ML20245E125
| ML20245E125 | |
| Person / Time | |
|---|---|
| Issue date: | 12/05/1988 |
| From: | Roberts J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Sakenas C Committee To Review Generic Requirements |
| Shared Package | |
| ML20245D207 | List:
|
| References | |
| FRN-54FR19379, RULE-PR-170, RULE-PR-50, RULE-PR-72 54-FR19379, AC-76-24, AC76-1-24, AC76-24, NUDOCS 8901100033 | |
| Download: ML20245E125 (24) | |
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'9 UNITED STATES i
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555
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-DEC 0 51988 MEMORANDUM FOR:
Cheryl A. Sakenas, Program Manager Committee to Review Generic Requirements FROM:
John P. Roberts, Section Leader Irradiated Fuel Section Fuel Cycle Safety Branch Division of Industrial and Medical Nuclear Safety, NMSS
SUBJECT:
PRESENTATION ON CERTIFICATE OF COMPLIANCE FOR A DRY SPENT FUEL STORAGE CASK I have enclosed twenty (20) copies of a draft Certificate of Compliance for the General Nuclear Systems, Inc., Model No. CASTOR V/21 dry spent fuel storage cask. This wil.1 be presented to CRGR on December 14, 1988.
If you have any questions please telephone me on (49-20608).
/___.
John P. Roberts, Section Leader
" Irradiated Fuel Section Fuel Cycle Safety Branch Division of Industrial and Medical Nuclear Safety, NMSS
Enclosure:
As stated cc:
R. M. Bernero R.'E. Cunningham G. L. Sjoblom B. M. Morris
... R.,Lahs.
W Di(WERiyPearso'n 1
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NRC Form XXX U.S. NUCLEAR REGULATORY COMMISSION
-( -89) 10 CFR 72 CERTIFICATE OF COMPLIANCE FOR ORY SPENT FUEL STORAGE CASKS la. CERTIFICATE NUMBER
- b. REVISION NUMBER
- c. PACKAGE IDENTIFICATION NUMBER 1000 0
USA /72-1000
- d. PAGE NUMBER e. TOTAL NUMBER OF PAGES 1
4
.2.
PREAMBLE This certificate is issued to certify that the cask and contents described in item 5 a) 2) below, meets the applicable safety standards set forth in Title 10, Code of Federal Regulations,-Part 72, " Licensing Requirements for the Independent Storage of Spent Nuclear Fuel _and High-Level Radioactive Waste."
3.
This Certificate is issued on the basis of a safety analysis report of the cask design.
- a. PREPARED BY (Name and Address)
- b. TITLE AND IDENTIFICATION OF REPORT OR APPLICATION General Nuclear Systems, Inc.
Topical Safety Analysis Report 220 Stoneridge Drive for the CASTOR V/21 Cask Columbia, SC 29210 Independent Spent Fuel Storage Installation (Dry Storage) (TSAR)
- c. DOCKET NUMBER 72-1000 4.
CONDITIONS This certificate is conditional upon fulfilling the requirements of 10 CFR 72, as applicable, and the conditions specified below, j
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5.
(a) Cask (1) Model No.:
CASTOR V/21 l
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Page 2 - Certificate No. 1000 - Revision No. 0 - Docket No. 72-1000
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l (2) Description The CASTOR V/21 cask is designed for the storage and shipment-of irradiated spent fuel assemblies.
The cask was designed to meet International Atomic Energy Agency international ~
specifications for Type B(U) packaging corresponding to Nuclear Safety Fissile Class I.
However, this certificate addresses spent fuel handling, transfer, and storage on a NRC-licensed j
nuclear reactor site but does not address any use or certifi-
'j cation of this cask design for offsite transport of spent fuel.
The CASTOR V/21 cask body consists of a thick-walled nodular iron casting.
The overall length is 4,886 mm (192.4 in), and the side wall thickness (without fins) is 379 mm (14.9.in).
The cross-sectional diameter of the cask body, which. weighs approximately 92.3 tonne (101.8 ton), is 2,400 mm (94.5 in).
The cask cavity has a diameter of 1,527 mm (60.1 in) and a length of 4,154 mm (163.5 in).
It holds a fuel basket and is designed to accommodate 21 PWR fuel assemblies..The loaded weight of the cask is about 106 tonne (117 ton).
Four trunnions are bolted on, two at the head end and two at the bottom end of the body.
