RBG-47505, Response to Request for Additional Information on License Amendment Request 2013-13

Response to Request for Additional Information on License Amendment Request 2013-13

ML14272A180
Person / Time
Site:	River Bend
Issue date:	09/23/2014
From:	Mashburn W Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML14272A185	List: ML14272A181 RBG-47505, Response to Request for Additional Information on License Amendment Request 2013-13
References
RBF1-14-0149, RBG-47505
Download: ML14272A180 (11)

Text

SEntergy Entergy Operations, Inc.

River Bend Station 5485 U.S. Highway 61N St. Francisville, LA 70775 RBG-47505 September 23, 2014 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Response to Request for Additional Information on License Amendment Request 2013-13 River Bend Station - Unit 1 Docket No. 50-458 License No. NPF-47

REFERENCES:

1. Entergy letter to NRC, dated July 29, 2013, License Amendment Request 2013-13, Heavy Load Movement Over Fuel Assemblies (Letter No. RBG-47382)

2. NRC letter to Entergy (via email), dated July 31, 2014, Request for Additional Information RBF1-14-0149

Dear Sir or Madam:

On July 29, 2013, Entergy Operations, Inc. (Entergy) submitted a request to allow the movement of heavy loads over fuel assemblies (Reference 1). During their review, the NRC staff determined that additional information is needed to complete the processing and approval of Entergy's request. The request for that information was transmitted to Entergy per Reference 2. Attachments 1 and 2 to this letter contain the requested information. The licensee commitment is summarized in Attachment 3.

If you have any questions on this matter, please contact Joey Clark, Manager -

Regulatory Assurance, at 225-381-4177.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 23, 2014.

Sincerely, Willis F. Mashburn Director - Engineering WFM/dhw Ac

RBG-47505 September 23, 2014 Page 2 of 2 : Response to Request for Additional Information : Calculation G13.18.2.7-116, Rev. 0 : Licensee Commitment cc: U. S. Nuclear Regulatory Commission Attn: Mr. Alan Wang MS 8-G14 One White Flint North 11555 Rockville Pike Rockville, MD 20852 (w/ Attachments 1 and 3 only)

Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 E. Lamar Blvd.

Arlington, TX 76011-4511 NRC Senior Resident Inspector River Bend Station Department of Environmental Quality Office of Environmental Compliance Radiological Emergency Planning and Response Section Ji Young Wiley P.O. Box 4312 Baton Rouge, LA 70821-4312 Public Utility Commission of Texas 1701 N. Congress Ave.

Austin, TX 78711-3326

Attachment 1 RBG-47505 Response to Request for Additional Information (8 pages)

For the purposes of addressing Request for Additional Information (RAI) items on Reference 1, a load drop analysis meeting the requirements of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" has been performed. This response includes the RBS spent fuel pool gate load drop analysis performed under calculation G13.18.2.7-116, Rev. 0, "Load Drop Calculation for spent fuel pool gates (FNS-GATE1 and FNS-GATE2)." That calculation is included as Attachment 2. The balance of this response addresses the compliance of the load drop analysis with NUREG-0612, Section 5.1 evaluation criteria and the Appendix A analysis guidelines.

Compliance with NUREG-0612 Section 5.1 Evaluation Criteria Evaluation Criterion I:

"Releases of radioactive material that may result from damage to spent fuel based on calculations involving accidental dropping of a postulated heavy load produce doses that are well within 10 CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses should show that doses are equal to or less than 1/4 of Part. 100 limits)."

RBS has adopted the criteria of Code of Federal Regulations, Title 10, Section 50.67, "Accident Source Term" with the design basis events analyzed in accordance with the requirements of Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000. The calculated 10CFR50.67 Total Effective Dose Equivalent (TEDE) limits for a fuel handling accident determined in accordance with Regulatory Guide 1.183 requirements are as follows:

Exclusion Area Boundary (EAB): 6.3 REM TEDE * (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration)

Low Population Zone (LPZ): 6.3 REM TEDE (30 day duration)

Control Room : 5 REM TEDE (30 day duration)

• Worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period As movement of the gates for the purposes of replacing the seals is typically scheduled to be performed during power operations, the dose analysis is based upon the assumption that the fuel impacted in the spent fuel pool has been subcritical for a minimum of 14 days (336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br />), which represents the minimum realistic refueling outage duration. The requirement for a minimum of 14 days decay time for recently irradiated fuel will be included in the fuel pool gate load movement procedure.

