ML063120248

Issuance of Amendment 247 Change of Containment Building Sump Buffering Agent from Trisodium Phosphate to Sodium Tetraborate

ML063120248
Person / Time
Site:	Fort Calhoun
Issue date:	11/13/2006
From:	Wang A NRC/NRR/ADRO/DORL/LPLIV
To:	Ridenoure R Omaha Public Power District
Wang A, NRR/DORL/LP4, 415-1445
Shared Package
ML063120239	List: ML063120248 ML063180546
References
TAC MD2864
Download: ML063120248 (13)

Text

November 13, 2006 Mr. R. T. Ridenoure Vice President - Chief Nuclear Officer Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

Post Office Box 550 Fort Calhoun, NE 68023-0550

SUBJECT:

FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE:

CHANGE OF CONTAINMENT BUILDING SUMP BUFFERING AGENT FROM TRISODIUM PHOSPHATE TO SODIUM TETRABORATE (TAC NO. MD2864)

Dear Mr. Ridenoure:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 247 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated August 21, 2006, as supplemented on September 6 and October 10, 2006.

The amendment changed the TSs to: (1) revise TS Section 2.3(4) to change the reactor containment building sump buffering agent from trisodium phosphate to sodium tetraborate and change the TS section title to Containment Sump Buffering Agent Specification and Volume Requirement, (2) revise TS 3.6(2)d to require a volume of sodium tetraborate that is within the area of acceptable operation, as shown in TS Figure 2-3, and (3) make an administrative correction to TS 3.6(2)d(i). The amendment allows Omaha Public Power District to replace the trisodium phosphate in the containment with sodium tetraborate. Changes were also made to the corresponding TS Bases. The TS changes are approved for Cycle 24 only, ending in the spring 2008 refueling outage.

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

Sincerely,

/RA/

Alan B. Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosures:

1. Amendment No. 247 to DPR-40

2. Safety Evaluation cc w/encls: See next page

PKG ML063120239 (Amendment: ML063120248, TS Pages: ML063180546 )

OFFIC NRR/LPL4/PM NRR/LPL4/LA ADES/CSGB/B ADES/SSIB OGC-NLO NRR/LPL4/BC E C NAME AWang LFeizollahi AHiser MScott BPoole DTerao DATE 11/8/06 11/13/06 11/6/06 11/8/06 11/13/06 11/13/06 OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 247 License No. DPR-40

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

A. The application for amendment by the Omaha Public Power District (the licensee), dated August 21, 2006, as supplemented on September 6 and October 10, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. 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 license 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, Renewed Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-40.

3. The license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-40 and Technical Specifications Date of Issuance: November 13, 2006

ATTACHMENT TO LICENSE AMENDMENT NO. 247 RENEWED FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace page 3 of the Renewed Facility Operating License No. DPR-40 with the attached page 3.

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 vertical lines indicating the areas of change.

REMOVE INSERT TOC - Page 8 TOC - Page 8 2.3 - Page 4 2.3 - Page 4 2.3 - Page 6 2.3 - Page 6 2.3 - Page 8 2.3 - Page 8 3.6 - Page 1 3.6 - Page 1 3.6 - Page 2 3.6 - Page 2 3.6 - Page 6 3.6 - Page 6 3.6 - Page 7 3.6 - Page 7

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 247 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285

1.0 INTRODUCTION

By application dated August 21, 2006, as supplemented on September 6 and October 10, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML062340039, ML062570173, and ML063070017, respectively), Omaha Public Power District (OPPD) requested changes to the Technical Specifications (Appendix A to Renewed Facility Operating License No. DPR-40) for the Fort Calhoun Station, Unit No. 1 (FCS).

The amendment proposed to revise the Technical Specifications (TSs) to replace references to trisodium phosphate (TSP) with sodium tetraborate (STB). Specifically, the proposed changes would: (1) revise TS Section 2.3.(4) to change the reactor containment building sump buffering agent from TSP to STB and change the TS section title to Containment Sump Buffering Agent Specification and Volume Requirement; (2) revise TS 3.6(2)d to require a volume of STB that is within an area of acceptable operation as shown in TS Figure 2-3; and (3) make an administrative correction to TS 3.6(2)d(i). Changes were also proposed to the corresponding TS Bases. The proposed TS change request was limited to one operating cycle, Cycle 24, ending in the spring 2008 refueling cycle. OPPD has informed the Nuclear Regulatory Commission (NRC) staff that it plans to submit another license amendment request supporting this change for operation during subsequent cycles.

The two supplements dated September 6 and October 10, 2006, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 30, 2006 (71 FR 51646).

2.0 REGULATORY EVALUATION

The NRC staff review addresses the impact of the proposed change from TSP to STB on containment sump performance, especially potential chemical effects impact on sump screen blockage and head loss.

