Text

Summary of 930520 Meeting W/Util in Rockville,Md Re Proposed Mod to Plant Sfsp
	ML20044H105
Person / Time
Site:	Millstone
Issue date:	06/01/1993
From:	Vissing G; Office of Nuclear Reactor Regulation
To:	; Office of Nuclear Reactor Regulation
References
	TAC-M86361, NUDOCS 9306070400
	Download: ML20044H105 (55)
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WASHINGTON, D C. 2055 90001 June 1, 1993

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Docket No. 50-336 LICENSEE:

Northeast Nuclear Energy Company FACILITY: Millstone Nuclear Power Station, Unit 2

SUBJECT:

SUMMARY

OF MEETING OF MAY 20, 1993, WITH REPRESENTATIVES OF NORTHEAST NUCLEAR ENERGY COMPANY CONCERNING A, PROPOSED MODIFICATION TO THE MILLSTONE, UNIT 2, SPENT FUEL STORAGE P0OL PAC N0. M86361)

INTRODUCTION On May 20, 1993, representatives of the NRC and the Northeast Nuclear Energy Company (the licensee) met in the NRC offices in Rockville, Maryland, to discuss a proposed modification to the Millstone, Unit 2, Spent Fuel Storage Pool (SFSP).

The request for amendment relating to this issue was submitted by letter dated May 14, 1993.

This meeting was for the purpose of further explaining the modification and identifying any immediate questions concerning the proposed change.

The modification is only in Region C where fuel assemblies meeting the burnup requirements of the Technical Specifications for this region are stored in a configuration that provides cell blocking devices in every fourth storage cell. The modification would remove the cell blocking devices to make room for additional storage capacity. To handle the criticality aspects such that the k effective would be less than.95 for any possible loading of fuel assemblies, the licensee has proposed to limit the enrichment and the minimum burnup of fuel assemblies to be stored in the region and/or providing additional poison in each assembly, as required. The additional poison would consist of three 304 stainless steel borated pins (Rodlets) in each fuel assembly to be placed in the opposite corners and the center of the control rod guide tubes of each fuel assembly, as required.

The attendance list is provided in Enclosure 1. provides the agenda and copies of the viewgraphs supporting the licensee's presentation.

The licensee provided a video that illustrated the testing conducted to verify the installation and removal of the Rodlets.

DISCUSSION The licensee indicated the need for the amendment is to provide the capability to fully off load the reactor core following the next refueling outage. The prior modification decreased the storage capacity in Region B by 40 fuel assemblies. That modification was made to correct for an error found in the original criticality calculations for storing fresh fuel in Region B.

Because of this decrease in storage capacity, the next refueling would be the last time that the reactor core could be fully off loaded. The addition of e\\

234 cells would provide full core off loading capability through 1998.

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a s June 1, 1993 A 1988 amendmeint'provided the licensee the capability to store 688 consolidated fuel boxes in Region C, even in cells with cell blocking devices.

If the licensee chooses to use the fuel consolidated method to extend the date at which time full core off loading capability would be reached, it could impact on their refueling schedule and on the possible capability of being accepted by the DOE fpr ultimate storage of spent fuel. Thus, the licensee chose the proposed method with rodlets placed in appropriate fuel assemblies as the method of extending the date at which time full core off loading would be reached.

The licensee indicated a need for the issuance of the proposed amendment to be approximately the middle of November 1993. This would allow for completing rodlet installation, removal of Region C cell blockers, moving fuel from other regions to Region C, receiving new fuel assemblies and early start for refueling and maintenance outage in May 1994.

The proposed modification utilizes the CASMO-3 computer code augmented by the KEN 0 V.a Monte Carlo Criticality Program in the analysis of the criticality aspects of the modification. This is different than the original method used by Combustion Engineering and, therefore, should not have the errors that were identified by the CE method.

In addition, Region C does not have Boraflex or any other poison and, therefore, would not involve criticality calculations involving Boraflex where the CE errors were discovered. The benchmarking for these computer codes that was identified for the application for Amendment 158, which applied to fuel assemblies in cells with Boraflex panels, is also applicable for the application of these codes to Region C where rodlets are used for poison.

Several cases were calculated including fuel assemblies with rodlets, fuel assemblies without rodlets, the areas adjacent to Regions A & 3, and the areas adjacent to the elevator. The staff indicated that it intended to audit or rerun the criticality calculations and to do this it would need the quantitative parameters necessary to provide that capability. The staff also indicated a need for the Holtech report on the criticality analysis.

Amendment 128 provided for the capability of storing consolidated fuel which consists of the storing of the fuel rods of two fuel assemblies into one box that could be placed in a cell. The licensee indicated that the seismic analysis for the fuel assemblies with rodlets was bounded by the seismic analysis for the consolidated fuel boxes.

