ML103200161

Lacrosse Amendment No. 71 Safety Evaluation

ML103200161
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	01/25/2011
From:	NRC/FSME/DWMEP/DURLD
To:
Hickman J, FSME/DWMEP,301-415-3017
Shared Package
ML103200116	List: ML103200120 ML103200161 ML103200166 ML103200175 ML103210493
References
TAC J00359
Download: ML103200161 (16)

Text

SAFETY EVALUATION RELATED TO AMENDMENT NO. 71 TO POSSESSION ONLY LICENSE NO. DPR-45 DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR DOCKET NO. 50-409

1.0 INTRODUCTION

By letter dated July 28, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092470459), as supplemented by letters dated August 7, 2009 (ADAMS Accession Nos. ML092310242 and ML092260299), May 19, 2010 (ADAMS Accession No. ML101550274), and email dated August 12, 2010 (ADAMS Accession No. ML103010320),

Dairyland Power Cooperative (DPC, or the licensee), requested amendment to the Technical Specifications (TS) for La Crosse Boiling Water Reactor (LACBWR), in support of the dry cask storage project at LACBWR. The supplements 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) staffs proposed no significant hazards consideration determination, which was noticed in the Federal Register on October 6, 2009 (74 FR 51326). Therefore, the staffs conclusions presented in the October 6, 2009, no significant hazards consideration determination are unchanged, and a revised notice is not required.

The licensee requested the changes to accommodate efficient dry cask storage system loading operations and reduce overall occupational dose to personnel during these operations.

Specifically, the licensee requested changes to the definition of Fuel Handling, TS 2.2 (fuel storage), TS 2.2.3 (drainage of Fuel Element Storage Well (FESW)), TS 4.1.1 (general fuel storage and handling requirements), TS 4.1.2 (limiting condition for operation of FESW), and TS 5.1.2.1 - TS 5.1.2.2 (surveillance requirements for FESW). The proposed changes would reduce the water coverage over spent fuel in the FESW, reflect inclusion of the cask pool in the scope of TS as part of an extended FESW, and provide editorial clarifications to clarify heavy load controls and the applicability of certain TS when spent fuel assemblies are not in the FESW or cask pool.

2.0 BACKGROUND

AND REGULATORY EVALUATION LACBWR was permanently shutdown on April 30, 1987, and reactor defueling was completed on June 11, 1987. The Decommissioning Plan (DP) was approved August 7, 1991, and it describes the current status of plant systems. The DP is considered the post-shutdown decommissioning activities report (PSDAR). The PSDAR public meeting was held on May 13, 1998. Since shutdown, the licensee has been conducting limited and gradual dismantlement activities. The licensee is currently planning and conducting a dry cask storage project, to move spent fuel from wet storage in the FESW to dry cask storage in an onsite independent spent fuel storage installation (ISFSI), under the general license provisions of 10 CFR Part 72, Subpart K. This Enclosure 2

current amendment request proposes changes to the 10 CFR Part 50 TS, in support of the LACBWR dry cask storage project.

The LACBWR FESW contains 333 spent fuel assemblies, stored in a two-tiered storage rack configuration. Because the LACBWR FESW is too small to fit a dry storage canister and transfer cask, the licensee plans to construct a cask pool in the area that previously housed the reactor pressure vessel (which was removed in June 2007), and the existing fuel transfer canal will be used to move the spent fuel from the FESW to the cask pool where it will be placed in a spent fuel storage canister. During fuel loading operations and when the fuel transfer canal gate is removed, the fuel transfer canal and cask pool will be hydraulically contiguous with the FESW and can be considered an extension of the FESW. The amendment request includes changes to the current 10 CFR Part 50 TS to reflect inclusion of the cask pool in the scope of the TS.

After canisters are loaded with spent fuel and the canister lid is placed on the canister to provide shielding, the water level in the cask pool will need to be lowered to allow removal of the canister and transfer cask and preparation of the canister for dry storage. The licensee plans to reinstall the fuel transfer canal gate between the loadings of the first 2 canisters, to maintain the FESW water level at the current TS level (16 feet of water above fuel) while fuel is located in the upper tier storage racks. However, to reduce occupational doses (obtained by personnel while uninstalling and reinstalling the fuel transfer canal gate) and for operational efficiency, the licensee does not plan to reinstall the fuel transfer canal gate after the fuel in the top tier storage racks in the FESW is moved to canisters. As the cask pool and FESW are hydraulically contiguous (with the fuel transfer canal gate removed), when the licensee lowers the water level in the cask pool to allow removal and preparation of the filled canister, the FESW water level will also be lowered. The licensee proposes to lower the water level in the FESW to just below the bottom of the fuel transfer canal gate, at a level of 11 feet, 6.5 inches of water above the spent fuel in the lower tier storage racks of the FESW. Therefore, the licensee is proposing a change in TS to lower the water level coverage above the fuel from 16 feet to 11 feet, 6.5 inches, to support its planned dry cask storage system (DCSS) loading operations. The amendment request describes how this lower water level impacts the thermal-hydraulic condition of the FESW, reactivity of the FESW, shielding and occupational dose, and accident analyses associated with the spent fuel and the FESW.

