ML19263C993

Decides to Pursue Reracking of Spent Fuel Storage Pool Instead of Mod Described in 780504 Submittal.Forwards Proposed Design Criteria for Reracking W/High Density Spent Fuel Racks & Requests NRC Comments by 790416

ML19263C993
Person / Time
Site:	Indian Point
Issue date:	02/27/1979
From:	William Cahill CONSOLIDATED EDISON CO. OF NEW YORK, INC.
To:	Stello V Office of Nuclear Reactor Regulation
References
FOIA-80-143 NUDOCS 7903200405
Download: ML19263C993 (10)

Text

• T William J. Cahill, Jr.

^*

, g ju hO My

• • V'"  ;) ,

'/ i p!l r, 'I I /.

Cor:sohdated E atson Con pany of New Ycrk lx 4 Irurg Place. New York, N Y 10003 /.v/

f */

Te;ephone (212) 400-3819 Tf February 27, 1979 / -

Re: Indian Point Unit No. 2 M Docket No. 50-247 Director of Nuclear Reactor Regulation ATTN: Mr. Victor Stello, Jr., Director Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Stello:

As discussed with your staff, we have decided to pursue reracking the Indian Point Unit No. 2 spent fuel stmrage pool instead of the modifications described in our May 4, 1978 submittal.

The Attachment to this letter presents proposed design criteria for reracking the Indian Point Unit No. 2 spent fuel pool with high density spent fuel racks. The present schedule is based on submittal to the NRC of a detailed design report by September, 1979. Con Edison requests that the NRC comment on the desian criteria presented in the Attachment by April'16, 1979.

Should you or your staff have any questions, please contact us.

Very truly yours, c l J cl ,

William J. C hill, Jr.

attach. Vice President

\

7903200906

s ATTACH?tE!!T PROPOSED DESIGN CRITERIA FOR RPRACKI?!G INDIAN POIIIT UNIT NO. 2 SPENT FUEL POOL Consolidated Fdison Company of New York, Inc.

• Indian Point Unit No. 2 Docket No. 50-247 February, 1979 ,

., ~.

• Table of Contents

1. Criticality Analysis 1.1 Criticality criteria 1.2 Calculational Methods

2. Thermal - Hydraulic Analysis 2.1 Spent Fuel Cooling Capability 2.2 Spent Fuel Assembly I! eat Renoval Capability

3. Mechanical Analysis 3.1 Design Criteria 3.2 Methods of Analysis -

4. Spent Fuel Storage Racks

5. Codes, Standards and Design Criteria I

1. ,

Criticality Analysis 1.1 Criticality Criteria The spent fuel storace racks vill be designed to limit Keff toi0.95. The criticality calculations will be based on a maxinum U-235 enrichment of 3.5 w/o. Credit will not be taken for horated water in the spent fuel pool during normal conditions.

1.2 Calc lational Methods Appropriate computer codes, will be used to obtain broad neutron cross sections for input into a diffusion theory, transport theory or Monte Carlo code for criticality calculations. If a dif fusion code is used in tile cal-culations, it will be benchmarked to a transport theory and/or Monte Carlo code.

1

2. Thermal - 11ydraulic Analysis 2.1 Spent Fuel Cooling Capability The adequacy of the spent fuel cooling systen will be determined. The impact of the proposed modification on the cooling capacity of the spent fuel cooling systen will be evaluated for spent fuel discharges to the spent fuel pool fron nornal refuelina operations and full core discharge. The decay heat, loads will be calculater". usina computer codes such as ORIGTli or by other accepted methods.

2.2 Spent Puel Assembly Feat Removal Capability The proposed spent fuel racks will be designed with adequate flow paths between the racks, pool walls, etc., to ensure natural enermal circulation of water in the spent fuel pool.

2

3. Mecnanical Aaalysis 3.1 Design Criteria Tne spent fuel racks will be designed to resist combined loadings of tne cead weight of the cells, the weight of the spent fuel assemblies and seisnic loads. The spent fuel racks will be designed to be Seismic Category 'I structures in accordance with Regulatory Guide 1.29. The spent fuel racks will be designed'in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section III or AISC Specification for Design, Fabrication anc Urection of Structural Steel for buildings, as des-cribed in the NRC Guidance document entitled " Review and Acceptance of Spent Fuel Storage and Handling Applications" (sent to all power reactor licensees by NRC letter dated April 14, 1978 and modified by URC letter dated January 18, 1979). In addition, the spent fuel pool structure will be evaluated to determine maximum stresses in the pool walls and floor.

3.2 netnod of Analysis A dynamic analysis of the spent fuel storage racks will be performed. All appreciable loads, including seismic will be considered. Stresses will be evaluated against code allowables for both normal load conditions & accident conditions. Load combinations presented in Section 3.8.4 of the Standard Rcview Plan will be used in the analysis.

3

4. Spent Fuel Storage Racks The spent fuel assemolies will be stored vertically in a more

. closely spaced lattice than tae presently installed racks.

It is estimatad tnat the spent fuel storage capacity will be increased to at least 840 assembly storage locations (depending on final selection of poison, material and vendor) .

