ML17188A412

Issuance of Amendment Nos. 203, 203, and 203 to Revise Technical Specifications to Incorporate Updated Criticality Safety Analysis

ML17188A412
Person / Time
Site:	Palo Verde
Issue date:	07/28/2017
From:	Siva Lingam Plant Licensing Branch IV
To:	Bement R Arizona Public Service Co
Lingam S, 301-415-1564
References
CAC MF7138, CAC MF7139, CAC MF7140
Download: ML17188A412 (46)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Robert S. Bement Executive Vice President Nuclear/

Chief Nuclear Officer Mail Station 7602 Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-2034 July 28, 2017

SUBJECT:

PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 -

ISSUANCE OF AMENDMENTS TO REVISE TECHNICAL SPECIFICATIONS TO INCORPORATE UPDATED CRITICALITY SAFETY ANALYSIS (CAC NOS. MF7138, MF7139, AND MF7140)

Dear Mr. Bement:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 203 to Renewed Facility Operating License No. NPF-41, Amendment No. 203 to Renewed Facility Operating License No. NPF-51, and Amendment No. 203 to Renewed Facility Operating License No. NPF-74 for the Palo Verde Nuclear Generating Station, Units 1, 2, and 3, respectively. The amendments revise the technical specifications (TSs) to modify TS requirements to incorporate the results of an updated criticality safety analysis for both new and spent fuel storage in response to Arizona Public Service Company's (the licensee's) application dated November 25, 2015, as supplemented by letters dated January 29, June 30, October 6, November 9, and November 23, 2016; and March 3 and May 24, 2017.

The amendments allow the licensee to install the NETCO-SNAP-IN neutron absorbing rack inserts into some spent fuel pool storage rack cells coupled with six classifications of fuel by initial enrichment, burnup, and decay time in six storage configurations for criticality control. Approval of these license amendments allows the licensee to meet the effective neutron multiplication factor (k-effective or ke11) criticality control requirements. Since the proposed spent fuel pool TSs include a reduction in the maximum weight percent (w%)

235Uranium (235U), the new fuel storage racks TSs are also revised to reflect the new maximum w% 235U.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Docket Nos. STN 50-528, STN 50-529, and STN 50-530

Enclosures:

1. Amendment No. 203 to NPF-41

2. Amendment No. 203 to NPF-51

3. Amendment No. 203 to NPF-74

4. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely,

~

If' ""'~

Siva P. Lingam, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ARIZONA PUBLIC SERVICE COMPANY, ET AL.

DOCKET NO. STN 50-528 PALO VERDE NUCLEAR GENERATING STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 203 License No. NPF-41

1.

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

A.

The application for amendment by the Arizona Public Service Company (APS or the licensee) 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 Mexico, Los Angeles Departmerit of Water and Power, and Southern California Public Power Authority dated November 25, 2015, as supplemented by letters dated January 29, June 30, October 6, November 9, and November 23, 2016; and March 3 and May 24, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission'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 rules and regulations of the Commission; 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 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.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C{2) of Renewed Facility Operating License No. NPF-41 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 203, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed operating license. APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 90 days of the date of issuance.

Attachment:

Changes to the Renewed Facility Operating License No. NPF*41 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: July 28, 2017

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ARIZONA PUBLIC SERVICE COMPANY ET AL.

DOCKET NO. STN 50-529 PALO VERDE NUCLEAR GENERATING STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 203 License No. NPF-51

1.

The Nuclear Regulatory Commission (the Commission) has found that A.

The application for amendment by the Arizona Public Service Company (APS or the licensee) 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 Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority dated November 25, 2015, as supplemented by letters dated January 29, June 30, October 6, November 9, and November 23, 2016; and March 3 and May 24, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission'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 rules and regulations of the Commission; 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 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.

2.

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

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 203, and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated into this renewed operating license. APS shall operate the facility in accordance with the Technical Specificaf1ons and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 90 days of the date of issuance.

Attachment:

Changes to the Renewed Facility Operating License No. NPF-51 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Dateoflssuance: July 28, 2017

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ARIZONA PUBLIC SERVICE COMPANY, ET AL.

DOCKET NO. STN 50-530 PALO VERDE NUCLEAR GENERATING STATION UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 203 License No. NPF-74

1.

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

A.

The application for amendment by the Arizona Public Service Company (APS or the licensee) 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 Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority dated November 25, 2015, as supplemented by letters dated January 29, June 30, October 6, November 9, and November 23, 2016; and March 3 and May 24, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission'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 rules and regulations of the Commission:

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 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.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Renewed Facility Operating License No. NPF-74 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 203, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed operating license. APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of the date of issuance and shaJJ be implemented within 90 days of the date of issuance.

Attachment:

Changes to the Renewed Facility Operating License No. NPF-74 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: July 28, 2017

ATTACHMENT TO LICENSE AMENDMENT NOS. 203 203 AND 203 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-41 NPF-51 AND NPF-74 PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 DOCKET NOS. STN 50-528, STN 50-529, AND STN 50-530 Replace the following pages of the Renewed Facility Operating Licenses Nos. NPF-41, NPF-51, and NPF-74, and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License No. NPF-41 REMOVE 5

INSERT 5

Renewed Facility Operating License No. NPF-51 REMOVE 6

INSERT 6

Renewed Facility Operating License No. NPF-74 REMOVE 4

INSERT 4

Technical Specifications REMOVE 3.7.17-1 3.7.17-2 3.7.17-3 3.7.17-4 4.0-2 4.0-3 5.5-19 INSERT 3.7.17-1 3.7.17-1a 3.7.17-2 3.7.17-2a 3.7.17-3 3.7.17-3a 3.7.17-4 3.7.17-4a 3.7.17-5a 3.7.17-6a 3.7.17-7a 4.0-2 4.0-2a 4.0-3 4.0-3a 5.5-19 5.5-19a 5.5-20a (1)

Maximum Power Level Arizona Public Service Company (APS) is authorized to operate the facility at reactor core power revers not in excess of 3990 megawatts thermal {100°/o power), in accordance with the conditions specified herein.

{2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 203, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed operating license.

APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

(3)

Antitrust Conditions This renewed operating license is subject to the antitrust conditions delineated in Appendix C to this renewed license.

(4)

Operating Staff Experience Requirements Deleted (5)

Post-Fuel-Loading Initial Test Program {Section 14, SER and SSER 2r Deleted (6)

Environmental Qualification Deleted (7)

Fire Protection Program APS shall implement and maintain in effect all provisions of the approved tire protection program as described in the Final Safety Analysis Report for the facility, as supplemented and amended, and as approved in the SER through Supplement 11, subject to the foJJowing provision:

APS may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

• The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Renewed Facility Operating License No. NPF-41 Amendment No. 203

( 1)

Maximum Power Level Arizona Public Service Company (APS) is authorized to operate the facility at reactor core power levels not in excess of 3990 megawatts thermal (100o/o power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 203, and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated into this renewed operating license.

APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

(3)

Antitrust Conditions This renewed operating license is subject to the antitrust conditions delineated in Appendix C to this renewed operating license.

(4)

Operating Staff Experience Requirements (Section 13.1.2, SSER 9)'

Deleted (5)

Initial Test Program (Section 14, SER and SSER 2)

Deleted (6)

Fire Protection Program APS shaJJ implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility, as supplemented and amended, and as approved in the SER through Supplement 11, subject to the following provision:

APS may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

(7) lnservice Inspection Program (Sections 5.2.4 and 6.6, SER and SSER 9)

Deleted

• The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Renewed Facility Operating License No. NPF-51 Amendment No. 203 (4)

Pursuant to the Act and 10 CFR Part 30, 40, and 70, APS to receive, possess, and use in amounts required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, APS to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 1 O CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level Arizona Public SeNice Company (APS) is authorized to operate the facility at reactor core power levels not in excess of 3990 megawatts thermal (100°/o power), in accordance with the conditions specified

herein, (2)

Technical Sgecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 203, and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated into this renewed operating license.

APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otheiwise stated in specific license conditions.

