ML17305A263

Safety Evaluation Supporting Amend 44 to License NPF-41

ML17305A263
Person / Time
Site:	Palo Verde
Issue date:	09/19/1989
From:	Office of Nuclear Reactor Regulation
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ML17305A262	List: ML17305A261 ML17305A263
References
NUDOCS 8910040045
Download: ML17305A263 (52)

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+p*y4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 44 TO FACILITY OPERATING LICENSE NO. NPF-41 ARIZONA PUBLIC SERVICE COMPANY ET AL.

PALO VERDE NUCLEAR GENERATING STATION UNIT 1 DOCKET NO.

STN 50-528

1.0 INTRODUCTION

By letter dated January 12, 1988 (Ref. 1) the Arizona Public Service Company (APS) on beha 1f of itse1 f, the Sa 1t River Project Agricul tura 1 Improvement and Power District, Southern California Edison

Company, El Paso Electric Company, Public Service Company of New Mexico, Los Angeles Department of Mater and Power, and Southern California Public Power Authority ( licensees),

requested changes to the Technical Specifications for the Palo Verde Nuclear Generating Station, Units 1, (Appendix A to Facility Operating License No. NPF-41.

In support of both the Technical Specification changes and Cycle 3 operation, the licensees submitted a

Reload Analysis Report by letter dated January 18, 1989 (Ref. 2).

By letters dated April 19 and 26, June 27, August 25, and September 11, 1989 (Refs. 3, 4, 5, 27 and 28), the licensees also provided clarifying information on the Reload Analysis Report.

The staff's evaluation of the reload analysis is presented in Section 2.0 through 5.0 below.

The evaluation of the specific change to the Technical Specification is presented in Section 3.0 below.

2.0 EVALUATION OF FUEL DESIGN 2.1

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No changes in the fuel mechanical design basis have occurred in the fabrication of the Batch E fuel.

A modification to the poison rod assembly design was incorporated into the Batch E fuel to improve the burnup capability of the poison rods.

The poison rod assembly's overall length was increased to be of equal length with the fuel rods.

The increased length provides greater internal void volume which enables higher burnups with poison rods with higher B-10 loadings, while reducing end-of-life internal pressure.

The staff has found Reference 4 acceptable where clad collapse analyses are not necessary for new Combustion Engineering manufactured fuel because of the absence of gaps between pellets.

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We find the above change to be a minor improvement which does not affect the mechanical design basis and, thus is acceptable.

2.2 3~33 3

The Cycle 3 thermal performance evaluation was based on the performance of a composite fuel pin of fuel batches B, C, D, and E.

The evaluation was performed using the NRC approved code FATES3A (Refs.

5 through 8) and a power history enveloping the power and burnup levels representative of the peak pin at each burnup interval from the beginning of cycle to the end of burnup (Ref. 5).

The peak pin burnup analyzed is in excess of that expected at the end of Cycle 3.

Based on this analysis, the internal pressure in the most limiting fuel rod will be 1,149.8 psia which is far below the reactor coolant pressure of 2,250 psia.

This satisfies the SRP requirements and is acceptable.

2.3

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2.3.1

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The Cycle 3 core will consist of 1 Batch B assembly, 52 Batch C, 80 Batch D, and 108 Batch E (new) assemblies.

The Cycle 3 loading is low

leakage, using previously burned assemblies in the periphery.

Thus, most of the Batch E assemblies are located throughout the core interior.

The expected Cycle 3 lifetime is 475 effective full power days.

The highest Batch E enrichment is 4.03 w/o U-235 which is lower than the 4.05 w/o U-235 for which the Palo Yerde facilities have been approved for fuel storage.

Comparison of characteristic physics parameters for Cycle 3

and Cycle 2 (the reference cycle) shows that the two cycles vary little from each other, and therefore Cycle 3, is acceptable.

2.3.2 Power Distribution Calculated all-rods-out relative assembly power densities were provided for the beginning, middle and the end of cycle.

Relative assembly power densities for rodded configurations were also presented.

The rodded configurations are those allowed by the power dependent insertion limit at full power.

The nominal axial peaking factors are estimated to range from 1.23 to 1.12 at the beginning and end of Cycle 3, respectively.

Augmentation factors have been eliminated from this cycle as discussed in Reference 9.

The methodology for the physics and power distribution calculations is based on ROCS-DIT (with the MC module)'hich has been approved by the NRC (Refs. 10,11).

These calculations, which are based on approved

methods, are acceptable.

2.3.3 Control Re uirements The most restrictive value of the shutdown margin occurs at the end of cycle under hot zero power conditions.

The minimum shutdown margin required to control the reactivity transient resulting from a steam line break is 6.5C dealt-k/k.

This shutdown margin is assured as discussed in

paragraph 2.5.3.

In addition sufficient boration capability and control element assembly worth with a stuck control element assembly exist to meet these shutdown requirements.

These results were derived with approved methods and incorporated conservative assumptions, therefore, the results are acceptable.

2.4 Thermal-H draulic Desi n

Steady state thermal-hydraulic analyses for Cycle 3 were performed using the approved code TORC (Ref. 11), the Combustion Engineering CE-1 critical heat flux correlation (Ref. 12) and the CETOP code described in Reference 13.

