NL-02-063, Updated Replacement of Technical Specifications for License Amendment Request (LAR 01-010) Regarding Spent Fuel Storage Pit Rack Criticality Analysis with Soluble Boron Credit

Updated Replacement of Technical Specifications for License Amendment Request (LAR 01-010) Regarding Spent Fuel Storage Pit Rack Criticality Analysis with Soluble Boron Credit

ML021220355
Person / Time
Site:	Indian Point
Issue date:	04/29/2002
From:	Dacimo F Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-02-063, TAC MB2989
Download: ML021220355 (16)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

Indian Point Energy Center LUJPVeW 296 Broadway, Suite 1 RO Box 249 Buchanan, NY 10511 0249 April 29, 2002 Re: Indian Point Unit No. 2 Docket No. 50-247 NL-02-063 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop 0-P1-17 Washington, D.C. 20555-0001

SUBJECT:

Updated Replacement Pages of Technical Specifications for License Amendment Request (LAR 01-010) Regarding Spent Fuel Storage Pit Rack Criticality Analysis with Soluble Boron Credit, Indian Point Energy Center, Unit No. 2 (TAC No. MB2989)

References:

1. Entergy Nuclear Operations, Inc. letter (NL-01-110) to NRC, "License Amendment Request (LAR 01-010) for Spent Fuel Storage Pit Rack Criticality Analysis with Soluble Boron Credit," dated September 20, 2001 Entergy Nuclear Operations, Inc. (ENO) hereby submits updated replacement pages for changes to the Indian Point Energy Center, Unit No. 2 (IP2) Technical Specifications (TS) requested by letter dated September 20, 2001 (Ref. 1). These updated pages remove the background shading from the added text, delete the text marked by strikeouts, and add the right margin revision bars required for issuance of the requested license amendment. These pages also contain other changes incorporated into the IP2 TS since the original Ref. 1 application.

Additionally, the updated replacement pages include the following editorial adjustments.

To support the addition of the proposed Insert A from Ref. 1 into TS 3.8.D.1, the action statement portion of the existing TS 3.8.D.1 has been renumbered as TS 3.8.D.2, and the existing TS 3.8.D.2 and 3.8.D.3 have been renumbered as 3.8.D.3 and 3.8.D.4, respectively. This will ensure the action statements are not confused with the required fuel storage configurations specified by the new paragraph insert. References to TS 3.8.D.2 in the existing TS 3.8.D.1 and in the associated Bases paragraph have also been renumbered to TS 3.8.D.3 accordingly. The reference to Figure 3.8-1 in the proposed TS 3.8.D.1 .d has been capitalized and Specification 1 of TS 5.4 has been aligned with the other paragraphs of that TS.

The Attachment to this letter provides the updated pages for the IP2 TS and the associated Bases, which can be directly used as replacement pages for license amendment issuance. A proposed page replacement instruction sheet is also provided.

O\C

NL-02-063 Page 2 of 3 The assessment submitted in Ref. 1 that concluded that the proposed changes to the TS did not involve a significant hazards consideration is not affected by the updated replacement TS pages submitted herewith in support of the application.

There are no commitments contained in this submittal.

Should you or your staff have any questions regarding this submittal, please contact Mr.

John F. McCann, Manager, Nuclear Safety and Licensing at (914) 734-5074.

I declare under penalty of perjury that the foregoing is true and correct.

Sincerely, Executed on 3v/O2-Fred Dacimo Vice President - Operations Indian Point Energy Center Unit No. 2 Attachment cc: See page 3

NL-02-063 Page 3 of 3 Hubert J. Miller Regional Administrator US Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. Patrick D. Milano, Senior Project Manager, Section 1 Project Directorate I Division of Licensing Project Management US Nuclear Regulatory Commission Mail Stop 0-8-2C Washington, DC 20555 NRC Senior Resident Inspector US Nuclear Regulatory Commission PO Box 38 Buchanan, NY 10511 Mayor, Village of Buchanan 236 Tate Avenue Buchanan, NY 10511 Mr. Paul Eddy NYS Department of Public Service 3 Empire Plaza Albany, NY 12223 Mr. William F. Flynn NYS ERDA Corporate Plaza West 286 Washington Ave. Extension Albany, NY 12223-6399

ATTACHMENT TO NL-02-063 TECHNICAL SPECIFICATION AND BASES PAGES ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT ENERGY CENTER UNIT NO. 2 DOCKET NO. 50-247

ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO. DPR-26 DOCKET NO. 50-247 Replace the following pages of the 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.