(3)

Antitrust Conditions This renewed operating license is subject to the antitrust conditions delineated in Appendix C to this renewed operating license.

(4)

Initial Test Program (Section 14, SER and SSER 2)

Deleted (5)

Additional Conditions The Additional Conditions contained in Appendix 0, as revised through Amendment No. 200, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Additional Conditions.

Renewed Facility Operating License No. NPF-74 Amendment No. 203

3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage Before SFP transition Spent Fuel Assembly Storage 3.7. 17 LCO 3.7.17 The combination of initial enrichment. burnup. and decay time of each fuel assembly stored in each of the four regions of the fuel storage pool shall be within the acceptable burnup domain for each region as shown in Figures 3.7.17 -1. 3.7.17-2. or 3.7.17-3. and described in Specification 4.3.1.1.

APPLICABILITY:

Whenever any fuel assembly is stored in the fuel storage pool.

ACTIONS CONDITION A.

Requirements of the LCD not met.

A. l REQUIRED ACTION

---

NOTE---------

LCO 3.0.3 is not applicable.

COMPLETION TIME Initiate action to Immediately move the noncomplying fuel assembly into an appropriate region.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial enrichment. burnup. and decay time of the fuel assembly is in accordance with Figures 3.7.17-1. 3.7.17-2. or 3.7.17-3. and Specification 4.3.1.1.

Prior to storing the fuel assembly in the fue 1 storage pool.

PALO VERDE UNITS 1.2.3 3.7.17-1 AMENDMENT NO..f..t.e.203

After SFP transition Spent Fuel Assembly Storage 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Spent Fuel Assembly Storage LCO 3.7.17 The combination of initial enrichment. burnup. and decay time of each fuel assembly shall be in compliance with the requirements specified in Tables 3.7.17-1 through 3.7.17-5.

APPLICABILITY :

Whenever any fuel assembly is stored in the fuel storage pool.

ACTIONS CONDITION A.

Requirements of the LCO not met.

A.1 REQUIRED ACTION

---

NOTE- ------- -

LCO 3.0.3 is not applicable.

COMPLETION TIME Initiate action to Immediately move the noncomplying fuel assembly into an appropriate region.

SURVEILLANCE REQUIREMENTS SR 3.7.17.1 SURVEILLANCE Verify by administrative means the initial enrichment. burnup. and decay time of the fuel assembly is in accordance with Tables 3.7.17-1 through 3.7.17-5. Figure 3.7,17-1.

and Spec Hi ca ti on 4. 3.1.1.

FREQUENCY

.Prior to storing the fuel assembly in the fuel storage pool.

PALO VERDE UNITS 1.2.3 3.7.17-la AMENDMENT NO. +/-2-9. 203

I*

I

~

3:

lE ci Figure 3.7.17-1 Before SFP transition Spent Fuel Assembly Storage 3.7.17 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Region 2

~ 10000-t-------"-;--t-----+-~~-+--~--+------j-----+-----,;'---~

Cl

i5 E

GI fl)

~

5000 +----~'--!------+----+-----+-----+----

e assumes ero decay time.

06-----~....-.__~~~---~~-+--.-,_.,...~.+-,__~---1------~~_...Jt-..-'

1.5 2.0 2.5 3.0 3.5 Initial Enrichment, weight %

4.0 4.5 5.0 4.80%

llmltJng enrichment PALO VERDE UNITS 1.2.3

' 3.7.17-2 AMENDMENT NO. ~.203

Fuel Region Fuel Region Fuel Region Fuel Region Fuel Region Fuel Region Notes*

1 2

3 4

5 6

Table 3.7.17-1 Fuel Regions Ranked by Reactivity After SFP transition Spent Fuel Assembly Storage 3.7.17 Highest Reactivity (See Note 2)

Lowest Reactivity

1. Fuel Regions are defined by assembly average burnup. initial enrichment1 and decay time as provided by Table 3.7.17-2 through Table 3.7.17-5.

2. Fuel Regions are ranked in order of decreasing reactivity, e.g.. Fuel Region 2 is less reactive than Fuel Region 1. etc.

3. Fuel Region 1 contains fuel with an initial maximum radially averaged enri chment up to 4. 65 wt% 235U.

No burn up is required.

4. Fuel Region 2 contains fuel with an initial maximum radially averaged enrichment up to 4. 65 wt% 235U with at least 16. 0 GWd/MTU of burnup.

5. Fuel Regions 3 through 6 are determined from the minimum burnup CBU) equation and coefficients provided in Tables 3.7.17-2 through 3.7.17-5.

6. Assembly storage is controlled through the storage arrays defined in Figure 3.7.17-1.

7. Each storage cel l in an array can only be populated with assemblies of the Fuel Region defined in the array definition or a lower reactivity Fuel Region.

1Initial Enrichment is the nominal 235U enrichment of the central zone region of fuel. excluding axial blankets. prior to reduction in 235U content due to fuel depletion.

If the fuel assembly contains axial regions of different ~

5U enrichment values. such as axial blankets. the maximum initial enrichment value is to be utilized.

PALO VERDE UNITS 1.2.3 3.7.17-2a AMENDMENT NO. -+/--2-9.203

I-

~

Figure 3.7.17-2 Before SFP t ransition Spent Fuel Assembly Storage 3.7.17 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Region 3 Cat decay times from 0 to 20 years) 45000,-----r---~r-------;------.-----..----~----.

ACCEP /~.BLE for Region 3

~ 25000 *t------J-----t-----t------,~YM-¥----+----+------1 ci

i E

i CD

>- 20000 -----1------i--------:~~"-----l-----+-----+----"---I

c E

Gl I/)

I/)

<(

ytime.

. os----.--+----,___,..--~---......-1---,.......,..-+---~-1-----~~

1.5 2.0 Decay lime J -o years PALO VERDE UNITS 1.2.3 2.5 3.0 3.5 4.0 4.5 5.0 Initial Enrichment, weight %

4.80%

limitirig

---5 years

-r-1 o years

-t!-15 years

__._ 20 years I

• h t

__,. ennc men 3.7.17-3 AMENDMENT NO. ~. 2 0 3

Fuel Region 3:

Table 3.7.17-2 After SFP transition Spent Fuel Assembly Storage 3.7.17 Burnup Requirement Coefficients Decay Coefficients Time (yr.)

A1 A2 Ai A4 0

-0.8100 6.5551

-2.9050

-21.0499 5

-0.9373 7.6381

-6.0246

-18.0299 10

-0.8706 6,8181

-3.1913

-21.0299 15

-0.7646

5. 6311 0.7657

-25.1599 20

-0.7233 5.1651 2.3084

-26.7499 Notes :

1. Relevant uncertainties are explicitly included in the criticality analysi s. For instance. no additional a 11 owance for burn up uncertai nty or enri chment uncertainty is required.

For a fuel assembly to meet the requirements of a Fuel Region. the assembly burnup must exceed t he "minimum burnup.. (GWd/ MTU ) given by the curve fit for the assembly "decay ti me"" and "ini tial enrichment." The specific minimum burnup (BU ) required for each fuel assembly is calculated from the foll owing equation :

BU = Ai

• En3 + Az

• En2 + A3

• En + A4

2. Initial enri chment. En. is the maxi mum radial average 235U enrichment.

Any En value between 2. 50 wt% 235U and 4. 65 wt% 235U may be used. Burnup credit is not requi red for an En below 2. 50 wt% 235U.

3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.

4. The 20-year coefficients must be used to calcul ate the minimum BU for an assembly with a decay time of greater than 20 years.