Gamma and neutron radiation is shielded by the cask iron wall of the cask body.
Also for neutron shielding, two concentric rows of axial holes in the wall of the cask body are filled with polyethylene rods.
The bottom and the secondary cover each have a slab of the same material inserted for the same purpose.
The cask is sealed, to maintain a helium atmosphere, with a multiple-cover system consisting of a primary lid and a secondary lid.
The primary lid is constructed of stainless steel.
The overall thickness is 290 mm (11.4 in).
It is fastened to the body with 44. bolts.
The primary lid has two penetrations, used for flushing and venting of the cask cavity as well as the performance of the leak test.
The flushing and venting connections are sealed with separate lids.
The secondary lid is also made of stainless steel.
The overall thickness is 90 mm (3.5 in).
It is bolted to the body.
A combination of multiple elastomer and metal seals for each lid provide leak tightness.
However, no credit is claimed in the TSAR (see Section 3.3.2.2) or given by NRC for elastomer seals for the 20 year storage period.
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Page 3 - Certificate No. 1000 - Revision No. 0 - Docket No. 72-1000 The fuel basket accepts the spent fuel assemblies and ensures that criticality will not occur.
In addition, it ensures exact positioning of the individual fuel assemblies.
It is of welded construction and is made either of stainless steel or stainless steel and borated stainless steel sections.
At the top end of the cask there is a flushing connection for rinsing, cleaning, and drying of the interior during loading and unloading procedures at the nuclear power plant.
The flushing channel runs inside the wall of the body; it has one end at the top and the other end at the bottom of the inside of the cask.
Gas intake and exhaust are via the valve in the primary lid.
The lid system is fitted with a leak-testing device, a pressure gauge, which is also a cask component classified as important to safety in Section 3.4 of the TSAR.
The gauge monitors the gas pressure in the interlid space between the primary and secondary lids.
This space is used for a gas barrier with an above atmospheric pressure maintained in it.
The inside of the cask, including the sealing surface, has a nickel coating for corrosion protection.
On tne outside, the
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cask is protected by an epoxy resin coating in the fin area and nickel coating elsewhere.
The internal heat-transfer medium is an inert gas (helium), which also serves to inhibit corrosion.
Impact limiters are attached at the top and bottom of the loaded CASTOR V/21 cask when it is transferred at L height greater than 15 inches from the reactor to emplacement on the concrete storage pad at the independent spent fuel storage installation.
One impact limiter desipa is used for both the top and bottom cask limiters.
It cons'sts of a ring of a dozen 9-inch lengths of 6-in diameter Schedule 80 stainless steel pipe contained between half-inch thick stainless steel plates.
A cask drop would crush the impacted pipe lengths between the steel plates reducino the impact load on the cask.
(3) Drawing The Model No. CASTOR V/21 dry spent fuel storage cask is described by drawings in Appendix 1 of the TSAR.
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(4) Basic Components The Basic Components of the Model No. CASTOR V/21 storage cask that are important to safety are listed in Table 3.4-1 of the TSAR.
6.
Cask fabrication activities shall be conducted in accordance with the reviewed and approved quality assurance program submitted with the TSAR.
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Page'4 -' Certificate No.;1000 - Revision No. 0,-_ Docket No. 72-1000'
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Notification of cask fabrication schedules shall be made in accordance
.with.the requirements'of 872.232(c), 10 CFR Part 72.
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Casks of.the Model NO. CASTOR V/21 authorized by this certificate are j
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hearby approved for general use by holders of 10 CFR Part 50' licenses for nucicar reactors at reactor sites under the general license issued
_ pursuant to $72.210, 10 CFR Part 72, subject to the conditions specified by $72.212 and the attached Conditions for Cask Use.
9.
Expiration date
, 2009.
FOR THE NUCLEAR REGULATORY COMMISSION By Lando Zech Chairman i
j REFERENCES Topical Safey Analysis Report for the CASTOR V/21 Cask Independent Spent Fuel l
Storage Installation (Dry' Storage), Revision 2A, June 1987.
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CONDITIONS FOR CASK USE CERTIFICATE OF COMPLIANCE 72-1000 i
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TABLE OF CONTENTS P_ age 1
- 1. 0 INTRODUCTION....................................................-
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1.1 General Conditions..........................................
A-1 1.2 Preoperational Conditions...................................
A-1 2.0 FUNCTIONAL AND OPERATING LIMITS...................................
A-2.