Each rodlet has a buoyant weight of-22 pounds. The staff indicated that we would need the seismic analysis for the condition of stored fuel assemblies with rodlets or at least the bounding analysis with explanations. Also needed was the material properties of the i

'rodlets to verify their structural adequacy, protection from corrosion and l

deterioration and poison capability.

The rodlets could not be inadvertently removed from the fuel assemblies once inserted because it would require the unique tool to remove them. They would be secured by the weight of the rodlets (22 pounds buoyant weight).

For surveillance purposes, the rodlets can be seen in position from above. Also, the licensee maintains surveillance through procedure controls. At infrequent j

intervals, samples of radlets would be removed for inspection.

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4-June 1, 1993 CONCLUSION h~

Since there was some lack of quantitative information in the May' 14, 1993, application for amendment, it is anticipated that there would be a number of -

requests for additional information. Also, for this request to stand alone for completeness, it was explained that requests and' responses should be formally submitted t.

'he record. No attempt will be'made to consolidate the 1

requests for additional information as identified by the individual review branches but will be forwarded to the licensee as soon as received 1from the reviewers on an individual basis. The staff should have their first round of questions prepared within the next 2 weeks. We may delay.the issuance of the Federal Reaister notice until the responses to the bulk of the RAls are received.

Original signed by:

Guy S. Vissing, Senior Project Manager Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

1.. Meeting Attendance List 2.

Copies of Viewgraphs of t

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une 1, 1993 CONCLUSION Since there was some lack of quantitative information in the May 14, 1993, application for amendment, it is anticipated that there would be a number. of requests for additional information. Also, for this request to stand alone for completeness, it was explained that requests and responses should be formally submitted for the record. No attempt will be made to consolidate'the t

requests for additional information as identified by the individual review branches but will be forwarded to the licensee as soon as received from the reviewers on an individual basis. The staff should have their first round of questions prepared within the next 2 weeks. We may delay the issuance of the Federal Reaister notice until the responses to the bulk of the RAIs are received.

Guy S. Vissing, Senior Project Manager Project Directorate I-4 i

Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

1.

Meeting Attendance List 2.

Copies of Viewgraphs of Licensee Presentation i

cc w/ enclosures:

See next page 9

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4 Millstone Nuclear Power Station Unit 2 CC:

Gerald Garfield, Esquire R. M. Kacich', Director Day, Berry and Howard Nuclear Licensing Counselors at Law Northeast Utilities-Service Company City Place Post Office Box 270 Hartford, Connecticut'06103-3499 Hartford, Connecticut 06141-0270 W. D. Romberg, Vice President J. P. Stetz, Vice President Nuclear Operations Services Haddam Neck.Plhnt Northeast Utilities Service Company Connecticut Yankee Atomic Power Company Post Office Box 270 362 Injun Hollow Road Hartford, Connecticut 06141-0270 East Hampton, Connecticut 06424-3099 Kevin McCarthy, Director Regional Administrator Radiation Control Unit Region I Department of Environmental Protection U.S. Nuclear Regulatory Commission State Office Building 475 Allendale Road Hartford, Connecticut 06106 King of Prussia, Pennsylvania 19406 Allan Johanson, Assistant Director First Selectmen Office of Policy and Development Town of Waterford Policy Development and Planning Division Hall of Records 80 Washington Street 200 Boston Post Road Hartford, Connecticut 06106 Waterford, Connecticut.06385 S. E. Scace, Vice President P. D. Swetland, Resident Inspector Millstone Nuclear Power Station Millstone Nuclear Power Station Northeast Nuclear Energy Company c/o U.S. Nuclear Regulatory Commission Post Office Box 128 Post Office Box 513 Waterford, Connecticut 06385 Niantic, Connecticut 06357 J. S. Keenan, Nuclear Unit Director Charles Brinkman, Manager Millstone Unit No. 2 Washington N.uclear Operations Northeast Nuclear Energy Company ABB Combustion Engineering Post Office Box 128 Nuclear Power Waterford, Connecticut 06385 12300 Twinbrook Pkwy, Suite 330 Rockville, Maryland 20852 Nicholas S. Reynolds Winston & Strawn G. H. Bouchard, Director 1400 L Strett, NW Nuclear Quality Services Washington, 9C 20005-3502 Northeast Utilities Service Company Post Office Box 270 Hartford, Connecticut 06141-0270 John F. Opeka i