The LACBWR 10 CFR Part 50 possession only license (POL) DPR-45, Section 2.B.(2), allows storage of spent fuel pursuant to the requirements of 10 CFR Part 70 and the limitations described in TS Section 2.2, Fuel Storage. The POL also requires that the facility be maintained in accordance with all TS, which includes Section 4/5.1, Fuel Storage and Handling. These TS are all relevant to spent fuel storage and handling at LACBWR, and their requirements must be met at all times in the FESW and cask loading area. Once a spent fuel storage canister has been loaded and closed, it ceases to be subject to the Part 50 TS requirements for spent fuel criticality and cooling.

The proposed changes to the TS are evaluated below.

3.0 TECHNICAL EVALUATION

The licensee proposes the following specific changes to the TS, and each change is evaluated below.

3.1 TS Section 1, DEFINITIONS The current definition of FUEL HANDLING states:

FUEL HANDLING shall be the movement of any irradiated fuel within the Containment Building. Suspension of FUEL HANDLING shall not preclude completion of movement of the fuel to a safe, conservative position.

The licensee proposes to modify the definition to read as follows:

FUEL HANDLING shall be the movement of individual spent fuel assemblies within the Reactor Building. Suspension of FUEL HANDLING shall not preclude completion of movement of a spent fuel assembly to a safe, conservative position. FUEL HANDLING, for the purposes of these Technical Specifications, does not include the movement of an NRC-certified spent fuel storage canister, transfer cask, or storage cask containing spent fuel in accordance with the dry cask storage systems 10 CFR 72 Certificate of Compliance.

The licensee proposes the changes in the first sentence of the definition to clarify that fuel handling applies to the movement of individual fuel assemblies and to correct the title of the location where the FESW is located and where DCSS loading operations will occur. The current term, Reactor Building, is used to clarify that containment integrity requirements are no longer applicable to the LACBWR facility, and this is the terminology used in the LACBWR DP. The licensee also proposes an editorial change of all references of fuel or irradiated fuel to be replaced with spent fuel to maintain consistency throughout the TS. These above editorial changes clarify the definition of fuel handling, and the changes are acceptable.

The licensees proposed addition of the third sentence is meant to clarify that the definition of fuel handling is not applicable to the movement of fuel assemblies in an NRC-certified spent fuel storage canister, transfer cask, or storage cask containing spent fuel in accordance with the DCSSs 10 CFR Part 72 Certificate of Compliance (CoC) and associated TS and Final Safety Analysis Report (FSAR). The licensee stated that the requirements applicable to FUEL HANDLING in the LACBWR TS pertain to the shielding provided for handling of individual fuel assemblies and the potential for dropping an assembly and releasing radioactive material from the dropped assembly into the water. When a spent fuel transfer cask and canister containing fuel assemblies are moved, the canister lid will be in place to provide the necessary shielding, as required by the 10 CFR Part 72 FSAR for the DCSS. The licensee stated that the transfer cask containing the canister will be lifted and moved with a single-failure-proof lifting system, making the probability of a cask drop a non-credible event. For these reasons, the LACBWR TS requirements for FUEL HANDLING do not apply to the movement of fuel assemblies within a spent fuel canister.

The staff reviewed the basis for this change provided by the licensee, as well as LACBWRs current license, TS, and regulations to determine its acceptability. In order to assure that the change in definition did not affect the conditions of the existing license and TS, the staff reviewed these documents for all references to FUEL HANDLING and evaluated the impact of the change.

The staff found no cases where the revised definition resulted in a relaxation of the requirements for protection of spent fuel in the current licensing basis. Individual fuel assemblies stored in the FESW, or placed in an open fuel storage canister or transfer cask are still subject to the

requirements of the POL, TS, and 10 CFR Part 50. Movement of a loaded spent fuel storage canister and transfer cask must still meet the requirements of TS 4.1.1.3 as modified by this amendment and evaluated below.

Based on the above considerations, the proposed changes to the definition of FUEL HANDLING are acceptable.

3.2 TS Section 2.2, FUEL STORAGE The licensee proposes to revise the title of TS Section 2.2 from FUEL STORAGE, to FUEL STORAGE WHILE IN THE FUEL ELEMENT STORAGE WELL.

The licensee proposes this change to clarify that the fuel storage requirements in this TS section apply only to the spent fuel assemblies while in the FESW storage racks, and that fuel located inside an spent fuel storage canister (approved under 10 CFR Part 72) in the cask pool awaiting transfer to the ISFSI is not considered in storage in the context of this TS section. This section of the TS addresses criticality control for the fuel in storage racks in the FESW; fuel restrictions (e.g., type of fuel that can be stored in FESW); requirements that the FESW is designed and maintained to prevent inadvertent draining of the FESW below an elevation of 679 feet mean sea level (MSL) and capacity of the FESW and maximum number of fuel assemblies that can be stored in the FESW. Since the requirements in this TS section are specific to the FESW and not applicable to the cask pool area, the staff agrees with the licensees interpretation of this section of the TS, and thus, the proposed change to the title of this TS section is acceptable.

3.2.1 TS Section 2.2.3, DRAINAGE The current TS 2.2.3 states:

The Fuel Element Storage Well is designed and shall be maintained to prevent an inadvertent draining of the well below elevation of 679 ft MSL.

The licensee proposes to change the TS to read as follows:

The Fuel Element Storage Well is designed and shall be maintained to prevent an inadvertent draining of the well below elevation of 679 feet MSL while spent fuel assemblies are in the Fuel Element Storage Well.