The methodologies described in References 10-12 with the statistical combination of uncertainties (Ref.

14) the core protection

system, the core operating limit system and the DNBR value of 1.24 assures that at the 95/95 confidence/probabi lity level that the hot rod will not experience DNB.

The 1.24 value includes all applicable penalties, such as the rod bow for burnups to 30,000 MWD/MTU, the.Ol DNBR for the HID-1 grids and the penalties specified in the statistical combination of uncertainties (Ref. 15-17).

The rod bow value used in the analysis is 1.7X DNBR, for burnups up to 30,000 MWD/MTU.

For burnups higher than 30,000 MWD/MTU sufficient margin exists to offset the rod bow penalty due to lower radial power peaks in these higher burnup assemblies and rods,

hence, the rod bow penalty is adequate for all anticipated burnups.

We conclude that the thermal-hydraulic design analyses were performed using approved codes and accounted for all applicable penalties,

and, therefore, are acceptable.

2.5 Safet Anal ses (Non-LOCA)

The design basis events considered in this safety analysis are classified in two groups:

The anticipated operational occurrences (moderate frequency and infrequent events) and the limiting fault events i.e.,

postulated accidents.

All events were evaluated with respect to four criteria:

fuel performance (centerline melt), reactor coolant system

pressure, loss of shutdown margin and offsite dose.

All events were reevaluated to assure that they meet their respective criteria for Cycle 3.

The limiting events for each criterion and those not bounded by the Cycle 2 values were reanalyzed.

The analytical methodology for the reanalyses are the same as for Palo Verde Unit 1 Cycle 2.

All of the methodologies used have been reviewed and approved by the NRC.

The following list includes the code, the purpose for which it was used in the analyses and the reference:

Code CESEC-I II CETOP-D TORC CENPD-183 HERMITE

~Por ose Plant response to non-LOCA events Hot channel and DNBR Pin DNBR and RCP shaft seizure Loss-of-f low methodo 1 ogy ana lysi s Core simulation for space-time kinetics Ref.

18 13 11, 19 20 21

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The input parameters for the analyses were comparable to those for the reference cycle.

Whenever the core protection system trip was evoked in the sequence the instrument channel response times assumed were conservative relative to the Cycle 3 Technical Specifications.

All of the events evaluated are bounded by the reference cycle.

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An ECCS analysis was performed for the limiting break size LOCA for Cycle 3 to demonstrate compliance with the requirements of 10 CFR 50.46.

The methodology is the same as for the Cycle 2 analysis (Ref. 23).

The analysis justifies a 13.5 Kw/ft peak linear heat generation rate.

For Cycle 3, since there have been no significant changes in hardware char-acteristics, only clad temperatures and oxidation are required in this reevaluation.

The code STRIKIN-II was used for this purpose (Ref. 24).

The performance data were generated with the FATES-3A fuel evaluation code (Refs.

6 and 7).

It was demonstrated that the double ended guillotine break with a discharge coefficient of 1.0 is the limiting size.

Similarly the limiting burnup, i.e., with the highest fuel stored

energy, was found to be 1000 MWD/MTU.

The ECCS analysis methods discussed above have been previously approved and are acceptable.

2.6.1 Lar e LOCA Anal sis The input data compared to the reference cycle were conservative.

The results for the limiting double ended guillotine break showed a peak clad temperature of 1944'F, peak clad oxidation of 5.4X and total core-wide oxidation less than.80%.'ll these values are within the required 10 CFR 50.46 limits of 2,200'F, 17.0X and 1.0X respectively, Therefor e, we find the large LOCA analysis results to be acceptable.

2.6.2 Small Break LOCA Anal sis Review of the Cycle 3 fuel and core data confirmed that the small break LOCA analysis results are bounded by the corresponding results of the reference cycle.

3.0 TECHNICAL SPECIFICATION CHANGES This section provides a summary of the proposed amendments to the Palo Verde Unit 1 Technical Specifications for the Cycle 3 operation.

A brief descr iption, justification and acceptability for each Technical Specification (TS) change is provided in the following.

TS Fi ure 3.1-1A:

The proposed change raises the required shutdown marg n

or co and hot shutdown conditions from 3.5X delta-k/k to 4.0%

delta-k/k to accommodate the requirements for inadvertent deboration.

This change is necessary to satisfy regulatory requirements and thus, is acceptable.

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The proposed changes increase the mons orang requency or ac up oron dilution detection to ensure that the time criteria for detection and correction of a boron di lution event remain the same as the reference cycle.

As such these proposed changes are acceptable.

TS Fi ures 3.1-3 and 3. 1-4:

The proposed changes revise the curves of e

ransien nser son smit lines.

These changes are required to make the Technical Specifications consistent with the Cycle 3 Safety analyses.

Thus, the proposed changes are acceptable.

TS Fi ure 3.2-1A:

The proposed change relaxes the azimuthal power tilt opera

>ng hami s with the core operating limit supervisory system in operation, to avoid lengthy delays in increasing power.

