LIC-17-0015, License Amendment Request (LAR) 17-03; Removal of Fort Calhoun Station, Unit 1 Dry Cast Loading Limits

License Amendment Request (LAR) 17-03; Removal of Fort Calhoun Station, Unit 1 Dry Cast Loading Limits

ML17160A405
Person / Time
Site:	Fort Calhoun
Issue date:	06/09/2017
From:	Fisher M Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR 17-03, LIC-17-0015
Download: ML17160A405 (23)

Text

Omaha Public Powe.rDistrict LIC-17-0015 June 9, 2017 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285 10 CFR 50.90 10 CFR 50.36

Subject:

License Amendment Request (LAR) 17-03; Removal of Fort Calhoun Station, Unit 1 Dry Cast Loading Limits

References:

1.

Letter from OPPD (T. Burke) to NRC (Document Control Desk), "Certification of Permanent Removal of Fuel from the Reactor Vessel," dated November 13, 2016 (LIC-16-0074) (ML16319A254)

2.

Letter from OPPD (M. J. Fisher) to NRC (Document Control Desk), "License Amendment Request (LAR) 17-01; Revised Fort Calhoun Station Technical Specifications to align to those requirements for decommissioning," dated March 31,2016 (LIC-17-0001) (ML17093A309)

In accordance with the provisions of 10 CFR 50.90, the Omaha Public Power District (OPPD), is submitting a request for an amendment to the Operating License for Fort Calhoun Station (FCS), Unit No. 1. The proposed amendment would delete Technical Specifications (TS) 2.8.3(6), Spent Fuel Cask Loading and associated Figure 2-11, Limiting Burnup Criteria for Acceptable Storage in Spent Fuel Cask; TS 3.2 Table 3-5(24), Spent Fuel Cask Loading; TS 4.3.1.3, Design Features associated with spent fuel casks; and portions of TS 3.2, Table 3-4(5), Footnote (4) on boron concentration associated with cask loading. The deletion of the TS sections will bring the FCS TS into conformance with 10 CFR 50.68(c) "Criticality accident requirements."

444 South 161h Street Mall Omaha, NE 68102-2247 EMitOIMfNT WITH fOUAl OPPOIITUNITY

U.S. Nuclear Regulatory Commission LIC-17-0015 Page 2 The proposed amendment would modify the TS to make the requested changes. The enclosure contains a description of the proposed changes, the supporting technical analyses, the No Significant Hazards Consideration determination and a clean and marked up version of the affected TS with their associated bases.

The proposed changes have been reviewed and approved by the Fort Calhoun Station Plant Operations Review Committee (PORC).

OPPD requests approval of the proposed license amendment by April 1, 2018, with the amendment to be implemented within 90 days of issuance. This is to coincide with the issuance of the amendment associated with FCS LAR 17-01, FCS Decommissioning TS (Reference 2).

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of Nebraska official.

There are no regulatory commitments contained within this letter.

If you should have further questions, please contact Mr. Bradley H. Blome, Director - Licensing and Regulatory Assurance, at 402-533-7270.

I declare under penalty of perjury that the foregoing is true and correct. Executed on June 9, 2017.

Mary J. Fisher Senior Director - Decommissioning Fort Calhoun Station MJF/epm

Enclosures:

OPPD's Evaluation of the Proposed Change c:

K. M. Kennedy, NRC Regional Administrator, Region IV J. S. Kim, NRC Project Manager R. S. Browder, NRC Senior Health Physicist, Region IV Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska

LIC-17-0015 Enclosure Page 1 OPPD's Evaluation of the Proposed Change License Amendment Request (LAR) 17-03:

Request to Eliminate Dry Cast Loading Limits from Plant Technical Specifications 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

Attachments: 1.

2.