The licensee proposes adding a statement while spent fuel assemblies are in the Fuel Element Storage Well to clarify that maintaining the FESW to prevent drainage below 679 feet is only applicable when spent fuel assemblies are stored in the FESW. This clarification will allow the licensee to drain the FESW after all fuel is removed and placed into the DCSS, so that the licensee may then decontaminate and dismantle the FESW and ancillary systems. The licensee also proposes editorial changes to this TS, to replace the ft abbreviation with the full word feet; and replace the term irradiated fuel with spent fuel for consistency with other sections of the TS.

The intent of this TS requirement is to ensure adequate water coverage over spent fuel assemblies is maintained for shielding purposes and for spent fuel cooling. Without fuel assemblies in the FESW, a draindown of water in the FESW below 679 feet would have no

consequences. The proposed change to TS 2.2.3 is acceptable because it maintains the current water level requirement until all fuel has been removed from the FESW.

In its review of this proposed TS change, the staff also considered prevention of inadvertent draining of the cask pool. The cask pool area is designed to be drained repeatedly in order to move empty spent fuel storage canisters and transfer casks into the reactor building and remove them after they have been loaded with fuel. However, prevention of inadvertent draining in the cask pool similar to that required for the FESW by TS 2.2.3 is important during spent fuel loading operations. The staff issued a request for additional information (RAI), dated March 11, 2010 (ADAMS Accession No. ML100540076), asking the licensee to describe the method of draining the cask pool and identify aspects of the cask pool design that would limit or prevent accidental drainage.

In its response, the licensee described the possible paths for accidental drainage of the cask pool area: two penetrations through the cask pool wall, the cask pool gate, and the transfer canal gate. The two penetrations in the lower pool each have double isolation valves and are connected to a water clean-up system which will be used to drain and clean the water in the cask pool. The cask pool gate will be constructed to be watertight, and the licensee will have the ability to monitor any leakage across the gate seal. The transfer canal gate connects the cask pool area to the FESW and will be removed while the cask pool is filled, connecting the two areas.

The FESW has been designed to prevent accidental draining and all penetrations below the top of the stored fuel have been capped.

The staff reviewed the licensees response and finds that the design of cask pool is adequate to prevent accidental drainage. Monitoring of the cask pool gate leakage and the cask pool/FESW water level will provide awareness of changing leakage rates and allow measures to be taken to address any leakage.

3.3 TS Section 4.1.1, GENERAL FUEL STORAGE AND HANDLING REQUIREMENTS 3.3.1 TS Section 4.1.1.1 The current TS 4.1.1.1 states:

Irradiated fuel assemblies shall be stored underwater in spent fuel storage racks that are positioned on the bottom of the Fuel Element Storage Well or in approved onsite dry spent fuel storage containers, or in an approved shipping cask.

The licensee proposes to change this TS to read as follows:

Spent fuel assemblies shall be stored underwater in spent fuel storage racks that are positioned on the bottom of the Fuel Element Storage Well or in an approved dry spent fuel storage cask.

The licensee proposes an editorial change to replace irradiated fuel with spent fuel for consistency throughout the TS. The licensee also proposes replacing the words in approved onsite dry spent fuel storage containers, or in an approved shipping cask with in an approved dry spent fuel storage cask to clarify that at LACBWR, spent fuel assemblies will only be stored in the FESW or in an approved dry spent fuel storage cask system selected for use at the LACBWR

ISFSI. The licensee does not consider the interim time period when the spent fuel assemblies will reside temporarily in the storage canister inside the transfer cask in the cask pool during fuel loading operations to be considered storage.

The staff agrees that for the purpose of this TS, that spent fuel residing temporarily in the storage canister inside the transfer cask in the cask pool during fuel loading operations, is not considered to be in storage. Spent fuel in a storage canister or transfer cask that is intact and closed per the requirements of the Part 72 CoC is adequately shielded and cooled, and therefore does not need to be stored underwater. The staff has reviewed these changes for any impact on the existing TS requirements and found them acceptable.

3.3.2 TS Section 4.1.1.2 The current TS 4.1.1.2 states:

During the handling of irradiated fuel assemblies that have been operated at power levels greater than 1 Mwt, the depth of water in the Fuel Element Storage Well shall be at least 2 feet above the active fuel, and only one fuel assembly will be moved at a time.

The licensee proposes to modify the TS to read as follows:

During the handling of spent fuel assemblies that have been operated at power levels greater than 1 Mwt, the depth of water in the Fuel Element Storage Well and the contiguous cask pool shall be at least 2 feet above the active fuel, and only one spent fuel assembly will be moved at a time.

The licensee proposes an editorial change to replace irradiated fuel with spent fuel for consistency throughout the TS. The licensee proposes to add the words and the contiguous cask pool to clarify that this requirement for 2 feet of water coverage over spent fuel assemblies during fuel handling applies as fuel is moved between the FESW and the cask pool. During cask loading operations when the FESW is hydraulically contiguous with the cask pool, the spent fuel assemblies are transferred underwater from the FESW through the fuel transfer canal to the storage canister located in the cask pool. The proposed change will make it clear that the minimum 2 feet water coverage above active fuel applies throughout the extended FESW (including the fuel transfer canal and cask pool). The staff finds that the revision to TS 4.1.1.2 does not reduce the requirements of the current TS and adds clarification of its applicability.