Mhen the core operating limit supervisory system is in operation, reactor operation within the analysis limits is assured, therefore, the proposed amendment is acceptable.

TS Fi ures 3.2-2 and 3.2-2A:

The proposed changes revise the DNBR limit curves or co dna sons o

EACs inoperable with COLSS inoperable.

These revisions are required to reflect cycle-specific parameter changes due to core loadings.

The changes are required to ensure that the Technical Specifications are consistent with the safety analyses for Cycle 3, and

thus, are acceptable.

4.0 STARTUP TESTING The licensee presented a description of the planned startup testing, which includes:

low power physics, ascension to power and procedures if accept-ance criteria are not met.

The objective of the testing is to verify that the core performance is consistent with the design and safety analyses.

The program conforms to the requirements of the ANSI/ANS-19.6.1, 1985 and supplements the normal surveillance requirements of the Technical Specifications (Refs.

25 E 26).

The low power physics tests include:

initial criticality, critical boron concentration, temperature reactivity coefficient, control element assembly reactivity worth and inverse boron worth.

The power ascension testing includes:

flux symmetry verification, core power distribution, shape annealing matrix, boundary point power correlation coefficient, radial peaking factors, control element assembly shadowing factor, reactivity coefficient at power and critical boron concentration.

These tests will provide reasonable assurance that the core has been loaded in accordance with the safety analysis assumptions.

They are therefore acceptable.

Should any of the startup tests reveal any unreviewed safety issues the NRC will be notified.

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A Lk 5.0 SUMI4ARY 6.0 We have reviewed the submitted information in support of the Palo Verde Unit I Cycle 3 operation.

The review covered fuels, physics, thermal hydraulics, accident and transient analyses, technical specification revisions and startup test procedures.

Based on the evaluations presented in the preceding sections we find the proposed reload acceptable.

CONTACT WITH STATE OFFICIAL 7.0 The Arizona Radiation Regulatory Agency has been advised of the proposed determination of no significant hazards consideration with regard to these changes.

No comments were received.

ENVIRONMENTAL CONSIDERATION The amendment involves changes in the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20.

The staff has determined that the amendment involves no significant increase in the amount, and no significant change in the type, of any effluent that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued proposed findings that the amendment involves no significant hazard consideration, and there has been no public comment on such finding.

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environ-mental assessment need to be prepared in connection with the issuance of the amendment.

8.0 CONCLUSION

The staff has concluded, based on the considerations discussed

above, that (I) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not,be inimical to the common defense and security or to the health and safety of the public.

We, therefore, conclude that the proposed changes are acceptable.

Principal contributor:

T. Chan Dated:

September 19, 1989

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REFERENCES Letter from D.B. Karner, Arizona Nuclear Power Project to USNRC, "Palo Verde Nuclear Generating Station Unit 1, Proposed Reload Technical Specification Changes,"

dated January 12, 1989.

2.

3.

5.

6.

7.

8.

9.

10.

12.

13.

14.

Letter from D.B. Karner, Arizona Nuclear Power Project to USNRC, "Palo Verde Nuclear Generating Station Unit 1, Submittal of the Reload Analysis Report," dated January 18, 1989.

Letter from D. B. Karner, Arizona Nuclear Power Project to USNRC, "Palo Verde Nuclear Generating Station-Unit 1, Submittal of Revised Reload Analysis Report," dated April 19, 1989.

Letter from D.B. Karner, Arizona Nuclear Power Project to

USNRC, "Applicability of RAR References,"

dated April 26, 1989.

Letter from W. F.

Conway, Arizona Public Service to USNRC, "Revised Reload Analysis Report Change Pages,"

dated June 27, 1989.

CENPD-139-P-A,"

C-E Fuel Evaluation Model," Combustion Engineering, dated July 1974.

CEN-161(B)-P, "Improvements in the Fuel Evaluation Model,"

Combustion Engineering, dated July 1981.

Letter from R.A. Clark (NRC) to A.E. Lundvall, Jr.

(BGEE), "Safety Evaluation of CEN-161 (FATES3)," dated March 31, 1983.

CENPD-153P, Rev.

1-P-A, "INCA/CECOR Power Peaking Uncertainty,"

Combustion Engineering, dated May 1980.

CENPD-266-PA, "The ROCS and DIT Computer Codes for Nuclear Design,"

Combustion Engineering, dated April 1983.

CENPD-161-PA, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," Combustion Engineering, dated April 1986.

CENPD-162-A, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Power Distribution" Combustion Engineering, dated September 1976.

CEN-160-S, Rev. 1-P, "CETOP Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Unit 2 and 3," Combustion Engineering, dated September 1981.

CEN-356-V-PA, Rev.

1-PA, "Modified Statistical Combination of Uncertainties,"

Combustion Engineering, dated May 1988.

15.

16.

17.

18.

19.

20.

21.

22.

23.

24.

25.

26.

27.

28.

CENPD-225-PA, "Fuel and Poison Rod Bowing" Combustion Engineering, dated June 1983

'etter from A.E. Scherer Combustion Engineering to D.G. Eisenhut NRC (Enclosure 1), "Statistical Combination of System Parameter Uncertainties in Thermal Margin Analyses for System 80," dated May 14, 1982.

