ML18044A894

Tech Spec Change Request Re Tech Specs 3.10 & 3.11 for Control Rod & Power Distribution Limits & in-core Instrumentation

ML18044A894
Person / Time
Site:	Palisades
Issue date:	05/14/1980
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
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ML18044A892	List: ML18044A891 ML18044A893 ML18044A894 ML19323C637
References
NUDOCS 8005160421
Download: ML18044A894 (28)

Text

ATTACHMENT Page Changes for Proposed Technical Specifications Change 8 0051 60,2.. \\

,, 3.10.L CONTROL ROD AND POWER DISTRIBUTION LIMITS (Contd) 3.10.~ Power Distribution Limits (Contd) satisfy the criterion.

Appropriate consideration shall be given to the following factors:

(1)

A flux peaking augmentation factor of 1.0, (2)

A measurement calculational uncertainty factor of 1.10, (3)

An engineering uncertainty factor (which includes fuel column shortening due to densification.and thermal expansion) of.1.03, and

( 4)

A the.rmal power measurement uncertainty factor of 1. 02.

b.

If the quadrant to core av~rage power tilt exceeds 15%, except for physics tests, then:

(1)

The linear heat generation rate shall promptly be demonstrated to be less than that specified in.Part a, or (2)

Immedi.ate action shall be initiated to reduce reactor power to 75%

or less of rated power.

c.

If the power in a quadrant exce_eds core average by 10% for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or if the power in a quadrant exceeds core average by 20% at any time, immediate action shall be initiated to reduce reactor power below 50% until the* situation is remedied.

d.

If the power in a quadrant exceeds-the core average by 15%.and if the linear heat generation rate cannot be demonstrated promptly to be within limits, then the overpower trip set point shall be reduced to 80% and the thermal margin low-pressure trip set point (PT. ) shall be rip increased by 400 psi.

e.

If the power in a quadrant exceeds core average by 5% for a period of 30 days, immediate action shall be initiated to reduce reactor power to 75%

or less of rated power.

f.

The part-length control rods will be completely withdrawn from the core (except for rod exercises and physics tests).

A

. g.

The calculated value of Fr.sl).all be limited to~ 1.45 (1.0 + 0.5 T*

(1 - P)), the calculated value of F shall be limited to~ 1.77 (1.0 +

r 0.5 (1 - P)), and the calculated value of F 6H shall be limited to r

< 1.66 (1.0 + 0.5 (1 - P)), where Pis the core thermal power in fraction of core rated thermal power (2530 MWt).

T

(*For the duration of Cycle-4 for H-fuel only, F for rods adjacent to r

the wide water gap shall be limited to 1.90 (1.0 + 0.5 (1 - P)).)

3-59

IN-CORE INSTRUNENTATION (Contd)

• .Specification (Contd)

. g.

a 10-hour peiiod) at least each t~o hours thereafter or the reactor power level shall be reduced to le*ss than 50~ of rated power (65% of rated power if no dropped or misaligned rods are present). If readings indicate a.local power level equal to or greater than the alarm set point, the action specified in 3.11.b shall be taken.

FA, FT and F AH shall be determined whenever the core power r

r r

distribution is evaluated.

If either Fr A' FrT or Frt:.H is.fourid to be in excess of the limit specified in Section 3.10. 3 (g), within six hours thermal power shall be reduced to less than that required to assure compliance. with. Section 3.10. 3 (g).

Basis A system of 45 in-core flux detector and thermocouple assemblies and a data display, alarm and record functions has been provided.

A four level, five level or six level system may be used. (l)(2)

The out-of-core nuclear instrumentat.ion. calibration includes:

a.

Calibration (axial and azimuthal) of the split detectors at initial reactor start-up and during the power escalation program.

b.

A comparison check with the in-core instrumentation in the event abnormal readings are observed on the out-of-core detectors during operation.

c.

Calibration check.during subsequent reactor start-ups.*

d.

Confirm that readings from the out-of-core split detectors are as expected.

Core power distribution verification includes:

a.

Measurement at initial reactor start-up to check that power distribution is consistent with calculations.

b.