The calculated spent fuel pool gate load drop analysis dose for each receptor is as follows:

Exclusion Area Boundary (EAB): 1.2155 REM TEDE * (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration)

Low Population Zone (LPZ): 0.16017 REM TEDE (30 day duration)

Control Room : 0.87328 REM TEDE (30 day duration)

• Worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period The spent fuel pool gate load drop analysis calculated doses are less than 25% of the calculated 10CFR50.67 total effective dose equivalent (TEDE) limits for a fuel handling accident determined in accordance with Regulatory Guide 1.183 requirements. The spent fuel pool gate load drop analysis doses are also bounded by the doses in current fuel handling accident (FHA) in the fuel building. Table 1 provides the regulatory acceptance criteria, FHA 1

in the fuel building, and the spent fuel pool gate load drop analysis dose results for comparison purposes.

Table 1 Dose Acceptance FHA in fuel building Gate Drop in fuel Receptor Criteria REM TEDE building REM TEDE REM TEDE EAB* 6.3 2.5735 1.2155 LPZ 6.3 0.33912 0.16017 CR 5 1.6790 0.87328

• Worst 2-hour period In summary, although 10CFR100 limits are no longer applicable to RBS, the postulated doses from a drop of a fuel building spent fuel pool gate are less than 25% of the established 10CFR50.67 regulatory limit and bounded by the current design and licensing basis FHA in the fuel building calculated doses. Thus, it is concluded that the requirements of this criterion are met.

Evaluation Criterion II "Damage to fuel and fuel storage racks based on calculations involving accidental dropping of a postulated heavy load does not result in a configuration of the fuel such that keff is larger than 0.95."

Based upon the analysis provided in calculation G13.18.2.7-116, the damage to the fuel storage racks as a result of a spent fuel pool gate load drop is bounded by the original load drop analysis. The original load drop analysis calculated a worst case keff of 0.914 considering the analyzed damage. As the damage analyzed in G13.18.2.7-116 is bounded by the original damage analysis, the worst case keff is also bounding. Thus, the resultant rack damage does not result in a keff greater than 0.95 and the requirements of this criterion are met.

Evaluation Criterion III "Damage to the reactor vessel or the spent fuel pool based on calculations of damage following accidental dropping of a postulated heavy load is limited so as not to result in water leakage that could uncover the fuel, (makeup water provided to overcome leakage should be from a borated source of adequate concentration if the water being lost is borated)."

The spent fuel pool, cask pool, and inclined fuel transfer system (IFTS) pool are provided with stainless steel liners to provide a leak-proof membrane that is resistant to damage from abrasion, impact and concrete floor and wall displacements. An analysis of the effect of the spent fuel pool gate load drop on the pool liners was performed. The results of the analysis conclude that none of the objects analyzed would result in perforation of the stainless steel liner beyond the required 25% margin. Thus, as the liner integrity will be maintained with the recommended margin during a postulated spent fuel pool gate load drop, there will be no water leakage that could uncover the fuel and providing makeup to accommodate leakage is not required. Thus the requirements of this criterion are met.

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Evaluation Criterion IV "Damage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental dropping of a postulated heavy load, will be limited so as not to result in loss of required safe shutdown functions."