CESSAR SSER'2 Section 4.4.6, "Statistical Combination of Uncertainties,"

Combustion Engineering.

CESEC, "Digital Simulation of a Combustion Engineering Nuclear Steam Supply System,"

Combustion Engineering enclosure 1-P to LD-82-001, dated January 6, 1982.

CENPD-206-P, "TORC Code Verification and Simplified Modeling Methods,"

Combustion Engineering, dated January 1977.

CENPD-183, "Loss of Flow, CE Method for Loss-of-Flow Analysis,"

Combustion Engineering, dated July 1975.

CENPD-188-A, "HERMITE, Space Time Kinetics," Combustion Engineering, dated July 1975.

CENPD-199-PA, Rev.

1P, "CE Setpoint Methodology," Combustion Engineering, dated July 1975.

CENPD-132-P, "Calculative Methods for the CE Large Break LOCA Evaluation Model," Combustion Engineering, dated August 1974.

Also Supplements 1 and 2 dated December 1974 and July 1975 respectively.

CENPD-135-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program,"

Combustion Engineering, dated April 1974.

Also Supplements 2P and 4P dated February 1975 an August 1975 respectively.

ANSI/ANS-19.6.1-1985, "Reload Startup Physics Tests for Pressurized Water Reactors."

CEN-319, "Control Rod Group Exchange Technique,"

Combustion Engineering, dated November 1985.

Letter from William F.

Conway, Arizona Public Service Co. to USNRC, "Revised Section 7 of Reload Analysis Report for Unit 1, Cycle 3,"

dated August 25, 1989.

Letter from William F.

Conway, Arizona Public Service Co. to USNRC, "Revision to Section 8 of Reload Analysis," dated September 11, 1989.

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May 23, 1989 Docket Nos.:

STN 0-528 STN 50-529, and S

-530 P:O S uT R D

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AJ~NX)MKNT IINO O < G TO NX mWX DHagan PDV Plant tl le EJordan Mr. William F.

Conway Executive Vice President Arizona Nuclear Power Project Post Office Box 52034

Phoenix, Arizona 85072-2034

SUBJECT:

ISSUANCE OF AMENDMENT NO. 43 TO FACILITY OPERATING LICENSE NO. NPF-41, AMENDMENT NO. 28TO FACILITY OPERATING LICENSE NO.

NPF-51 AND AMENDMENT NO.

17 TO FACILITY OPERATING LICENSE NO.

NPF-74 FOR THE PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 (TAC NOS. 71148, 71149 AND 71150)

The Comofssfon has issued the subject Amendments, which are enclosed, to the Faci lsty Operating Licenses for Palo Verde Nuclear Generating Station, Units 1, 2, and 3.

The Amendments consist of changes to the Technical Specff1catfons (Append1x A to each license) fn response to your-applicatfon transmitted by letter dated November 9, 1988.

A copy of the related Safety Evaluation is also enclosed.

A Notice of Issuance wil'I be included in the Comisslon's next regular bi-weekly Federal

~Re later notice.

Sincerely,

/s/

The Amendments revise Palo Verde Nuclear Generating Station (PVNGS) Technical Specif1cation Sect1on 3/4.4.5, "Reactor Coolant System Leakage,"

by changing the operability requirements of the containment radioactivity monitoring systems and the assocfatea Action Statement.

Enclosures:

1.

Amendment No. 4q to NPF-41 2.

Amendment No. 28 to NPF-51 3.

Amendment No. 17 to NPF-74 4.

Safety Evaluation cc:

See next page

• See previous concurrence Terence L. Chan, Senior Project Manager Project Directorate V

Div1sfon of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulat1on

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UNITED STATES t

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 May 23, 1989 Docke~ ttos.:

STN 50-528, STN '50-529, and STta 50-530 t!r. 'Hilliam F.

Conway ExecI tive Vice Presiaent Arizona nuclear Power Pro'ect Post Office Box 52034

Phoenix, Arizona 85072-2034 Dear ter.

Conway:

SUBJECT:

ISS4ANCE OF AMENDMENT NO.

43 TG FACILITY OPERATING LICENSE NO. HPF-41, AMENDMENT HO.

28TO FACILITY OPERATING LICENSE NO.

NPF-51 AND AMENDMENT NO.

17 TO FACILITY OPERATING LICENSE NO.

NPF-74 FOR THE PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 (TAC NOS. 71148, 71149 AND 71150)

The Commission has issued the subject Amendments, which are enclosed, to the Facility Operating Licenses for Palo Verde Nuclear Generating Station, Units 1, 2,

and 3.

The Amendments consist of changes to the Technical Specifications (Appendix A to each license) in response to your application transmitted by letter dated November 9, 1988.

The Amendments revise Palo Verde Huclear Generating Station (PVNGS) Technical Specification Section 3/4.4.5, "Reactor Coolant System Leakage,"

by changing the operability requirements of the containment radioactivity monitoring systems alld the associated Action Statement.

A copy of the related Safety Evaluation is also enclosed.