Executive Vice President, Nuclear Connecticut Yankee Atomic Power Company 4

Northeast Nuclear Energy Company i

Post Office Box 270 i

Hartford, Connecticut 06141-0270

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May 20. 1993 Proposed Soent Fuel Storaae Pool Meetina With NNECO Millstone 2 l

HAME ORGANIZATION Guy Vissing DRPE/NRR/NRC

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Larry Wagner SRXB/NRR/NRC John Minns DRSS/NRR/NRC Ann Hodgdon

-0GC/NRC Kris Parczewski DE/NRR/NRC Hans Ashar DE/NRR/NRC Ed Reis OGC/NRC~

John Stolz DRPE/NRR/NRC D

David Repka Winston & Strawn/ Northeast-Utilities George Betancouct NU/ Nuclear Fuel Joe Parillo NU/ Nuclear' Analysis..

Robert Borchert NNEC0/ Millstone 2 Reactor Engineer-John Riley

NNEC0/ Millstone 2 Ei.gineer i

Steven-Unikewicz NU/ Project Services'- MP2 Stan. Turner Holtec -International-

-Rihal Young

-NU Nuclear Licensing 1

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Spent Fuel Pool Rodlet Modification j

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Presentation to:

1 U. S. Nuclear Regulatory Commission Staff

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May 20,1993,10:30 am One White Flint North Rockville, MD

rJs A-Attendees J. W. Riley Engineering Manager NNECO Unit No. 2 R. A. Borchert Reactor Engineer-NNECO Unit No.- 2 ~

1 J. J. Parillo

. Sr. Engineer, Nuclear Analysis NUSCO-l G. N. Betancourt Sr. Engineer, Reactor Performance.

NUSCO 3

S. M. Unikewicz Project Engineer, Project Services NUSCO R. H. Young, Jr.

Sr. Licensing Engineer NUSCO

- j S. T. Turner Chief Physicist.

. Holtec, Inc.

_j D. A. Repka Counsel

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Meeting Overview H. Young o

Technical Introduction J. Riley L o

Criticality and Design Analysis J. Parillo o

Mechanical Design, Analysis, and Tooling G. Betancourt t

o implementation B. Borchert i

o Requested Communication Protocol

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Conclusion J. Riley.

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Meeting Overview i

1 Present comprehensive SFP Rodlet Technical Information o

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Address all NRC Staff Technical questions i

Resolve NRC Staff Questions to the maximum extent possible o

Justify NNECO proposed No Significant Hazards Consideration (NSHC) o determination 1

i Leave the NRC Staff with a through understanding of all facets of this o

proposal

' Request a Communication Protocol to resolve technical questions and-o facilitate the review process l

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Technical Introduction-o Millstone Unit No. 2 is currently in Cycle 12.

Unit No. 2 will lose full core core offload capability following Cycle 13 o

startup (estimated to be August 1994).

.i NNECO wishes to maintain full core offload reserve capacity (as with all NU o

units).

l This License Amendment will reclaim 234 storage locations, resulting in 3 o

more cycles (6 years) of full core offload capacity.

.. i NNECO has evaluated this proposal and determined it to be SAFE.

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o NNECO needs this Amendment as an ' interim measure', since Consolidation / Compaction has been demonstrated, but not yet uplto full.'

production capability.

y NU has concern regarding ' Department of Energy acceptance.of o

consolidated fuel.

NNECO requests Fall 1993 approval to implement before the 1994 outage.

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NNECO proposes a NSHC Determination !

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Criticality and Design Analysis Purpose of Proposed Technical Specification Change

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Changes to Allow Removal of 234 Region C Cell Blockers o

o Criticality Analysis and Safety Analysis 1

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The purpose of the proposed Technical Specification change is to authorize o

the additional storage of 234 fuel assemblies in the Millstone Unit No. 2 Spent Fuel Pool.

Review current Millstone 2 Unit No. Spent Fuel Pool layout.

o The proposed Technical Specification change will remove 234 cell blocking o

devices from Region C, allowing 234 existing fuel storage locations to be utilized.

The proposed Technical Specification changes are needed to compensate o

for the removal of the 234 cell blocking devices by maintaining the required shutdown margin in the Spent Fuel Pool and ensuring that radiological dose consequences of postulated accidents do not increase.

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Changes to Allow Removal of 234 Region C Cell Blockers Must maintain s; 0.95 Ker, under normal and accident conditions in the spent fuel o

pool Approximately 9% reactivity is to be held down by poison to allow limiting fuel o

assemblies to be stored in Region C Negative reactivity insertion will be accomplished by using solid stainless steel rods o

with the following characteristics:

2 weight percent (w/o) natural boron in stainless steel 0.87" OD 152.5" long Weigh 26 pounds each Solid with no welds Machined at top to add a small countert> ore for handling and machined at bottom to make a rounded bottom 2400 total poison pins o

For comparative purposes:

0.87" OD stainless steel pin 0.95" max OD of control rod finger 0.966" min ID of guide tube in dashpot region 1.035" nominal ID of guide tube over most of fuellength Each fuel assembly will use 3 poison pins to obtain the required reactivity effect.

o The 3 poison pins must be placed in the center guide tube and any 2 diagonally opposite guide tubes.