PALO VERDE UNITS 1.2.3 3.7.17-3a AMENDMENT NO. +/-J&, 203

Before SFP transiti on Spent Fuel Assembly Storage 3.7.17 Figure 3.7. 17-3 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT for Region 4 Cat decay times from 0 to 20 years) 45000 ---1---------1----+----~--

ACCE TABLE fo Region 4 35000t------+---------11-----+----+-_,:.....~~.....-~----+-------1 I-E 30000-+----+----__,1------+-+--+.~;,£-----1-----1---*~-t c

E

~

E 20000~---1-----4--+-....4'-ti"'----l-----+----+--~

• ..__~

~

co

~

JS

~ 20000+----!-----,.<i-7¥:-H----+-----+-----+----+--~*;--~

(I)

(I)

<(

Note:

time.

o-1-.--;...--+----.......,..---......+--.----l--.----.---1-----...--+-~---~

1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial Enrichment, weight %

4.80%

D 1i I I

limiting ecay 1me

~Oyears

--syears

-.-1oyears

-ll!-15years

-2oyeara

---

~.. ennchment PALO VERDE UNITS 1.2.3 3.7.17-4 AMENDMENT NO. ~. 203

Fuel Region 4:

Table 3.7.17-3 After SFP transition Spent Fuel Assembly Storage 3.7.17 Burnup Requirement Coefficients Decay Coefficients Time (yr.)

A1 k_

A3 A4 0

0.0333

-2.1141 27.4985

-41. 8258 5

-0. 2105 0.2472 19.7919

-34.2641 10 0.0542

-2.5298 28.0953

-41. 7092 15

0. 3010

-5. 0718 35.6966

-48.5494 20 0.4829

-6.9436

41. 3118

-53.6182 Notes:

1. Relevant uncertainties are explicitly included in the criticali ty analysis.

For instance. no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to 'meet the requirements of a Fuel Region. the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time" and "i nitial enrichment... The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation:

BU = Ai

• En3 + A2

• En2 + A3

• En + A4

2. Initial enrichment. En. is the maxi mum radial average 235U enri chment.

Any En value between 1. 75 wt% 235U and 4. 65 wt% 235U may be used.

Burnup credit is not required for an En below 1. 75 wt% 235U.

3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.

4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.

PALO VERDE UNITS 1.2.3 3.7.17-4a AMENDMENT NO. +/-2-9,203

Fuel Region 5:

Table 3.7.17-4 After SFP transition Spent Fuel Assembly Storage 3.7.17 Burnup Requirement Coefficients Decay Coefficients Time (yr.)

A1 A2 k

A4 0

0.1586

-3.0177 28.7074

-39.8636 5

-0.2756 1.3433 14.5578 *

-26.4388 10

-0.2897 1.3218 14.6176

-26.4160 15

-0.0736

-0.9107

21. 2118

-32.1887 20 0.1078

-2.7684

26. 6911

-36.9873 Notes:

1. Relevant *uncertainties are expli citly included in the criticality analysis.

For instance. no additional a 11 owance for burn up uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a Fuel Region. the assembly burnup must exceed the "mi ni mum burnup" (GWd/MTU) given by the curve fit for the assembly "decay time " and "initial enrichment." The specific minimum burnup (BU) required for each fuel assembly is calculated from the fol lowing equation :

BU = A1

• En3 + Az

• En2 + A3

• En + A4

2. Initial enrichment, En. is the maximum radi al average 235U enrichment.

Any En val ue between 1. 65 wt% 235U and 4. 65 wt% 235U may be used.

Burnup credit is not required for an En below 1. 65 wt% 235U.

3. It is acceptable to linearly interpolate between calculated BU limits based on decay ti me.

4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.

PALO VERDE UNITS 1.2.3 3.7.17-5a AMENDMENT N0. 203

Fuel Region 6:

Table 3.7.17-5 After SFP transition Spent Fuel Assembly Storage 3.7.17 Burnup Requirement Coefficients Decay Coefficients Time (yr.)

A1 A2.

A3 A4 0

0.4890

-6.7447 42.7619

-49.3143 5

0.5360

-6. 9115 41.1003

-46. 6977 10 0.4779

-6.1841 37.6389

-43.0309 15 0.4575

-5.8844 35.8656

-41.0274 20 0.3426

-4.7050

31. 8126

-37.2800 Notes:

1. Relevant uncertainties are explicitly included in the criticality analysis.

For instance. no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a Fuel Region. the assembly burn up must exceed the "ini ni mum burn up.. ( GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment...

The specific minimum burnup (BU) required for each fuel assembly is calculated from the following equation:

BU = Ai

• En3 + Az

• En2 + A3

• En + A4

2. Initial enrichment. En. is the maximum radial average 235U enrichment.

Any En value between 1. 45 wt% 235U and 4. 65 wt% 235U may be used.

Burnup credit is not required for an En below 1. 45 wt% 235U.

3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.

4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decay time of greater than 20 years.

PALO VERDE UNITS 1.2.3 3.7.17-6a AMENDMENT NO. 20.3

After SFP transition Spent Fuel Assembly Storage 3.7.17 Array A Figure 3.7.17-1 Allowable Storage Arrays Two Regi on 1 assemblies (1) checkerboarded with two blocked cell s (X).

The Regi on 1 assemblies are each in a cell with a stainless steel L-insert. No NETCO-SNAP-IN inserts are credited.

Array B Two Region 1 assemblies (1) checkerboarded with two cell s containing trash cans (TC). The Region 1 assemblies are each in a cell with a stainless steel L-insert.

Every cell without a stainless steel L-insert must contain a NETCO-SNAP-IN insert.

Array C Two Region 2 assemblies (2) checkerboarded with one Regi on 3 assembly (3) and one blocked cell (X ). The Region 2 assemblies are each in a cell with a stainless steel L-insert. The Region 3 assembly is in a cell containing a NETCO-SNAP-IN insert.

Array D One Region 2 assembly (2) checkerboarded with three Regi on 4 assemblies (4). The Regi on 2 assembly and the diagonally located Region 4 assembly are each in a storage cell with a stainless steel L-insert. The two storage cells without a stainless steel L-insert contain a NETCO-SNAP-IN insert.

Array E Four Region 5 assemblies (5).

Two storage cells contain a stainless steel L-insert.

One cell contains a NETCO-SNAP-IN insert.

One storage cell contains no insert.

Array F Four Region 6 assemblies (6 ). Two storage cell s contain a stainless steel L-insert. The other two cell s contain no inserts.

Notes:

1. The shaded locations indicate cells which contain a stainless steel L-insert.

2. A blocked cell (Xl contains a blocking device.

1 x

1 TC 2

3 2

4 5

5 6

6

3. NETCO-SNAP-IN inserts must be ori ented in the same direction as the stainless steel L-inserts.

4. NETCO-SNAP-IN inserts are only located in cells without a stainless steel L-insert.

x 1

TC 1

x 2

4 4

5 5

6 6

5. Any cell containing a fuel assembly or a TC may instead be an empty (water-filled) cell in all storage arrays.

6.

Any storage array location designated for a fuel assembly may be replaced with non-fissile materi al.

7.

Interface requirements: Each cell is part of up to four 2x2 arrays and each cell must simultaneously meet the requirements of all those arrays of which it is a part.

PALO VERDE UNITS 1.2.3 3.7.17-7a AMENDMENT NO. 203 I

Before SFP transition Design Features 4.0 4.0 DESIGN FEATURES (continued ).

4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shal l be maintained with:

a.

Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.80 weight percent;

b.

keff < 1.0 if fully flooded with unborated water.

which includes an al lowance for biases and uncertainties as described in Section 9.1 of the UFSAR;

c.

keff s 0.95 if fully flooded with water borated to 900 ppm. which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR.

d.

A nominal 9.5 inch center-to-center distance between adj acent storage cell locations.

e.

Region 1:

Fuel shall be stored in a checkerboard (two-out-of-four) storage pattern.

Fuel that qualifies to be stored in Regions 1. 2. 3. or 4 in accordance with Figures 3.7.17-1. 3.7.17-2. or 3.7.17-3. may be stored in Region 1.

f.

Region 2:

Fuel shal l be stored in a repeating 3-by-4 storage pattern in which Region 2 (two-out-of-twelve) assemblies and Region 4 (ten-out-of-twelve) assemblies are mixed as shown in Section 9.1 of the UFSAR.

Only fuel that qualifies to be stored in Regions 2. 3. or 4. in accordance with Figures 3.7.17-1. 3.7.17-2. or 3.7.17-3. may be stored in Region 2.

g.