7 2.1 Fuel To Be Stored At ISFSI..................................
A-2 2.2 GNS CASTOR V/21 O ry Storage Cas k............................
A-4 2.3 Limiting Condition - Handling Height........................
A-6 2.4 Ory Storage Cask Surface Contamination......................
A-6 2.5 Dry Storage Cask Internal Cover Gas.........................
A-7 2.6 Limiting Condition - Pressure Switch........................
.A-7
- 3. 0 SURVEILLANCE REQUIREMENTS........................................
A-7 3.1 Cask Seal Testing...........................................
A-9
- 3. 2 Ca s k Co ntami na ti o n..........................................
A-9 3.3 ~0ose Rates..................................................
A-10 3.4 Cask Interlid Pressure (CASTOR V/21)........................
A-10 3.5 A'. arm System.................................................
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- 1. 0 INTRODUCTION These Conditions for Cask Use govern the safety of the receipt, possession and storage of irradiated nuclear fuel at an Independent Spent Fuel Storage Installation and the transfer of such irradiated nuclear fuel to and from a Nuclear Power Station and its Independent Spent Fuel Storage Installation.
1.1 GENERAL CONDITIONS 1.1.1. Operating Procedures i
Written operating procedures shall be prepared for cask handling, movement, i
emplacement, surveillance and maintenance.
1.1.2 Quality Assurance Activities at the ISFSI shall be conducted in accordance with the requirements i
of Appendix B, 10 CFR Part 50, 1.2 PREOPERATIONAL CONDITIONS The user shall not allow the initial loading of spent nuclear fuel in the Model No. CASTOR V/21 cask until such time as the following preoperational license conditions are satisfied:
1.
A training module shall be developed for the Station Training Program establishing an ISFSI Training and Certification Program which will cover tP.e following:
a.
Cask Design (overview) b.
ISFSI Facility Design (overview) c.
ISFSI Safety Analyr,is (overview) d.
.Cuel loading and cask handling procedures and abnormal procedures e.
Certificate of Compliance (overview).
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'A training' exercise (Dry Run) of cask loading.and' handling actidities.
shall be held which shall' include but not be limited to:
n a.
Moving cask in and out of spent fuel pool area.
b.
Loading a' fuel assembly (using dummy assembly).
c.
Cask sealing and cover gas backfilling operations.
d.
Moving cask to and placing it on the storage pad.
e.
Returning the cask to the reactor.
f.
Unloading the cask assuming fuel cladd'ing failure.
g.
Cask decontamination.
2.0 FUNCTIONAL AND OPERATING LIMITS 2.1 FUEL TO BE STORED AT ISFSI 2.1.1 Specification The spent nuclear fuel to be received and stored at the ISFSI in CASTOR V/21 casks shall meet the following requirements:
(1) Only irradiated 14 x 14, 15 x 15 and 17 x 17 PWR fuel assemblies with Zircaloy-fuel rod cladding may be used.
Total assemblies per cask 5, 21.
(2) Maximum initial enrichment shall not exceed 2.2 weight percent U-235 for fuel stored in the stainless steel basket reviewed and found l
acceptable.
Maximum initial enrichment shall not exceed 3.5 weight percent U-235, for fuel stored in the borated stainless steel basket reviewed and found acceptable.
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f(3) Maximum assembly average burnup shal'1 not exceed 35,000 megawatt-days _-
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.per metric ton. uranium.and' specific power shall not exceed 35 KW/KG.
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~(4) Maximum heat generation rate shall not exceed 1 kilowatt per. fuel assembly.
~(5)'; Fuel shall have cooled a minimum of 5 years after reactor discharge "and' prior to storage in the ISFSI.
(6)' Fuel shall be. intact unconsolidated fuel.
Partial fuel assemblies, that is,. fuel assemblies from which fuel pins are missing must"not be
. stored unless' dummy fuel pins are used.to displace an amount of water-equalito that displaced by the orignal pins.
.(7) Fuel assemblies' known or suspected or suspected either to hav'e gross
-cladding defects or to have structural defects'sufficiently severe to adversely affect fuel handling and transfer. capability shall not be loaded-into the_ cask for storage.
(8) A procedure shall be developed'for the documentation'of the character-
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izations. performed to select spent fuel to be stored in the casks.
Such. procedure shall include independent verification of fuel assembly-selection by an individual.other than the original individual making the selection.