The only safe shutdown function potentially affected by a postulated spent fuel pool gate load drop is spent fuel pool cooling. Due to the piping configuration, there is limited potential for a load drop to affect the piping in both divisions of spent fuel pool cooling. However, to ensure no loss of safe shutdown function with respect to spent fuel pool cooling, an assessment of the potential damage that could result from impact of a dropped gate on the spent fuel pool cooling supply and return piping located in the spent fuel pool was performed. The analysis demonstrates that the impact of the gate will be below the minimum spent fuel pool water level, thus any damage resulting in penetration of the pipe wall would not prevent the accomplishment of the fuel pool cooling function. Further analysis to assess the potential damage was performed. The calculated strike velocity associated with the impact of the gate on the pipe would result in penetration of the pipe wall less than the wall thickness of the piping. A structural evaluation modeling the piping as a simply supported beam was performed The use of this model is reasonable given that ductility is primarily a function of material properties and general configuration and as the gate will absorb a portion of the impact energy. The evaluation determined that the piping has adequate ductility to accommodate the impact with only denting-type deformation damage at the point of impact.

Given that the worst case condition of penetration of the pipe wall occurs below minimum water level does not result in the loss of spent fuel cooling, and further analysis indicates that the damage to the piping is not anticipated to result in perforation, the requirements of this criterion are met.

Compliance with NUREG-0612, Appendix A, "Analyses of Postulated Load Drops" The pool gate load drop analysis was performed evaluating the drop of three major elements involved in the movement of the fuel building spent fuel pool gates:

" Fuel pool gate and rigging. The nominal weight of the gate is 1600 lb., and the estimated total weight of the gate and rigging including intermediate beam is 2000 lb.

For this analysis, the estimated 2000 lb. total weight of the gate and rigging was conservatively utilized for the weight of the gate itself.

• The intermediate lifting beam. The intermediate lifting beam has an approximate weight of 175 lb.

• An alternate lifting beam. The weight of the alternate lifting beam used in the analysis is 200 lb.

The total weight of the elements utilized in this analysis is 2375 lb. Approximately 2500 lb.was the weight value used for inclusion in Section 9.1.2.3.3 of the RBS Updated Safety Analysis Report (USAR). Given that the weight of the elements in the analysis is within 5% of the approximate 2500 lb. USAR change weight value and the substantial conservatisms included in the analysis, no change to the previously submitted analyzed weight of "approximately 2500 Ib" is required.

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In addition to the analysis including the weight of the rigging and beams with the weight of the gate, each of the listed elements was assumed to independently strike the spent fuel in order to maximize the number of bundles struck by the objects, thereby increasing potential fuel damage. The maximum number of damaged rods determined in the analysis of the drop of each element was combined to calculate the overall maximum fuel damage associated with the load drop. The total number of damaged rods was then increased by approximately 25%

to provide design margin.

The following is a discussion of the pool gate load drop analysis relative to the General Considerations provided in NUREG 0612, Appendix A "Analysis of Postulated Load Drops",

Section 1. This analysis applies to the fuel building pool gates only and is not applicable to the pools and gates located in the reactor building.

1) The load is dropped in an orientation that causes the most severe consequences.

The pool gate load drop analysis evaluates the maximum spent fuel damage for the drop of the gate element for two impact sequence cases. The first case assumes the impact orientation sequence as shown on Figure 1.

Figure 1 Position 1 Position 2 Position 3 Impact 1 - Bottom edge of gate strikes stored spent fuel (Position 1).

Impact 2 - Side edge of gate strikes stored spent fuel (Position 2).

Impact 3 - Face of gate strikes stored spent fuel (Position 3).

The second case assumes the following impact orientation sequence:

Impact 1 - Bottom edge of gate strikes stored spent fuel (Position 1).

Impact 2 - Face of gate strikes stored spent fuel (Position 3).

Given that fuel damage could be affected by the number of bundles involved in the impact of each element, each impact orientation sequence includes damage assessed for the condition of the racks 100% full of spent fuel and for the condition of the racks 50% full of spent fuel. The analysis of the racks 50% full was included as this case increases the force 4

applied to each impacted fuel bundle. Note that the analysis results indicate that the cases with the racks 100% full are bounding.

The largest fuel damage resulting from the evaluated cases was utilized in the determination of the total number of fuel rods damaged from the load drop. The mass of the slings, shackles, hoists, and intermediate beam were added to the mass of the gate for analysis purposes.

The current rigging configuration contains an intermediate beam that is attached to the spent fuel pool gate. Although the weight of the intermediate beam was included in the weight of the gate and rigging, the intermediate beam was also analyzed as a separate dropped element with a weight of 175 lb.