Region 3:

Fuel shall be stored in a four-out-of-four storage pattern.

Only fuel that quali fies to be stored in Regions 3 or 4. in accordance with Figures 3.7.17-2 or 3.7.17-3. may be stored in Region 3.

PALO VERDE UNITS 1.2.3 4.0-2 AMENDMENT NO. ~. 203

4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality After SFP transition Design Features 4.0.

4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.65 weight percent;

b.

keff < 1.0 if fully flooded with unborated water.

which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR;

c.

keff ~ 0.95 if fully flooded with water borated to 1600 ppm. which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR.

d.

A nominal 9.5 inch center-to-center distance between adjacent storage cell locations.

e.

Fuel assemblies are classified in Fuel Regions 1-6 as shown in Tables 3.7.17-1 through 3.7.17-5.

(continued)

PALO VERDE UNITS 1.2.3 4.0-2a AMENDMENT NO. ~. 203

Before SFP transition Design Features 4.0 4.0. DESIGN FEATURES (continued)

h.

Region 4:

Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2 (two-out-of-t welve) assemblies and Region 4 (ten-out-of-twelve) assembl ies are mixed as shown in Section 9. 1 of the UFSAR.

Only fuel that quali fies to be stored in Region 4 in accordance with Figure 3.7.17-3 shall be stored in Region 4.

4.3.1.2 The neY.1 fuel storage racks are designed and shal l be maintained with:

,4.3.2 Drainage

a.

Fuel assemblies having a ~aximum radially averaged U-235 enrichment of 4.80 weight percent:

b.

keff ~ 0.95 if fully flooded with unborated water.

which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR:

c.

keff ~ 0. 98 if moderated by aqueous foam. which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR; and

d.

A nominal 17 inch center to center distance between fuel assemblies placed in the storage racks.

The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 137 feet - 6 inches.

4. 3. 3 Capacity The spent fuel storage pool is designed and shal l be maintained with a storage capacity limited to no more than 1329 fuel assemblies.

PALO VERDE UNITS 1. 2.3 4.0-3 AMENDMENT NO. ~.203

After SFP transition Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

4.3.2 Drainage

a.

Fuel assemblies having a maximum radially averaged U-235 enrichment of 4.65 weight percent:

b.

keff 5 0.95 if fully flooded with unborated water.

which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR :

c.

ketf 5 0. 98 if moderated by aqueous foam. which includes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR; and

d.

A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center distance between fuel assemblies placed in the storage racks.

\\

The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 137 feet - 6 inches.

4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuel assemblies.

PALO VERDE UNITS 1.2:3 4.0-3a AMENDMENT NO. +2-B. 203

Before SFP transition Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)

4.

In Regulatory Guide 1.129. Regulatory Position 3.

Subsection 5.4.1. "State of Charge Indicator." the following statements in paragraph (d) may be omitted:

"When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements. the battery is near full charge.

These measurements shal l be made after the initial ly high charging current decreases sharply and the battery voltage rises to approach the charger output voltage."

5.

In lieu of RG 1.129. Regulatory Position 7, Subsection 7.6. "Restoration." the* following may be used: "Following the test. record the float voltage of each cel l of the string."

b.

The program shall include the following provisions:

1.

Actions to restore battery cells with float voltage

< 2.13 V;

2.

Actions to determine whether the float voltage of the remaining bat tery cel ls is ~ 2. 13 V when the float voltage of a battery cell has been found to be

< 2.13 V;

3.

Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;

4.

Limits on average electrolyte temperature. battery connection resistance. and battery terminal voltage; and

5.

A requirement to obt ain specific gravity readings of all cells at each discharge test. consistent with manufacturer recommendations.

PALO VERDE UNITS 1.2.3

5. 5-19 AMENDMENT NO. +/--9J, 203

After SFP transition Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued) 5.5.20

4.

In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge. These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage."

5.

In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string."

b.

The program shall include the following provisions:

1.

Actions to restore battery cells with float voltage < 2.13 V; 2..

Actions to determine whether the float voltage of the remaining battery cells is ;:: 2.13 V when the float voltage of a battery cell has been found to be < 2.13 V;

3.

Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;

4.

Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and

5.

A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

Not Used PALO VERDE UNITS 1,2,3 5.5-19a AMENDMENT NO. ~. 203

5.5 Programs and Manuals (continued)

After SFP transition Programs and Manuals 5.5 5.5.21 Spent Fuel Storage Rack Neutron Absorber Monitoring Program Certain storage cells in the spent fuel storage racks utilize neutron absorbing material that is credited in the spent fuel storage rack criticality safety analysis to ensure the limitations of Technical Specifications 3.7.17 and 4.3.1.1 are maintained.

In order to ensure the reliability of the neutron absorber material, a monitoring program is provided to confirm the assumptions in the spent fuel pool criticality safety analysis.

The Spent Fuel Storage Rack Neutron Absorber Monitoring Program shall require periodic inspection and monitoring of spent fuel pool test coupons and neutron absorber inserts on a performance-based frequency, not to exceed 10 years.

Test coupons shall be inspected as part of the monitoring program.

These inspections shall include visual, B-10 areal density and corrosion rate.

Visual in-situ inspections of inserts shall also be part of the program to monitor for signs of degradation. In addition, an insert shall be removed periodically for visual inspection, thickness measurements, and determination of retention force.

PALO VERDE UNITS 1,2,3 5.5-20a AMENDMENT NO. ~. 203

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 203, 203, AND 203 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-41 NPF-51 AND NPF-74 ARIZONA PUBLIC SERVICE COMPANY ET AL.

PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 DOCKET NOS. STN 50-528, STN 50-529, AND STN 50-530

1.0 INTRODUCTION

By letter dated November 25, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15336A087), as supplemented by letters dated January 29, June 30, October 6, November 9, and November 23, 2016; and March 3 and May 24, 2017 (ADAMS Accession Nos. ML16043A361, ML16182A519, ML16286A240, ML16321A002, ML17167A215, ML17062B036, and ML17144A376, respectively), Arizona Public Service Company (the licensee) requested changes to the technical specifications (TSs) for Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (PVNGS). The amendments would revise the TSs to include the results of an updated criticality safety analysis for both new and spent fuel storage. The proposed changes include changes to TS 3.7.17, "Spent Fuel Assembly Storage,"

and TS 4.3.1, "Criticality," and would add a new program: TS 5.5.21, "Spent Fuel Storage Rack Neutron Absorber Monitoring Program." This license amendment request (LAA) is being submitted to correct non-conservative errors that were identified in the current licensing basis that occurred as a result of License Amendment No. 125 (ADAMS Accession No. ML003689038), which increased the storage capacity of the fuel pool by allowing credit for soluble boron and decay time in the criticality safety analysis. Since the proposed spent fuel pool (SFP) TSs include a reduction in the maximum weight percent (w0/o) 235Uranium (U), the new fuel storage racks (NFSR) TSs are also proposed to be revised to reflect the new maximum w0/o 235U.

The supplemental letters dated June 30, October 6, November 9, and November 23, 2016; and March 3 and May 24, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on April 5, 2016 (81FR19644).

2.0 REGULATORY EVALUATION

The NAG staff reviewed the proposed revision to the TSs against the regulations and guidance described below.

2.1 Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.36(c)(4), "Design features," requires "Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section."

2.2 Paragraph 50.68(b)(1) of 10 CFR requires, "Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water."

2.3 Paragraph 50.68(b)(2) of 10 CFR requires, "The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used."

2.4 Paragraph 50.68(b)(3) of 10 CFR requires, "If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used."

2.5 Paragraph 50.68(b){4) of 10 CFR requires, "If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water."

2.6 Section 50.120 of 10 CFR, "Training and qualification of nuclear power plant personnel."

2.7 General Design Criteria (GOG) 62, "Prevention of criticality in fuel storage and handling,"

of 10 CFR Part 50, Appendix A, states that "[cJritica!ity in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations."

2.8 NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" (SAP), Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling Review Responsibilities," Revision 3, March 2007 (ADAMS Accession No. ML070570006).