Remove Pages Insert Pages ix ix 3.8-4 3.8-4 3.8-5 3.8-5 3.8-6 3.8-6 Figure 3.8-1 (DELETED) Figure 3.8-1 Figure 3.8-2 Figure 3.8-2 Figure 3.8-3 Figure 3.8-3 Figure 3.8-4 Figure 3.8-5 5.4-1 5.4-1

LIST OF FIGURES Title Figure No.

Reactor Core Safety Limit-Four Loops In Operation 2.1-1 PORV Open Pressure 3.1 .A-1 Maximum RCS Pressure: OPS Inoperable and 1 Charging Pump Capable of Injecting into the RCS 3.1 .A-2 Maximum RCS Pressure: OPS Inoperable and 2 Charging Pumps Capable of Injecting into the RCS 3.1 .A-3 Maximum RCS Pressure: OPS Inoperable and 3 Charging Pumps Capable of Injecting into the RCS 3.1 .A-4 Maximum RCS Pressure: OPS Inoperable and Start of 1 RCP with SGs 40°F Hotter than RCS 3.1 .A-5 Maximum RCS Pressure: OPS Inoperable and Start of 1 RCP with SGs 100°F Hotter than RCS 3.1 .A-6 Reactor Coolant System Heatup and Leak Test Limitations Applicable for the First 25 EFPY 3.1 .B-1 Reactor Coolant System Cooldown Limitations Applicable for the First 25 EFPY 3.1.B-2 Spent Fuel Storage Rack Layout 3.8-1 Region 1-1 Limiting Fuel Burnup versus Initial Enrichment 3.8-2 Region 1-2 Minimum Number of IFBA Rods versus Initial Enrichment 3.8-3 Region 2-1 Limiting Fuel Burnup versus Initial Enrichment 3.8-4 Region 2-2 Limiting Fuel Burnup versus Initial Enrichment 3.8-5 Required Hot Shutdown Margin versus Reactor Coolant Boron Concentration 3.10-1 DELETED 4.3-1 Map Defining Unrestricted Areas for Radioactive Gaseous and Liquid Effluents 5.1-1 Amendment No. ix

not be moved on or above El. 95' in the Fuel Storage Building. Additionally, loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool shall not be moved over spent fuel in the spent fuel pit. The weight of installed crane systems shall not be considered part of these loads.

2. The spent fuel storage pit water level shall be maintained at an elevation of at least 93'2". In the event the level decreases below this value, all movement of fuel assemblies in the spent fuel pool storage pit and crane operations with loads over spent fuel in the spent fuel pit shall cease and water level shall be restored to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

D. The following conditions are applicable to the spent fuel pit anytime it contains fuel:

1. The spent fuel racks are categorized as either Region 1-1, 1-2, 2-1, or 2-2 as specified in Figure 3.8-1. Fuel assemblies to be stored in the spent fuel storage racks are qualified based on burnup, enrichment, and cooling time as specified in Figures 3.8-2 through 3.8-5.

a. Storage of fuel in Region 1-1 is restricted to assemblies that meet the burnup criteria in Figure 3.8-2.

b. Storage of fuel in Region 1-2 is restricted to assemblies with enrichments

< 4.5 weight percent (w/o) or < 5.0 w/o with a minimum number of Integral Fuel Burnable Absorbers (IFBA) rods as specified in Figure 3.8-3. Fuel that meets the criteria for Region 1-2 may be stored in Region 1-1 in a checkerboard (1 out of 2 cells with every other cell left vacant) loading configuration.

c. Storage of fuel in Region 2-1 is restricted to assemblies that meet the burnup criteria in Figure 3.8-4.

d. Storage of fuel in Region 2-2 is restricted to assemblies that meet the burnup criteria in Figure 3.8-5. Fuel that meets the criteria for Region 2-1 may be stored in Region 2-2 in "peripheral" cells. As shown in Figure 3.8-1, the peripheral cells are located along the west wall and are separated by at least 3 cells.