.(9) Prior to insertion of a spent fuel assembly into a cask, the identity of the assembly will be verified by an individual other than the one who previously identified the assembly.
2.1.2 Basis The. design criteria and subsequent safety analysis assumed certain characteristics and limitations for the fuels that are to be received and stored.
Specification 2.1.1 assures that these bases remain valid by defining I
the type of spent fuel, maximum initial enrichment, irradiation history, maximum thermal heat generation, and minimum post irradiation cooling time.
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x, 2;2L GNSI' CASTOR V/21 DRY STORAGE CASK 2.2.1 Specification-The'GNSI CASTOR V/21 Dry Storage Casks used to store spent-nuclear fuel at an ISFSI shall.have the operating limits shown in Table 2-1.
2.2.2 Basis-The design criteria and subsequent safety analysis of the GNSI CASTOR V/21 assumed certain characteristics.and operating limits for the use of the casks.
This specification assures that those design criteria are not exceeded.
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s Table 2-1L GNSI CASTOR V/21 OPERATING LIMITS p
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- Operating ' Limit-
[!['Y l Max. Lifting Height with'a Non--
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-Redundant Lifting. Device withLimpact limiters.
5'
' without impact limiters 15"
- Dose - Rate..
' 2 m Distance-
'f 10 mrem /hr
Surface 1 200 mrem /hr o
- Ca'sk. Tightness
- (Standard'He-Leak Rate)_
Primary Lid Seal.
1 10.s mbar1/s
'. Secondary Lid ' Seal-1 10 6 mbar 1/s Max. - Specific Power of One.
1.0 kW.
Fuel Assembly Helium Pressure Limit (Cask Cavity) 800 i 100 mbar Pressure during Cask Drying (Cask Cavity)
-5 3 mbar (holding for 10 min.)
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2.3~ LIMITING CONDITION - HANDLING HEIGHT
- 2. 3.1'.. Specification g-This specification applies to handling of a cask being used for spent fuel storage outside of the Fuel Building and Crane Enclosure Building.
a.
The CASTOR V/21 dry storage cask shall not be handled at a height of greater than 15 inches without an impact limiter.
i b.
With the impact-limiter the CASTOR V/21 dry storage cask shall not.be handled at a height greater than 5 feet.
2.3.2 Basis The drop analyses performed for the CASTOR V/21 dry storage cask requires that an impact limiter be used for postulated cask drop incidents on the ISFSI storage pad for drops greater than 15 inches up to 62 inches without sustaining unacceptable damage to the storage cask and fuel basket.
This limiting condi-tion ensures that the handling height limits will not be exceeded at the storage pad or in transit to and from the reactor.
2.4 DRY STORAGE CASK SURFACE CONTAMINATION 2 4.1 Specification Removable contamination on the dry storage cask shall not exceed 1000 dis / min /
2 from a sources.
100 cm2, from p, y sources and 20 dis / min /100 cm 2.4.2 8 asis i
Compliance with this limit ensures that the decontamination requirements of l
49 CFR 173.443, will be met over the lifetime of the cask in storage.
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2.5 DRY STORAGE CASK INTERNAL COVER GAS E
2.5.1 Specification p
The dry storage casks shall be. backfilled with helium.
2.5.2 Basis The. thermal analysis performed for the dry storage casks. assumes the use' of helium as a cover gas.
In addition, the use of an inert gas (helium) is to
. ensure'long-term. maintenance of fuel clad integrity.
2.6' LIMITING CONDITIONS - PRESSURE SWITCH 2.6.1 Specification The pressure switch ~used to monitor the leak. tightness of the CASTOR _V/21 dry storage cask shall have the performance characteristics shown in Table 3.3-6' of the TSAR.
3.0 SURVEILLANCE REQUIREMENTS
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Requirements for surveillance of-various radiation levels, cask internal
. pressure, contamination levels, cask seal leak rates, and fuel related param-eters are contained in this section.
These requirements are summarized in
. Table 3-1'from details contained in Section 3.1 through 3.6.
Specified time intervals may be adjusted plus or minus 25 percent to accommodate norinal test schedules.
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--LTable 3-l SURVEILLANCE REQUIREMENTS
SUMMARY
!J tSection Quantity or Item
-Period E
3.1.11 Cask? Seal: Testing L-l3.2.11 Cask Contamination cL~
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3.3.1 Dose Rates (Cask surface or'up to 2 meters.