A Notice of issuance will be included in the Comnission's next regular bi-weekly Federal

~Re ister notice.

Since

Enclosures:

1.

Amendment No. 43 to NPF-41 2.

Amendment No. 28 to NPF-51 3.

Amendment No. 17 to NPF-74 4.

Safety Evaluation cc:

See next page T rence L. Chan, Senior Project Manager Project Directorate V

Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation

0 I

Mr. William F.

Conway Arizona Nuclear Power Project CC:

Mr. William F.

Conway Arizona Nuclear Power Project Executive Vice President Post Office Box 52034

Phoenix, Arizona 85072-2034 Arthur C.

Gehr, Esq.

Snell 5 Wilmer 3100 Valley Center

Phoenix, Arizona 85073 Charles R. Kocher, Esq. Assistant Council James A. Boeletto, Esq.

Southern Ca Iifornia Edison Company P. 0.

Box 800

Rosemead, Ca 1 ifornia 91770 Mr. Tim Polich U.S. Nuclear Regulatory Commission HC-03 Box 293-NR Buckeye, Arizona 85326 Regional Administrator, Region V

U.

S.

Nu c 1 ear Regu 1 a tory Commi s s ion 1450 Maria Lane Suite 210 Walnut Creek, California 94596 Mr. Charles B. Brinkman Washington Nuclear Operations Combustion Engineering, Inc.

12300. Twinbrook Parkway, Suite 330 Rockvil le, Maryland 20852 Mr. Charles Tedford, Director Arizona Radiation Regulatory Agency 4814 South 40 Street

Phoenix, Arizona 85040 Chairman Maricopa County Board of Supervisors ill South Third Avenue

Phoenix, Arizona 85003 Palo Verde

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t UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ARIZONA PUBLIC SERVICE COMPAVY, ET AL.

OOCKET NO.

STN 50-528 PALO VEROE NUCLEAli GENERATING STATION UNIT NO..I htIENDNENT TO FACILITY.CPERATING LICENSE Amendment Nv 43 License No.

NPF-41 1.

The Nuclear Regu Iatory Commfssion {the COII3Ifssfon) has founa that:

A.

The applicatfon for amerIament, dated November 9, 1988 by the Arizona Public Service Company (APS) on behalf of itself and the Salt River Project Agricultural ImproveIIIent and Power Dfstrfct, El Paso Electric

Company, Southern California Edison

Company, Public Service Company of New Mexico, Los Angeles Oepartment of Mater and Power, and Southern California Public Power Author'ity (licensees),

complies with the standards arId requirements of the Atomic Energy Act of 1964, as amended

{the Act) and the CoIIefssfon's.regulations set forth in 10 CFR Chapter I; 8.

The facflsty will operate fn conformity with the applfcauon, the provisions of the Act, and the regulations of the Comnf ssfon; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and {fi) that such activities will be conducted in coapl>ance with the CoIImfssfon's regulations; G.

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

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

2.

According ly, the license is amended by changes to the Technical Specifications as indicated fn the enclosure to thfs license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-41 fs hereby amended to read as follows:

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Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.43, and the. Environmental Protection Plan contained in Appendix 8, are hereby incorporated into this license.

APS shall operate the facility in accordance with the Technical Specifications and the Environmenta 1 Protection Plan.

3.

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

Enclosure:

Changes to the Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION 4

ting'fd/1 George R Knighton,.6irector Project Directorate V

Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation

0 1

ENCI OSURE TO LICENSE AMENDMENT AMENDMENT NO.

43 TO FACILITY OPERATING LICENSE NO.

NPF-41 DOCKET NO.

STN 50-528 Replace the following page of the Appendix A Technical Specifications with the enclosed page.

The revised page is identified by Amendment number and contains vertical lines indicating the areas of change.

Also to be replaced is the following overleaf page to the amended page.

Amendment Pa e

Overleaf Pa e

3/4 4-18 3/4 4-17

3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE t

LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.5.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a.

Either the containment atmosphere gaseous radioactivity or containment atmosphere particulate radioactivity monitoring system, and b.

The containment sump level and flow monitoring system.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a ~

b.

With either/or both the containment atmosphere gaseous radioactivity and containment atmosphere particulate radioactivity monitors INOPERABLE, operation may continue for up to 30 days provided the containment sump level and flow monitoring system is OPERABLE and gaseous and/or particulate grab samples of the containment atmosphere are obtained at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and analyzed within the subsequent 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With the containment sump level and flow monitoring system INOPERABLE, operation may continue for up to 30 days provided the containment atmosphere gaseous radioactivity monitoring and the containment atmosphere particulate radioactivity monitoring systems are OPERABLE; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.5. 1 The leakage detection systems shall be demonstrated OPERABLE by:

a.

Containment atmosphere gaseous and particulate monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, b.

Containment sump level and flow monitoring system-performance of CHANNEL CALIBRATION at least once per 18 months.

PALO VERDE - UNIT 1 3/4 4-18 AMENDMENT NO. 43

TABLE4.4-2 ICl m

Rl C7m 1ST SAMPLE INSPECTION STEAM GENERATOR TUBE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION C

Sample Size Result Action Required Result Action Required Result Action Requii'ed A minimum of S Tubes per S. G.