Subsequent. checks during operation to insure that J:>OWer distribution is consistent with calculations.

c.

Indication of power distribution in the event that abnormal situations.

occur during reactor operation.

If the data logger for the in~core readout is not in operation for more than two hours, power will be reduced to provide margin between the actual peak linear heat generation rates and the limit and the in-core readings will be manually collected at the te.rminal blocks in the control room utilizing a suitable signal detector.

If this is not feasible with the 3-66

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XN*Nf *80*18(NP)

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tees. AND THERMAL-HYDRAULIC ANALYSIS FOR -THE PALISADES RELOAD H DESIGN APRIL 1980 RICHLAND, WA 99352 ElS{ON NUCLEAR COMPANY, Inc.

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/mb ECCS AND THERMAL-HYDRAULIC ANALYSIS FOR THE PALISADES RELOAD H DESIGN Prepared By M. J. Ades J. C. Cherng XN-NF-80-18 (NP)

Issue Date:

04/15 /80 EJ){ON NUCLEAR COMPANY, Inc.

NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was <lerived through research and development programs sponsored by Exxon Nuclear Company, Inc.

It is being sub-mitted by Exxon Nuclear to. the USN RC as part of a technical contri-

• bu ti on to facilitate safety analyses by. licensees of the USN RC which utilize Exxon Nuclear-fabricated. reloa<l fuel or other technical services provided by Exxon Nuclear for liaht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief.

The information contained herein may be used by the USN RC in its review of this report, and by licensees or applicants before the USN RC which are customers of Exxon Nuclear in their demonstration of comoliance with the USNRC's regulations.

Without derogating from the foregoing, neither Exxon Nuclear nor any oerson acting nn its behalf:

A.

Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method,.or process disclosed in this document will not infringe privately owned rights; or B.

Assumes any liabilities with respect to the use of, or for darrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.

XN-NF-FOO, 766

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-i-XN-NF-80-18 (NP)

TABLE OF CONTENTS Section

1. 0 INTRODUCTION ANO

SUMMARY

................ 1 2.0 ECCS ANALYSIS.......*.......... ~... ~. 3 3.0 THERMAL-HYDRAULIC ANALYSIS

4.0 REFERENCES

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-ii-LIST OF TABLES Table l. l Design Differences between Reload E/G and H.. * * * * * * *

• 2.1 Palisades H-fuel ECCS Analysis Parameters 2.2 Heatup Analyses Results for the H-fuel Design 2.3 Comparison of Peak Clad Temperature Results for

3. 1 E/G and H-fuel Designs..... ~.......

Thermal-Hydraulic Design Conditions and Performance............

XN-NF-80-18 (NP)

Page 2

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-iii-LIST OF FIGURES Figure 2.1 Cladding Surface Temperature During Heatup for H-fuel. (Axial Peak Power Location X/L=0.6, Skewing Factor= 1.0).......... 10 2.2 Cladding Surface Temperature During Heatup for H-fuel (Axial Peak Power Loc~tion X/L=O. 8, Skewing Factor = 1.1)

............... 11 XN-NF-80-18 (NP)

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1 XN~NF-80-18 (NP) 1. 0 INTRODUCTION AND

SUMMARY

This report documents LOCA/ECCS and thermal-hydraulic analyses for the Exxon Nuclear Company (ENC) Batch. H fuel design for the Palisades nuclear power plant. These analyses support a maximum allowable linear heat generation rate limit of 15.28 Kw/ft for the Palisades H design, which is the same as previously established for the ENC E/G design(l).

The allowable assembly radial peaking factor of 1.45.at full core power (2530 MWt) is also unchanged from the analyses for the E/G design.

Increased local peaking for the wide gap edge rods in the Palisades H design has been considered. Thus for the present H-fuel analysis, a wide gap cor~er ~od local peaking factor df.l.31 is incorporated i~ the analyses versus a wide gap corner rod local peaking of 1.. 22 in the E/G analyses.

Limits on interior rod local peaking remain unchanged.

The mechanical design differences between the Palisades H design and the prior E/G design are shown in Table 1.1.