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Dose Rates (Fence)
- Q 3.4.1 Pressure Switch Parameters P&L 3.5.1
- Alarm System' A
'P - Prior to cask loading.
.L - During loading operations:
iQ - Quarterly
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'3.1' CASK SEAL TESTING 3.1.1 Specification
' Prior to storage, the cask must be. properly sealed by testing as specified in
- Section 10.2.2.1 of the TSAR.
'3.1.2 Basis The safety analysis of leak tightness of the cask as discussed in the topical report is based on the seals being leak tight to 10 6 mbar 1/s.
This check is
.done to ensure-compliance with this design criteria.
~3.2 CASK-CONTAMINATION 3.2.1 Specification i
After cask loading and prior to moving the cask to the' storage pad, the cask
- shall be swiped to ensure that removable surface contamination levels are less 2
2 than'1000 dis / min /100 cm, from p, y emitting sources and 20 dis / min /100 cm.
from a emitting sources.
3.2.2 Basis This surveillance requirement will ensure compliance with the decontamination requirements of 49 CFR 173.443 during storage in the ISFSI.
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- 3'.3' DOSE RATES i3.1. Specification
-The following dose rate measurements shall be made for the ISFSI:
. a.
Cask Surface Gamma and Neutron' Dose Rates:
After completion of cask:
loading,' gamma and neutron measurements shall be taken on the outside l..
surface (or within 2 meters of the cask surface).
The combined gamma and neutron dose rates shall be less than the surface dose rate.
stated in Table 2-1 (or the specified rate at a distance of up to 2 meters from the cask surface).
b.
Dry Cask ISFSI Boundary:
Doses shall be determined by measurement.at' the Dry Cask ISFSI site fence and shall be made at least quarterly -
or after any cask-movements to demonstrate compliance with i
920.105(b)(2), 10 CFR Part 20.
.3.3.2' Basis These measurements are necessary to assure compliance with the cask specifications and that the dose rates at the security fence meet Part 20 limits as additional casks are placed in storage.
l 3.4 CASK INTERLID PRESSURE (CASTOR.V/21) l-l 3.4.1 Specification The cask interlid pressure shall be monitored by use of a pre'ssure switch having
-l the characteristics described in Table 3.3-6 of the TSAR.
The switching pressure shall be factory set at 4 bar for the inter 11d space, and a functional test shall be performed during cask preparation.
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- i 3.4.2 Basis A'i LThis' specification requires the interlid space to be maintained to detect any, possible leakage of either cask seal.
3.5 ALARM SYSTEM L3.5.1 Specification An alarm system to which all of the pressure switches are connected shall be-installed at the storage site and functionally tested annually to ensure proper operation of the system.
3.5.2 Basis The alarm system must be capable of alerting. surveillance personnel of-possible cask seal failure and must permit identification of the specific cask indicating a seal failure, i
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'CRGR BRIEFING-DECEMBER 114, 1988 STORAGE OF SPENT NUCLEAR FUEL'IN NRC-APPROVED STORACE CASKS ~-AT NPP SITES RESPONSE TO CRGR RECOMMENDATIONS BASES FORl GENERAL AND SPECIFIC' LICENSING FOR ONSITE.
CASK ~ STORAGE
. REG GUIDES 3,62 AND 3,61 l.-
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RESPONSE TO CRGR RECOMMENDATIONS-
\\4 S 72.216 REPORTING REQUIREMENTS PR MODIFIED TO-REFER TO NEW SS'50'72(B)(2)(VII)(1).
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REFERENCES TO NRC, I.E.' AUTHORITY ASSURING COMPLIANCE RI AND-SITE-RELATED INSPECTIONS DELETED RECORDS ON CASK REPAIRS ELIMINATION-OF DISCUSSION DESCRIBING NRC USE OF LICENSEE NOTICE OF FIRST USE TO ESTABLISH INDEPENDENT RECORD REVISED LANGUAGE IN S72.236 ELIMINATION OF " TECHNICAL SPECIFICATION" TERMINOLOGY CLARIFICATION OF (F), HEAT REMOVAL CAPABILITY W/0 l-ACTIVE COOLING 1.