C-1 C-2 C-3 None Plug detective tubes and inspect additional 2S tubes in this S. G.

Inspect all tubes in this S. G., plug de-fective tubes and inspect 2S tubes in each other S. G.

Notification to NRC pursuant to II50.72

{b)(2) of 10CFR Part 50 C1 C-2 C-3 AII other S. G.s are C-1 Some S. G.s C-2 but no additional S. G. are C-3 Additional S. G. is C-3 N. A.

None Plug defective tubes and inspect additional 4S tubes in this S. G.

Perform action for C-3 resul t o f Iir st sample None Perform action for C2 result of second sample Inspect all tubes in each S. G. and plug defective tubes.

Notification to NRC pursuant to I'150.72 (b){2) of IOCFR Part 50 N. A.

C-I C-2 C-3 N. A.

N. A.

N. A.

N. A.

None N. A.

N. A.

N. A.

N. A.

Pliig defective tubes Perform action for C-3 result ot first saniple 3

gg Where N is the number of steam, generators in the unit, and n is tlie number of steam generators iiisliected during an inspection

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 ARIZONA PU5L'C SERVICE COtlPANY ET AL.

DOCKE1 hG.

STN 50-529 PALO VERDE NUCLEAR GENERATING STATION UNIT NO.

2 ANENGhIENT 1O-FACILITY OPERATING LICENSE Amendment No. 28 License No.

NPF-51 1.

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

A.

The application for amendment, dated November 9, 1988 by the Arizona Public Service Company (APS) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric

Company, Southern California Edison Company, Public Service Company of New hIexico, Los Ange1es Depar tment of Mater and Power, and Southern California Public Power Authority (licensees),

complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations set forth in 10 CFR Chapter I; 8.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as inaicated in the enclosure to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-51 is hereby amended to read as follows:

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{2)

Techaical.S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.28 and the Environmental Protect>on Plan containeo in Appendix 8, are hereby incorporated into this license.

APS shall operate the faci lity in accordance with the Technical Specifications ana the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COYBISSION

Enclosure:

Changes to the Technical Specifications

.~~a.

ec'eorge llFKnighton, erector Project Directorate Y

Division of Reactor Projects III, IY, V and Special Projects Office of Nuclear Reactor Regulation

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ENCLOSURE TO LICENSE AMENDMENT AMENDMENT NQ..

TO FACILITY OPERATING LICENSE.NO.

NPF-51 DOCKET NO.

STN 5C-529 Replace the following page of the Appendix A Technical Specifications with the enclosed page.

The revised page is iaentified ty Amendment number and contaisn vertical lines indicating the areas of change.

Also to be replaced is the following overleaf page to the amended page.

Amendment Pa e

Overleaf P~ae 3/4 4-18 3/4 4-17

REACTOR COOLANT SYSTEM 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE

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LEAK'AGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4. 5. 1 The following Reactor Coolant System leakag'e detection systems shall be OPERABLE:

a.

Either the containment atmosphere gaseous radioactivity or containment atmosphere particulate radioactivity monitoring system, and b.

The containment sump level and flow monitoring system.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a.

With either/or both the containment atmosphere gaseous radioactivity and containment atmosphere particulate radioactivity monitors INOPERABLE, operation may continue for up to 30 days provided the containment sump level and flow monitoring system is OPERABLE and gaseous and/or particulate grab samples of the containment atmosphere are obtained at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and analyzed within the subsequent 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the containment sump level and flow monitoring system INOPERABLE, operation may continue for up to 30 days provided the containment atmosphere gaseous radioactivity monitoring and the containment atmosphere particulate radioactivity monitoring systems are OPERABLE; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.4. 5. 1 The leakage detection systems shall be demonstrated OPERABLE by:

a.

Containment atmosphere gaseous and particulate monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified.in Table 4.3-3, b.

Containment sump level and flow monitoring system-performance of CHANNEL CALIBRATION'at least once per 18 months.

PALO VERDE - UNIT 2 3/4 4-18 AMENDMENT N0.28

TABLE 4.4-2 STEAN GENERATOR TUBE INSPECTION 1ST SAMPLE INSPECTION 2NO SAMPLE INSPECTION 3RO SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum'ot S Tubes per S. G.

C-1 C-2 C-3 Plug defective tubes and inspect additional 2S tubes in this S. G.

Inspect all tubes in this S. G., plug de-fective tubes and inspect 2S tubes ln each other S. G.

Notification to NRC pursuant to f50.72 (b)(2) of 10 CFR Part 50 N. A.

C-1 C-2 C-3 All other S. G.s are C-1 Some S. G.s C-2 but no additional S. G. are C-3 Additional S. G. is C-3 N. A.

None Plug defective tubes and inspect additional 4S tubes in this S. G.

Perform action for C-3 result of first sample None Perform action for C-2 result of second sample Inspect all tubes in each S. G. and plug defective tubes.

Notification to NRC pursuant to $ 50.72 (b)(2) of 10 CFR Part 50 C-1 C-2 C-3 N. A.