In addition to the mechanical design differences of T~ble 1.1, the neutronics design of the Palisades H reload fuel is slightly different from the E/G reload fuel (e.g. fuel rod enrichments, poison rods.) The present LOCA/ECCS and thermal-hydraulic analyses account for the mechanical design differences, as well as the increased local peaking for the wide gap rods in the H-fuel design.

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XN-NF-80-18 {NP)

Table l. l

• Design Differences between Reloads E/G and H Design Component Fuel Pellet Pellet Diamet~r (in.)

Dish Volume (%)

Cladding Outside Diameter (in.)

Fuel Rod He Fill Pressure (psia)

Reload E/G

.3505 2.0

.415 Reload H

.35 1.0

.417

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3 XN-NF-80-18 (NP) 2.0 ECCS ANALYSIS

2. l LOCA PERFORMANCE

SUMMARY

The perfonnance of the Palisades H fuel design in a postulated loss-of-coolant accident (LOCA) has been evaluated.

• The nominal total peaking FQ limit of 2.76 that previous ECCS analyses had detennined for ENC Batch E/G fuel has been applied to the H-fuel design, and has been found to be acceptable relative to 10 CFR 50.46 criteria. This FQ limit corresponds to a maximum allowed linear heat generation rate (LHGR) of 15.28 Kw/ft for the Palisades H design.

The calculations were performed for the Palisades 0.6 DEG/PD limiting break(2).

• The calculations account for the me.chanical design *differences from E/G fuel noted in Table 1.1, and incorporate an ECCS limiting rod local peaking factor of 1. 31.

An axial power peak location sensitivity calculation was also performed.

This calculation confinned the continued applicability of the ECCS allowable LHGR as a function of axial power peak location established previously for E/G fuel (1). *The present analysis is sufficient for pellet exposures.of approximately 30,000 MWD/MTM for the Palisades H design fuel.

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4 XN-NF-ao..:1a (NP) 2.2 MODELS AND ASSUMPTIONS RELAP4-EM/HOT CHANNEL and TOODEE2 heatup analyses were made to determine the performance of Batch H fuel during a LOCA.

The HOT CHANNEL and TOODEE2 calculations require the following boundary conditions:

The upper and lower plenum fluid conditions {pressure and enthalpy) versus time during blowdown.

Normalized power versus time.

e The EOBY and BOCREC event times.

Reflood rate and saturation temperature versus time during the refill and reflood time periods.

ECCS subcooling during reflood.

These boundary conditions are provided by the Palisades, 2530 MWt core

. power, 0.6 DEG/PD limiting break calculation for _E:NC Batch E/G fuel(3)

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.. l As in the previous break spectrum analysis for Palisades e ax1a location relative to the bottom of the active core for the reference case in the present analysis is at X/L=0.6.

The present axial power power peak peak location sensitivity study also considered the most limiting case of the previous. E/G-fuel axial sensitivity study(l) where the maximum LHGR has been reduced by 16% and peak*ed at X/L=0.8, and the axial power distribution has a 1.1 skewing factor.

The key parameters used* in the analysis are summarized in Table 2.1.

With the exception that the limiting rod local peaking has been increased from 1.22 to 1.31, the peaking factors identified in Table 2.1 are the same as i dent i fi ed in Reference 1.

• The analysis is i ri accordance with

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ENC's WREM-II PWR ECCS Evaluatio~ Model( 2' 4' 5). The HOTCHANNEL calculations were performed with ENC's Version 26A of the RELAP4 code as have pri8r analyses for Palisades.

The current MAY79 TOODEE2 code was used to calculate the limiting rod heatup transient.

Fu~l parameters in the present analysis correspond to beginning-of-1 ife conditions.

The recent Nuclear Regulatory Commission (NRC) Clad Swelling and Rupture Model(G) has been applied in the present TOODEE2 heatup calculations for the Palisade~.H design.

The present.analysis includes consirleration of uncertainties in the hoop stress at the time of clad rupture in the application of the NRC model.

This is done by consideration of rod internal pressure uncertainties in accordance with t~a models det~iled in References 7, 8, 9, and as approved in Reference 10.

