ML20149J936

Forwards Suppl Info for Review of Exemption Request from Requirements of 10CFR70.24(a), Criticality Accident Requirements, Submitted on 961122

ML20149J936
Person / Time
Site:	Browns Ferry
Issue date:	07/21/1997
From:	Abney T TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M97301, TAC-M97302, TAC-M97303, NUDOCS 9707290157
Download: ML20149J936 (9)

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1 Tennessee Valley Authority, Post Office Box 2000. Decatur, Alabama 35609-2000 l July 21, 1997 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 - REQUEST FOR ADDITIONAL INFORMATION REGARDING EXEMPTION FROM THE REQUIREMENTS OF 10 CFR 70.24 (TAC NOS. M97301, M97302, AND M97303)

This letter provides supplemental information for the review of i the exemption request from the requirements of 10 CFR 70.24 (a),  !

, " Criticality Accident Requirements", which was submitted l November 22, 1996. The NRC request for additional information was dated July 7, 1997.

Specific exemptions from Section 70.24 were previously granted  ;

in the construction phase special nuclear material (SNM) )

licenses for each unit (SNM-1268, SNM-1434, SNM-1511). These i exemptions were, however, not carried forth when the Part 50 l operating licenses were issued. As discussed in the November 22, 1996 submittal, TVA previously requested an exemption from ,

the subject criticality monitoring requirements in a letter from l R. L. Gridley, TVA, to NRC dated August 31, 1987. TVA was later 9707290157 DR 970721 ' 6T3G)

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U.S. Nuclear Regulatory Commission Pa'ge 2 July 21, 1997 notified by NRC (reference: letter from R. A. Hermann, NRC, to S. A. White, TVA, dated May 11, 1988) that an exemption request was not required. Without prejudicing TVA's position regarding the need for a specific exemption to 10 CFR 70.24, an exemption request was submitted in the referenced November 22, 1996, letter.

10 CFR 70.24(d) anticipates that licensees may request relief l from the requirements of Section 70.24, in whole or in part, if good cause is shown. 10 CFR 70.24(c) states that holders of Part 50 operating licenses are exempt from 10 CFR 70.24 (b) provisions. Therefore, only an exemption to Section 70.24 (a) is being requested.

Responses to the seven NRC requested items are provided in the Enclosure. TVA continues to assert the requested exemption is appropriate for the same reasons as for the exemption granted in the original SNM licenses and as justified in the November 22, 1996 submittal. An accident criticality monitoring system was not and is not necessary at BFN Units 1, 2, and 3.

There are no commitments in this letter. If you have further questions, please contact me at (205) 729-2636.

Sdncerely, A

' eh Manager of censing and I ustry Affai s Subscr ed and swornjto before me on this lot d jaf 6ulu 1997.

Q1 atAca .

CLys &

Notary Public My Commission Expires My Co.MistJon Expires 1WOf/da Enclosure cc: see page 3

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I U.S. Nuclear Regulatory Commission Page 3 July 21, 1997 1

i Enclosure cc (Enclosure):

Chairman Limestone County Commission )

310 West Washington Street i Athens, Alabama 35611 l

l Mr. Mark S. Lesser, Branch Chief i U.S. Nuclear Regulatory Commission l Region II Atlanta Federal Center  :

601 Forsyth St., Suite 23T85 l Atlanta, Georgia 30303 j 1

NRC Resident-Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr. Joseph F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Fike Rockville, Maryland 20852 Dr. Donald E. Williamson State Health Officer Alabama State Department of Public Health 434 Monroe Street Montgomery, Alabama 36130-3017 i

ENCLOSURE i

TENNESSEE VALLEY AUTHORITY I BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 EXEMPTION REQUEST FROM 70.24 (a)

CRITICALITY ACCIDENT MONITORING REQUIREMENTS I

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -

DATED JULY 7, 1997

'Below are responses to the seven NRC criteria provided in the j subject request.for additional information (RAI) on the TVA  !

exemption request from the requirements of 10 CFR 70.24 (a), l Criticality Accident Requirements". TVA's request was initially j submitted November 22, 1996. The NRC RAI was dated July 7, j 19.4 / . In the RAI, NRC requested that TVA verify the following i criteria were met. I NRC Item 1 Plant. procedures do not permit more than three boiling water reactor fuel assemblies to be in storage or transit between their shipping cask or storage rack at one time.

i TVA Response l

. Plant procedures adequately address the storage and transit of new reactor fuel bundles. A summary description of new fuel i handling activities and-related procedural controls is provided  ;

I below, i

New fuel bundles are transported and received in NRC approved packaging (commonly referred to as shipping containers).

Package design for the shipping containers ensures that a

j. geometrical criticality safe configuration is maintained during transport, handling, and storage.