The containment sump (also known as the emergency recirculation sump) is part of the emergency core cooling system (ECCS). Every nuclear power plant is required by Section 50.46 of Title 10 of the Code of Federal Regulations (10 CFR) to have an ECCS to mitigate a design-basis accident. Paragraph 50.46(a) of 10 CFR states in part, that each pressurized light-water nuclear power reactor ... must be provided with an [ECCS] that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section. Paragraph 50.46(b)(5) of 10 CFR, Long-term cooling, states After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

In addition, the NRC staff utilized the following regulatory guidance in performing this review:

• NUREG-0800, Section 6.5.2, Containment Spray as a Fission Product Cleanup System, which states, in part, that long-term iodine retention may be assumed only when the equilibrium sump solution pH, after mixing and dilution with the primary coolant and ECCS injection, is above 7,

• Regulatory Guide 1.82, Revision 3, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, Section 1.1.2, which states, in part, that debris that could accumulate on the sump screen should be minimized, and

3.0 TECHNICAL EVALUATION

Under loss-of-coolant-accident (LOCA) conditions, buffering agents must be added to the pool of water in the containment building to increase the pH to a value of greater than 7. Buffering is primarily required to reduce the release of iodine fission products from the containment pool to the containment atmosphere as iodine gas, in order to control the radiological consequences of the accident. Maintaining a pH above 7 prevents significant amounts of iodine, released from fuel failures and dissolved in the recirculation water, from converting to a volatile form and evolving into the containment atmosphere. In addition to dose considerations, raising the pH in the post-LOCA containment pool to a value greater than 7 reduces the general corrosion rate of some structural materials (e.g., steel) and inhibits stress corrosion cracking in austenitic stainless steel. The current FCS TS 2.3(4) requires a TSP volume that is a function of the reactor coolant system critical boron concentration over an operating cycle as shown in TS Figure 2-3.

NRC-sponsored testing at the Argonne National Laboratory (ANL) indicated that substantial head loss can occur if sufficient calcium phosphate is produced in a sump pool and transported to a pre-existing fiber bed on the containment sump screen. Consequently, the ECCS flow and containment spray system flow could be reduced by the increased head loss across the sump screen while in the post-LOCA recirculation phase. The NRC staff communicated these results to the industry in Information Notice (IN) 2005-26, Results of Chemical Effects Head Loss Tests in a Simulated PWR [Pressurized-Water Reactor] Sump Pool Environment, dated September 16, 2005, and its associated Supplement 1 to IN 2005-26, Additional Results of Chemical Effects Tests in a Simulated PWR Sump Pool Environment, dated January 20, 2006.

Currently, FCS has both a potential source of calcium (calcium silicate insulation) and a source of phosphate (TSP buffer) that could produce a calcium phosphate precipitate in a postulated LOCA. OPPD requested this amendment to minimize the potential for exacerbating sump screen blockage due to a potential chemical interaction between TSP and certain calcium-containing insulation material used in containment to form calcium phosphate. Therefore, this OPPD amendment switches the buffering chemical from TSP to STB to remove the phosphate source from containment, thereby reducing the amount of precipitate that may be formed in a postulated post-LOCA containment pool.

3.1 Replacement of TSP with STB The NRC staff reviewed the licensees regulatory and technical analyses related to the impact of the proposed change from TSP to STB on containment sump performance, particularly the potential impact from chemical effects on sump screen blockage and head-loss aspects of design-basis accidents. For the purpose of this safety evaluation, the issue of chemical effects involves interactions between the post-LOCA PWR containment environment and containment materials that may produce corrosion products, gelatinous material, or other chemical reaction products capable of affecting sump screen head loss. Information regarding these analyses was provided by OPPD in Section 5.0 of its August 21, 2006, submittal.

The OPPD evaluation determined that STB is an acceptable alternative to TSP based on industry testing of buffers outlined in WCAP-16596-NP, Evaluation of Alternative Emergency Core Cooling System Buffering Agents and through plant-specific application of the chemical model developed in WCAP-16530-NP, Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191. The OPPD analysis, based on the aforementioned industry buffer tests and chemical model, determined that switching from TSP

to STB would maintain the appropriate post-accident pH control while reducing the amount of precipitate by approximately 70 percent. The NRC staff has not completed its review of WCAP-16530-NP and has issued a request for additional information (ADAMS Accession No. ML062440433). The NRC staff did not perform a detailed review of WCAP-16596-NP because the NRC staff found that the licensee provided adequate information to support this amendment request without reference to this WCAP. Although the NRC staff review of WCAP-16530-NP model is not complete, for the purpose of this amendment, the NRC staff considers the predicted aluminum concentration to be consistent with aluminum corrosion test data. The model predicted the dissolved aluminum concentration in the FCS postulated post-LOCA containment pool would be 22 parts per million (ppm). As discussed later in this section, even if the predicted FCS aluminum concentration amount is doubled, it would not be expected that this would affect head loss in the STB buffered environment.