The racks will be manufactured of type 304 stainless steel and will rest on the stainless steel liner plate at cne bottom of the spent fuel storage pool.

The design and fabrication of the spent fuel storage racks shall comply witn:

(1) The requirements pertaining to fuel storage racks contained in ANSI M18.2.

(2) The requirements pertaining to spent fuel storage racks contained in ANSI N210.

(3) The requirements contained in 10CFREJ Appendix A General Design Cri?.erion 62.

(4) Tne structural design snall be in accordance with S RP 3. 8. 4 (see Section 5 Codes, Standards and Design Criteria).

(5) Spent fuel storage racks will be designed as a Seismic Category I structure as designated in Regulatory Guide 1.29.

4

~ . -

5. Codes, Standards and Design criteria The abbreviations listed below have the follouina

• meanings:

AISC - Dnerican Institute of Steel Construction AISI - American Iron and Steel Institute AMS - American Nuclear Society ANSI - Anerican I'ational Standards Institute ASME - 3merican Society of Mechanical Engineers ASMT - American Society for Mondestructive "esting ASTM - American Society for "esting and Materials AWS - American Felding Society CFR - Code of Federal Regulations R.G. - Regulatory Guide of the USNEC SRP - Standard Review Plans of the USNRC SSPC - Steel Structures Painting Council USNRC- United States Muclear Regulatory Corr 1ssion When and if any of the follouine codes, tecnnical or trade standards, reculations and soecifications, such as AISC, ANSI, ASME, AS T:'. , SRD, AFS, USNRC neculatory Guides, Federal Specifications shall be enployed durinc this pronosed project the latest edition and latest addenda thereto, to the extent applicable to this projecy shall be used:

AISC Specification for the Desian, Fabrication, and Prection of Structural Steel for Buildings 5

w AISC Code of Standard Practice for Steel Buildings and Bridges

. ANS 5-1 Proposed ANS Standard Decay Heat Re-lease Following Shutdown of Uranium Fueled Thermal Reactors ANSI N16.1 Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors ANSI N18.2 Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants and Supplement M18.2A-1975 ANSI N4 5. 2. -197 1: QA program requirements for Nuclear Power Plants ANSI N45.2.1-1973: Cleaning of Fluid Systems and Associated Components During the Construction Phase

, of Nuclear Power Plants ANSI N45.2.9 Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants ANSI N210 Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants ANSI N45.2.12 Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants 6

ANSI N45.2.13 Quality Assurance Requirements for Control of Procurement of Equipment, Materials and Services for Nuclear Power Plants ASME B&PV Code Section II, Material Specification, Parts A, B and C and all addenda ASME B&PV Code Section V, Nondestructive Examination and all addenda ASME B&PV Code Section VIII, Division 1, Pressure Vessels and all addenda ASNT (SNT-TC-1A) Recommended Practice for Nondestruc-tive Testing, Personnel Qualification and Certification ASTM-A262 Recommended Practices for Detecting Susceptibility to Intergranular At-tack in Stainless Steel AWS D.l.1 Structural Welding Code USNRC Guidance document entitled " Review and Acceptance of Spent Fuel Storace and Handling Applications " (sent to all power reactor licensees by USNRC letter dated April 14, 1978 and modified by USNRC letter dated January 18, 1979).

ASME B&PV Code Section IX, Welding and Brazing Qualifications and all addenda.

7

. ~ _ . _ -

SSPC-SP-10 Surface Preparation Specification, Near-White Blast Cleaning SSPC-Vis-1 Surface Preparation Specification, Pictorial Surface Preparation Stan-dards for Painting Steel Structures R.G. 1.13 Spent Fuel Storage Facilities Design Basis R.G. 1.25 Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors R.G. 1.29 Seismic Design Classification R.G. 1.31 Control of Stainless Steel Welding R.G. 1.48 Design Limits ,tnd Loading Combination for Seismic Category I Fluid Systens Components R.G. 1.50 Control Preheat Temperature for Weld-ing of Low Alloy Steel R.G. 1.70 Standard Format and Content of Safety Analysis Reports for Muclear Power Plants R.G. 1.71 Welder Qualification for Areas of Limited Accessibility R.G. 1.92 Combining Model Responses and Spatial Components in Seismic Response Analy-sis 8

R.G. 3.15 Standard Format and Content of License Applications for Storage Only of Un-

, irradiated Reactor Fuel and Associated Radioactive Material R.G. 3.4 Nuclear Criticality Safety in Operation with Fissionable Materials Outside Reactors R.G. 8.8 Information Relevent to Maintaining Occupational Radiation Exposure as i.ow as Practicable (Nuclear Reactors)

SRP 3.7.2 Seismic System Analysis SRP 3.8.4 Other Category I Structures SRP 9.1.2 Spent Fuel Storage SRP 9.1.3 Spent Fuel Pool Cooling and Cleanup System 10CFR50, General Design Criterion 62, Preven-Appendix A tion of Criticality in Fuel Storage

. and Handling 9