Mark-up of Technical Specification and associated Bases Pages Clean Technical Specification and associated Bases Pages

LIC-17-0015 Enclosure Page 2 1.0

SUMMARY

DESCRIPTION The Omaha Public Power District (OPPD) hereby requests an amendment to Fort Calhoun Station (FCS), Unit No. 1 Renewed Facility Operating License No. DPR-40 to delete Technical Specifications (TS} 2.8.3(6), Spent Fuel Cask Loading and associated Figure 2-11, Limiting Burnup Criteria for Acceptable Storage in Spent Fuel Cask; TS 3.2 Table 3-5(24), Spent Fuel Cask Loading; TS 4.3.1.3, Design Features associated with spent fuel casks; and portions of TS 3.2, Table 3-4(5), Footnote (4) on boron concentration associated with cask loading.* The deletion of the TS sections will bring the FCS TS into conformance with 10 CFR 50.68(c)

"Criticality accident requirements."

2.0 DETAILED DESCRIPTION On November 13, 2016, OPPD notified the NRC that all fuel has been permanently removed from the FCS reactor vessel and placed into the FCS spent fuel pool (Reference 6.1 ). The 10 CFR Part 50 license for FCS no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vesseL In preparation for the eventual removal of all fuel from the spent fuel pool OPPD is proposing to revise the listed sections of the FCS TS to comply with 10 CFR 50.68(c) as currently written. On April10, 2006 (Reference 6.2), the NRC issued amendment 239 to the FCS TS to bring the station into full compliance with 10 CFR 50.68 as written on that date.

10 CFR 50.68 was amended by 71 FR 66648, on Nov. 16, 2006. The NRC added section (c) which reads as follows:

(c)

While a spent fuel transportation package approved under Part 71 of this chapter or spent fuel storage cask approved under Part 72 of this chapter is in the spent fuel pool:

(1)

The requirements in§ 50.68(b} do not apply to the fuel located within that package or cask; and (2)

The requirements in Part 71 or 72 of this chapter, as applicable, and the requirements of the Certificate of Compliance for that package or cask, apply to the fuel within that package or cask.

The addition of this section to 10 CFR 50.68 eliminates the need for FCS TS 2.8.3(6), Spent Fuel Cask Loading and associated Figure 2-11, Limiting Burn up Criteria for Acceptable Storage in Spent Fuel Cask; TS 3.2 Table 3-5(24), Spent Fuel Cask Loading; TS 4.3.1.3, Design Features associated with spent fuel casks; and portions of TS 3.2, Table 3-4(5), Footnote (4) on boron concentration associated with cask loading.*

• In conjunction with FCS LAR 17-01 (Reference 6.6), TS 3.2, Table 3-4(5) and its associated Footnote (4) will be deleted in its entirety.

Ll C-17 -0015 Enclosure Page 3

3.0 TECHNICAL EVALUATION

Federal Register I Vol. 71, No. 221 I Thursday, November 16, 20061 Rules and Regulations 66648, NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150-AH95 Criticality Control of Fuel within Dry Storage Casks or Transportation Packages in a Spent Fuel Pool (Reference 6.3} in section I. Background explained:

On March 23, 2005, the NRC issued Regulatory Issue Summary (RIS) 2005-05 addressing spent fuel criticality analyses for spent fuel pools under 10 CFR 50.68 and Independent Spent Fuel Storage Installations (ISFSI) under 10 CFR Part 72. The intent of the RIS was to advise reactor licensees that they must meet both the requirements of 10 CFR 50.68 and 10 CFR Part 72 with respect to subcriticality during storage cask loading in spent fuel pools.

The rulemaking then further explained:

The regulations, as currently written, create an unnecessary burden for both industry and the NRC, of performing two different analyses with two different sets of assumptions for the purpose of preventing a criticality accident, with no associated safety benefit. This burden is considered unnecessary because the conditions which could dilute the boron concentration within a transportation package or dry storage cask (hereinafter "package or cask') in a spent fuel pool, and cause fuel damage with the release of radioactive material, are highly unlikely.