In order to address ongoing technical issues related to the potential for chemical products to affect the sump screen or downstream components, the NRC staff has sponsored several test programs. The initial investigation of potential chemical effects was accomplished in a NRC-nuclear industry jointly sponsored Integrated Chemical Effects Test (ICET) program conducted by the Los Alamos National Laboratory at the University of New Mexico. A total of 5 tests were conducted in the ICET program, with each test representing a different set of plant materials and post-LOCA containment pool environments. Potential chemical effects from the combination of TSP and calcium silicate, which is relevant to the FCS plant-specific conditions, were evaluated in ICET #3. During this test, a calcium phosphate precipitate was observed in the test tank fluid within 30 minutes of the initiation of the spray solution containing TSP.

Since chemical products had formed in some of the environments during the ICET program, the NRC subsequently sponsored testing at ANL to evaluate the impact of these chemical products on head loss. The ANL tests were performed in a vertical head-loss loop using a flat test screen covered with simulated debris from containment materials. For the ICET #3 type environment (borated water, TSP buffer) relevant to FCS, the tests included measurement of head loss across the test screen that had been covered with calcium silicate and fiberglass insulation. Test results are provided in the attachment to IN 2005-26 and Supplement 1 to IN 2005-26. These results demonstrated that substantial head loss can occur if sufficient calcium phosphate is produced in a containment pool and is transported to a sump screen along with fibrous insulation debris.

Along with the calcium silicate insulation-TSP environment (i.e., ICET #3), both the ICET program and the ANL Head Loss Test program evaluated a STB buffered environment (i.e.,

ICET #5). In contrast to the ICET #3 environment with calcium silicate and TSP, the ICET #5 STB environment did not produce a calcium phosphate precipitate since there is no phosphate source. In the STB environment, however, corrosion of aluminum is important since aluminum-containing precipitate formed in the ICET #5 test solution after it had cooled from the 140 " F test temperature to ambient temperature. The peak dissolved aluminum concentration measured in the ICET #5 test solution was approximately 55 ppm.

In addition to differences in the chemical products formed in TSP and STB environments and the timing of their formation, there were distinct differences in how these precipitates affected head loss during testing at ANL. Although chemical species are discussed in terms of fluid volume (i.e., ppm) within the ANL test loop, it is important to scale chemical effects test results based on a product mass per unit screen area basis. In the TSP-buffered environment with an approximate 12 millimeter Nukon fiberglass bed, significant head loss was consistently measured on a horizontal flat screen with a calcium silicate loading of 0.7 kilograms per meter2.

For calcium silicate insulation additions as low as 0.5 grams per liter, the equivalent dissolved calcium concentration was shown to reach a level that produced significant pressure drop within a few hours. In contrast to the TSP environment head-loss behavior, the STB environment with a 50 ppm dissolved aluminum concentration (approximately twice the dissolved aluminum concentration predicted by the WCAP-16530-NP model for FCS) produced little or no increase in head loss above the baseline head loss in tests of 11 and 19 days duration. Since 100 ppm dissolved aluminum produced significant head loss in the STB environment, long-term bench-top tests were performed to determine a threshold concentration for precipitate formation. In

these bench-top tests, small amounts of precipitate were initially observed at a 60 ppm dissolved aluminum concentration.

Table 1 provides a comparison of results from NRC-sponsored tests with the ICET #3 type environment (i.e., TSP with dissolved calcium) and ICET #5 type environment (i.e., containing a STB buffer).

Table 1. Comparison of ICET and Head-loss Test Results with TSP and STB Buffers.

Test Buffer/Materials Result ICET #3 TSP-calcium silicate Calcium phosphate precipitate observed within and fiberglass 30 minutes of TSP addition. Significant amounts of precipitate deposited on stainless steel mesh insulation bags within the test tank.

ICET #5 STB-fiberglass No visible precipitate at the test temperature.

Precipitate observed in the test solution after cooling to ambient temperature.

ANL Head TSP-dissolved Significant head loss across a fiber-covered screen Loss, ICET #3 calcium from calcium section consistently occurred when the dissolved environment silicate insulation or calcium concentration was 25 ppm or greater.

by CaCl2 solution ANL Head STB with dissolved Long-term head loss test with fiber bed (or fiber bed Loss, ICET #5 aluminum from with some cal-sil insulation) showed no significant environment, aluminum nitrate head-loss increase with a 50 ppm dissolved and bench-top solution aluminum concentration. Visual indications of small tests amounts of precipitate in bench-top testing at 60 ppm. An increase in head loss was measured at a 70 ppm concentration and a substantial head loss was observed at a 100 ppm dissolved aluminum concentration.