2.9 Jn accordance with SAP Section 9.1.2, "New and Spent Fuel Storage," Revision 4, March 2007, the review should ensure that there are no potential mechanisms that will:

(1) alter the dispersion of boron carbide (84C) in the neutron attenuation panels, and/or

{2) cause physical distortion of the tubes retaining the stored fuel assemblies (ADAMS Accession No. ML070550057).

2.10 SAP Section 13.5.1.1 "Administrative Procedures - General," Revision 1, December 2011 (ADAMS Accession No. ML112730402).

2.11 SAP Chapter 18.0 "Human Factors Engineering," Revision 2, March 2007 (ADAMS Accession No. ML070670253).

2. 12 NUREG-0711, "Human Factors Engineering Program Review Model," Revision 3, November 2012 (ADAMS Accession No. ML12324A013).

2.13 NRG Interim Staff Guidance (ISG) DSS-ISG-2010-01 "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools," Revision O (ADAMS Accession No. Mll 10620086).

3.0 TECHNICAL EVALUATION

3.1 Proposed Change The LAA proposes six different storage "arrays" (aka configurations) and six different fuel "regions" (aka categories) for the SFP. This is an increase in complexity from the current four different storage configurations and four different fuel categories. The maximum enrichment is proposed to be decreased from 4.80 to 4.65 w0/o 235U for both the SFP and NFSRs. The storage configurations must also comply with the interface requirements. The proposed changes to TS 3.7.17 and TS 4.3.1 impose the storage requirements reflecting the new SFP and NFSR criticality analyses.

3.2 Method of Review The licensee submitted Westinghouse Report, WCAP-18030-P, Revision 1, "Criticality Safety Analysis for Palo Verde Nuclear Generating Station, Units 1, 2, and 3,"1 documenting the PVNGS criticality analysis. The review was performed consistent with Section 9.1.1 of NUREG-0800.

On September 29, 2011, the NRG staff issued ISG DSS-ISG-2010-1. The purpose of the ISG is to provide updated review guidance to the NRC staff to address the increased complexity of recent SFP license applications. The NRC staff used ISG DSS-lSG-2010-1 for the review of the current application.

1 A publicly available version (WCAP-18030-NP, Revision 1) is under ADAMS Accession No. ML 16321 AOD3.

The NRC staff issued an internal memorandum on August 19, 1998, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," containing guidance for performing the review of SFP nuclear criticality safety (NCS) analysis. This memorandum is known colloquially as the "Kopp Memo" (ADAMS Accession No. ML003728001 ), named after the author. While the Kopp Memo does not specify a methodology, it does provide some guidance on the more salient aspects of an NCS analysis, including computer code validation. The guidance is germane to boiling-water reactors and pressurized-water reactors {PWRs), borated and unborated.

3.3 SFP NCS Analysis Review 3.3.1 Computational Methods For the criticality calculation, the licensee used SCALE Version 6.1.2, with the 238 group Evaluated Nuclear Data File, Version 7 (ENDF/B-Vll) neutron cross section library. For the depletion calculation to determine the spent fuel isotopics, the licensee used the two-dimensional PARAGON code with an Evaluated Nuclear Data File, Version 6 (ENDF/B-VI) neutron cross section library. These computer codes and the nuclear data sets with them have been used in many NCS analyses, are industry standards, and are considered acceptable.

3.3.2 Criticality Code Validation The purpose of the criticality code validation is to ensure that appropriate code bias and bias uncertainty are determined for use in the criticality calculation. The ISG DSS-ISG-2010-1 references NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," January 2001 (ADAMS Accession No. ML050250061). The NRC staff used NUREG/CR-6698 as guidance for the review of the code validation methodology presented in the application.

NUREG/CR-6698 states, in part, that In general, the critical experiments selected for inclusion in the validation must be representative of the types of materials, conditions, and operating parameters found in the actual operations to be modeled using the calculationaJ method. A sufficient number of experiments with varying experimental parameters should be selected for inclusion in the validation to ensure as wide an area of applicability as feasible and statistically significant results.

SCALE 6.1.2 was used in both the code benchmark analysis and the fuel storage analysis, and includes the control module CSAS5 and the foffowing functional modules: BONAM!, CENTRM, PMC, and KENO V.a (KENO). The licensee periormed the validation of the CSAS5 sequence by comparing KENO calculated effective neutron multiplication factor (k-effective or ket1) values with several different sets of critical configurations. The licensee determined and applied separate sets of bias and bias uncertainty based on the specific storage conditions. The sources of critical configurations included the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE, 2009), NUREG/CR-6361, "Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages," March 1997 (not publicly available) and NUREG/CR-6979, "Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data," September 2008 (ADAMS Accession No. ML082880452).

The IHECSBE is an appropriate source of information for the critical experiment models.

Critical experiments from NUREG/CR-6361 contain important features such as soluble boron and poisoned fuel rods. The use of the HTC experiments documented in NUREG/CR-6979 is important to cover the actinide distribution of burned fuel. These experiments use fuel that was specifically tailored to match this actinide distribution.

Fission product keff validation was identified by the licensee as a validation gap. The analysis determined a bias to accommodate this validation gap using a method consistent with current practice. The NRC staff finds that the approach used to determine the bias and the derived values are appropriate.

NUREG/CR-6698 outlines the basic elements of validation, including identification of operating conditions and parameter ranges to be validated, selection of critical benchmarks, modeling of benchmarks, statistical analysis of results, and determination of the area of applicability. The license performed the validation of SCALE Version 6.1.2, with the 238 group ENDF/B-Vll neutron cross section library consistent with NUREG/CR-6698.

Based on above, the NRG staff finds that the information supporting the code validation is acceptable.

3.3.3 Fuel Assembly Selection Section 3.1, "Reactor Description," and Section 4.3, "Fuel Design Selection," of WCAP-18030-P provide information on fuel assembly selection. The licensee analyzed current, future, and all legacy fuel assembly designs used or expected to be used at PVNGS to establish the limiting fuel assembly design. Westinghouse supplied the STANDARD fuel assembly for Cycles 1 through 4 for all three units. Beginning with Cycle 5, Westinghouse has supplied the Value Added Pellet fuel assembly designs. The licensee also included lead use assemblies that have been used at PVNGS. The licensee included a Next Generation Fuel (NGF) design for potential future use.

The licensee performed a fuel assembly reactivity comparison and concluded that the NGF design, as described in Table 3-6, "Palo Verde Operation with IFBA [Integral Fuel Burnable Absorber] and NGF," in WCAP-18030-P, would be limiting throughout the life of the fuel assembly as compared to legacy and current fuel designs. The licensee then used the NGF design tor the rest of the analysis.

Based on the noted design considerations, the NRG staff accepts the selection of the NGF design, as described in Table 3-6, as the reference assembly design.

3.3.4 Depletion Analysis Section 4 of WCAP-18030-P provides information on the depletion analysis. To take credit for the reduction in reactivity due to fuel burnup, the spent fuel composition should be based on an appropriate depletion analysis with proper assumptions regarding depletion uncertainty, depletion parameters, and axial burnup profiles.

3.3.4.1 Depletion Uncertainty The licensee used the two-dimensional PARAGON code to calculate the isotopic composition of the spent fuel as a function of fuel burnup, initial feed enrichment, and decay time. The NRG staff has approved the use of PARAGON for PWR depletion calculations as a part of the approval of WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON."2 The uncertainty in the k611 introduced by the depletion isotopic uncertainty was addressed by applying the 5 percent of the reacf1vity decrement from depletion as an uncertainty component in the determination of the maximum kett. The NRG staff finds that this uncertainty treatment is consistent with ISG DSS-ISG-2010-1.

3.3.4.2 Depletion Parameters The ISG DSS-ISG-2010-1 provides guidance that depletion simulations should be periormed with parameters that maximize the reactivity of the depleted fuel assembly. The ISG DSS-ISG-2010-1 references NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," February 2000 (ADAMS Accession No. ML003688150), which discusses the treatment of depletion parameters. GeneraJJy, for PWR fuel, conservative depletion parameters are those that increase the production rate of plutonium (Pu) from increased fast neutron capture in 238U. Enhanced plutonium production and the concurrent diminished fission of 235U due to increased plutonium fission has the effect of increasing the reactivity of the fuel at discharge and beyond.

