ML072750397

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Summary of Meeting with Public Service of Colorado (PSC) Regarding a License Application for Dry Spent Fuel Storage Under 10 CFR Part 72 at PSCs Ft. St. Vrain Nuclear Power Plant Site
ML072750397
Person / Time
Site: Fort Saint Vrain, 07200009  Xcel Energy icon.png
Issue date: 10/23/1989
From: Roberts J
NRC/NMSS/IMNS
To: Rouse L
NRC/NMSS/IMNS
References
NUDOCS 8911060023
Download: ML072750397 (58)
v • d • e


Text

QQC 2 3 1989 Docket No. 72-9 MEMORANDUM FOR:

Leland C. Rouse, Chlef Fuel Cycle Safety Branch Division of Industrial and Medical Nuclear Safety FROM:

John P. Roberts, Section Leader Irradiated Fuel Section Fuel Cycle Safety Branch Division of Industrial and Medical Nuclear Safety

SUBJECT:

MEETING WITH PUBLIC SERVICE OF COLORADO (PSC)

DATE/TIME:

October 13, 1989; 8:30 a.m.

LOCATION:

Room 14B-11, One White Flint North Building, Rockville, MD ATTENDEES:

See enclosure 1 PURPOSE:

To discuss a license application for dry spent fuel storage under 10 CFR Part 72 at PSC's Ft. St. Vrain nuclear power plant site.

DISCUSSION:

PSC introduced the topics to be covered (see enclosure 2) and described the situation at Ft. St. Vrain, which is to be decommissioned, and the proposed location for a modular vault dry store (MVDS) independent spent fuel storage installation (ISFSI) onsite.

FW Energy Applications, Inc., in concert with General Electric Company of the United Kingdom (GEC) have received NRC staff approval for a topical report (TR) for an MVDS design to store light water reactor spent fuel (Project No. M-46).

GEC described the modifications for the Ft. St. Vrain site to the design of the MVDS to accommodate storage of high temperature gas reactor (HTGR) spent fuel.

The proposed ISFSI capacity may be as much as 1482 HTGR spent fuel blocks. Six vault modules would be required to store the fuel and associated material (neutron sources and reflector blocks).

PSC then discussed its proposed schedule (see enclosure 3).

Under Part 72, Section 72.40(b), construction at site before a finding that issuance of the proposed license with any appropriate conditions to protect environmental values is the action called for may be grounds for denial of a license.

Thus, construction at the site is not appropriate before requirements of the National Environmental Policy Act are met (i.e, as implemented through NRC regulations.)

89116O0023 891023 PDR ADOCK 05000267 Y

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. NRC regulations and guidance were discussed with PSC a license application (see enclosure 4). NMSS staff proposed change in 10 CFR Part 51, Section 51.23 (54 1989).

outlining its approach to noted the Commission's FR 39765, September 28, PSC proposed another meeting to discuss their proposed application.

Participants agreed to meet on December 8, 1989, at the Nuclear Regulatory Commission headquarters. The meeting ended.

Original Signed by John P. Roberts John P. Roberts, Section Leader Irradiated Fuel Section Fuel Cycle Safety Branch Division of Industrial and Medical Nuclear Safety cc: K. Heitner w/o enclosures

Enclosures:

1) Attendance List
2) Ft. St. Vrain and MVDS Descriptions 33 ISFS'I Project Summary Schedule 4 Applicable Regulatory Guidance Used for ISFSI Distribution:

NRC File! Center Docket 72-9 (50-26,71 PDR LPDR NMSS R/F IMIF R/F IMSB R/F JRoberts LCRouse FBrown Beveridgie/Cornell 1-23 FSturz JSchneider SRuffin KCLeu JStokley (JR/MEMO TO LELAND ROUSE)

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NAME :J bert :,C DATE :10/)S/89

1/0A0/89 OFFICIAL RECORD COPY

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~wi OCT 2 31 za S PSC/NRC MEETING October 13, 1989 NAME Organization Ken Heitner NRC/NRR Mike Niehoff PSC Chris Ealing GECA/ESL B. K. Agarwal FWEA R. E. Farrell NRC/SRI Fritz Sturz NRC/NMSS John P. Roberts NRC/NMSS T. F. Westerman NRC/RIV Peter B. Erickson NRC/NRR John Holt GEC. A. ESL Mick Blackbourn GEC. A. ESL Charles F. Nash FWEA Mike Lehr PSC Erian Dyck PSC Mike Holmes PSC Norman Wagner NRC/NRR Steve Ruffin NRC/NMSS

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  • Fuel Containerisation at Reactor
  • Uncontaminated Fuel Route (Normal Operations)
  • HTR Fuel
  • Storage of Greater than Class C Wastes
  • Small Storage Capacity
  • Total Independence from on-site Reactor
  • Abillity to handle individual Fuel Assemblies during off-normal events