There is no requirement for orientation of poison pins from one fuel _ assembly to o

adjacent fuel assemblies. Any fuel assembly orientation is permissible in the spent fuel pool, as long as the 3 poison pins have been correctly installed in the correct guide tubes. This has been verified by the criticality analysis.

The poison pins are sized lengthwise to ensure they do not protrude above the top o

of the fuel assembly guide tube. This ensures no possibility of inadvertent removal.

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o The poison pins axially completely shadow the top of active fuel height. The poison pins do not necessarily shadow all of the fuel at the bottom. The criticality analysis bounds this by assuming that a conservative 3" of the active fuel height at the bottorp is unshadowed by the poison rods.

This was found to be acceptable.

o New Technical Specification Figure 3.9-1 B shows required burnup vs. enrichment for fuel with poison, pins.

o New Technical Specification Figure 3.9-1 A shows required,burnup vs. enrichment for fuel with no poison pins.

Much larger bumups are required for fuel assemblies to qualify for storage in o

Region C with no poison pins.

These unpoisoned fuel assemblies provide locations which allow for storage of spare /old control rods, plugs, etc.

Compare relationship of existing burnup vs. enrichment curve for 3 out of 4 Region o

C storage, to the 2 proposed bumup vs. enrichment curves for 4 out of 4 Region C storags.

Approximately 16% of the fuel in the spent fuel pool qualifies for storage in Region o

C with no poison pins.

Summary: To allow the removal of the 234 cell blockers:

Most fuel assemblies will need poison pins, as well as larger burnups than currently required to qualify for storage in Region C. Proposed new Technical-Specification Figure 3.9-1B shows required burnup vs. enrichment for fuel with poison pins.

Some fuel assemblies will not need poison pins, but will qualify for storage in Region C by having much larger bumups than currently required. Proposed new Technical Specification Figure 3.9-1 A shows required burnup vs. enrichment for fuel without poison pins.

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Criticality Analysis and Safety Analysis The criticality-analysis verified that s 0.95Kg,, is maintained under normal and o

accident conditions in the Spent Fuel Pool.

These calculations use the CASMO-3 computer code augmented by the three o

dimensional NITAWL-KENO Sa Model with the 27 group SCALE cross-section set.

Benchmarks for these codes were transmitted as part of license amendment #158:

application.

Criticality Analysis - Normal Conditions o

Under normal conditions in the Spent Fuel Pool, the criticality analysis takes no credit for boron in the water. The criticality analysis also use the most reactive fuel design that is present in the Spent Fuel Pool, o

The results of the criticality analysis shows Kerr =.9402 for Region C poisoned fuel and Kg,, =.9459 for Region C fuel with no poison. These values include all uncertainities and tolerances. The proposed Technical Specification burnup vs.

enrichment curves correspond to these Kep, values.

i The existing Technical Specifications allow consolidated fuel in Region C. The o

criticality analysis verified that this remains acceptable.

Region C interfaces with Region A and Region B. The criticality analysis verified o

that this interface was not adversely affected by the changes in Region C.

The criticality analysic confirms that the Spent Fuel Pool K,, does not increase t

o with a fresh 4.5 w/o U-235 fuel assembly in the new fuel elevator with Region C fully loaded with limiting fuel.

The criticality analysis confirms that rotation of poisoned fuel assemblies did not o

alter the Spent Fuel Pool K,, within the statistical accuracy of the KENO t

calculations.

i 19

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Table 1 Spmary of Criticality Safety Calculations for Alternative Storage Arrangements in Millstone Unit 2 Item Reaion C Reaion C (With Poison Pins)

(No Poison Pins)

Calculated K, 0.9252 0.9243 Temperature 150*F 150*F Calculational Method CASMO-3 CASMO-3 Bias 0.0000 0.0000 1

Uncertainties:

Uncertainty in Bias 1 0.0024 1 0.0024 Rodlet Diameter 1 0.0012 NA Rodlet B-10 loading 1 0.0008 NA Enrichment t 0.0035 1 0.0054 00, Density-t 0.0054 2 0.0039 l

Lattice Spacing 1 0.0052 1 0.0065 SS Wall Thickness 1 0.0024 1 0.0025 Uncertainty in Depletion 1 0.0119 1 0.0175 Calculations"8 Statistical Average 1 0.0150 1 0.0201 Axial Burnup Distribution"'

O.0 1 0.0015 Calculated Reactivity 0.9252 0.9258 1 0.0150 1 0.0201 Maximum reactivity 0.9402 0.9459 (1)

Evaluated for 3% enriched Westinghouse fuel at 25 MWD /KgU for Region C with poison pins and 35 MWD /KgU for Region C with no poison pins. Other enrichments and burnups evaluated for appropriate values, all yielding the same maximum reactivity (K,).