In addition to aluminum-based precipitates, calcium-based precipitates can also form in the STB environment. Two potential sources of calcium in the FCS containment are calcium silicate insulation and concrete. The OPPD analysis using the WCAP-16530-NP chemical model did not predict any calcium-based precipitates. Given the limited duration of this amendment, the NRC staff determined that it did not need to perform a detailed review of the OPPD reference material that developed the threshold amount of dissolved calcium needed to form a precipitate in STB. A more detailed review would be performed for an amendment seeking a permanent change to STB. NRC-sponsored testing at ANL included a single test with calcium silicate insulation in the STB environment and much lower head losses were observed compared to corresponding tests with TSP. The ANL tests, however, were not performed over the full range of calcium-silicate loadings that might be of interest.

Along with the above evaluations based on chemical effects testing, the NRC staff performed an independent calculation to verify that the amount of STB required by the FCS TS would be sufficient to raise the equilibrium pH in the post-LOCA containment pool to a value greater than 7.

The NRC staff confirmed that the post-LOCA containment pool pH will be maintained above pH 7 with STB.

The NRC staff has reviewed the licensees safety analyses of the impact of this amendment to replace the TSP buffer with STB, for a period of one cycle, and has determined that this replacement would not result in an accident initiator nor create a new or different kind of accident than those previously analyzed. As currently required, the post-LOCA containment pool pH will be maintained above a pH of 7 with STB. The NRC staff has concluded that the proposed change decreases the risk of sump screen blockage following a design-basis accident, since the

amount of chemical precipitate produced in the post-LOCA containment pool will be substantially reduced and the timing of any precipitate that may form should be delayed compared to the precipitate that would form with the TSP buffer. Therefore, the NRC staff finds the amendment requesting replacement of TSP with STB to be acceptable.

3.2 TS Changes 3.2.1 TS 2.3(4)

As discussed above, the NRC staff has determined that the reactor containment building sump buffering agent can be changed for one cycle from TSP to STB and, therefore, the changes proposed for TS Section 2.3(4) are acceptable. In addition, the change of TS Section 2.3(4) Title to Containment Sump Buffering Agent Specification and Volume Requirement, is editorial in nature and, therefore, is acceptable.

3.2.2 TS 3.6(2)d As discussed above, the NRC staff has determined that the reactor containment building sump buffering agent can be changed for one cycle from TSP to STB and, therefore, the changes proposed for TS Section 3.6(2)d are acceptable. The surveillance requirement of TS 3.6(2)d is revised to require a volume of STB that is within the area of acceptable operation of Figure 2-3.

The NRC staff has determined that Figure 2-3 provides the pH control limits necessary to maintain the post-LOCA containment pool to a pH value greater than 7 and, therefore, this TS change is acceptable.

In addition, the licensee identified a typographical error in TS 3.6(2)d. An extraneous i was removed from the first section. The NRC staff agrees that this change is editorial in nature and, therefore, is acceptable.

3.2.3 Bases Changes The licensee included in its application the revised TS Bases to be implemented with the TS change. The NRC staff finds that the TS Bases Control Program is the appropriate process for updating the affected TS Bases pages and, therefore, has no comment on the proposed Bases changes.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration and there has been no public comment on such finding (71 FR 51646; published on August 30, 2006). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

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) 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.

Principal Contributors: P. Klein K. Parczewski Date: November 13, 2006

Ft. Calhoun Station, Unit 1 cc:

Winston & Strawn Mr. David J. Bannister, Manager ATTN: James R. Curtiss, Esq. Fort Calhoun Station 1700 K Street, N.W. Omaha Public Power District Washington, DC 20006-3817 Fort Calhoun Station FC-1-1 Plant P.O. Box 550 Chairman Fort Calhoun, NE 68023-0550 Washington County Board of Supervisors P.O. Box 466 Mr. Joe L. McManis Blair, NE 68008 Manager - Nuclear Licensing Omaha Public Power District Mr. John Hanna, Resident Inspector Fort Calhoun Station FC-2-4 Adm.

U.S. Nuclear Regulatory Commission P.O. Box 550 P.O. Box 310 Fort Calhoun, NE 68023-0550 Fort Calhoun, NE 68023 Mr. Daniel K. McGhee Regional Administrator, Region IV Bureau of Radiological Health U.S. Nuclear Regulatory Commission Iowa Department of Public Health 611 Ryan Plaza Drive, Suite 400 Lucas State Office Building, 5th Floor Arlington, TX 76011-4005 321 East 12th Street Des Moines, IA 50319 Ms. Julia Schmitt, Manager Radiation Control Program Nebraska Health & Human Services R & L Public Health Assurance 301 Centennial Mall, South P.O. Box 95007 Lincoln, NE 68509-5007 April 2006