2. In the event any fuel assembly is found to be stored in a configuration other than specified, immediate action shall be initiated to:

a. Verify the spent fuel storage pit boron concentration meets the requirements of Specification 3.8.D.3, and Amendment No. 2-8-4

b. Return the stored fuel assembly to the specified configuration.

3. At all times the spent fuel storage pit boron concentration shall be at least 2000 ppm. With the boron concentration less than this value, all fuel movement within the spent fuel storage pit shall cease and immediate action shall be initiated to restore the boron concentration to at least the minimum specified. The required boron concentration shall be verified by chemical analysis at the frequency specified on Table 4.1-2.

4. During operations described in Specification 3.8.8, the spent fuel storage pit boron concentration shall be at least equal to that required in Specification 3.8.8.2. With the boron concentration less than the specified value either:

a. Isolate the spent fuel storage pit from the refueling cavity, or

b. Take actions required by Specification 3.8.B.12.

E. Specification 3.0.1 is not applicable to the requirements of Specification 3.8.

Basis The equipment and general procedures to be utilized during refueling are discussed in the FSAR. Detailed instructions, the above-specified precautions, and the design of the fuel-handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety('). Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumentation. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.

The shutdown margin requirements will keep the core subcritical. During refueling, the reactor refueling cavity is filled with borated water. The minimum boron concentration of this water is the more restrictive of either 2000 ppm or else sufficient to maintain the reactor subcritical by at least 5% Ak/k in the cold shutdown condition with all rods inserted. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. Periodic checks of refueling water boron concentration ensure the proper shutdown margin. The specifications allow the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

In addition to the above safeguards, interlocks are utilized during refueling to ensure safe handling.

An excess weight interlock is provided on the lifting hoist to prevent movement of more than one Amendment No. 3.8-5

fuel assembly at a time. The spent fuel transfer mechanism can accommodate only one fuel assembly at a time.

The 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay time following plant shutdown and the 23 feet of water above the top of the reactor vessel flanges are consistent with the assumptions used in the dose calculations for fuel-handling accidents both inside and outside of the containment. The analysis of the fuel handling accident inside and outside of the containment takes no credit for removal of radioactive iodine by charcoal filters.

The requirement for the fuel storage building charcoal filtration system to be operating when spent fuel movement is being made provides added assurance that the offsite doses will be within acceptable limits in the event of a fuel-handling accident. The additional month of spent fuel decay time will provide the same assurance that the offsite doses are within acceptable limits and therefore the charcoal filtration system would not be required to be operating.

The spent fuel storage pit water level requirement in Specification 3.8.C.2 provides approximately 24 feet of water above fuel assemblies stored in the spent fuel storage racks.

The fuel enrichment and burnup limits in Specification 3.8.D.1, the partial credit taken for Boraflex panels, and the boron requirements in Specification 3.8.D.3 assure the limits assumed in the spent fuel storage safety analysis will not be exceeded. The analysis (Ref. 2) takes credit for the amount of Boraflex predicted to be available through 2006.

The requirement that at least one RHR pump and heat exchanger be in operation ensures that 0

sufficient cooling capacity is available to maintain reactor coolant temperature below 140 F, and sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two RHR pumps and heat exchangers operable when there is less than 23 feet of water above the vessel flange ensures that a single failure will not result in a complete loss of residual heat removal capability. With the head removed and at least 23 feet of water above the flange, a large heat sink is available for core cooling, thus allowing adequate time to initiate actions to cool the core in the event of a single failure.

References (1) FSAR Section 9.5.2 (2) Northeast Technology Corporation Report NET-1 73-01, "Criticality Analysis for Soluble Boron and Burnup Credit in the Con Edison Indian Point Unit No. 2 Spent Fuel Storage Racks."