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.c REG GUIDE 3,61 - STANDARD FORMAT AND CONTENT FOR a
A TSAR FOR A SPENT FUEL DRY STORAGE CASK l
HISTORY INITIATED IN 1983, ISSUED FOR COMMENT 1986 i
GUIDE PROCESSING HALTED IN-1986 PENDING PART 72 REVISION CURRENTLY 4TH PMU AT BRANCH LEVEL - BEING REVISED TO INDICATE THAT ALTERNATIVES EXIST REGARDING NEED FOR SITE-SPECIFIC INFORMATION 1
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g-REG GUIDE 3.6] --CONTENTS-u
. PRINCIPLE DESIGN CRITERIA
. ENVIRONMENTAL' CONDITIONS AND' NATURAL PHENOMENA SAFETY PROTECTION SYSTEMS --
CONFINEMENT, EQUIPMENT
& INSTRUMENTATION,
.j CRITICALITY SAFETY, j
RADIOLOGICAL j
i PROTECTION, FIRE 8 EXPLOSION l
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.-DECOMMISSIONING
" CASK" RELATED' EVALUATIONS AND DESCRIPTIONS-1 STRUCTURAL, THERMAL, SHIELDING, CRITICALITY, CONFINEMENT l-OPERATING PROCEDURES, CONTROLS AND LIMITS l
I RADIATION PROTECTION l
ACCIDENT ANALYSIS QUALITY ASSURANCE
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' REG-GUIDE 3;62 --STANDARD FORMAT AND CONTENT-FOR-THE'SAR FOR'ON-SITE STORAGE 0F SPENT FUEL STORAGE CASKS-SAME HISTORY AS RG 3.61-J-
SPECIFICALLY MAKES POINT THAT MUCH INFORMATION IN REACTOR FSAR AND CASK' TSAR C0NTENTS'OF GUIDE
. TOPICS COVERED IN NPP FSAR AND CASK TSAR 6
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&?$~f of UNITED STATES I
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'g NUCLEAR REGULATORY COMMISSION 4
j WASHINGTON, D. C. 20555 1
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3 January 5, 1989 l
1 MEMORANDUM FOR:
Victor Stello, Jr.
Executive Director for Operations FROM:
Edward L. Jordan, Chairman Committee to Review Generic Requirements
SUBJECT:
MINUTES OF CRGR MEETING NUMBER 152 1
- The Committee to Review Generic Requirements (CRGR) met on Wednesday, December 14, 1988 from 10 a.m.-5 p.m.
A list of attendees for this meeting-is attached (Enclosure 1).
The following items were addressed at the meeting:
1.
L. Shao (NRR) and P. T. Kuo (NRR) presented for CRGR review a draft bulletin related to thermal stresses in pressurizer piping.
The Committee recommended in favor of issuing the proposed bulletin subject to several clarifications (to be coordinated with the CRGR staff).
This matter is discussed in Enclosure 2, 2.
M. Malsch (0GC), S. Crockett (OGC, and J. Wilson (RES) briefed _the Committee on staff actions in resolving public comments on the 10 CFR Part 52 i
rulemaking.
The briefing was focused on the resolution of some of the key issues addressed by public comments and which were previously of concern to the Committee, such as, prototype testing, and the scope and detail of designs for certification.
A copy of the briefing slides used by the staff are included as Enclosure 3.
i The Committee identified the issue of prototype testing as one area which needs additional clarification in the rule.
3.
B. Grimes (NRR), W. Brach (NRR), and H. Clausen (NRR) presented for CRGR consideration a proposed Advance Notice of Proposed Rulemaking (ANPR).
The ANPR solicits public comment on a list of issues related to the procurement of products for use at nuclear power plants.
The Committee had no objection with issuing the ANPR, with only minor clarifications.
A copy of the briefing slides used by the staff are included as Enclosure 4.
4.
W. Houston (RES), W. Beckner (RES), L. Soffer (RES), and J. Ridgely (RES) presented for CRGR review proposed enhancements aimed at improving the severe accident performance of Mark I containments.
The Committee did not complete their review of the Mark 1 package due to some unresolved concerns with several technical issues.
The staff also requested a subsequent review meeting to address the method of implementation.
This matter is discussed in Enclosure 5.
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. 1 5.
B. Morris (RES), W.'Lahs (RES), and J. Roberts (NMSS). presented for CRGR review;a Certificate of Compliance which will be incorporated in'10 CFR Part 72 rulemaking and two implementing Regulatory Guides, 3.6.1, Standard i
Format and Content for a Topical Safety Analysis Report for a Spent Fuel-I Dry Storage Cask, and 3.6.2, Standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks.
The Committee recommended in favor of issuing the proposed rule with the.
incorporated certificates.of compliance.