The rulemaking explained the actions needed by licensees in section XI. Backfit Analysis:

However, for those licensees who have amended their 10 CFR Part 50 license to comply with 10 CFR 50.68 and have included minimum fuel bumup limits, and choose to take advantage of this voluntary relaxation of requirements, they must request removal of the previously amended portions of the 10 CFR Part 50 technical specifications as a conforming change consistent with the amended rule.

This amendment request is written to seek the relaxation indicated in the rulemaking as indicated above.

FCS TS 2.8.3(6}, associated Figure 2-11, changes that modified TS 3.2, Tables 3-4 and 3-5, and a new subsection 4.3.1.3 in Design Features 4.3.1 were added to the FCS TS (Reference 6.2} to address concerns identified in Regulatory Issue Summary (RIS) 2005-05, "Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations" (Reference 6.4 ). The RIS expressed an expectation that a license amendment request be submitted to add a TS to the site's Part 50 license restricting the minimum burn up of fuel assemblies loaded in dry casks in accordance with the criticality analysis guidance contained in 1 0 CFR 50.68. On January 30, 2007, a new revision to 10 CFR 50.68 became effective. The revision added paragraph (c) which states that the requirements in 10 CFR 50.68(b) do not apply to the fuel located within a dry cask when the dry cask is in the spent fuel pool. Therefore, TS 2.8.3(6) associated Figure 2-11, and changes that modified TS 3.2, Tables 3-4 and 3-5, and subsection 4.3.1.3 in Design Features 4.3.1 will be deleted.

LIC-17 -0015 Enclosure Page4 Since the FCS license no longer authorizes use of the facility for power operation or emplacement or retention of fuel into the reactor vessel as provided in 10 CFR Part 50.82(a)(2)

(Reference 6.1 ), there will be no placement of new fuel within the Spent Fuel Pool. This aspect of the decommissioning plant requirements also supports the justification for the removal of the above sections.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.1.1 In accordance with 10 CFR 50.68, Criticality accident requirements, paragraph (c), while a spent fuel storage cask approved under 10 CFR Part 72 is in the spent fuel pool: 1) the requirements of 10 CFR 50.68(b) do not apply to the fuel located within the cask; and 2) the requirements of 10 CFR 72 and the requirements of the Certificate of Compliance for the cask apply to the fuel within the cask. Therefore, the deletion of the associated TS, Figure, surveillance requirement and supporting material is consistent with the allowances in 10 CFR 50.68.

4.2 Precedent 4.2.1 Reference 6.5, Arkansas Nuclear One, Unit No. 2 - Issuance of Amendment Re: Revisions to Technical Specifications to Support Partial Re-Rack and Revised Loading Patterns in the Spent Fuel Pool (TAC No.

MD4994 ), removed a similar requirement in this amendment.

4.3 No Significant Hazards Consideration The Omaha Public Power District (OPPD) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change simply bring the stations technical specifications into compliance with the current version of 10 CFR 50.68. There is no change to probability or consequences of any accident previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

LIC-17-0015 Enclosure Page 5

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not alter any, safety limits, or safety analysis assumptions associated with the operation of the plant. The proposed change does not introduce any new accident initiators, nor does the change reduce or adversely affect the capabilities of any plant structure or system in the performance of its safety function.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change does not alter the manner in which safety limits or limiting safety system settings are determined. The safety analysis acceptance criteria are not affected by the proposed change. The proposed change does not change the design function of any equipment assumed to operate in the event of an accident.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, OPPD concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

LIC-17-0015 Enclosure Page 6

5.0 ENVIRONMENTAL CONSIDERATION

A review of the proposed amendment has determined that the proposed amendment would not change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1.

Letter from OPPD (T. Burke) to USNRC (Document Control Desk) "Certification of Permanent Removal of Fuel from the Reactor Vessel," dated November 13, 2016 (LIC-16-007 4) (ML16319A254) 6.2.

Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure) "Fort Calhoun Station, Unit No. 1 - Issuance of Amendment RE: (TAC No. MC8860)," (Amendment 239) dated April10, 2006 (ML061000606) 6.3.

Federal Register I Vol. 71, No. 221 I Thursday, November 16, 2006 I Rules and Regulations 66648, NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150-AH95 Criticality Control of Fuel within Dry Storage Casks or Transportation Packages in a Spent Fuel Pool 6.4.

USNRC Regulatory Issue Summary (RIS) 2005-05, Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations, dated March 23, 2005 (ML043500532) 6.5.

U.S. NRC (A. B. Wang) to Entergy Operations, Inc. (T. G. Mitchell), "Arkansas Nuclear One, Unit No. 2-Issuance of Amendment Re: Revisions to Technical Specifications to Support Partial Re-Rack and Revised Loading Patterns in the Spent Fuel Pool (TAC No. MD4994)," dated September 28, 2007 (ML072620412) 6.6.

Letter from OPPD (M. J. Fisher) to NRC (Document Control Desk), "License Amendment Request (LAR) 17-01; Revised Fort Calhoun Station Technical Specifications to align to those requirements for decommissioning," dated March 31, 2016 (LIC-17-0001) (ML17093A309)

LIC-17-0015 Enclosure, Attachment 1 Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 Mark-up of Affected Technical Specification pages and Associated Bases

[Word-processor mark-ups using "double underline/strikeout" feature for "new text/deleted text" respectively

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueling 2.8.3 Refueling Operations-Spent Fuel Pool 2.8.3(6)

DELETED Spent Fuel Cask Loading Applicabilitv Applies to storage of spent fuel assemblies whenever any fuel assembly is located in a spent fuel cask in tho spent fuel pool. The provisions of Specification 2.0.1 for Limiting Conditions for Operation are not applicable.

Objective To minimize the possibility of an accident occurring during REFUELING OPERATIONS that could affect public health and safety.

Specification (1)

The spent fuel pool boron concentration shall be :::1£ 800 ppm, and (2)

Tho combination of initial enrichment and burnup of each spent fuel assembly located in a spent fuel storage cask in the spent fuel pool shall be within the acceptable burnup domain of Figure 2 11.

Required Actions (1)

'.AJith the spent fuel pool boron concentration <800 ppm, suspend REFUELING OPERATIONS in*lolving spent fuel cask loading immediately, aRd (2)

Restore spent fuel pool boron concentration to > 800 ppm immediately.

(3)

'A'ith the requirements of the LCO 2.8.3(6)(2) not met, initiate action to remove the noncomplying fuel assembly from the spent fuel cask immediately.

2.8-Page 14 Amendment No. 239, XXX

TECHNICAL SPECIFICATIONS Intentionally Blank

~--

.. :*=:_.,

!. 30000 +----1

_-I

-. -~-

. 0: 25000. +....;.._-~1--...;.._~+-_;,....~~---ho'~--+----____;-+------i

.:)

~ -

. i 20000 + ~

-....;......;.___;_~--~~~~~~llr-+~--:--~-,.....--+-::--:---:--!

-u:l

-*<_*~

~

~- tsooa

• +

• __,....,...;...~.__,_-....,...;...~~~~,......;.----~---~__.........,..--r---~

• - *10000:***+*. ~,;....;..;.;_~.;.;...;....---iF-~;;_,;,;;.;.;;;...;............. ~:-~

- ~- ;**.

1.5 2.0

• . 2.5

.. 3:0.... 3.5 4:0

_INITIAL ENRiCHMENT, wt% U~235 LIMITING BURNUP CRITERIA.

~OR

,A,C~E -PT:A~'-£ *sT.OAACE.IN *.