Amendment No. 3.8-6

9n PP 9Jý 97 DF DL DJ GTON22 DC DE CP CF CM CK CK CH CH CF CF CD Cc BN BN BL BK I BJ BH BG BF BE BD BC AM AK AH AF AD AB PERIPHERAL CELL Figure 3.8-1 Spent Fuel Storage Rack Layout Amendment No.

40 30 20 10 0

1 2 3 4 5 Initial Enrichment, w/o U2..

Figure 3.8-2 Region 1 Limiting Burnup versus Initial Enrichment Amendment No.

25 20 Acceptable for Storage in Region 1-2 or Checkerboard

-a 0

15 Loading in Region 1-1 m

U 0

-~10 E

z Not Acceptable for Storage in Region 1-2 or Checkerboard Loading in Region 1-1 5

I 0

4.5 4.6 4.7 4.8 4.9 5 235 Initial Enrichment, w/o U Figure 3.8-3 REGION 1 MINIMUM NUMBER OF IFBA RODS VERSUS INITIAL ENRICHMENT Amendment No.

0 Years Cooling 60 5 Yr Cooling 10 Yr. Cooling 15Yr Cooling 50 20 Yr. Cooling t Al Acceptable for D Storage in Region 2-1 7 I 40 0

Storage inRgo ~ -

61 30 7-

/

L. 7 7

m "I

20 Accetabegofor Ntotag 10 "I,

I I . I 0 I I 1 2 3 4 5 Initial Enrichment, w/o U235 Figure 3.8-4 Region 2 Limiting Burnup versus Initial Enrichment Amendment No.

40 0 Cooling Time 5 Years Cooling

-- 10 Years Cooling

---

15YearsCooling

. 20 Years Cooling 30

,777

- 7

-7 D Acceptable for Storage in Region 2-2 7 7

(.9 20

,,/

CO 13 10

// Not Acceptable for Storage in Region 2-2

/

0 I I I I I II I I 1 2 3 4 5 Initial Enrichment, w/o U23 .

Figure 3.8-5 Region 2 Limiting Burnup versus Initial Enrichment Amendment No.

Table 4.1-2 Frequencies for Sampling Tests Maximum lime Check Frequency Between Tests

1. Reactor Coolant Samples Gross Activity (1) 5 daystweek (1) 3days Radiochemical (2) Monlhly 45 days E Determninalion Semi-annually (3) 30weeks Trhium Activity Weeldy (1) 10 days F,CI&O2 Weekly 10 days

2. Reactor Coolant Boron Boron Concenlration Twice/week 5 days

3. Refueling Water Storage Boron Concenlralion Monthly 45 days Tank Water Sample

4. Boric Add Tank Boron Concentralion Twice/week 5days

5. DELETED

6. DELETED

1. Accumulator Boron Concentration Monthly 45 days

8. Spent Fuel Pit Boron Concentralion Weeky 10 days

9. Secondary Coolant Iodine-131 Weeky (4) 10 days

10. Containment Iodine Iodine-131 and Continuous When NA*

Particulate Monitor Parliculate Activily Above Cold Shutdown (5) or Gas Monitor or Gross Gaseous Activity Amendment No. (Page I of 2)

5.4 FUEL STORAGE Applicability Applies to the capacity and storage arrays of new and spent fuel.

Objective To define those aspects of fuel storage relating to prevention of criticality in fuel storage areas.

Specifications

1. The spent fuel pit structure is designed to withstand the anticipated earthquake loadings as a Class I structure. The spent fuel pit has a stainless steel liner to ensure against loss of water.

2.A. The new fuel storage rack is designed so that it is inpossible to insert assemblies in other than an array of vertical fuel assemblies with a sufficient center-to-center distance between assemblies to assure Kf < 0.95, even if unborated water were used to fill the pit and with fuel assemblies containing a maxinmum enrichment of 5.0 weight percent U-235, and poisons, if necessary to meet the K, limit.

2.B. The spent fuel storage racks are designed and their loading maintained within the limits of Technical Specification 3.8.D.1, such that Kf < 1. 0 in unborated water and Kf, < 0.95 with credit for soluble boron and with the fuel assemblies containing a nmaxinm enrichment of 5.0 weight percent U-235 (or equivalent reactivity).

Amendment No- 5.4-1