Therefore, the staff finds the proposed change appropriate and acceptable.

3.3.3 TS Section 4.1.1.3 The current TS 4.1.1.3 states:

With the exception of a shipping cask or transfer cask, the core spray bundle, the transfer canal shield plug and the other waste processing components and fixtures weighing less than 50 tons that are located and used within the storage well, no objects heavier than a fuel assembly shall be handled over the Fuel Element Storage Well.

The licensee proposes to completely replace this TS with the following:

No object heavier than 25 tons shall be handled over spent fuel assemblies located in the Fuel Element Storage Well or cask pool. Lifting and movement of a fuel-loaded storage canister and transfer cask shall be performed using the single-failure-proof cask handling crane lifting system meeting the guidance in NUREG-0612, Section 5.1.6. Lifting and movement of objects over spent fuel assemblies located in the Fuel Element Storage Well or cask pool shall be performed in accordance with the LACBWR NUREG-0612 commitments and the dedicated project heavy load control plan.

The proposed new TS 4.1.1.3 removes references to obsolete or unused equipment, limits the acceptable load to 25 tons, and establishes that all heavy load lifts will be done in accordance with the licensees commitments to NUREG-0612. As stated by the licensee, the limit of 25 tons provides a safety factor of 2 for lifts performed using the Reactor Building polar crane, which has a capacity of 50 tons. The worst-case analyzed drop is of a 50-ton cask onto spent fuel assemblies.

The staff reviewed the proposed new TS 4.1.1.3 against the current licensing basis for acceptability. The new TS conservatively limits the weight of objects being moved over the spent fuel pool to 25 tons without exceptions, ensuring that the analyzed 50 ton cask drop remains bounding for all heavy load activity. NUREG-0612 is specifically referenced, providing clarification that the guidance contained therein must be used for heavy load lifts. The licensees proposed changes reflect that the use of a single-failure-proof cask handling crane will be used to protect the spent fuel assemblies while being moved while inside the canister and protects the FESW and cask pool from heavy load drops beyond the plants licensing basis. TS 4.1.1.3, as proposed, is more limiting than the current TS 4.1.1.3 and provides greater assurance that heavy load lifts will be performed safely and that the accident analysis will continue to be bounding.

Therefore, the staff finds the proposed new TS 4.1.1.3 acceptable.

3.4 4.1.2, FUEL ELEMENT STORAGE WELL The licensee has proposed multiple changes to the limiting conditions for operation (LCO) and the surveillance requirements (SR) in TS 4.1.2. The proposed changes are shown below with deleted text in strikeout and new text in bold.

4.1.2 FUEL ELEMENT STORAGE WELL AND CASK POOL LIMITING CONDITION FOR OPERATION

=======================================================

Note This LCO does not apply to the cask pool if the spent fuel storage canister lid is in place in the canister or if there are no spent fuel assemblies in the cask pool.

The Fuel Element Storage Well (FESW) and cask pool shall meet the following requirements.

a. The Fuel Element Storage Well and cask pool water level shall be at least

16 11 feet, 61/2 inches above any irradiated spent fuel assembly stored in the spent fuel storage racks or in a spent fuel storage canister in the cask pool, and

b. Water in the storage well and cask pool shall be maintained at a temperature 150°F.

APPLICABILITY: At all times. While spent fuel assemblies are in the FESW or the cask pool.

ACTIONS

a. With water level less than required by the LCO, 16 feet above any irradiated fuel stored in the Fuel Element Storage Well storage racks, take immediate action to restore water level and suspend all operations involving FUEL HANDLING.

b. With water temperature in the storage well or cask pool above 150°F, take actions to reduce water temperature to 150°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and suspend any evolutions all operations involving FUEL HANDLING.

SURVEILLANCE REQUIREMENTS

=======================================================

Note SR 5.1.2.1 and 5.1.2.2 do not apply to the cask pool if the spent fuel storage canister lid is in place in the canister or if there are no spent fuel assemblies in the cask pool.

5.1.2.1 The Fuel Element Storage Well and cask pool water level and FESW System water temperature shall be monitored verified at least once per 24 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

5.1.2.2 The Fuel Element Storage Well and cask pool water level indication channel shall be calibrated (CHANNEL CALIBRATION) at least once per 18 months.

The impacts of the proposed changes to TS 4.1.2 are discussed and evaluated below.

3.4.1 Fuel Element Storage Well Thermal Hydraulic Impact Once the fuel in the upper tier spent fuel storage racks in the FESW is removed and placed into the DCSS, the licensee desires to no longer reinstall the fuel transfer canal gate during cask loading operations, to reduce occupational doses (obtained by personnel while uninstalling and reinstalling the fuel transfer canal gate) and for ease in operations. As the cask pool and FESW are hydraulically contiguous (with the fuel transfer canal gate removed), when the licensee lowers the water level in the cask pool to allow removal and preparation of the filled cask, the FESW water level will also be lowered. The FESW water level will be lowered to just below the bottom of the fuel transfer canal, which is at elevation 680 feet, 5 inches. Specifically, the licensee proposes to reduce the water level requirement in LCO 4.1.2, to 11 feet, 6.5 inches of water above

spent fuel assemblies stored in the spent fuel storage racks in the FESW or in an open spent fuel canister in the cask pool. As the top of spent fuel in the lower tier spent fuel storage racks in the FESW is at elevation 668 feet, 2.5 inches, this proposed minimum water level corresponds to an elevation of 679 feet, 9 inches in the FESW.