Mdodular Vault Dry Store GENERAL ARRANGEMENT

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  • Civil Structure excluding steelwork roof structure
  • Fuel Storage Container
  • FSC Handling Machine
  • Charge Face Structure eStorage Wells

MVDS Fort St Vrain Outline Operating Sequence l Container Loading

  • Receive Transfer Cask at Transfer Bay
  • Raise Transfer Cask into MVDS
  • Position Transfer Cask into Load/Unload Port
  • Remove Transfer C:ask Closure
  • Fix Container Shield Plug Handling Adaptor
  • Position Load/Unk)ad Port Isolation Valve (I.V.)
  • Remove Container Shield Plug
  • Position Contaner Handling Machine on l.V.
  • Raise Container into Handling Machine
  • Position Container Handling Machine on Charge Face lIV.
  • Lower Container into Storage Vault
  • Remove Container Handling Machine (empty)
  • Replace Charge Face Shield Plug

I MVDS Fort St Vrain Outline Operating Sequence

Preparation for Container Loading
  • Position Isolation Valve (I.V.) on Charge Face
  • Remove Charge Face Shield Plug
  • Position Container Handling Machine on I.V.
  • Raise empty Container into Handling Machine
  • Move Container Handling Machine To Transfer Cask Load/Unload Port
  • Lower empty.Container into Transfer Cask
  • Remove Container' Handling Machine
  • Position Container' Shield plug into Container Lid
  • Remove Load/Unload Port l.V.
  • Remove Container Shield Plug Handling Adaptor
  • Replace Transfer C~ask Closure
  • Raise Transfer Casck from Load/Unload Port
  • Position Transfer Cask on Trailer

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[-Design Parameters

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Parameter FSV TSAR Heat Load/Storage Location Decay Period Fuel Source -

Normal Assembly Gamma/Storage Location Neutron/Storage Location Fuel Source - Neutron Source Assembly Gamma/Storage Location Neutron/Storage Location Ambient Temperatures Flood Levels Seismic Ground Acceleration Spectrum Tornado Missile Tornado Maximum Velocity Snow Loading 0.7 kw (average)

A^^n 140 day 1.52 x 1015 MeV/sec 2.08 x 106 N/sec 1.52 x 1015 MeV/sec 5.61 x 1010 N/sec

-32'F to 102'F 6ft 0.1 Reg. Guide 1.61 Site specific 300 mph 30 psf 1kw 5 year 9.32 x 10'5 MeV/sec 5.17 x 106 N/sec it

-20' to 100 F Site specific 0.25 Reg. Guide 1.61 NUREG 0800 11 Reg. Guide 1.76 (360 mph) 100 psf

MVDS Fort St Vrain Design Codes

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ANS 57.9-1984 for General Codes ASME III - Class ND for Storage Container

£ ACI 349-85N for Civil Structure NUREG 0554-1979 for Container Handling Hoist

V MVDS Fort St. Vrain SAFETY CONSIDERATIONS

  • Fuel Storage Environment
  • Potential Release from Fuel
  • Impact (Dropping) Incidents
  • Fuel Containment
  • Containment Monitoring
  • Sub-Criticality
  • Tornado Missiles
  • Fuel Heat Removal

MVDS Fort St. Vrain SAFETY CONSIDERATIONS Fuel Storage Environment

  • Air storage environment
  • No fuel degradation at storage temperature!
  • Wide margin on allowable fuel temperatures
  • Air storage experience at IDAHO

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MVDS Fort St. Vrain SAFETY CONSIDERATIONS Potential Release from Fuel

  • Failed Fuel Fraction
  • Fission Gas Generation by Hydrolysis of Uranium Carbide
  • Available water limited to atmospherics moisture
  • No particulate release

MVDS Fort St. Vrain SAFETY CONSIDERATIONS Impact (Dropping) Incidents

  • Building Crane not High Integrity No radiological hazard from bounding impacts

- Impact heights limited by design and operating procedures

  • Bounding MVNDS Impacts

- Drop of Fuel Handling Machine onto Charge Face Structure

- Drop of Transfer Cask into Transfer Cask Bay (no impact limiter)

  • Fuel Handling Machine Hoist is High Integrity

- No drop of Fuel Storage Container into vault considered

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MVDS Fort St. Vrain SAFETY CONSIDERATIONS Fuel Containment

  • Single barrier containment
  • Carbon steel with corrosion protection finish
  • Bolted container lid with two metal'0' rings
  • Interspace leak checking capability for seals
  • Provision for leak checking container body

MVDS Fort St. Vrain SAFETY CONSIDERATIONS Containment Monitoring

  • No significant radiological release potential from fuel container
  • No retention of inert storage environment - air storage
  • Thus:

Intermittent sampling of containment leak tightness during storage period

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W MVDS Fort St. Vrain SAFETY CONSIDERATIONS Sub-Criticality