New fuel shipments are stored'on the refuel floor in the approved shipping containers until.the bundles are inspected and placed in the spent fuel pool. Inspection involves removing-t individual fuel bundles from the shipping container, placement

( in the new fuel inspection stand and inspection, installation of

! fuel channels, and then storage in the spent fuel pool pending i use in the reactor. Handling of both new fuel and irradiated l fuel is carefully controlled by site fuel handling procedures.

Strict limits are established for the maximum number of fuel bundles allowed out of approved storage locations at any given i.

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time as' delineated in Site Standard Practice ( S S P ) - 12 .12, " Fuel Receipt, Storage, and Use". SSP-12.12 also identifies approved storage areas for new and irradiated fuel.

Specific limitations re.;arding the number of new fuel assemblies allowed out of storage specified in SSP-12.12 are as follows.

There are two approved new fuel handling areas located on the refuel floor used for processing of new fuel for the Units 1 and 2 fuel pools, and the Unit 3 fuel pool respectively. These two fuel handling areas are separated by over 100 feet. Within these areas, no more than two fuel bundles are allowed out of approved storage locations at any given time. Therefore, at most, four fresh bundles can be out of approved storage i locations simultaneously (if fuel receipt / inspection activities were in progress on two units). We believe this procedural limitation is consistent with the NRC criteria of no more than 1 three bundles out of storage locations since the fuel handling areas are physically separated.

In the individual fuel pools and reactor cavities, no more than three fuel bundles are allowed out of storage locations.

SSP-12.12 provides that a specific evaluation must be performed if more than three bundles are desired to be out of storage locations, and for the removal and storage of individual fuel rods from fuel bundles. These provisions are also considered consistent with the NRC criteria.

NRC Item 2 The requirement is met that k-effective not exceed 0.95, at a 95% probability, 95% confidence level with the fresh fuel storage racks filled with fuel of the maximum U-235 enrichment and flooded with pure water.

TVA Response The maximum design basis k-effective ( ko r r ) for the new fuel storage racks is 0.95 (flooded condition) as specified in Chapter 10.2 of the Updated Final Safety Analysis Report (UFSAR) and Section 5.5.A of the BFN Technical Specifications (TS).

The new fuel storage racks at BEN were designed by General Electric (GE). Subcriticality is ensured by not allowing the k-infinity of any bundle designed by GE to have a k-infinity l larger than that used in the analysis of record as defined in General Electric Standard Application for Reactor Fuel (GESTAR),

NEDE-24011-P-A. Additional detail on this methodology is E-2 I

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provide'd in Section 3.5 of NEDE-24011-P-A-13 and is consistent with NRC Criterion 2. l NRC Item 3 l The requirement is met that k-effective not exceed 0.98, at a 95% probability, 95% confidence level with the fresh fuel i storage racks filled with fuel of the maximum U-235 enrichment l and flooded with moderator at the (low) density corresponding I to optimum moderation. l TVA Response As noted above, the current BFN licensing basis for the new fuel storage racks is described in Chapter 10.2 of the UFSAR and Section 5.5.A of the BFN TS, and does not include requirements for a hypothesized optimum moderator configuration analysis.

Likewise, TVA has not determined whether the conditions necessary to create such a configuration are credible at BEN.

Therefore, TVA has not previously attempted to demonstrate conformance with this particular criteria. This general issue has, however, been previously considered in GE Service Instruction Letter (SIL) 152, Criticality Margins for Storage of New Fuel, which provided a number of recommendations to further l reduce the remote probability of a criticality occurrence i associated with an optimum moderator configuration.

As discussed in the November 22, 1996 exemption request, the new fuel storage area is not currently in use at BEN since the shipping containers provide a more convenient means for temporary storage prior to inspection and placement in the spent fuel pool.

Also, direct placement in the fuel pool reduces the number of fuel moves that would be required if the new fuel storage racks were used.

l However, to preserve flexibility for future activities, the new fuel storage racks are included in the scope of the exemption request. Prior to these racks being utilized for storage of new fuel, it will be necessary to revise SSP-12.12 to reinstate the new fuel racks as a storage location. In support of the procedure revision, TVA is agreeable to evaluating the optimum moderator hypothesis as a prerequisite for use of the new fuel racks for storage. This evaluation would consist of an analysis of an optimum moderator configuration, or implementation of administrative or physical barriers similar to those recommended in SIL 152. A change to the BFN UFSAR is being processed to track this action prior to using the new fuel racks. Since, E-3

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however, use of the new fuel racks for storage is considered a contingency option and there is no current need for using this area, TVA does not intend to pur.tue an optimum moderator evaluation at this time. l NRC Item 4 )

The requirement is met that the k-effective not exceed 0.95, at a 95% probability, 95% confidence level with the spent fuel storage racks filled with fuel of the maximum U-235 enrichment and flooded with pure water.