The staff reviewed the proposed change with respect to the licensees ability to maintain adequate FESW cooling with a reduced water level. The FESW cooling system takes suction from the FESW via a 6 inch penetration at elevation 679 feet and pumps the water through a cooler and back to the FESW, using either of two available storage well pumps. With spent fuel stored in only in the lower storage racks, the minimum required water level would be at elevation 679 feet, 9 inches, which is the minimum required level of water (9 inches above the pipe centerline at 679 feet) to prevent vortexing of air into the FESW pump suction piping. The applicant has also stated that water level will generally be maintained between elevation 680 feet and 680 feet, 5 inches, which would provide adequate submergence for the FESW cooling system suction line.

The licensee performed a test to measure the actual heat-up rate of the FESW with all cooling and circulation isolated, and the results are documented in LACBWR Technical Report LAC-TR-137.

Over 15 days, the temperature of the FESW temperature increased to only 114°F. The licensees extrapolation of the data indicated that the temperature would have stabilized at approximately 150°F. In the event of a loss of FESW cooling (e.g., if there was air entrainment and the FESW cooling system pumps became inoperable), makeup water is available from the Demineralized Water System or the Overhead Storage Tank. In consideration of the very slow heat-up rate, the staff finds that there is sufficient time to provide makeup water and initiate actions to ensure that FESW temperature does not exceed the 150°F limit required by TS LCO 4.1.2.b. The staff has reviewed the information submitted by the licensee and finds that the proposed reduced water level will be sufficient to maintain FESW cooling, and thus the proposed change to LCO 4.1.2 is acceptable.

The staff reviewed the licensees proposed changes to SR 5.1.2.1. The licensee stated that FESW water level and temperature are continuously monitored and alarmed by various level transmitters and temperature thermocouples. The licensee proposes to replace the word monitored with verified to more accurately reflect the purpose of the SR to verify that the LCO is met. The staff believes this editorial change clarifies the purpose of the SR and is acceptable.

The current SR 5.1.2.1 requires verification of the FESW water level and temperature every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee proposed adding the cask pool to SR 5.1.2.1, so that cask pool water level and temperature will also be verified. The staff believes this proposed change to extend the surveillance requirement to the cask pool is appropriate and acceptable.

The licensee proposes changing SR 5.1.2.1 to increase the frequency of verification of water level and temperature in the FESW and cask pool, from once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The addition of a more frequent surveillance of water level and temperature is appropriate because any loss of inventory from the FESW during lowered water level operations/conditions would leave less time for corrective actions to be taken before the FESW cooling system is affected. Verifying the extended FESW water level and temperature on a more frequent basis ensures that any problem causing a loss of FESW level or increase in temperature would be detected early. Rather than applying the increased verification frequency only during lowered water level conditions, the licensee determined that, from a human factors standpoint, it was prudent to change the surveillance frequency at all times to every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. A 12-hour frequency

also aligns with the 12-hour operating shifts employed at LACBWR. The staff evaluated the proposed change to SR 5.1.2.1 within the current licensing basis. The proposed reduction of the minimum required water level results in a decreased margin for accidental leakage. A more frequent verification of the water level will result in earlier detection of significant leakage from the FESW and cask pool. As discussed previously, the FESW is designed to prevent drainage below the elevation of 679 feet and leakage from the cask pool is addressed by double isolation on penetrations and monitoring of leakage past the cask pool gate. The staff finds that the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> frequency for SR 5.1.2.1 is acceptable and will ensure that any leakage is detected before water level decreases below the required minimum water level. Therefore, the increased frequency of the verification of FESW and cask pool water level and temperature is appropriate, and the staff finds the licensees proposed changes to SR 5.1.2.1 acceptable.

3.4.2 Fuel Element Storage Well Reactivity Impact The staff reviewed and evaluated the FESW reactivity impact of lowering the water level in the FESW, per the proposed changes to TS 4.1.2. In the submittal, the licensee evaluated the reactivity impact of lowering the water level of the FESW. The licensee stated that any reduction in the water coverage below the normal water level will increase neutron leakage, and therefore, decrease the reactivity of the FESW system.

In its evaluation, the licensee credits the neutron absorbing material, Carborundum, in the spent fuel racks to maintain the effective neutron multiplication factor (Keff) less than or equal to 0.95 (to ensure the FESW is maintained in a subcritical state). Carborundum is used in the spent fuel storage racks for the absorption of neutrons, and degradation of the Carborundum in the racks could result in an increase in the reactivity of the spent fuel configuration. In order to ensure that the neutron absorbing material will be able to maintain its functionality during movement of fuel assemblies and work in the FESW during the cask loading operations, and until the fuel is removed from the racks, the staff evaluated the past surveillance history of the material and the Operations Procedure, OP-58-02, Irradiated Fuel Element Storage Rack Poison Material Surveillance Program. OP-58-02 contains the procedures to perform the testing on the coupon samples including visual testing and determining weight loss. OP-58-02 also indicates that a weight loss of 10% or less is acceptable and any weight loss greater than 10% would require an evaluation of the effect on the shutdown margin of the FESW by the Reactor Engineer.