SPENT FUEL Cil\\SK 4.5 2.8-Page 15 Amendment No. ~. XXX

TECHNICAL SPECIFICATIONS LIMITING CONDITIONS FOR OPERATION 2.8 Refueling Bases (Continued) 2.8.3(6) DELETED Spent Fuel Cask Loading (1)

Soluble Boron The basis for the 800 ppm minimum boron concentration requirement during spent fuel cask loading operations is to maintain the kett in the cask system less than or equal to 0.95 in the e*1ent a mis loaded unirradiated fuel assembly is located anywhere in tho cask with up to 31 other fuel assemblies meeting the burnup and enrichment requirements of LCO 2.8.3(6)(2). This boron concentration also ensures the kon in the cask system will be less than or equal to 0.95 if an unirradiated fuel assembly is dropped in the space between the spent fuel racks and tho cask loading area during cask loading operations next to a spent fuel assembly. A mis loaded or dropped unirradiated fuel assembly at maximum enrichment condition, in the absence of soluble poison, may result in exceeding the design effective multiplication factor. Soluble boron in the spent fuel pool 'Nator, for "Nhich credit is permitted during spent fuel cask loading operations, assures that the effective multiplication factor is maintained substantially less than the design basis limit.

This LCO applies whenever a fuel assembly is located in a spent fuel cask submerged in the spent fuel pool. The boron concentration is periodically sampled in accordance with Specification 3.2. Sampling is performed prior to movement of fuel into the spent fuel cask and periodically thereafter during cask loading operations, until the cask is removed from tho spent fuel pool.

The provisions of Specification 2.0.1 for Limiting Conditions for Operations are not applicable. If moving fuel assemblies 'Nhile in MODES 4 or 5, LCO 2.0.1 would not specify any actions. If moving fuel assemblies in MODES 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

'.A/hen "immediately" is used as a completion time, the required action should be pursued without delay and in a controlled manner. Suspension of refueling operations shall not preclude completion of movement of a component to a safe, conservative position.

2.8 Page 28 Amendment No. 257 TSBC 09 006 0

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueling Bases (Continued) 2.8.3(6)

Spent Fuel Cask Loading (Continued)

(2) Burnup vs. Enrichment References The spent fuel cask is designed for subcriticality by use of neutron absorbing material. The restrictions on the placement of fuel assemblies within the spent fuel pool, according to Figure 2 11, and the accompanying LCO, ensure that tAe-keff of the spent fuel pool always remains < 0.95 assuming the pool to be flooded with borated w-ater and <1.0 assuming the pool is flooded with unboratod Yl-ater, in accordance with 10 CFR 50.68(b)(4 ).

/\\ spent fuel assembly may be transferred directly from the spent fuel racks to the spent fuel cask provided an independent verification of assembly burn ups has been completed and the assembly burnup meets the acceptance criteria identified in Figure 2 11. If any fuel assembly located in the spent fuel cask is not in accordance with Figure 2 11, immediate action must be taken to remove the non complying fuel assembly from the spent fuel cask and return it to the spent fuel rack.

The provisions of Specification 2.0.1 for Limiting Conditions for Operations are not applicable. If moving fuel assemblies *.vhile in MODES 4 or 5, LCO 2.0.1

'A'eUid not specify any actions. If moving fuel assemblies in MODES 1, 2, or 3, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown. '.'\\'hen "immediately" is used as a completion time, the required action should be pursued *.vithout delay and in a controlled manner.

(1)

USAR Section 9.5 (2)

USAR Section 9.10 (3)

USAR Section 14.18 2.8 Page 29 Amendment No. 257 TSBC 09-001-0 TSBC-09-006-0

TECHNICAL SPECIFICATIONS TABLE 3-4 (Continued)

MINIMUM FREQUENCIES FOR SAMPLING TESTS Type of Measurement Sample and Analysis and Analysis Frequency

1.

Reactor Coolant (Continued)

(c) Cold Shutdown (1) Ch~oride 1 per 3 days (Operating Mode 4)

(d) Refueling Shutdown (1) Chloride 1 per 3 days<3l (Operating Mode 5)

(2) Boron Concentration 1 per 3 days<3l (e) Refueling Operation (1) Chloride 1 per 3 days<3l (2) Boron Concentration 1 per 3 days<3l

2.