  • Fully flooded vault considered
  • Partial water densities considered
  • No credit for burn-up
  • Special analyst sub-contractor using US - NRC approved codes

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q MVDS Fort St. Vrain SAFETY CON1SIDERATIONS Tornado Missiles

  • Site specific bounding missile
  • Fir plank-12" x 4" x 12'
  • Velocity 151 mnph H < 6' 188 mph 6' 255 mph 50' H <50'

< H < 125'

W MVDS Fort St. Vrain SAFETY CONSIDERATIONS Fuel Heat Removal

  • Passive self-regulating natural thermo-syphon cooling system
  • Insensitive to partial cooling blockages
  • Adequate time to recover from total cooling blockages
  • Insensitive to wind direction and velocity
  • No provision to monitor inlet or outlet cooling air temperature

-a ETF FRI: LATMaP BEn TUFM ISESI a

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RG 3.62 Standard Format and Content For The Safety Analysis Report For Cnsite Storage of S4ent Fuel Storage Casks (February 1989) o RG 4.2 Preparation of EnvizOramntal Reports For Nuclear Pwer Stations (July 1976) o RG 3.50 Guidarme on Preparirg a License Application to Store Spent Fuel in an Imdqperdent Srent Fuel Storage Installation (September 1989) o PG 3.60 esign of an Indreperdent Spent Fuel Storage Installation (Dry Storage)

(Mardh 1987)

FSC's; LlaEma APEP;CATI CONENTS 10CFR License Application License Application Attachkts:

72 Subpart B Safety Analysis Report ISMI Deommissionirq Plan Emerger~ry Plan Envit vntal Report Tchical Specifications QA Program Rrysical Security Plan Design for Physical Patecti=i Sa~feguards onti ry Plan Pers=mel Taining Program 72.24 72.30 72.32 72.34 72.26 72 Subpart G 72.181 72.182 72.183 72 Subpart I o

Licenise Application Attachments will remve applicable information from the 50 Progras/Planas

PSC's ISFSI SAFETY Ale ANAL R

IT' o

Part 72.24 requires that a Safety Analysis Report be sulbitted for an ISFSI and cntains the minkmm, information that must be included in the Safety Analysis Report o

RG 3.62 will be used as a guide in preparing the Safety Analysis Report o

Part 72.18 allows the incorporation of infobiation, by reference, frn doemwnts previously submitted to the NI Final Safety Analysis Report Ervirmntal Report Final Ervirormzntal Statement Topical Safety Analysis Report

V Ak'4 PSC's ISFSI EN1VIRFN RECIR o

Part 72.34 regaires that an Eriviramental Report be submitted that meets the regurit of Subpart A of 10CFR51 o

Part 51.61 requires that an Ernrimetal Report be submitted with the ISFSI License Application o

Part 72.18 allaws the inoozporaticn of infbnuation, by reference, fma documents previously submitted. to the NRC Final Safety Analysis Report Enviramiental Report Final Environmental Statement Tipical Safety Analysis Report ISFSI Safety Analysis Report o

Part 51.23 makes a generic determination of no significant envirnimental impacts followirn cessation of reactor operation o

A Regulatory Guide specifically for an ISFSI Erivirormental Report does not ezdst o

PSC's objective is that an Environcental Assessnmnt of "no significant impact" be issued

S PARE 50 vs RAW 72 FUEL EZNDM o

Part 72 requires the licensee provide for safety and emergency systems and c ilities for manitorirrg, hardlirg, and retrieving the spent fuel under nral accident cm-iditicris.

O Fuel Handling at the Reactor Eluildirg will be performed urder the Part 50 icense with licensed operators o

Fuel Handling at the ISFSI will be perfoz21ed under the Part 72 license without the rirent for licensed cperators o

Part 72.190 requires trained and certified persomel operate equipment important to safety at the ISFSI o

RG 3.50 states that ISMSI cperators are not required to be licensed

Storage of 7he 3tal O aRflectr Blocks an Neuton. Sounres under Part 72 o

PSC is proposin1 that the Metal Clad Reflector Blocks and Neutrc SIroes will be stored under Part 72 as "radioactive material asiated with spent fuel stwages o

Based on activation analysis the Metal Clad Reflector Blocks have been classified as Greater Than Class C Radioactive Waste, per Part 61 o

Currently no repositories are available for Greater Than Class C Waste

- I SP=T KIEL SKYPPfl CLSE o

Prcposed drae to the shitpprg cask innr liner material and inner lid o

Will not result in a reduction in the safety margin for a droped cask in the Reactor Building o

Final Safety Analysis Report analysis for a drcped cask in the Reactor Buildirg takes no credit for the liner as a containnent barrier o

7he transportation route between the Reactor uildin and the ISFSI will not be on any public highways

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