2.3 ANALYSIS RESULTS The results of the final TOODEE2 heatup calculations are given in Table 2.2.

The upper bound on fuel rod internal pressure uncertainties has been incorporated into these heatup results. Table 2.2 shows that a margin exists between the calculated PCT and the limiting PCT of 2200°F for both axial peaking locations at X/L=0.6 and X/L=0.8. The corresponding limiting rod clad temperature heatup transients as calculated using TOODEE2 are given in Figure$ 2. l and 2.2 respectively.

Table 2.3 shows a comparison of calculated peak clad temperature (PCT) results.for E/G fuel in Reference l with the present results for the H-fuel design.

In comparing these results, it is noted that the PCT's calculated for E/G fuel were determined using the ENC clad swelling and

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XN-NF-80-18 {NP) rupture model, while those for H-fuel used the NRC model(6).

In addition, the ~-fuel design has a slightly larger fuel rod diameter than that of the E/G designs.

In spite of these design and analysis differences, the calculated PCT 1 s for both designs are approximately equivalent.

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7 Table 2. 1 Palisades H-fuel ECCS Analysis Parameters Reactor Power at 102%, MWt Reactor Pressure, psia Heat Release in Fuel Limiting Break Hot Assembly Radial Peaking Hot Rod Local Peaking, Fi Engineering Factor Reference' Case X/L Skewing Factor FQ Top Skewed Axial Power Profile Case X/L Skewing Factor F.

Q 2580.6 2060.

97.5%

0.6 DEG/PD

.1. 45

1. 31

1. 03 0.6 1.0 2.76 0.8

1. 1 2.32 XN-NF-80-18 (NP)

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8 Table. 2.2 Heatup Analyse~ Results for the H-Fuel Design Axial Power Peak Location, X/l Skewing Factor Total Peaking~ fQ Peak Clad Temperature (PCT), °F Max. Local Zr/H20 Reaction, I Hot Rod Burst Time, sec.

Hot Rod Burst Location, ft.

Rupture Pressure, psid Flow Blockage, %

Time of PCT' S5!C.

PCT Location, ft.

Max. Zr/H20 Reaction location, ft.

Heatup Rate at Rupture - 0c;sec.

0.6 1.0 2.76 2067.

6.8 0.8

1. 1 2.32 2183.

11. 1 XN-NF-80-18 (NP)

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Axial Power Peak Location, X/l 0.6 0:8

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9 XN-NF-80-18 (NP)

Comparison of Peak Clad Temperature Results for E/G and H-Fuel Designs Skewing Factor G-Fuel H-fuel Ft=l.22 F i=L 31 1.0 2081 2067

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r 12 XN-NF-80-18 (NP) 3.0 THERMAL-HYDRAULIC ANALYSIS The ENC Reload H fuel is designed to be compatibl~ with the Palisades Reactor core and with the existing fuel.

The thermal-:hydraulic design criteria for ENC reload fuel at Palisades are:

The maximum fuel temperature at 115% overpower shall not exceed the fuel melting temperature.

The minimum DNBR shall be greater than or equal to 1.30 at 115% of rated power based on the W-3 correlation (or an accepted equivalent) plus correction factors which have been accepted by the NRC for the purpose of licensing the fuel design described herein.

The cladding temperature at nominal operating conditions (based on crud-free surface) shall be less than:

850°F internal surface 675°F external surface 750°F volume average (local)

The fuel assemblies must be thermally and hydraulically compatible with the existing fuel and the reactor core during the design*

life of the fuel.

The thermal-hydraulic analysis for the Palisades H fuel was performed in a manner consistent with conditions reported in licensing data provided for operation of the Palisades plant at 2530 MWt(l,ll).

The thermal-hydraulic design conditions for this analysis are shown in Table 3.1, which reflects the increased local peaking for the wide gap edge rods.

The resulting performance of Palisades H fuel falls within the thermal-hydraulic design criteria. Therefore, the insertion of the Batch H reload fuel into the Palisades reactor is acceptable.