SIRWTank Boron Concentration M

3.

Concentrated Boric Boron Concentration w

Acid Tanks

4.

Sl Tanks Boron Concentration M

5.

Spent Fuel Pool Boron Concentration See Footnote 4 below

6.

Steam Generator Slowdown Isotopic Analysis for Dose W<5l (Operating Modes 1 and 2)

Equivalent 1-131 (1) Until the radioactivity of the reactor coolant is restored to ~1 ~-tCi/gm DOSE EQUIVALENT 1-131.

(2) Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since reactor was subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

(3) Boron and chloride sampling/analyses are not required when the core has been off-loaded.

Reinitiate boron and chloride sampling/analyses prior to reloading fuel into the cavity to assure adequate shutdown margin and allowable chloride levels are met.

(4) Prior to placing unirradiated fuel assemblies in the spent fuel pool or placing fuel assemblies in a spent fuel cask in the spent fuel pool, and weekly when unirradiated fuel assemblies are stored in the spent fuel pool, or every 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> when fuel assemblies are in a spent fuel storage cask in the spent fuel pool.

(5) When Steam Generator Dose Equivalent 1-131 exceeds 50 percent of the limits in Specification 2.20, the sampling and analysis frequency shall be increased to a minimum of 5 times per week. When Steam Generator Dose Equivalent 1-131 exceeds 75 percent of this limit, the sampling and analysis frequency shall be increased to a minimum of once per day.

3.2-Page 8 Amendment No. 28,67,86,124,133,152 172,188,239, 257, 289, XXX

TECHNICAL SPECIFICATIONS TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS

17.

DELETED

18.

Shutdown Cooling

19.

Refueling Water Level

20.

Spent Fuel Pool Level Test

1.

Verify required shutdown cooling loops are OPERABLE and one shutdown cooling loop is IN OPERATION.

2.

Verify correct breaker alignment and indicated power is available to the required shutdown cooling pump that is not IN OPERATION.

Verify refueling water level is ~ 23 ft. above the top of the reactor vessel flange.

Verify spent fuel pool water level is ~ 23 ft.

above the top of irradiated fuel assemblies seated in the storage racks.

21.

Containment Penetrations Verify each required containment penetration is

22.

Spent Fuel Assembly Storage

23.

P-T Limit Curve in the required status.

Verify by administrative means that initial enrichment and burn up of the fuel assembly is in accordance with Figure 2-10.

Verify RCS Pressure, RCS temperature, and RCS heatup and cooldown rates are within the limits specified by the P-T limit Figure(s) shown in the PTLR.

Frequency S (when shutdown cooling is required by TS 2.8).

W (when shutdown cooling is required by TS 2.8).

USAR Section Reference Prior to commencing, and daily during CORE ALTERATIONS and/or REFUELING OPERATIONS inside containment.

Prior to commencing, and weekly during REFUELING OPERATIONS in the the spent fuel pool.

Prior to commencing, and weekly during CORE ALTERATIONS and/or REFUELING OPERATIONS in containment.

Prior to storing the fuel assembly in Region 2 (including peripheral cells).

This test is only required during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing. While these operations are occurring, this test shall be performed every 30 minutes.

24.