For fuel and moderator temperatures, NUREG/CR-6665 recommends using the maximum operating temperatures to maximize Pu production. Westinghouse report WCAP-18030-P describes the fuel temperature used in the analysis in Section 4.2.2.2, "Fuel Temperature."

WCAP-18030-P describes the moderator temperature used in the analysis in Section 4.2.3.2, "Axial Moderator Temperature Profile Selection." The analysis used a proprietary method for determining the fuel and moderator temperature used during the depletion modeling portion of the analysis. The fuel and moderator temperature used should bound the fuel during its use in the PVNGS cores.

NUREG/CR-6665 recommends using a conservative cycle average boron concentration for the depletion analysis. The licensee used a boron concentration that bounds past cycle average boron concentrations and should bound future cycle average boron concentrations for all three units.

WCAP-18030-P treated specific power and operating history consistent with ISG DSS-ISG-2010-1.

To date, PVNGS has used discrete and integral burnable poisons. The discrete burnable poisons were solid Al20 3*84C rods that replaced fuel rods. The integral burnable poisons included Erbia and Gadolinia where the poison is incorporated in the fuel pellet itself and the IFBA where the poison is applied as a coating to the fuel pellet. The licensee performed calculations to determine which burnable poison resulted in the most reactive fuel; post-irradiation in the reactor core. The licensee then used that poison for the rest of the analysis.

The licensee addressed rodded operation in WCAP-18030-P, Section 5.7. Rodded operation was not included in the analysis as the licensee stated that it has not operated with rods inserted nor does it intend to operate with rods inserted. Choosing to not include rodded operation is acceptable, but future significant rodded operation is outside the bounds of this review.

~A publicly avaNable version {WCAP-16045NP-A) is under ADAMS Accession No. ML042250322.

3.3.4.3 Axial Burnup Profiles At the beginning of life, a PWR fuel assembly will be exposed to a near-cosine axial-shaped flux, which will deplete fuel near the axial center at a greater rate than at the ends. As the reactor continues to operate, the cosine flux shape will flatten because of the fuel depletion and fission-product buildup that occurs near the center. Near the fuel assembly ends, burnup is suppressed due to leakage. If a uniform axial burnup profile is assumed, then the burnup at the ends is over predicted. Analysis discussed in NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis," March 2003 (ADAMS Accession No. ML031110292) has shown that, at assembly burnups above about 10 to 20 gigawatt-days/metric ton uranium, this results in an underprediction of k611. Generally the underprediction becomes larger as burnup increases. This is what is known as the "end effect." Proper selection of the axial burnup profile is necessary to ensure that ke11 is not underpredicted due to the end effect.

The ISG DSS-ISG-2010-1 recommends using the appropriate axial burnup profiles from NUREG/CR-6801, but arrows for the use of site-specific axial burnup profiles. The NCS analysis documented in WCAP-18030-P did not use the axial burnup profiles from NUREG/CR-6801; instead, it used site-specific profiles. A description of how the axial burnup profiles were derived is provided in Section 4.2.3.1 of WCAP-18030-P. The licensee used past and current core design information along with predicted future core designs to derive numerous axial burnup profiles. Those profiles were then binned according to burnup. The licensee evaluated those to determine the limiting profile in each bin. For the criticality analysis the licensee used the limiting profile from each bin along with a uniform burnup profile for each scenario, taking the most reactive of the two cases for the analysis.

3.4 Criticality Analysis 3.4.1 Calculation of K-effective PVNGS has three SFPs, one in each unit, that are identical in layout. Each SFP contains a

• single rack design and each SFP is surrounded by a concrete wall with a stainless steer liner.

The SFP storage racks are made up of individual modules, each of which is an array of fuel storage cells. The storage racks are comprised of 17 modules: twelve 8x9 arrays, four 8x12 arrays, and one 9x9 array. The storage racks are stainless steel honeycomb structures with rectangular fuel storage cells compatible with fuel assembly materials and the spent fuel borated water environment. The fuel assembly spacing of a nominal 9.5 inches center-to-center distance between adjacent storage cell locations is a minimum value after allowances are made tor rack fabrication tolerances and predicted deflections resulting from a safe shutdown earthquake.

The SFP racks contain a stainless steel L-insert in every other cell location as shown in the PVNGS Updated Final Safety Analysis Report, Figure 9.1-5, to help center the new fuel assemblies within this space. The stainless steel L-inserts are offset 11/ 16 inch from the cell wall.

A NETCO-SNAP-IN rack insert will be instaJJed in some cells with a fuel assembly depending on the particulars of the storage array. A NETCO-SNAP-JN@ will not be installed in a cell that contains a stainless steel L-insert.

The licensee proposes six different storage c9nfigurations with six different fuel categories for the SFP. The analysis treats each storage configuration as independent from the others calculating separate biases and uncertainties for each, and then imposes interface restrictions when one storage configuration would abut another.

The manufacturing tolerances of the storage racks and fuel assemblies contribute to the reactivity. Determination of the maximum keff should consider either (1) a worst-case combination with mechanical and material conditions set to maximize keff or (2) a sensitivity study of the reactivity effects of tolerance variations. If used, a sensitivity study should include all possible significant tolerance variations in the material and mechanical specifications of the racks.

Section 5.2.3.1.1, "Manufacturing Tolerances," of WCAP-18030-P describes how the licensee's analysis evaluated the manufacturing tolerances of the storage racks and fuel assemblies. To determine the delta k (Lik) associated with a specific manufacturing tolerance, the licensee used KENO to calculate the kett for the reference condition and the ke11 for the perturbed case. The reference condition is the condition with nominal dimensions and properties. All tolerance perturbations were applied in the direction that increases reactivity relative to the nominal condition. The applicant calculated a separate set of tolerance uncertainties for each burnup/enrichment combination included in the loading curve. The list of manufacturing tolerances of the storage racks and fuel assemblies is consistent with ISG OSS-JSG-2010-1 and recent criticality analyses. Based on the considerations discussed above, the NRC staff finds the applicant's treatment of manufacturing tolerances acceptable.

The criticality analysis should account for the temperature corresponding to the highest reactivity. The licensee performed the base criticality analysis at a nominal SFP temperature and evaluated the reactivity effect of higher and lower temperatures. If a higher or lower SFP temperature resulted in a higher reactivity, a bias was added to the nominal case. The NRC staff accepts the licensee's treatment of the temperature bias.

Some fission products are gases or volatile. A portion of these fission products migrate from the location where they were generated. If these fission products are to be credited in the criticality analysis, their potential to migrate needs to be considered. The licensee evaluated these fission products in a manner consistent with current NRC guidance.

Depending on several factors the fuel assembly grids may expand over the course of their utilization in the reactor in a way that increases the pitch between the fuel rods. lf this occurs the interstitial moderation of the fuel assemblies will increase. Since PWR fuel assemblies are typically under moderated, this would increase the post-irradiation reactivity of the fuel assemblies. To address potential grid expansion, the licensee determined a site-specific bias.

3.4.2 Normal Conditions Normal conditions in the SFP include more than just the static condition where the fuel assemblies reside in the storage cells. While the static condition was used to set the loading requirements captured in the technical specifications, other normal conditions/activities within the SFP need to be considered to ensure a more reactive situation does not occur. The licensee considered numerous normal conditions as described in WCAP-18030-P, Sections 5.4.1 through 5.4.5.

Of note within these sections is that the licensee has several fuel assemblies that have open lattice locations where fuel pins were removed and not replaced, and reserved the option to have more in the future. An open lattice location increases the interstitial moderation of the fuel assemblies. Since PWR fuel assemblies are typically under mcx:lerated this would increase the post~irradiation reactivity of the fuel assemblies. In response to NRC requests for additional information (RAls) the licensee established controls to limit the placement of fuel assemblies with open lattice locations. These controls are articulated in the licensee's Jetter dated May 24, 2017. The NRG staff has reviewed the controls and finds them acceptable to provide reasonable assurance of compliance with the regulations for the fuel assemblies with open lattice locations. The NRC considers the licensee's incorporation of the controls for fuel assemblies with open lattice locations into the appropriate PVNGS fuel handling procedures to be an implementation requirement of this license amendment.