This matter is discussed in-.
i
'In accordance-with the E00's July 18, 1983 directive concerning " Feedback and Closure of CRGR Reviews," a written response is required from the cognizant office to report agreement'or disagreement with the CRGR recommendations in these minutes.
The response, which is required within five working days after receipt of these minutes, is to be forwarded to the CRGR Chairman and if there is disagreement with CRGR recommendations, to the EDO for decisionmaking.
Questions concerning these minutes should be referred to Cheryl Sakenas-(492-4148).
j briginal 5igne'd IIs E O krdan Edward L. Jordan, Chairman Committee to Review Generic Requirements-
Enclosures:
As stated cc/w enclosures:
Commission (5)
SECY Office Directors Regional Administrators CRGR Members Distribution: w/o enc.
i Central File CRGR SF (w/cnc.)
i PDR (NRC/CRGR)
M. Taylor (w/ enc.)
S. Treby E. Jordan (w/ enc.)
l W. Little J. Heltemes (w/ enc.)
M. Lesar J. Conran (w/ enc.)
l P. Kadambi (w/ enc.)
C. Sakenas (w/ enc.)
,A 0FC
- CRGR: AE00
- DD:AE00
- C:C 6R D:
NAME :CSakenas
- CJHeltemes rd n
.. _ _ _ _.. _ _ _ _............ _......, '../____._......._____...:_............:.__..___.....:.____....
- pf/jj/8)f DATE ~ :12/ /88:jr :12/ /88 i
/
0FFICIAL RECORD COPY
K Attendance List CRGR Meeting No. 152-F
.CRGR Members-J.' Goldberg -
R. Bernero
'D; Ross~.
C. Paperiello.
E.' Jordan 1
F. Miraglia (for J. Sniezek)/J.' Sniezek (partial)'
n NRC Staff
'C.
Sakenas-C.:J. Heltemes'
.g J. Conran-C. Rossi
-l
.L.
Shao-J. Richardson C. Liang "S.-Hou P.'Kadambi 1T.~Chan 1
P. Kuo G.'Mizuno M. Malsch S. Crockett.
J. Wilson b
-D. Scaletti
.s M.-Clausen B. Grimes W. Brach-A. Thadani.
W.'Hodges
. C. - Tinkler -
^L. Soffer W.'Beckner J. Ridgely D. Houston J. Murphy M. Taylor T. Cox l
S..Treby
.J. Telford L; Rouse W. Pearson l
- 8. Morris l
J. Roberts i
]
W. Lahs
.I a
, to the. Minutes of CRGR Meeting No. 152 Regulatory Guides 3,6.1, and 3.6.2, and a Certificate of Compliance Topic B. Morris.(RES), W. Lahs (RES), and J.' Roberts (NMSS) presented.for CRGR review two draft' regulatory guides which address the format and content for safety analysis reports related to:the onsite dry cask storage of spent fuel and the certificate of compliance which will be incorporated in-10 CFR Part 72 rulemaking.
A copy of the slides used by the staff is attached to this enclosure.
Background
The Committee previously met to review the-10 CFR Part 72 rulemaking on November 9, 1988 at CRGR Meeting No. 150.
The Committee requested, at that time, that additional material be supplied for review.
The material supplied included:
1.
Draft Reg. Guide 3.61, Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask.
2.
Draft Reg. Guide 3.62, Standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks.
3.
Draft Certificate of Compliance for the General Nuclear Systems, Inc.
Model No. CASTOR V/21 dry spent fuel storage cask.
Conclusions / Recommendations As a result of their review of this matter, including discussions with the staff at this meeting, the Committee made the following recommendations:
1.
In section 2.0, Functional and Operating Limits, of the Certificate of Compliance several concerns were raised.
a.
2.1.1(7) - clarify what is meant by gross cladding defects to assure j
that licensees understand this requirement.
b.
2.1.1(9) - clarify when the initial identity of the assembly is determined to make it clear that it is immediately prior to cask loading.
2.
Under section 2.4, Dry Storage Cask Surface Contamination, and 3.2, Cask Contamination, the radioactivity limits should be consistent with trans-portation' limits and correct the basis section of 3.2 to be consistent with the language used in section 2.4.
The staff requested waiver of CRGR review of the three other certificates of compliance.
The Committee recommended in favor of forwarding the proposed rule and accompanying certificates of compliance provided the recommended changes are included in all four certificates.
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