In a letter dated May 19, 2010, the licensee provided additional information on Carborundum and the results of past surveillances and testing of the sample coupons of Carborundum material in the FESW. The licensee noted that the Carborundum, as manufactured, contained approximately 64 weight percent of boron carbide (B4C). The neutron and gamma flux to the sample coupon tree was maximized every outage by the placement of the coupon tree in the area of highest flux. The visual inspection results of the sample coupons showed that the surface appearance was very similar to the archive material, the B4C matrix was tightly adhered to the backing material, and the surface of the sample material was more friable than the archive material. The surveillance results are trended and monitored. In 2005, there were coupons tested that had results outside the OP-58-02 procedure acceptance criteria (greater than 10 weight percent loss of the sample coupon). In response, more coupons were pulled and tested, and these results met the acceptance criteria.

The licensee provided additional information on the calculations to determine that a Keff of less than or equal to 0.95 is maintained. The minimum boron-10 (B-10) areal density required for the

manufactured racks is 0.024g/cm2. The as-built areal density of the B-10 is nominally 0.038g/cm2. The assumed areal density in the criticality analysis is 0.022g/cm2. Assuming the worse case abnormal configuration with no burn-up credit, the Keff upper limit was calculated to be 0.9275 (using as-built areal densities). When a B-10 areal density of 0.0195g/cm2 is assumed, the calculated Keff is 0.9325. This is below the regulatory limit of Keff less than or equal to 0.95. The results of the 1997 neutron attenuation test were that the areal density was greater than that required in the criticality analysis.

The staff reviewed the licensees additional information and found it to be acceptable. The nominal as-built areal density, 0.038g/cm2, is much greater than the amount assumed in the criticality analysis, 0.022g/cm2. Using the nominal as-built areal density, the Keff was calculated to be 0.9275. This is much less than the regulatory value of 0.95 and substantial margin is available. In addition, if the areal density were to degrade approximately 50 weight percent, the licensees calculated Keff is 0.9325, which is still below 0.95. Therefore, the degradation acceptance criteria of 10 weight percent loss in the licensees surveillance program (in OP-58-02) is conservative since the criticality analysis minimum areal density can have at least 50 weight percent loss and still meet the regulatory limit. The maximum degradation (12 weight percent),

seen in the 2005 surveillance testing, is bounded by the approximately 50 weight percent loss Keff calculation. While the 12 weight percent loss result did not meet the acceptance criteria, the licensee took prompt corrective actions to test additional samples and disposition the original result. In addition, the licensee performed acceptable visual inspections, performed neutron attenuation testing in 1997 (which yielded acceptable results), and the latest surveillance results in 2008 met the acceptance criteria. Therefore, the staff has reasonable assurance that the surveillance program in place for the neutron absorbing materials is sufficiently robust to identify any degradation of material well before it can challenge safety margins or regulatory limits.

Taking into account the dispositioned 12 weight percent loss sample and some staff uncertainties about the testing method of the sample coupons (e.g., consistency of water drying methods and handling techniques), the amount of margin available and the results of the testing support a reasonable assurance determination that the degradation is unlikely to exceed 50 weight percent in the next few years and challenge the criticality analysis. Therefore, the staff has reasonable assurance that the neutron absorbing material will be able to perform its intended safety function during cask loading operations, maintaining the Keff less than or equal to 0.95 (and maintaining subcriticality of the FESW) until all of the fuel is removed from the FESW storage racks.

3.4.3 Fuel Element Storage Well Shielding Impact The staff reviewed and evaluated the impact of lowering the water level in the FESW, as proposed in TS 4.1.2, on shielding and occupational doses.

The licensee used a special test procedure STP-58-01, Perform Radiation Survey of FESW at Canal Gate Level, to obtain underwater radiation measurements to assess potential occupational doses for the proposed operations. In response to a staff RAI, the licensee provided additional information (by letter dated May 19, 2010) on the manufacturer, model, and capabilities of the radiation survey instrumentation used for the radiation surveys in STP-58-01.

A Thermo Scientific underwater detection system model FHZ312 with an FH40 GX display unit is used for the underwater radiation surveys. The licensee described the radiation calibration checks and protocols, where the systems operability and calibration were checked before and after its use. The description indicated that the radiation instrumentation range is 10 mrem/hr to

10,000 rem/hr and is designed for operation at the water depth of 20 meters. NRC staff concludes that adequate radiation instrumentation is available to make the radiation measurements using STP-58-01.

The licensee elected to use a measured radiation dose rate rather than a calculated dose rate to analyze the expected occupational doses from planned operations. Using STP-58-01, the licensee provided measured radiation dose rates, which appear to staff to be adequate to demonstrate the dose rates expected from spent fuel assemblies at the proposed reduced water level.

The licensee used the measured radiation dose rates to estimate occupational doses that may be accrued during the: (1) fuel loading operations per the current TS (where the licensee would need to continually remove and reinstall the fuel transfer canal gate throughout loading operations to maintain the current 16 feet minimum water coverage above fuel, per TS 4.1.2.a.); and (2) fuel loading operations per the proposed changes to the TS (to lower the minimum water coverage in the FESW to 11 feet, 6.5 inches above fuel, which would eliminate the need for the licensee to reinstall the fuel transfer canal gate after all of the fuel assemblies in the upper tier racks are removed).