20 i

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l The criticality analysis confirms that the reactivity of Region C fuel assemblies with o

and without poison pins is nearly the same, therefore, they may be interchanged or intermixed in any desired configuration in Region C.

Criticality Analysis-Accident Conditions Criticality accident conditions considered:

o The misloading or dropping of a fresh 4.5 w/o U 235 fuel assembly into Region C, surrounded by limiting Region C fuel.

Dropped fuel assembly on top of spent fuel racks. -

Omitting the poison rodlets from 1 fuel assembly, otherwise qualified for Region C.

Spent Fuel Pool temperatures greater than 150*F.

Cask drop event into Region C.

The limiting event by far was found to be the misloading or dropping of a fresh 4.5 w/o U-235 fuel assembly into Region C, surrounded by limiting Region C fuel.

With credit for soluble boron in the water, Kep, was verified to be less than 0.95.

The Millstone Unit No. 2 Technical Specifications currently require 800 ppm boron in the Spent Fuel Pool water when fuel is moved, or a cask is on the refueling floor. The Millstone Unit No. 2 Technical Specifications require Spent Fuel Pool temperature to be limited to 140*F.

Use of 150*F is conservative, as the breakpoint between normal and accident conditions.

Safety Analysis (FSAR)

There are 2 events considered in the FSAR that involve the Spent Fuel Pool, the o

dropped fuel assembly event and the cask drop event.

The radiological consequences of the dropped fuel assembly event are unchanged, o

since the number of fuel pin failures is independent of whether the fuel assembly does or does not have poison pins. If the number of failed fuel pins does not change, the dose consequences are unchanged.

To ensure that the radiological consequences of the cask drop event are bounded o

by the current analysis, the decay time requirement of fuel within the target area of the cask drop has been increased from 120 days to 1 year.

21

e

.d Criticality and Safety Analysis Summary f

. l t

I The criticality analysis shows that Km is less than 0.95 under normal and accident conditions, and the radiological dose consequences of this' j

proposed Technical Specification change are bounded by the current FSAR safety analysis.

+

' k i

W 5

4 b

t 22

e Mechanical Design, Analysis, and Tooling

~~

o Commercial Nuclear Applications of Borated Stainless Steel (B-SS) o Qualification Standard for Borated Stainless Steel o

Mechanical Design Attrib'utes o

Engineering Anaiysis, Evaluations, and Determinations o

Manufacture and Test Requirements r

o Material Properties o

Surveillance Inspection E

o Tooling Overview o

Security Considerations o

Mechanical Summary b

i i

t I

]

a Commercial Nuclear Applications of B-SS Domestic indian Point Unit No. 3 Docket 50-286 (3/22/78) 837 cells o

i t

o indian Point Unit No. 2 Docket 50 247 (1/11/82).

1248 cells i

Foreign o

30 Nuclear Plants in 7 countries Germany, Austria, Finland, Hungary, Spain, Brazil, Korea o

o 1977 through 1992 ~37,000 cells

> 2,000 Tons Future Fuel separator baskets in cask technology for Utilities, Vendors, and.

o Department of Energy Spent Fuel Pool regionalized configuration o

l 1

u 1

24'

a a

4 O'

Qualification Stancard for B-SS t

Electric Power Research Institute (EPRI) o Initiates Study RP 2813 1988_

Material, NDE, Testing, and Standards o

Published Report TR 100784 1992 American Society for Testing and Materials (ASTM) 1988 o

Specification Approved ASTM A 887-88 (currently 89) 3 Standard Specification for Borated Stainless Steel Plate, Sheet, and Strip for Nuclear Applications American Society of Mechanical Engineers (ASME) 1993 o

ASME Code Case N 510 Structural Application of Borated Stainless Steel pending Board Approval for Certification i

T 25

j f

Mechanical Design Attributes t

'i i

B-SS Rodlet Mechanical Dealgn configuration F

Similar in Shape, Size, and Weight to a Control Element Assembly (finger) o l

r Comparison i

B-SS Rodlet QEA l

Length 152.5 inches 150.75 inches j

Diameter 0.870 inches 0.948 inches Tip Spherical Spherical-Weight 75-78 lbs. (3 rodlets) 80-90 lbs. (CEA assembly)._