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13 XN-NF-80-18 TABLE 3.1 THERMAL-HYDRAULIC DESIGN CONDITIONS AND PEFORMANCE Reactor Conditions Core power ( MWt)

Total reactor flow rate (Mlb/hr)

Active core flow rate (Mlb/hr)

Coolant inlet temperature (°F)

Core pressure (psia)

Power Distribution Assembly radial peaking, FR Pin power peaking (for interior rods), FRxFt Pin power peaking (for narrow gap edge rods), FRxFt Pin power peaking Axial peaking, F a (for wide gap edge rods), FRxFt (at 0.6 of active fuel height)

Engineering factor, F e

Total peaking factor Thermal-Hydrauli~ Performance Hot assembly flow factor MDNBR Fuel center temperature (°F)*

Clad outer surface temperature Clad inner surface temperature (oF)**

(oF)**

Volumetric averaged clad temperature (°F)**

At 115% of rated power.

,, At nomi na 1 power.

Design 2910 121. 7 114.4 542.5 2010,

Use, reproduction, transmittal or disclosure of the above information is subject to the restriction on the first or title page of this document.

Nominal 2530 121. 7 114.4 537.5 2060 l.45 l.66 l.77

l. 90 l.41 l.03 2.76 0.97

> 1. 3 4767 668 803 736

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14 XN-NF-80-18 (NP)

4.0 REFERENCES

1. - Exxon Nuclear Comapny, Analysts of Axial Powe'r Distribution Limits for the Palisade_s Nuclear Reactor at 2530 MWt~

• XN-tff-78:-16, June 9, 1978.

2.

Exxon_Nuclear Company, Exxon Nuclear Company WREM-Based Gerieri~

PWR ECCS Evaluation Model Update ENC WREM-Il, XN:-76-27:

a.

July 1976

b.

Supplement l, September 1976

c.

Supplement 2, November 1976

3.

Exxon Nuclear Company, LOCA Analysis for Palisades at 2530 MWt using the ENC WREM-II PWR ECCS Evaluation Model~ XN-NF-77-24, July 19, 1977.

4.

Exxon Nuclear Company,- Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model, XN-75~41:

a

• Vo 1 ume I,

Ju 1 y 1 9 7 5

b.

Volume II, August 1975

c.

Volume III, Revision 2, August 1975

d.

Supplement 1, August 1975

e.

Supplement 2, August 1975

f.

Supplement 3, August 1975

g.

Supplement 4, August*1975

h.

Supplement 5, Revision 1, October 1975

i.

Supplement 6, October.1975 j.-

Supplement 7, November 1975

• 5.

Exxon Nuclear Company, Revised Nucleate Boiling Lockout for ENC WREM-Based ECCS Evaluation.Models, XN-76-44, September 1976.

6.

D.A. Powers and R.O. Meyers, "Cladding Swelling and Rupture Models for LOCA Analysis, Draft NUREG-0630, February 12, 1980.

7.

Exxon Nuclea~ Company~ Flow Blockage and Exposure Sensitivity For D. C. Cook Unit l Reload Fuel Using ENC WREM-11 Model,.

XN-76-51, Supplement l,. January 1977.

8.

Exxon N~clear Company, Flbw'Blockage and Exposure Sensitivity Study for D.C. Cook Unit l Reload Fuel Us in ENC ~~REM-II Model, XN-76-51, S_upplement 2, January 978 proprietary

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I 15 XN-NF-80-18 (NP) 4.0 *REFERENCES (contd.)

9.

Exxon Nuclear Company, Flow Blockage and Exposure Sensitivity Study for D. C. Cook Unit Reload Fuel Using ENC WREM-.11 Model XN~NF-76-51, Supplement 3, April 1978 {proprietary).

10~ Supplement l to Safety Evaluation Report on the Exxon Nuclear Company WREM-Based Generic PWR-ECCS Evaluation Model Update ENC-WREM-11 for Conformance to Requirements of Appendix K *to 10 CFR 50 by the Office of Nuclear Reactor Regulation, May 1978.

Tl.

Exxon Nuclear Comp.any, Steady-:State Thermal-Hydraulic and Neutronics Analysis Of th~ Palisades R~actor for Ooeration at 2530 MWt, XN-NF-77-22, July 1977.