3.4.3 Abnormal Conditions Section 5.6, "Accident Description," of WCAP-18030-P presents the abnormal conditions considered in the analysis. The licensee considered the following abnormal conditions:

misloaded assembly, inadvertent removal of a NETCO-SNAP-IN rack insert, spent fuel temperature outside of operating range, seismic event, and dropped and misplaced fresh assembly.

The licensee appropriately identified the limiting abnormal condition and the amount of soluble boron necessary to maintain kett less than or equal to 0.95. The analysis of abnormal conditions is consistent with ISG DSS-ISG-2010-1 and recent criticality analyses. Based on the considerations discussed above, the NRC staff finds the applicant's treatment of abnormal conditions acceptable.

3.4.4 New and Interim Fuel Storage Racks and Fuel Transfer Equipment In addition to the new analysis and storage requirements for the SFP, the licensee pertormed new criticality analyses for the NFSRs, interim fuel storage rack (IFSR), and fuel handling equipment (FHE). The NFSRs are outside the SFP and are intended to be dry. Therefore, they are subject to 10 CFR 50.68(b)(2) and (3) while the IFSR and FHE are in the SFP and subject to 10 CFR 50.68(b)(4).

The licensee used essentially the same methods tor analyzing the NFSR, IFSR, and FHE as it used for the SFP. The analysis for the NFSR, JFSR, and FHE all showed significant margin to the regulatory limit and are, therefore, acceptable.

3.5 Review of Maxus Neutron Absorber Material Provided by the Licensee The proposed change would credit NETCO-SNAP-rN rack inserts made from Maxus neutron absorber material for criticality control in the SFP. The inserts would be installed in individual SFP storage rack cells in order to ensure that the requirements of TS 3.7.17 are maintained.

NETCO-SNAP-IN rack inserts have previously been approved by the NRG staff and installed in other plants. However, this is the first requested use of the Maxus neutron absorber material in an operating U.S. nuclear power plant.

Maxus is composed of a 5000 series aluminum alloy cladding, which is tightly bound to an inner core of 84C mixed with 1000 series aluminum alloy. The Maxus sheets are produced using powder metallurgy techniques similar to those used to produce Baral material, which has been widely used in US and international SFPs for over 40 years. The material qualification report for the Maxus material indicates that Maxus has a reduced porosity when compared to Baral. The report states that this reduction in porosity coupled with increased bonding of the cladding, reduces the possibility of water intrusion and blistering of the clad material.

During the manufacturing of the inserts, coupons will be cut from the material sheets and Boron-10 areal density will be measured. The licensee will, therefore, have a record of the as-built areal density of each insert before it is installed in the SFP.

3.5.1 Neutron Absorber Monitoring Program The licensee has proposed a new program in TS 5.5.21, "Spent Fuel Storage Rack Neutron Absorber Monitoring Program." This TS establishes a monitoring program that will test coupons and inserts on a periormance based frequency, not to exceed 10 years. A key aspect of the program is that the testing will require Boron-1 O areal density measurements of the coupons.

Details of the monitoring program are described below.

Each SFP will contain a coupon tree, which will be installed prior to the installation of the inserts.

In addition to being installed in the SFPs prior to the insert installation, the trees will be placed near the center of the active fuel region. This will maximize the coupons' exposure to the fuel and ensure that they experience an environment and time in service that bounds that of the inserts.

The coupons themselves are cut from the identical material used to manufacture the inserts.

Each tree contains 48 general coupons, 24 galvanic couple coupons, and 24 bend coupons.

The bent coupons are used to simulate stresses on the actual inserts, and bi-metallic coupons simulate galvanic corrosion that may occur with stainless steel, inconel, and zircaloy in the SFP.

Prior to installation of the coupon tree, each general and galvanic coupon receives a pre-characterization, which includes visual, dimensional, density, and areal density testing.

The coupons will be removed from the SFP, examined, and compared to the initial coupon condition, on a prescribed schedule described below.

~---------

-~----------

After 10 Years with Coupon Type First 10 Years Acceptable Performance General 2 coupons every 2 years 2 coupons every 4 years Bend 1 coupon every 2 years 1 coupon every 4 years Galvanic couples -

304L stainless 1 couple every 6 years Zircafoy 1 couple every 6 years lnconel 718 1 couple every 6 years In response to an NRC RAI, the licensee clarified that after the initial testing it may elect to remove coupons from one pool rather than removing coupons from all three pools for each inspection. This peliormance-based approach would be contingent on the initial testing of coupons providing similar results for all three pools. Jn addition, an assessment of the conditions in each pool would be needed to justify the similarity of the exposure conditions for the coupons and inserts in each pool. If the licensee can demonstrate that the pools experience similar chemistry, temperature, flow, and exposure, then the NRC staff finds it reasonable to sample the pools on a rotational basis rather than testing all three during each surveiJJance.

The licensee also stated that if this approach is used it will establish a program to ensure that each pool is sampled on an established periodic basis.

The licensee's acceptance criteria for the coupon surveillances establish limits for changes in dimension, weight, density, corrosion, visual anomalies, and Boron-10 areal density. The bend coupons include an acceptance criterion for bending stress.

In addition to the coupon testing, a portion of the full length inserts will be subjected to in situ visual inspection and removal for detailed inspection of wear pertormance. The in situ visual inspections will be pertormed by camera to monitor for bubbling, blistering, corrosion pitting, cracking, or flaking. An emphasis will be placed on examining for edge or corner defects. The inspections on removed inserts will measure thickness at several locations along the insert lengths and compare them to pre-service thickness measurements to identify wear. The fu!!

length insert inspections will be pertormed on the following frequency:

fter 10 Years w-it_h __ _

Inspection Type First 10 Years ceptable Performance

---

+-----~------

In situ Removal 3.6 Human Factors

~nserts ever~:_years__

nserts every 4 years 1 insert every 1 O years

---

*--------------~

3.6.1 Description of the Spent Fuel Storage Rack Neutron Absorber Monitoring Program The LAA dated November25, 2015, proposes to revise TS and TS Bases to include the results of an updated criticality safety analysis for both new and spent fuel storage. Changes to TS 3.7.17 and TS 4.3.1 would update the TS to reflect the new SFP configuration and the limits of the new criticality analysis (see Attachments 1 and 2 of the LAR).

The LAR also proposes a new TS 5.5.21, which implements a Spent Fuel Storage Rack Neutron Absorber Monitoring Program. The program includes periodic performance-based inspection of SFP test coupons and neutron absorber inserts. The program is based upon protecting the limits described in TSs 3. 7.17 and 4.3.1.1. Details of how the program will be implemented are described in LAA Section 3.1.4, "NETCO-SNAP-IN@ Rack Inserts." Methods used include pre-test and post-test measurements of the NETCO-SNAP-JN neutron absorbers and test coupons via visual inspection, high resolution photography, neutron attenuation, stress relaxation, blister and pit characterizations, and measurements of thickness, length, width, dry weight, and density. Inspection characterizations are documented in Tables 1 through 5. The tables provide acceptance/rejection criteria for various coupon properties and related inspection periodicity.

The LAA explains that the purpose of this proposed revision is to correct a non*conservatlve TS approved by License Amendment No. 125. The addition of TS 5.5.21 is intended to reduce the likelihood of the criticality accidents in the SFP described in LAA Section 3.1.3, "Spent Fuel Pool Criticality Analysis." The licensee describes several human periormance enhancements intended to reduce the probability of a fuel mislead event in LAA Section 3.1.5, "Spent Fuel Poof Configuration Control (Human Performance Enhancements)."