DELETEQ Spent Fuel Cask Loading Verify by administrative means that initial Prior to placing the fuel assembly in a spent enrichment and burn up of the fuel assemblv fuel cask in the scent fuel oeoh i~ in '=11""-f"'n.rrl'.:llnl"".o.. \\_uitb_____l;i.n.ttr.a ') _ _1 i

• ~~"l,.--rlt8 3.2-Page 14 Amendment No. 138, 169, 188, 246, 250.257, 274,

2-7-7,-289, XXX

TECHNICAL SPECIFICATIONS 4.0 DESIGN FEATURES (Continued)

c. A nominal 8.6 inch center to center distance between fuel assemblies placed in Region 2, the high density fuel storage racks,

d. A nominal 9.8 inches (East-West) by 10.3 inches (North South) center to center distances between fuel assemblies placed in Region 1, the low density fuel storage racks,

e. New or partially spent fuel assemblies with a discharge burn up in the "acceptable domain" of Figure 2-10 for "Region 2 Unrestricted" may be allowed unrestricted storage in any of the Region 2 fuel storage racks in compliance with Reference (1 ).

f.

Partially spent fuel assemblies with a discharge burn up between the "acceptable domain" and "Peripheral Cells" of Figure 2-10 may be allowed unrestricted storage in the peripheral cells of the Region 2 fuel storage racks

'n compliance with Reference (1 ).

g. New or partially spent fuel assemblies with a discharge burn up in the "unacceptable domain" of Figure 2-10 will be stored in Region 1 in compliance with Reference (1 ).

4.3.1.2 The new fuel storage rack is designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent,

b. A nominal 16 inch center to center distance between fuel assemblies placed in the storage rack.

4.3.1.3 DELETED The spent fuel casks aro designed and shall be maintained with:

a. Fuel assemblies having a maximum U 235 enrichment of 4.5 weight percent,

~

0 if fully flooded with unboratod water, *.vhich includes an aiiO\\vance for <

1. uncertainties as described in Section 9.5 of the USAR,

&.---kett < 0.95 if fully flooded with borated water> BOO ppm, whish includes an allowance for uncertainties as described in Section 9.5 of the USAR,

d. A nominal 9.075 inoh center to center distance between fuel assemblies plaoed in the spent fuel oask,

e. Spent fuel assemblies with a combination of discharge burn up and initial average assembly enrichment in the "acceptable" range of Figure 2 11.

4.0 - Page 2 Amendment No. 236,239, 24 0, XXX

L1 C-16-0099 Enclosure, Attachment 2 Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 Clean Affected Technical Specification pages and Associated Bases

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueling 2.8.3 Refueling Operations - Spent Fuel Pool 2.8.3(6)

DELETED 2.8-Page 14 Amendment No. ~. XXX

TECHNICAL SPECIFICATIONS Intentionally Blank 2.8-Page 15 Amendment No. 239, XXX

TECHNICAL SPECIFICATIONS LIMITING CONDITIONS FOR OPERATION 2.8 Refueling Bases (Continued) 2.8.3(6)

DELETED References (1)

USAR Section 9.5 (2)

USAR Section 9.10 (3)

USAR Section 14.18 2.8-Page 28 Amendment No.~. XXX TSBC 09-001-0 TSBC-09-006-0

TECHNICAL SPECIFICATIONS TABLE 3-4 (Continued)

MINIMUM FREQUENCIES FOR SAMPLING TESTS Type of Measurement Sample and Analysis and Analysis Frequency

1.

Reactor Coolant (Continued)

(c) Cold Shutdown (1) Chloride 1 per 3 days (Operating Mode 4)

(d) Refueling Shutdown (1) Chloride 1 per 3 days(3l (Operating Mode 5)

(2) Boron Concentration 1 per 3 days(3l (e) Refueling Operation (1) Chloride 1 per 3 days(3l (2) Boron Concentration 1 per 3 days(3l

2.

SIRW Tank Boron Concentration M

3.

Concentrated Boric Boron Concentration w

Acid Tanks

4.

Sl Tanks Boron Concentration M

5.

Spent Fuel Pool Boron Concentration See Footnote 4 below

6.

Steam Generator Slowdown Isotopic Analysis for Dose W(S)

(Operating Modes 1 and 2)

Equivalent 1-131 (1) Until the radioactivity of the reactor coolant is restored to ~1 1-LCi/gm DOSE EQUIVALENT 1-131.