The license provided tables showing occupational dose estimates for the two cases - operations per the current TS with 16 feet minimum water coverage above fuel, and operations per the proposed changes to the TS with minimum water coverage of 11 feet, 6.5 inches above fuel.

The tables provided information on the task performed, expected dose rate per task, time for task completion, number of personnel involved, and the expected dose per fuel transfer canal gate cycle. The information indicated that the dose estimates are approximately 296 person-rem for 16-feet water coverage and 7 canal gate cycles, and approximately 169 person-rem for 11 feet, 6.5 inches water coverage and 3 canal gate cycles. The license concluded that the reduced water coverage permits a reduction in the canal gate cycles, resulting in less occupational dose to workers overall. The licensees proposed change to TS 4.1.2 to reduce the minimum water coverage above fuel in the FESW, results in reducing occupational dose by an estimated 127 person-rem during fuel loading operations. Therefore, the staff finds the licensees proposed change to TS 4.1.2 to be acceptable, from a shielding and dose perspective.

3.4.4 Accident Analysis Impact 3.4.4.1 Spent Fuel Handling Accident As described in Section 9.2 of the DP, the spent fuel handling accident (FHA) postulates a fuel assembly falling from the hoist into the FESW. It is assumed that the fuel cladding ruptures releasing the activity contained in the gap between the solid fuel pellets and the fuel pin cladding.

The gap activity is assumed to be released over a two hour period with no credit for dilution or filtration. The licensee considered the potential for changes to the current licensing basis (CLB) dose consequence FHA during periods of reduced water coverage, per the proposed change to TS 4.1.2. The licensee has determined that there will be no impact on the accident as described in Section 9.2 of the DP. The licensee asserts, and the NRC staff agrees, that because no fuel movements will occur while the water level in the FESW is lowered to slightly below the transfer canal bottom elevation, there is reasonable assurance that an FHA could not occur during the period of reduced water cover. In addition, the fuel transfer bridge will have to be parked over the FESW to allow the cask handling crane to lift and move the transfer cask/canister assemblage

into and out of the cask pool thereby eliminating the possibility of fuel assembly movement using the fuel transfer bridge.

The NRC staff reviewed the CLB FHA, as documented in section 9.2 of the DP, and found that the licensee used conservative assumptions to evaluate the potential dose consequences. The analysis assumes the release of 431.4 curies of krypton-85 (Kr-85) representing 30 percent of the total Kr-85 activity in the damaged fuel as of October 1987. Conservatively, the licensee did not credit the considerable reduction in the Kr-85 inventory as a result of radioactive decay since 1987. In addition, since the release of Kr-85 is independent of the amount of water cover, any reduction in the assumed amount of water cover, as proposed in this amendment request, would have no impact on the CLB FHA analysis regarding the release of Kr-85.

In a letter dated May 19, 2010, the licensee provided additional information describing the basis for considering only the release of Kr-85 in the FHA dose consequence analysis. The licensee stated that because the reactor was permanently shut down over 20 years ago, the principal fission gas remaining for any potential fuel damage accident at LACBWR is Kr-85. Other krypton and xenon radioisotopes that were produced during plant operation have decayed to insignificant levels. Halogen radionuclides, such as radio-bromines and radio-iodines have also decayed to insignificant levels, with the exception of iodine-129 (I-129) which has a half life of approximately 15.9 million years. The licensee computed the total I-129 inventory in the FESW to be less than 0.4 curies. The licensee calculated a thyroid dose of less than 1 rem as a result the postulated release of the entire inventory of I-129 and therefore determined that the postulated release of I-129 from an FHA would result in an insignificant thyroid dose to onsite personnel and to members of the public.

The NRC staff noted some differences in the assumptions used by the licensee and the currently accepted regulatory guidance for determining the dose consequence from the postulated release of I-129 in an FHA. While the licensee used a very conservative gap fraction of 30 percent for I-129, current guidance as stated in Regulatory Guide (RG) 1.195, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors, indicates a gap fraction of 5 percent for halogens other than iodine-131. The licensee assumed a decontamination factor (DF) of 50 for an 11 foot depth of water cover over the fuel assemblies. Using current guidance from RG 1.195, the NRC staff has determined that an appropriate DF for an 11 foot depth of water cover over the fuel assemblies would be on the order of 20. Notwithstanding the stated differences in assumptions, the NRC staffs confirmatory dose analysis, using the methods described in RG 1.195, indicates that the calculated dose from an FHA, including the contribution from I-129, would be well within the dose limits of 10 CFR Part 100, Reactor Site Criteria, as stated in 10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance. Therefore, the NRC staff concludes that the reduced water coverage limits proposed in TS 4.1.2 will ensure that the dose consequences of a fuel assembly drop will continue to remain within the applicable limits and are therefore acceptable.

3.4.4.2 Shipping Cask or Heavy load Drop into FESW This accident, as described in Section 9.3 of the DP, postulates a shipping cask or other heavy load falling into the FESW. The licensee considered the potential for changes to the CLB dose consequence shipping cask or heavy load drop into the FESW accident, as a result of the reduced water coverage in the proposed change to TS 4.1.2. For this accident, it is postulated that all 333

spent fuel assemblies located in the FESW experience fuel pin clad damage releasing the Kr-85 activity contained within the fuel pin gap. As in the case of the FHA, since the release of Kr-85 is independent of the amount of water cover, any reduction in the assumed amount of water cover, as proposed in this amendment request, would have no impact on the CLB shipping cask or heavy load drop into the FESW analysis regarding the release of Kr-85.