For conservatism and margin, an additional alternate beam of a larger size and length than the intermediate beam was added to the analysis. This alternate beam was assumed to weigh 200 lb. and was analyzed as a separate dropped element. Although the weight of the alternate beam was not specifically analyzed either as an option to the intermediate beam or in addition to the intermediate beam in the total weight of the gate and rigging, adequate margin exists in the radiological analysis to accommodate its use in either.

scenario. This is because the total weight including gate, rigging, intermediate beam and alternate beam would be approximately 2200 lbs. representing a 10% increase in weight.

As discussed below, the calculated number of damaged rods was increased by approximately 25% which exceeds the damage that would be experienced by a 10%

weight increase.

The potential exists for the intermediate beam and the alternate beam to impact the stored spent fuel in a location separate from the gate. As a result, each beam was independently analyzed for a load drop on the spent fuel. The analysis of the drop of each beam element was performed assuming two impact orientations:

Impact 1 - Impact of the end of the beam on spent fuel.

Impact 2 - Impact of the side of the beam on spent fuel.

The worst case geometry for the drop of each beam was determined and maximum damaged number of fuel rods for each of these elements included in the overall fuel damage total.

The total number of damaged rods was calculated based upon the sum of:

• The maximum number of rods damaged from the drop of the gate.

• The maximum number of rods damaged from the drop of the intermediate beam.

• The maximum number of rods damaged from the drop of the alternate beam.

The total number of damaged rods was increased by approximately 25% to provide design margin.

2) The fuel impacted is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> subcritical (or whatever the minimum that is allowed in facility technical specifications prior to fuel handling).

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RBS Technical Requirements Manual TR 3.9. 10, "Decay Time," requires that the reactor be subcritical for greater than or equal to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in Mode 5 during movement of irradiated fuel in the reactor pressure vessel. However, the fuel building spent fuel pool gate seal replacement being analyzed is typically performed during non-refueling outage conditions.

On this basis, the spent fuel pool gate load drop analysis assumes that the fuel impacted in the fuel building spent fuel pool has been subcritical for a minimum of 14 days (336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br />),

which represents the minimum realistic refueling outage duration. The requirement for a minimum of 14 days decay time for recently irradiated fuel will be included in the fuel pool gate load movement procedure.

3) The load may be dropped at any location in the crane travel area where movement is not restricted by mechanical stops or electrical interlocks.

The gate movement pathway involved in this load drop analysis includes the spent fuel pool, the gate opening area between the spent fuel pool and cask pool, and the cask pool located in the fuel building. This pathway was previously documented in License Amendment Request 2013-13 Attachment 3.

The fuel damage assessment is based upon potential load drops in the spent fuel pool, as this is the only location where spent fuel is stored in the fuel building during gate seal replacement.

The pool liner damage assessment is based upon potential load drops in the spent fuel pool, cask pool, and the gate opening area between the spent fuel pool and cask pool. The postulated load drops in each area are based upon the location and elevation of each element during the course of gate movement.

The concrete structure of the cask pool including floor and intermediate shelf was previously analyzed for drops of 125 ton or larger shipping casks. The floor of the cask pool and spent fuel pool are both nominal 4 feet thick reinforced concrete and form part of the fuel building foundation. The cask pool shelf is also 4 feet thick reinforced concrete.

The gate opening area in the pool wall is over 4 feet wide and approximately 18 feet thick (deep). The consequences of a drop of a pool gate or beam elements in the spent fuel pool, gate opening area, or cask pool portion of the load movement pathway is bounded by the concrete structural analysis provided in the cask drop analysis.

Damage to the fuel storage racks is also postulated and evaluated in the fuel building gate movement load drop analysis.

As maintaining spent fuel pool cooling is classified as a safe shutdown function, an assessment of the effect of the impact of the gate on the spent fuel cooling piping located in the spent fuel pool was performed. Based upon the load pathway, the postulated scenario was a drop of the gate when the gate was located: 1) at the northern-most point of the opening between the spent fuel pool and cask pool, which is the closest location to the spent fuel pool cooling return piping; and 2) at the southern-most point of the opening between the spent fuel pool and the inclined fuel transfer system (IFTS) pool, which is the closest location to the spent fuel pool cooling suction piping. The gate would first impact the spent fuel and storage racks as shown in Figure 1 Position 1. Then the gate would impact the cooling piping during the transition to either Position 2 or 3 as shown in Figure 1.