TVA Response The design basis for the spent fuel storage racks is described in Chapter 10.3 UFSAR and section 5.5.B of the BFN TS. As indicated in Section 10.3.5.1, the maximum allowed design k a for the spent fuel pool is 0.95.

The spent fuel storage racks at BFN were designed by GE.

Subcriticality is ensured by not allowing the k-infinity of any bundle designed by GE to have a k-infinity larger than that used l in the analysis of record as defined in GESTAR. Additional l detail on this methodology is provided in Section 3.5 of l

NEDE-24011-P-A-13 and is consistent with NRC Criterion 4.

NRC Item 5 The quantities and forms of special nuclear material, other than nuclear fuel, such as sources and detectors, that are stored i onsite in one area, is less than that necessary for a critical I mass. )

TVA Response A summary of the special nuclear material (SNM) inventory being stored onsite, other than nuclear fuel, was provided in the November 22, 1996, submittal and is summarized below. l 1

The largest single amount of non-fuel SNM stored in the same l area is in the form of six Fuel Loading Chambers (FLCs) which each contain approximately 2 grams of U-235. The quantity of SNM specified to be enough for a critical mass in Section 1.1 of Regulatory Guide 10.3, " Guide for the Preparation of Applications for Special Nuclear Material Licenses of Less than Critical Mass Quantities", is 350 grams of U-235, 200 grams of U-233, and 200 grams of Pu-239. Clearly, the quantity of SNM E-4 i

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in the'FLCs is far below the amounts for which criticality monitoring would be of concern. FLCs are only needed for neutron monitoring during core reloads following lengthy I refueling outages and thus are seldom needed. The six FLCS in storage are kept for this contingency and TVA has no need to possess more than this number.

The total quantity of SNM (U-235) in the form of new and used incore detectors is very small and is currently less than 0.5 grams. BEN also has several sources containing very small amounts (~ 0.2 grams) of plutonium-239. Thus, the net total amount of non-fuel SNM, including the FLCs, is far below the 2

Regulatory Guide 10.3 values for that necessary for achieving a critical mass. The geometry of the SNM forms (small quantities in multiple individual detectors) is also not conducive to support the formation of a critical configuration.

1 NRC Item 6 Radiation monitors, as required by General Design Criterion 63, are provided in fuel storage and handling areas to detect  ;

excessive radiation levels and to initiate appropriate safety l actions.

1 T/A Response Area radiation monitors (ARMS) are located on each unit in the vicinity of the new fuel storage vault, spent fuel storage pool, new fuel handling area of the refueling floor, and near the new fuel storage vault at the next lower elevation. ARMS are permanently installed gamma sensitive radiation monitors designed to detect abnormal radiation levels in plant areas where radioactive material may be present, stored, or handled. These monitors have local area alarm and control room annunciation capability. The alarm function serves to warn operating personnel of equipment malfunctions causing increased radiation levels, and also serves to provide a general radiation hazard warning to plant personnel if abnormal radiation levels occur in the plant area. With regard to personnel safety, the function of the ARM system is included in general employee training, and employees are instructed to immediately vacate the vicinity upon ARM alarms. A description of the operating characteristics of the ARMS is provided in Section 7.33 of the UFSAR, and the specific locations of the ARMS, including the subject ARMS mentioned above, is provided in Table 7.13-2 of the UFSAR. TVA considers that these radiation monitors satisfy the objectives of the GDC at BFN in providing a means to detect excessive radiation E-5

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actions.

Additionally, as described in Section 7.12.5 of the UFSAR, the radiation level at each spent fuel pool is monitored by redundant reactor building zone ventilation monitors that provide both a ventilation isolation function and a control room annunciation in l

the event of abnormal radiation levels. For personnel safety, it is also routine radiological control practice to provide portable radiation monitors for all work activities that involve the potential for high exposures. For instance, remote radiation monitors with local alarm capabilities are mounted on the refueling bridge during_ fuel transfer activities to provide an  ;

early warning to workers in the event of an unexpected problem i that could result in increased radiation levels. Similar steps I are taken for work in the spent fuel pool or reactor cavity involving irradiated fuel or components, or for other potentially ,

hazardous activities as deemed appropriate by good radiological I controls practices.

NRC Item 7 The maximum nominal enrichment is 5 wt%.

1 TVA Response Currently, there is no specific fuel enrichment limitation on fuel bundles' applicable to the new fuel racks or spent fuel racks at BFN. Rather, the criticality limits are ensured by applying design criteria on the calculated k-infinity of  ;

individual fuel bundles as discussed in the responses to items 2 and 4, above.

At present, General Electric supplies all fuel for BEN. The GE fabrication license for special nuclear material limits them to supplying fuel assemblies containing no more than 5 wt% U-235.

Therefore, the maximum nominal enrichment of new fuel assemblies at BFN is less than 5 wt% uranium-235.

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