N. A.

N. A.

N. A.

N. A.

None Plug defective tubes Perform action for C-3 result of first sample N. A.

N. A.

< Where N is the number of steam generators in the unit, and n is the number of steam generators inspected a

during an inspection

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I UNITED STATES t

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C, 20555 ARIZONA PUBLIC SERVICE Cot PAr;Y

.ET AL.

DOCKET Ilb.

STN 5G-530 PALO YEROE NUCLEAR GENERATING STATION UNiT NO.

3 AIlENDIIENT TO FACIL;'TY OPERAT! NG LICENSE Ameridment No.17 License No.

NPF-74 1.

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

A.

The application for amendment, dated November "9, 1988 by the Arizona Publ1c Serv1ce Coiiipany (APS) on behalf of itself arid the Salt River Project Agr1cultural Improvement and Power District, El Paso Electric Company, Southern California Edison Company, Pub11c Service Company of New tlexfco, Los Angeles Department of Mater and

Power, and Southern Ca11fornia Public Power Authority (licensees),

complies with the standards and requfrements of the Atomic Energy Act of 1954, as amended (the Act) and the Collmifssfon's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity w1th the application, the provisions of the Act, and the regulations of the Coamifssion; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and

( 11) that such activities will be conducted in compliance with the Commission's regulations; D.

The 1ssuance of this amendment will r<ot be inimical:to the correon

,defense and security or to the health and safety of the public; E.

The issuance of this amendrent is

>n accordance with 10 CFR Part 51 of the Comissfon's regulations and all applicable requirements have been satisf1ed.

2.

Accord1ngly, the license 1s amended by changes to the Technical Specifications as indicated fn the enclosure to this license amendment, and paragraph 2.C(2) of Facility Operating L1cense No.

NPF-74 1s hereby amended to read as follows:

I (2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 17 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license.

APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/

1 aACg &Pep>>

George WYKnighton, Pirector Project Directorate V

Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

Changes to the Technical Specifications

e ENCLOSURE TO LICENSE AMENDMENT AMENDMENT NO.

1".

TO FACILITY OPERATING LICENSE NO. NPF-74 DOCKET NO.

STN 56-530 Peplace the following page uf the Appendix A Technical Specifications with the enclosea page.

The revised page is identified by Amendment number and contains vertical lines indicating the areas of change.

Also to be replaced is the following overleaf page to the amended page.

Amendment.Pa e

Overleaf Pa e

REACTOR COOLANT SYSTEM 3/4.4.5

.REACTOR COOLANT SYSTEM LEAKAGE e

LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.5.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a.

Either the containment atmosphere gaseous radioactivity or containment atmosphere particulate radioactivity monitoring system, and b.

The containment sump level and flow monitoring system.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With either/or both the containment atmosphere gaseous radioactivity and containment atmosphere particulate radioactivity monitors INOPERABLE, operation may continue for up to 30 days provided the containment sump level and flow monitoring system is OPERABLE and gaseous and/or particulate grab samples of the containment atmosphere are obtained at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and analyzed within the subsequent 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the containment sump level and flow monitoring system INOPERABLE, operation may continue for up to 30 days provided the containment atmosphere gaseous radioactivity monitoring and the containment atmosphere particulate radioactivity monitoring systems are OPERABLE; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS

4. 4. 5. 1 The leakage detection systems shall be demonstrated OPERABLE by:

a ~

b.

Containment atmosphere gaseous and particulate monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, Containment sump level and flow monitoring system-performance of CHANNEL CALIBRATION at least once per 18 months.

PALO VERDE - UNIT 3 3/4 4-18 AMENDMENT NO. 17

TABLE 4.4-2 ICl C7m I

1ST SAMPLE INSPECTION STEN GENERATOR TUBE INSPECTION 2NO SAMPLE INSPECTION 3RD SAMPLE INSPECTION C

I Sample Site A minimum of S Tubes per S. G.

C-2 C-3 Action Required Plug defective tubes and inspect additional 2S tubes in this S. G.

Inspect all tubes in this S. G., plug de.

fective tubes and inspect 2S tubes in each other S. G.

'otification to NRC pursuant to tt50.72 ib)i2) of 10 CFR Part 50 Result N. A.

C-1 C-2 C-3 All other S. G.s are C-1 Some S. G.s C-2 but no additional S. G. are C-3 Additional S. G. is C-3 Action Required None Plug defective tubes and inspect additional 4S tubes in this S. G.

Perform action for C-3 result of first sample None Perform action for C-2 result of second sample Inspect all tubes in each S. G. and plug defective tubes.

Notification to NRC pursuant to I't50.72 (b){2) of 10 CFR Part 50 Result C-1 C-2 C-3 N. A.

N. A.

N. A.

Action ftequired None N. A.

N. A.

Plug defective tubes Perform action for C-3 result of first sample 3 N Where N is the number of steam generators in the unit. and n is the number of steam generators inspected during an inspection

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UNITED STATES t

NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT HO.

43 TG FACILITY OPEFATIHG LICENSE NO.

HPF-CI AMENDMENT NO.

28 TO FACILITY OPERATING LICENSE HO.

NPF-51 AND AHENDHEHT N0.17 TO FACILITY OPERATING LICENSE HO.