(2) Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since reactor was subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

(3) Boron and chloride sampling/analyses are not required when the core has been off-loaded. Reinitiate boron and chloride sampling/analyses prior to reloading fuel into the cavity to assure adequate shutdown margin and allowable chloride levels are met.

(4) Prior to placing unirradiated fuel assemblies in the spent fuel pool and weekly when unirradiated fuel assemblies are stored in the spent fuel pool.

(5)When Steam Generator Dose Equivalent 1-131 exceeds 50 percent of the limits in Specification 2.20, the sampling and analysis frequency shall be increased to a minimum of 5 times per week. When Steam Generator Dose Equivalent 1-131 exceeds 75 percent of this limit, the sampling and analysis frequency shall be increased to a minimum of once per day.

3.2 - Page 8 Amendment No. 28,67,86,124,133,152 172,188,239, 257, 289, XXX

TECHNICAL SPECIFICATIONS TABLE 3-5

17.

DELETED

18.

Shutdown Cooling

19.

Refueling Water Level

20.

Spent Fuel Pool Level MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Test

1.

Verify required shutdown cooling loops are OPERABLE and one shutdown cooling loop is IN OPERATION.

2.

Verify correct breaker alignment and indicated power is available to the required shutdown cooling pump that is not IN OPERATION.

Verify refueling water level is ~ 23 ft. above the top of the reactor vessel flange.

Verify spent fuel pool water level is ~ 23 ft.

above the top of irradiated fuel assemblies seated in the storage racks.

Frequency S (when shutdown cooling is required by TS 2.8).

W (when shutdown cooling is required by TS 2.8).

USAR Section Reference Prior to commencing, and daily during CORE ALTERATIONS and/or REFUELING OPERATIONS inside containment.

Prior to commencing, and weekly during REFUELING OPERATIONS in the the spent fuel pool.

21.

Containment Penetrations Verify each required containment penetration is Prior to commencing, and weekly during CORE ALTERATIONS and/or REFUELING OPERATIONS in containment.

22.

Spent Fuel Assembly Storage

23.

P-T Limit Curve

24.

DELETED in the required status.

Verify by administrative means that initial enrichment and burn up of the fuel assembly is in Prior to storing the fuel assembly in Region 2 (including peripheral cells).

accordance with Figure 2-10.

Verify RCS Pressure, RCS temperature, and RCS heatup and cooldown rates are within the limits specified by the P-T limit Figure(s) shown in the PTLR.

3.2-Page 14 This test is only required during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing. While these operations are occurring, this test shall be performed every 30 minutes.

69 188. 246, 250,257, 274, Amendment No. 138* 1 277. 289, XXX

TECHNICAL SPECIFICATIONS 4.0 DESIGN FEATURES (Continued)

c. A nominal 8.6 inch center to center distance between fuel assemblies placed in Region 2, the high density fuel storage racks,

d. A nominal 9.8 inches (East-West) by 10.3 inches (North South) center to center distances between fuel assemblies placed in Region 1, the low density fuel storage racks,

e. New or partially spent fuel assemblies with a discharge burn up in the "acceptable domain" of Figure 2-10 for "Region 2 Unrestricted" may be allowed unrestricted storage in any of the Region 2 fuel storage racks in compliance with Reference (1 ).

f.

Partially spent fuel assemblies with a discharge burn up between the "acceptable domain" and "Peripheral Cells" of Figure 2-1 0 may be allowed unrestricted storage in the peripheral cells of the Region 2 fuel storage racks in compliance with Reference (1 ).

g. New or partially spent fuel assemblies with a discharge burn up in the "unacceptable domain" of Figure 2-10 will be stored in Region 1 in compliance with Reference (1 ).

4.3.1.2 The new fuel storage rack is designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent,

b. A nominal 16 inch center to center distance between fuel assemblies placed in the storage rack.

4.3.1.3 DELETED 4.0- Page 2 Amendment No. 236,239, 240, XXX