In addition to the evaluation contained in the DP, the licensee considered the postulated release of I-129 from a heavy load drop into the FESW. The licensee calculated a thyroid dose of less than 1 rem as a result the postulated release of the entire inventory of I-129 and therefore determined that the postulated release of I-129 from a heavy load drop into the FESW would result in an insignificant thyroid dose to onsite personnel and to members of the public.

The NRC staff performed a confirmatory dose analysis using the methods described in RG 1.195 to assess the dose consequence from the release of I-129 due to a heavy load drop accident in the FESW. The results indicate that the calculated dose would be well within the limits of 10 CFR 100.11. Therefore, the NRC staff concludes that the reduced water coverage limits proposed in TS 4.1.2 will ensure that the dose consequences of a shipping cask or heavy load drop into the FESW will continue to remain within the applicable limits and are therefore acceptable.

The NRC staff notes that the discussion of the accident analyses in the DP references the 10 CFR Part 100 limits of 25 rem whole body and 300 rem thyroid for the FHA and the shipping cask or heavy load drop, rather than the current acceptance criteria of well within 10 CFR Part 100 limits. The Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (NUREG-0800, Formerly issued as NUREG-75/087) (SRP) Section 15.7.4, Radiological Consequences of Fuel Handling Accidents, defines well within to be 25 percent or less of the Part 100 limits, or 6 rem whole body and 75 rem thyroid.

In a letter dated May 19, 2010, the licensee provided additional information explaining that the LACBWR plant was first licensed in the late 1960s, well before the SRP was issued. The SRP was first issued in 1975 as NUREG-75/087 and subsequently reissued as NUREG-0800 in 1981.

Because LACBWR was licensed before the SRP existed, the full 10 CFR Part 100 dose limits are documented as the acceptance criteria for radiological accidents in the LACBWR DP, Sections 9.2 and 9.3, rather than the fractions of those dose limits published in the current revision of the SRP. To address the NRC staffs concern, the licensee performed a comparison of the calculated accident event doses to the well-within SRP criteria. The licensees evaluations, as well as the NRC staffs confirmatory calculations, demonstrate that the estimated doses for the radiological accidents described in Section 9.2 and 9.3 of the DP, meet the well-within dose criteria in SRP 15.7.4. In addition, the licensees evaluations, as well as the NRC staffs confirmatory calculations, demonstrate that the reduced water coverage limits in the proposed changes to TS 4.1.2 will provide reasonable assurance that the dose consequences for the radiological accidents described in Section 9.2 and 9.3 of the DP, will continue to meet the well-within dose criteria in SRP 15.7.4 and are therefore acceptable.

3.4.5 Technical Specification Changes The licensee proposed various modifications to TS 4.1.2 including editorial changes, expansion of the TS applicability to the cask pool, a decrease in the required height of water above fuel stored in the FESW and cask pool, and an increase in the surveillance frequency for FESW water level.

The licensee proposes several editorial changes to this TS section: (1) replace irradiated fuel with spent fuel for consistency throughout the TS; (2) change monitored to verified to more accurately reflect the purpose of the surveillance requirements to verify that the LCO is met; (3) change any evolutions to all operations to make the language in the two action items in the TS consistent; (4) replace 16 feet above any irradiated fuel stored in the Fuel Element Storage Well storage racks with required by the LCO to simply refer to the LCO not being met rather than repeating the LCO. These editorial changes are improvements and clarifications to this TS section, and thus, are acceptable to the staff.

In the proposed modifications, a reference to the cask pool is inserted into the conditions, applicability, actions, and surveillance requirements of TS 4.1.2. The references to the cask pool explicitly capture that the requirements of the TS extend to the cask pool area. Additionally, statements are added to clarify that the LCO and surveillance requirements do not apply to the cask pool if the canister lid is in place on a spent fuel storage canister or if there is no spent fuel in the cask pool. The staff reviewed these proposed modifications for their impact on the TS.

These statements extend the requirements of the TS to the cask pool area, and the staff finds this acceptable. As discussed above, with the spent fuel storage canister lid in place, the canister and spent fuel assemblies stored inside it are subject to the Part 72 CoC. Therefore, the licensee has accurately restricted the applicability of TS 4.1.2 in the cask pool to periods when spent fuel assemblies are present in an open spent fuel storage canister.

4.0 STATE CONSULTATION

In accordance with NRC's regulations, the Wisconsin State official was notified of the proposed issuance of this amendment, on October 9, 2009 and November 23, 2010. The State official had no comment.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment involves changes to the LACBWR License and Technical Specifications. The 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. NRC has previously issued a proposed finding that the amendment involves no significant hazards consideration (74 FR 51326; October 6, 2009), and there has been no public comment on such finding. NRC staff has made a final determination that the proposed amendment does not involve a significant hazards consideration. 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 this amendment.

6.0 CONCLUSION

NRC 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 NRC=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: K. Banovac D. Cunanan E. Davidson J. Hickman J. Parillo E. Wong T. Youngblood Date: January 25th, 2011