 68

. p ocoq'o, UNITED STATES

+!_,

NUCLEAR REGULATORY COMMISSION s.,

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.. p WASHINGTON, D. C. 20555

/

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 68 T0' PROVISIONAL OPERATING LICENSE NO. 20 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET NO. 50-255 l

l-l l

Date:

December 8,1981 l

l

(

TABLE OF CONTENTS SECTION SECTION TITLE NUMBER PAGE INTRODUCTION ---------------------------------------------------1.0 1

DISCUSSION ---------------------------------------------------- 2.0 1

NOT AT I O N - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- - - - - - - - - - - - - - - - - - - - - - 3. 0 2

-Acronyms ---------------------------------------------------- 3.1-2 Notation Specific to this Report ---------------------------- 3.2 3

TS CHANGES ---------------------------------------------------- 4.0 4

Motivation for Making TS Changes ---------------------------- 4.1 4

-Evaluation of TS Changes ------------------------------------ 4.2 5-Findings of Review of TS Changes ---------------------------- 4.3

--- 10 CYCLE 5 FUEL DESIGN ------------------------------------------- 5.0

--- 10 Fuel Mechanical Design -------------------------------------- 5.1

--- 10 Cladding Creep Collapse ------------------------------------- 5.2

--- 13 Fission Gas Release ------------------------------------------5.3

--- 13 Cycle 4 Fuel Failures --------------------------------------- 5.4

--- 13 Gadolinium Fuel Demonstration Program ----------------------- 5.5

--- 14 Advantageous Properties of Gadolinium as a BP ------------- 5.5.1 '--- 14 Neutronic Computations With Gadolinium Fuel --------------- 5.5.2 --- 15 Other Properties of Gadolinium Fuel ----------------------- 5.5.3 --- 16 Approval of Gadolinium Fuel Program '----------------------- 5. 5. 4 --- 16 BASIS FOR COMPARING CYCLE X PARAMETERS WITH REFERENCE ANALYSIS PARAMETERS:

BRIEF REVIEW OF PERTINENT TOPICAL R E PO R T S - - - - --- - - - - - -- - - - - -- -- -- -- - - - - - - - - - - - - - - - - - - - - - - - -- -- - 6. 0

--- 16 Original 2530 MWT DBE Analyses ------------------------------ 6.1

16 TS LHR Limitation Prior to Cycle 3 and Resolution of This Limitation -------------------------------------------- 6.2

--- 19 NewTSLHRLimi$s: TS Figure 3.23-3:

Transient Analyses with ;;/L 150%; TS Figure 3.23-1:

LOCA Ana lys e s wi th ::/L 1 6 0% ------------------------------------ 6. 3

--- 21 i

Reference 2530 MWT Transient Analyses ----------------------- 6.4

--- 24 i

e TABLE OF CONTENTS (continued)

SECTION SECTION TITLE NUMBER PAGE Computation of Assemblywise Pin Radial Peaking p

Factor, (F 9, For Cycle X ---------------------------------- 6.5

--- 25 r

Criteria For Determining Which Transient Events Require Reanalysis for Cycle X ----------------------------- 6.6

--- 26 Incongruity Between Transient Analysis Input Peaking Factors and TS Peaking Factors --------------------- 6.7

--- 28 A

Anomalous Values of F in the Cycle 5 Safety Report --------- 6.8

--- 30 r

Reference LOCA Analysis ------------------------------------- 6.9

--- 31 Fue l Expo s u r e Sens i ti v i ty ----------------------------------- 6.10

--- 32 LOCA Analysis Inputs ---------------------------------------- 6.11

--- 33 Cycle 4 Batch H Fuel LOCA Analysis -------------------------- 6.12

--- 34 Incongruity Between LOCA Analysis Input Peaking Factors and TS Peaki ng Factors ----------------------------- 6.13

--- 34 ANALYSIS OF DBEs OTHER THAN LOCA ------------------------------ 7. 0

--- 36 Uncontrolled Rod Withdrawal -------------------------~-------- 7.1-4

--- 46 Control Rod Drop -------------------------------------------- 7.5-6

--- 46 Four Pump Coastdown ----------------------------------------- 7.7

--- 47 Locked Rotor ------------------------------------------------ 7.8

--- 47 Reduct, ion in Feedwater Enthalpy ----------------------------- 7.9-10 --- 48 Increased Feedwater Flow ------------------------------------ 7.11

--- 48 Excessive Load ---------------------------------------------- 7.12-13--- 49 Loss of Load ------------------------------------------------ 7.14-15--- 49 Loss of Feedwater ------------------------------------------- 7.16

--- 49 Steam Line Break -------------------------------------------- 7.17-18--- 49 Single Rod Withdrawal --------------------------------------- 7.19-20--- 51 Rod Ejection ------------------------------------------------ 7.21-22--- 51 LOCA ANALYSIS ------------------------------------------------- 8.0

--- 52 Cycle 5 LOCA Analysis Input Parameters Compared with Reference LOCA Analysis Input Parameters ------------------- 8.1

--- 52 Features of Batch I Fuel Which in Cycle 5 Facilitate Meeting the Power Peaking TS Criteria Based on the Re f e rence LOCA Anal ys i s ------------------------------------ 8. 2

--- 53 Two. Types of Batch I Fuel --------------------------------- 8.2.1

--- 53 Pinwise Power Flattening Within Batch I Assemblies -------- 8.2.2

--- 53 11

TABLE OF CONTENTS (Continued)

SECTION SECTION TITLE NUMBER PAGE Pinwise Power Flattening Within Spare Rod Batch I Assemblies ----------------------------------------------- 8.2.3

--- 54 Calculated LOCA Power Peaking Factors and TS LOCA Powe r Pea k i ng Facto rs ------------------------------------ 8. 2. 4 --- 54 Conservatisms in Cycle 5 Over Reference LOCA Analysis ------- 8.3

--- 54 Larger Clad Diameter Pins in Batch I Fuel ----------------- 8.3.1 --- 54 A

Lower TS {F 3 in Cycle 5 ---------------------------------- 8.3.2

--- 55 r

STARTUP PHYSICS TESTING PROGRAM FOR CYCLE 5 ------------------- 9.0

--- 55 ENVIRONMENTAL IMPACT OF CYCLE 5 ------------------------------- 10. 0

--- 56 CONCLUSION ---------------------------------------------------- 11.0

--- 56

~ REFERENCES ---------------------------------------------------- 12.0

--- 56 I

iff

_

ML20039D554

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