.i l

.j t

i t

i Note:

1 Mechanical design of B-SS Rodlet requires Specialized Tootina for installation and.

Removal

.l 2

.i l

'l 26 1

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RODLET END CONFIGURATION AND DESIGN ORAVN BY: JJG 05/18/93 DVG FILE NAME: ROOLET CADO GRIO SIZE:

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O

9 4

Engineering Analysis, Evaluations, and Determinations Design Basis Accidents considered in Final Safety Analysis Report (FSAR) o Fuel Drop Accident o

Cask Drop Accident Seismic / Structural Qualifications l

o Fuel Assembly I

o Fuel Rack o

Fuel Pool Structure o

Fuel / Rack / Pool Interface l

l o

Normal / Abnormal Storage Conditions l

o Drop Qualifications I

o Heavy Loads Thermal Hydraulles / Pool Cooling Quai?lcations o

Fuel Assembly o

Fuel Rack i

o Fuel Pool Engineering Determinations Design Basis Accidents are unaffected by the B-SS Rodlets o

Structural aspects, Drop Qualifications, and Heavy Loads are unaffected by o

virtue of the weight of the rodlet; No Safety Significance (continuos to

' bounded by License Amendment 128)

Thermal Hydraulic aspects are unaffected due to design configuration of the o

rodlet and Pool Cooling Qualification (continues to be bounded by License Amendment 128) 28

2' Manufacture and Test 1 Requirements Northeast Utilities SP-ME 864, Revision 01 4

Poisoned (Borated Stainless Steel) Rodlets for Spent Fuel Storage QA Category 1.

l Manufacture Tests / Inspections j

Specified Boron Content

- Design Dimensions Dispersion of Boron

- Rodlet End Configuration B,o vs B,3 Split -

4

- Boron Requirements Continuity of Stainless Steel

- Quantity of Rodlets ASTM A 887-89 Standard Specification for Borated Stainless Steel Plate, Sheet, and Strip o

for Nuclear Applications Chemical Composition and Test / Inspection Requirements o

ASTM A 484-91 Standard Specification for General Requirements for Stainless Steel and o

Heat Resisting Bars, Billets, and Forgings o

Manufacture and Test / Inspection Requirements

- 29L

7 Material Properties

- (

Reference:

EPRLReport TR 100784).

Mechanical Attributes

- Ultimate Strength-(Tensile)-

- Yield Strength---(Tensile)

- Elongation------- (Tensile)

- Ductility

- Bend Test-------- (Fracture toughness ASTM E 399, E 813) -

- Impact Test

- Hardness

- Specific Gravity

- Specific Heat

- Thermal Conductivity

- Thermal Expansion

- Modulus of Elasticity

- Density Corrosion Resistance

- A 262 B Sulfuric Acid

- A 262 C Nitric Acid

- A 262 E Intergranular

- Nacl Spray

- 2,000 ppm H B0 (Immersion, Crevice, Purged, Galvanic) -

3 3

Radiation Resistance

- Neutron Fluence (1x10 neutrons /cm')-

- Gamma Fluence (does not alter material propeities)-

Material Determination Mechanical: All attributes are armp+ahle for structural application Corrosion:

No susceptibility to corrosion in spent fuel pool environment Radiation:

No change in properties that affect structural application Note: NNECO application of Borated Stainless Steelis non-structural as

s:

i; i

Surveillance Inspection t

1 Conalatent with the guidance provided in:

j NRC Regulatory Guide 1.13

" Spent Fuel Storage Facility Design Basis" i

Appendix A Nuclear Criticality Safety Paragraph 5(c)

Use of Neutron Absorbers in Storage Rack Design '

q 1978 NRC "Overall Technical' Position for Review and Acceptance of Spent Fuel

[

Storage and Handling Applications" l

Section ill Nuclear and Thermal Hydraulic Considerations ~-

Paragraph 1.1 (e) Normal Storage

[

Paragraph 1.5 (I)

Accc-*ance Criteria for Criticality 1

.i I

' NNECO prefers to avoid Technical Specification provisions for these:

1.

Initially confirm the presence of the neutron abisorber in the rodlet via the tests and inspections associated with the QA Category _I manufacturing s

program (CMTR).

Initially confirm the presence of the rodlets in tho' designated. fuel 2.

assemblies via tho' strict adherence to the Unit. administrative control procedures and the double verification and documentation of the installation.

i 3.