The January 29, 2016, supplement to the LAA clarifies the types of potential fuel mislead events, including the possibility of an inadvertent removal of an insert. The licensee explains that the effect of this error is bounded by other of the potential misleading errors described in the LAR.

The June 30, 2016, RAI response clarifies the intent of a vague acceptance criterion labeled "evidence of visual indications of performance inhibitors" on LAA Tables 2, 3, 5. This criterion was not intended to be a specific acceptance criterion; rather, the licensee is using this to prompt operators to identify additional abnormalities not listed in the other more specific criteria.

NRC Staff Evaluation

The licensee describes a coherent and logical program that periodically and methodically tests the neutron absorbers and test coupons using a diverse variety of methods. The program provides assurances that mislead events will be prevented when possible and that the effects will be mitigated in the event that a misload event does occur. Therefore, the NRC staff finds this treatment to be acceptable.

3.6.2 Operating Experience Review (OER)

Section 3.1.4 of the LAR indicates that the monitoring program will be based upon:

manufacturer recommendations, current industry operating experience, Nuclear Energy Institute guidance, and NRC safety evaluations for other plants. Jn addition, Section 3.1.5 indicates that OER for fuel misleading events will be used to inform the training program.

The licensee further elaborates on the details of the OER that was conducted in the RAI response dated June 30, 2016.

NRC Staff Evaluation

The licensee has conducted an OER that is consistent with the guidance. In addition, the licensee plans to continue to monitor relevant industry operating experience to ensure that the program remains effective. Therefore, the NRG staff finds this treatment to be acceptable.

3.6.3 Training and Qualifications Section 3.1.5 of the LAA indicates that there will be training for use of move sheets {an administrative control process) and training for positions during fuel movement. This training will include operating experience related to fuel mislead events. Training will be provided for arr individuals operating the spent fuef handling machine and those acting as independent verifiers.

Training will also be provided to those who generate or revise move sheets.

NRC Staff Evaluation

The licensee has described several areas of the training program that will be updated to reflect the proposed changes. This training will be updated to included industry operating experience.

The process also relies on several administrative and other controls to prevent errors such as the independent generation of move sheets for movement of fuel and blocking devices. These controls typically utilize standard industry practices, which do not need additional training, but supplement the training for processes described in the application. Therefore, the NRC staff finds this treatment to be acceptable.

3.6.4 Task Analysis Section 3.1.5 of the LAR identifies several detailed "Human Periormance Enhancements" which will be used to reduce the likelihood of a multiple fuel misread error. These include:

Control of Move Sheet Generation Control of Fuel Movement Use of Blocking Devices Confirmation of Configuration Control Section 3.1.5 afso contains plans for mitigating a misload event in case the preventative techniques listed above fail.

NRC Staff Evaluation; Although the licensee does not provide a description of a formal task analysis in the LAA, it is clear from the submittal that administrative controls described in Section 3.1.5 have considered error prevention at several subsequent steps of the fuel movement process.

The licensee provides details for each of the areas above that use standard human periormance techniques such as independent verification and work package generation by independent parties (for move sheets and blocking device move sheets) amongst other techniques listed in the section.

The description is thorough and should minimize the probability of a misfoad event. In the event that the several actions listed above fail and a mislead event does occur, the licensee provides a reasonable plan for mitigation (via boron concentration management).

The NRC staff finds that the licensee plan for error prevention and mitigation is thorough and has several barriers in place to ensure that fuel mislead errors do not occur. Therefore, the staff finds this treatment to be acceptable.

3.6.5 Design Implementation Section 3.3 "Spent Fuel Pool Transition Plan," of the LAR indicates that the implementation of this amendment will occur over the course of 2 years. During that time, there will be two sets of TS, one labeled "Before SFP Transition" and the other labeled "After SFP Transition." The transition from the "before" to the "after" TS will occur by module with the Shift Manager indicating that the transition is complete with an entry in the control Jog. At that time, the module will be considered "transitioned" and the "After SFP Transition" TS will then apply to that module. When all modules have been transitioned, the licensee will submit an administrative TS change to remove the "Before SFP Transition" TS pages. The final step includes performing a 100 percent pool verification that confirms several parameters of importance including the appropriate installation of inserts and blocking devices.

The supplement dated January 29, 2016, describes the analyses used to identify potential performance shaping factors that may affect appropriate loading. The licensee used both the Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) method and the Electric Power Research Institute Cause-Based Decision Tree Method to analyze fuel loading both prior to the transition period and during the transition period. No performance shaping factors were found to have significant impacts prior to the transition period.

The licensee will enact additional administrative controls during the transition period, including adding an additional independent verification of fuel move sheets and the suspension of other coincident fuel movement (such as dry cask loading or fuel receipt). The licensee also indicates that the levels of soluble boron are such that even if a multiple mislead event was to occur that reactivity would remain within allowable specifications.

NRC Staff Evaluation

The licensee considers errors that may occur during the transition period and has devised appropriate safeguards to prevent errors from occurring during this period. In addition to the use of administrative controls intended to minimize errors, the licensee uses soluble boron in a way that would mitigate the effects of any potential errors that do occur. Therefore, the NRC staff finds this treatment to be acceptable.

3.7 Summary Based on its review of the licensee's coupon sampling program and material qualification tests, summarized above, the NRC staff finds that the NETCO-SNAP-IN rack inserts made from Maxus neutron absorber are compatible with the environment of the PVNGS SFPs. Also, the staff finds that the proposed surveillance program, which includes visual, physical, and confirmatory tests, is capable of detecting potential degradation of the rack insert material that could impair its neutron absorption capability. Therefore, the staff concludes that the use of NETCO-SNAP-IN rack inserts made from Maxus neutron absorber at PVNGS is acceptable.

The licensee submitted the LAR to address the non-conservatisms in the SFP NCS analysis of record and the associated TSs. The LAR is supported by Westinghouse report WCAP-18030-P, Revision 1, documenting the criticality analysis for PVNGS new and spent fuel storage. The proposed changes to TS 3.7.17 and TS 4.3.1 impose the storage requirements reflecting the new SFP, NFSR, IFSR, and FHE criticality analysis.

The NRC staff reviewed the analysis to ensure that the assumptions and analytical techniques used were adequately substantiated to conclude at a 95 percent probability, 95 percent confidence level, that the regulatory requirements will be met Based on the discussions in Section 3 of this safety evaluation, the NRC staff finds reasonable assurance that the licensee will comply with the applicable regulatory requirements. Therefore, the NRC staff concludes that the proposed TS changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Arizona State official was notified of the proposed issuance of the amendments on July 6, 2017. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on April 5, 2015 (81FR19644). 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 environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission 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) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Kent Wood, NRR/DSS/SNPB Matthew Yoder, NRR/DE/ESGB Brian Green, NRR/DRNAPHB Date: July 28, 2017

SUBJECT PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3-ISSUANCE OF AMENDMENTS TO REVISE TECHNICAL SPECIFICATIONS TO INCORPORATE UPDATED CRITICALITY SAFETY ANALYSIS (CAC NOS. MF7138, MF7139, AND MF7140) DATED JULY 28, 2017 DISTRIBUTION:

PUBLIC LPL4 r/f RidsACRS_MailCTR Resource RidsNrrDorl Resource RidsNrrDorllp14 Resource RidsNrrLAPBlechman Resource RidsNrrOd Resource RidsNrrPMPaloVerde Resource RidsNrrDssSnpb Resource RidsRgn4Mai1Center Resource ADAMS A ccess1on o.:

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NAME SLingam PBlechman w/comments SWeerakkody AJohnson DATE 07/24/17 07/13/17 08122116 11/07/16 OFFICE NRR/DSS/SNPB/BC*

NRR/DE/EICB/BC**

NRR/DSS/STSB/BC (A)** OGG NLO JWachutka I subject to NAME Rlukes MW ate rs JWhitman e-mail comments DATE 06/30/17 07/14/17 07/18/17 07/21/17 OFFICE NRR/DORULPL4/BC NRR/DORL/LPL4/PM NAME RPascarelli Slingam DATE 07128117 07/28/17 OFFICIAL AGENCY RECORD