HPF-74 ARIZONA PUBLIC SERVICE COMPANY ET AL.

PALO VERDE NUCLEAR GENERATING STATION UNITS 1

2 AND 3 DOCKET NOS.

STN 50-528 STH 50-529 AND STN 50-530

1.0 INTRODUCTION

By letter dated November 9, 1988 the Arizona Public Service Company (APS) on behalf of itself, the Salt River Project Agricultural Improvement and Power District, Southern California Edi'son

Company, El Paso Electric Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Author ity ( licensee's),

requested changes to the Technical Specifications for the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Appendix A to Facility Operating License Nos.

NPF-41, NPF-51, and NPF-74, respectively).

The proposed changes would revise Technical Specifications (TS) Section 3/4.4.5, "Reactor Coolant System Leakage" by changing the operability requirements of the containment radioactivity monitoring systems.

The action statement is also revised to reflect this change.

2.0 DISCUSS IOH AND EVALUATION The existing TS'Section 3/4.4.5 for each of the Palo Verde licenses identifies three systems which comprise the Reactor Coolant System Leakage Detection System (RCSLDS):

a.

containment atmosphere particulate radioactivity monitoring system b.

containment atmosphere gaseous radioactivity monitoring

system, and c.

containment sump level and flow monitoring'ystem The existing TS Action Statement permits continued operation for up to 30 days if any one of the thr ee monitoring systems becomes inoperable, provided

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that grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gaseous and/or particulate radioactivity monitoring system is inoperable.

This Action Statement requires the plant to shutdown in the event both the gaseous and particulate radioactivity monitoring systems are inoperable, or if either o>>e of the radioactivity monitoring systems and the containment sump level anc flow monitoring system are inoperable.

In actuality, there are two independent systems which comprise the RCSLOS.

The containment atmosphere gaseous monitor and the containment atmosphere particulate monitor share a

common sample point, sample line, isolation

valves, sample fan, radiation monitor package and power supply.

There are two monitoring systems, one look>ng at a particulate filter assembly and the other at a gas chamber.

Should one of the co>mon components in the system fail, both systems will become inoperable.

With the present technical specifications, the unit is required to shut down in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

However, plant shutdown is unnecessary because adequate capability remains to detect primary system leakage.

The containment sump monitoring capabil>ties are available and containment atmosphere a>rborne radioactivity levels will be determined using grab samples.

To eliminate this unnecessary shutdown requirement, the licensees propose to revise the operability requirements of the three systems which comprise the RCSLDS to accurately reflect the systems configuration, by requiring the containment sump level and flow monitoring system and either of the two containment atmosphere radioactivity monitors to be operable.

In conjunction with this change, the licensees propose to revise the Action Statement to permit continued operation for up to 30 days in the event either/or both containment atmosphere particulate radioactivity and containment atmosphere gaseous monitors are inoperable in order to allow repair or replacement of inoperable components.

The proposed Action Statement also requires more frequent sampling (i.e., once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> rather than once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) than the present requirement.

As such we find this change to be acceptable.

The proposed Act)on Statement for inoperable containment sump level and flow monitoring system would permit continued operation for 30 days to allow for'epair or replacement of inoperable components if both the gaseous and particulate radioactivity monitoring systems are operable.

This is the same requirement as the present technical specifications.

Grab samples would not be required because adequate leakage detection is provided by the operable radioactivity monitor without the grab samples.

We find this, change acceptable'.

The proposed change eliminates unnecessary plant shutdowns because of inoperable components coaeon to the containment atmosphere gaseous and particulate radioactivity monitoring systems.

It also eliminates an unnecessary sampling procedure when at least one containment radioactivity monitoring system is available for leak detection.

Further, the compensatory measure of grab sampling is improved due to the increased sample frequency and prompt analysis requirement.

Therefore, we find the proposed technical specification changes to be acceptable.

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3.0

.CONTACT WITH STATE OFFICIAL

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The Arizona Radiation Regulatory Agency has been advised of the proposed

-eterminatfon of no sigrifffcant hazards consideration with regard to r.hese changes.

No comments were received.

4.0 ENVIRONtlEllTAL CONSIDERATION The amendments involve changes in the >nstallatfon or use of facility components located withirr the restricted area as defined irr 10 CFR 20.

The staff has determined that the amendments

',nvolve no significant increase in the amount, and no significant change in the type, of any ef ',uent that may be released offsite and that there is no significant increase

>n >ndfvfdual or cumulative occupational radiation exposure.

The Commission has previously issued proposed findings that the amendments involve no significant hazard consideration, and there has been no public comment on such finding.

Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environ-mental assessment rreed to be prepared irr connect>on with the issuance of the amendments.

5.0 CONCLUSION

The staff has concluded, based on the considerations discussed

above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, (2) such activities will be conducted in compliance wfth the Commission's regulations, and (3) the issuance of the amendments will not be infmfcal to the common defense and security or to the health and safety of 7he public.

We, therefore, conclude that the proposed changes are acceptable.

Principal contributor:

T. Chan pated:

allay 23, 1989

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