Institute a material surveillance of 1% of the rodlets.st 5 year intervals to:

perform a visual inspection for material degradation. Criteria will address j

provisions for responding to any material abnormalities encountered.-

' 31

1 x

p^

j 4

1 Installation and: Removal: Tooling General Design d'rit'erla i

1 Feature Purpose Simple Ease of operation Fail Safe Industrial and Nuclear Safety Reliable Cyclic operation for > 2,000 installations i

Lightweight No Heavy Loads and operator safety Portable Security (access to tools controlled)'

Installation Tool (all manufactured components Stainless Steel)

]

Conical wedge that actuates axially within a double actin 0' pneumatic cylinder (with) in-line compression spring)_ to drive _ 3' hardened stainlessLsteel balls radially outward to engage the counterbored tooling hole in the rodlet top.

q Safety Features: The~ compression. spring provides a mechanical backup.to maintain lift capacity in event of loss of air; Design Safety Factor of 5:1; Capable of engagement in 0.010" overbore; Control. alve handle protected under guard.'

v Inspection and Testing: Test gages provide for performance chechs; Functional' l

and dimensional inspection of critical attributes; 125 lb. load test of compression l

spring; 180 lb. load test of tool in normal operation; functional load test of all-i mock up components 3/18/93; Full scale "in pool" test of all tooling components

(

using full size B-SS rodlets in the SFP 4/8/93 performed under Automated Work Order with Safety Evaluation per 10CFR50.59.

J installation Funnel (Stainless Steel)

Stainless Steel design adapted from lower section of CEA handling tool. it is 1

~

designed to. maintain alignment on the storage rack walls and key into the fuel-

]

assembly upper end fitting for a centered flush fit in any installed orientation. Only

1 3 diagonal opening exist on the funnel top, positively. ensuring proper. rodlet j

orientation.

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Security Consicerations

~.

Borated Stainless Steel Rodlets are SECURE to maintain Region C sub crhical 1 They cannot be inadvertently or accidently removed because:

1.

Removal, like installation, is administratively contro!!ed under a unit quality procedure.

2.

Once installed, the B-SS Rodlets are " captured and contained" inside

(~ 1 inch below) the guide tube and therefore not subject to being

" snagged" during routine SFP activity.

3.

Special tooling is required to remove them.

4.

Special tooling is unique to only this application.

5.

Special tooling is controlled by the Unit Reactor Engineer.

6.

Storage requirements for tooling will mandate restricted access.

Note:

These criteria are consistent with those which were previously determined to be acceptable by the NRC Staff Safety Evaluation, Docket 50-285, Ucense Amendment 133, dated October 12,1990.

Mechanical Evaluation Summary Engineering has determined that the non-structural application of the Borated Stainless Steel Rodlets in the Millstone Unit No. 2 Region C Spent Fuel is mechanically Safe and Acceptable.

39

i

..J Implementation 3

4 o

Existing Fuel Assembly qualification

-r i

o Proposed Fuel Assembly qualification i

i s

o Rodlet installation Scenario i

s

}

?

t t

t

. i t

.i 9

V

.i

+

e Region C Technical Specification Requirements l

~-

Existing Requirernents All fuel assemblies'to be placed in Region C must be verified to be within the a.

enrichment and burn-up limits of Figure 3.9-1 by checking the assembly's design and burn-up documentation.

b.

The contents of each consolidated fuel storage box to be placed in Region C must -

be verified to be within the enrichment and burn-up limits of Figure 3.9-3 by checking the design and burn-up documentation for the storage box contents.

Proposed Requirements a.

All fuel arsemblies to be placed in Region C must satisfy either:

(1)

Fuel assembly enrichment and burn-up are within the limits of Figure 3.9-1 A by checking the assembly's design and burn-up documentation, or (2)

Fuel assembly enrichment and burn-up are within the limits of Figure 3.9-1 B by checking the assembly's design and burn-up documentation, AND borated stainless steel poison pins are installed in the assembly's center guide tube and in two diagonally opposite guide tubes.

b.

The contents of each consolidated fuel storage box to be placed in Region C must be verified to be within the enrichment and burn-up limits of Figure 3.9-3 by checking the design and burnup documentation for the storage box contents.

i f

41

4 e

Existing Region C-Qualification Process FA discharged to SFP Region A or B j

U~

Obtain FA initial enrichment i

Key y

FA = Fuel Assembly SFP = Spent Fuel Pool B rn4sp Obtain FA measured B/U

,"R V

Calculate allowable B/U based on initial enrichment V

i is measured N

FA is.N.QI qualified for Reg'on C allowable B/U

?