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4) Credit may not be taken for spent fuel pool area charcoal filters if hatches, wall, or roof sections are removed during the handling of the heavy load being analyzed, or whenever the building negative pressure rises above (-)1/8 inch (-3 m) water gauge; No credit is taken for operation of the fuel building charcoal filtration system or the control room charcoal filtration system.

5) Analyses that rely on results of Table 2.1-1 or Figures 2.1-1 or 2.1-2 for potential offsite doses or safe decay times should verify that the assumptions of Table 2.1-2 are conservative for the facility under review. X/Q values should be derived from analysis of on-site meteorological measurements based on 5% worst meteorological conditions.

The analysis does not rely on NUREG 0612 Table 2.1-1 or Figures 2.1-1 or 2.1-2. The gate load drop scenario fuel damage analysis is based upon the methodology for determining fuel damage in the design basis fuel handling accident in the fuel building. The number of rods damaged for the gate load drop scenario was specifically determined based upon the geometry of the dropped elements (gate, intermediate lifting beam, and alternate lifting beam). The offsite dose analysis for the gate load drop utilizes the RBS design basis X/Q values used in the current analysis for the fuel handling accident in the fuel building and meets all requirements of Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000.

6) Analyses should be based on an elastic-plastic curve that represents a true stress-strain relationship.

The analysis for the fuel damaged in the load drop assumes elastic impact on the fuel by the dropped element until all kinetic and potential energy is expended. The damage to the fuel is based upon plastic compression failure of the rods located in the impacted bundles.

The analysis for the effect of the load drop on the pool liner is based upon plastic deformation (penetration) of the liner. Analysis of impact damage to the pool racks and effect on the keff is based upon plastic deformation of the racks resulting from the dropped load. Assessment of the impact on the spent fuel pool cooling piping is based upon elasto-plastic target response.

7) The analysis should postulate the "maximum damage" that could result, i.e., the analysis should consider that all energy is absorbed by the structure and/or equipment that is impacted.

All energy in the dropped gate and beam elements is assumed to be absorbed by the spent fuel. No credit is taken for energy absorbed in the elements being dropped. Similarly, in the analysis of the liner and racks, all energy is assumed to be absorbed by the liner and racks without crediting any energy absorbed by the dropped object. In the analysis of the spent fuel pool cooling piping, all energy is assumed to be absorbed by the piping.

8) Loads need not be analyzed if their load paths and consequences are scoped by the analysis of some other load.

The concrete structural analysis for a shipping cask drop in the cask pool envelopes the consequences of a drop of any of the elements involved in the spent fuel pool gate load movement. The areas involved in the gate movement pathway include the spent fuel pool, 7

gate opening area between the spent fuel pool and cask pool and the movement to the laydown area on the fuel building 113' elevation floor.

Analysis regarding perforation of the liner is performed at representative locations in the movement pathway.

9) To overcome water leakage due to damage from a load drop, credit may be taken for borated water makeup of adequate concentration that is required to be available by the technical specifications.

River Bend does not utilize borated water in the spent fuel pool. Analysis was performed to determine the potential extent of perforation of the stainless steel spent fuel pool, gate opening area, and cask pool liners. Based upon the results of the impact analysis, the existing liner thickness far exceed the potential depth of liner perforation due to a drop of the gate, intermediate beam or alternate beam elements. As the liner maintains its integrity following a postulated gate movement load drop event, the leak-proof barrier function of the liner is maintained and no water leakage will occur. Consequently, no water makeup due to damage resulting from this postulated load drop is required.

10) Credit may not be taken for equipment to operate that may mitigate the effects of the load drop if the equipment is not required to be operable by the technical specifications when the load could be dropped.

There is no credit taken for any equipment to mitigate the effects of the load drop.

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