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16 ECCS AND THERMAL-HYDRAULIC ANALYSES FOR.

THE PALISADES RELOAD H DESIGN Distribution M. J. Ades J. c. Cherng J. N. Morgan G. F. Owsley Consumers Power Company (-40.)

Document Control (5)

XN-NF-80-18 (NP) 04/lS/80

,.. J' e

STATE OF Washington COUNTY OF Benton

)

)

e A F *'F *I D A V I T SS.

I, James N. Morgan, being duly.sworn,*hereby say and de.pose:

l.

I am Manag~r *.Licensing a*n*d Safety Engineering, for Exxon Nuclear Company, Inc., (

11 ENC") and as such I am-authorized to execute this Affidavit.

2.

I am familiar with ENC 1 s deta i 1 ed document control sys tern and policies which govern the protection and control of information.

3.

I am fa~iliar with the doc~ment XN-NF-80-lB(P), entitled 11 ECCS and Thermal-Hydraulic Analyses for the Palisades Reload H Design, 11 referred to as "Document".

Informati6n contained in this Document has been classi~~ed by ENC as ~ropriet~.ry i~ acc,0 1

r.~~n-~e with the cont.roi' system and

-* f.

policies establ:is.hed by ENC.fort.he control. and protection of information.

.

• 4... The Document conta*i ns information of a proprietary and

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qrnfidential nature and. i.s of the. type-.c~st~marily held in confidence by ENC and not made available to the public.

Based on my experience, I am aware that other.companies regard information of the kind contained in the Document as being.pro pr) etary and tonfi denti a 1.

5.

The Document *has.been made available to the United States Nuclear Regulatory Commission in confidence, with the request that the information contained in the Document not be disclosed or divulged.

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6.

The Document contains information which is vital to a competitive advantage of ENC and would be helpful to competitors of ENC when competing with ENC,

7.

The information contained in the Document is con.sidered to be proprietary by ENC because it reveals certain distinguishing aspects of design and safety~nalysis and fu~l design Which secure competitive economic advantage to ENC for fuel management and safety analysis opti-.

mi zati on and improved marketability, and includes. info~mation utilized by ENC in its bus:iness which affords ENC an opportuniti to obtain a competitive advantage over its compe~itors who do not or may not know or use the information c6ntained in ~he Document.

8.

The disclosure of the propri~tary information contained in the Document to a competitor would permit the competitor to reduce its expenditure of money and. manpower and to improve its competitive position by giving it extremely valuable.insights into ENC's design and safety analy~-is and fuel design procedures, and would result in substantial harm to the competitive position of ENC.

9.

The Document contains proprietary information which is held in confidence by ENC and.i.s not available in public sources.

10.

I~ accordance with ENC's policies governing the protection and control of information, proprietary information contained in the Document has been made a'vailable,- on a limited basis, to others outside ENC only a.s required and under suitable agreement providing for non-

.disclosure and limiied use of the information.

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• , 11. :. '.ENC *policy requ_ir~s th(lt** proprietary information be kept in a secured file or area and distributed on a need~to-know basis.

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Checks :are made routinely to'as*sure the *p'olicy procedures are being met.

12.

Th.is Document pr,ovides information which reveals fuel design and analyses methods.. devel6ped by ENC over the past several years.

ENC has investe.d.millions_ of do:llars_and many man-year.s of effort in related fu~l design arid. safety analysis 'method development... Assuming a competitor had available the same backg~ound data and incentives as ENC, the competit6r might, at a minimum, develop the information for the same expenditure of m~npower and ~oney a~ ENC.

13.

Ba'Sed on my experience in*the industry, I do not believe that the background data and iflcentives of ENC's competitors are s~f ficiently similar to the correiponding background data and incentives of ENC to reasonably expect such competitors. would be in a position to duplicate ENC's proprietary information contained in the Document.

THAT the statements made here.inabqve are, to the best of my knowl.edge, information, and belief, truthful and complet~.

FURTHER AFFIANT SAYETH NOT.

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