J Y

U FA is qualified for Region C v

RE performs dual verification of j

qualification V

i Visually verify FA SerialNumber U

U Place qualified FA END into Region C

(

y i

4P_ -

a e

e Region C. Qualification Process FA discharged to SFP g

Region A or B

.j 1I Obtain FA initial enrichment Key-II FA = Fuel Assembly SFP = Spent Fuel Pool Obtain FA measured B/U B/U = Bum up FIE = Reactor Engineer If Calculate allowabte B/U Note: Calculate allowable B/U based on initial enrichment EabQulpoison pins (Figure 3.9-1 A)

If Is sured Calculate allowable B/U Note: Calcutate allowable B/U m

based on initial enrichment -

& poison pins anowable B/U7 Wure 3.MB)

(Fig. 3.9-1 A)

II Y

lT is rneasured b.

B/U >

FAis qualified for Region C allowable B/U?

without poison pins (Fig. 3.9-1 B) 37 If RE performs dualverification of FAis qualified for Region C with m

qualification poison pins y

1 I Visually verify FA FAis blQIqualified for Region C Serial Number t

I install poison pins as required and i

perform dual verification of installation i

1 I g

r i

Placequalified FA END 1

i into Region C L

J j

43

l Rodlet Installation Scenario 1

)

Under existing Technical Specifications:

1.

Move all fuel assemblies that meet the requirement for storage in Region C without poison pins to the 4 rack modules at the East end of the SFP.

/

2.

Move all fuel assemblies that meet the requirement.for storage in l

Region C with poison pins to the remainder of Region C.

3.

Remove all fuel assemblies from Region C that will ngt qualify for ' storage in Region C under the proposed Technical Speedcation. These fuel i

assemblies will be placed into Regions A or B storage racks.

ll 4.

Insert poison pins into Region C fuel assemblies that require poison pins.

under Plant Design Change Record, per internal Safety Evaluation and in j

accordance with 10CFR50.59.

i l

After proposed SFP Technical Specifications are approved and issued:

{u j

5.

Remove all cell blocking devices from Region C q

i i

s Key implementation Point:

i

~

Coordinate the approval / issuance date of the License Amendmert subsequent to'

.i completion of steps 1 through 4 above.

.j i

?l

%i

.o

ni a

/

e Requested Communication Protocol j

~

Notify Staff Project Manager of questions during Staff technical review o

/

o Staff PM notify NNECO Ucensing Engineer via telephone 9

o NNECO will respond as quickly as possible

?

i o

Telephone conference resolutions (RAI as necessary)

Requested approval / issue date in mid-November 1993 ( ~ 6 months) o l

'f i

1 Coordinate the exact issue date as NNECO proceeds with implementation ~

o iI i

)

P I

I s

1-

- a Conclusion Proposed Uconse Amendment is Necessary, as presented...

o o

NNECO Analyses has determined it to be Safe...

o Proposed NSHC determination o

Request Staff issuance mid-November 1993 We will respond, as necessary... A S A P 1 o

s i

4 4

4(o -

f. O j,

Significant Target Milestones May 1993:

License Amendment submitted to NRC June 1993:

Material Release for rodlets October 1993:

PDCR approved; Early Start rodlet installation mid-November:

Requested Ucense Amendment approval November 1993:

All material onsite; continue rodlet installation December 1993:

Complete rodlet installation January 1994:

Romove Region C cell blockers l

February 1994:

Move fuel in Region C; prepare for outage March 1994:

Receive New Fuel Assemblies May 1994:

Early Start for Refuel and Maintenance Outage August 1994:

Early Finish for Refuel and Maintenance Outage l

r.

}cr s'

DISTRIBUTION:

Docket File NRC & Local PDRs PDI-4 Memo File TMurley/FMiraglia JPartlow SVarga

1 JCalvo GVissing SNorris OGC EJordan LKopp NWagner JMinns AHodgdon KParczewski HAshar EReis

~

JRajan GBidinger JStolz ACRS (10)

VMcCree LDoerflein, RGI T

y F'

I

.i

..

v t e Letter Sequence Meeting
TAC:M86361, (Approved, Closed)
Initiation Request, Request, Request Acceptance... Administration Meeting Questions 27 May 1993 2 June 1993 15 June 1993 Answers Results Approval Other: B14688, Forwards Nonproprietary & Proprietary Info Extracted from Seismic & Structural Calculations Performed by C-E to Support 930514 LAR Re Spent Fuel Pool Mods,In Response to NRC 931026 Rai.Proprietary Info Withheld (Ref 10CFR2.790), ML20045D404
MONTHYEAR1993-06-10 [Table View]

ML20044H105

Text

SUBJECT:

SUMMARY

Enclosures:

Enclosures:

Reference:

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