ML20097A546

Forwards Rev 2 to 920327 Application for Amends to Licenses DPR-77 & DPR-79,consisting of TS Change 92-01,increasing Spent Fuel Storage Capacity,Per 920511 Discussions W/Nrc

ML20097A546
Person / Time
Site:	Sequoyah
Issue date:	05/28/1992
From:	Burzynski M TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-TVA-SQN-TS-92-01, CON-TVA-SQN-TS-92-1 NUDOCS 9206030020
Download: ML20097A546 (9)

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Terine*see Vailey Authority 1101 Myket Street, Chattarega Tennessee 37402 MAY 281992 TVA-SQN-TS-92-01 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC_ 20555-Gentlemen:

In the Matter of )- Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 l

- SEQUOYAH NUCLEAR PLANT (SQN) - REVISION 2 TO DSQUEST FOR LICENSE AMENDMENT TO TECHNICAL SPECIFICATION (TS) CHANGE 92 SPENT-POOL STORAGE CAPACITY-INCREASE The enclosed pages reflect revisions to Enclosure 3 of the subject licensing amendment request submitted on March 27, 1992. Actual changes

'to each page are reflected by revision bars. Please make the appropriate changes as indicated below. .

1. Page 1: Added discussion to address the storage of additional fuel in the spent-fuel pool in regard to the' potential accident scenarios which were considered.

2. Page 2: Added discussion to-describe the effect upor. the additional

= fuel stored-in the spent-fuel storage racks in the event of a s:ismic event.

.3. Pages 2 and 3: Added discussion to describe the effect of additional fuel stored in the spent-fuel pool would have in the event cooling flow was: lost in the spent-fuel pool.

These revisions were discussed with members of your staff on May 11, 1992, and do not.have a--significant effect-on any previous analysis or

-' calculation performed.

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U.S. Nuclear Regulatory Commission

-Page 2 MAY 281992 Please direct questions concerning this issue to C. R. Davis at (615) 751-7509.

Sincerely,

'[//,

f J. Burzynski Manager Nuclear Licensing and Regu'atory Affairs Enclosure cc (Enclosure):

Mr. D. E. LaBarge, Project Manager U.S. Nuclear Regulatory Commission Onc White Flint North 11555 Rockville Pike Rockville, Maryland 20852 Mr. Michael H. Mobley, Director (w/o Enclosures)

Division of Radiological Health T.E.R.R.A. Building 150 9th Avenue, N Nashville, Tennessee 37203 NRC Resident inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 Mr. B. A. Wilson, Project Chief U.S. Nuclear Regulatory Commiss-ion Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323

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ENCLOSURE.3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET.NOS. 50-327 AND 50-328 (TVA-SQN-TS.92-01)

DETERMINATIOtt OF NO SIGNIFICANT HAZARDS CONSIDERATIONS 4

, Enclosure 3 SIG!J1 FICA!1T llAZARDS INALUATIOli TVA has evaluated the proposed technical specification (TS) changes and has determined that they do not represent a significant hazards consideration based on criteria established in 10 CFR $0.92(c).

Operation of Sequoyah in accordance with the proposed amendment will nota r

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

The following patential scenarios were considered: D

1. A spent.;uel assembly drop.

1 2. Drop of the tJansfer canal gate or the divider gate in the spent-fuel

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a pool.

l 3. A seinmic event.

4. Logie r,f cooling flow in the spent-fuel pool.

i 5. Installation activities.

The effect at additional spent-fuel pool storage cells fully loaded with ,

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[uel on the first four potential accident scenarios listed above has been B

reviewed. ::t was concluded that af ter installation activities have been completed, t.hn presence of additional fuel in the pool does not increase i the probability 7f occurrence of these four events.

With regard t > installation activities, the existing Sequoyah TOs prohibit londu in excess of 2100 pounds from travel over fuel assemblies

( in the storage pool and require the associated crane interlocks and

& physical otops be periodically demonstrated operablo. During I installation, racks and associated handling tools will be moved over the

. npent-fuel pool but movement over fuel will be prohibited. All r

insta11at. ion vork in the spent-fuel-pit area will be controlled and

. performed in u rict accordance with specific written procedures.

IIRC regulationn provide that, in lieu of providing a single fallute-proof crana rystem, the control of hoavy loads guidelines can be satisfied by establishlag that the potential for a heavy load drop is extremely

[ emall. Storage rack movements to be accomplished with the Sequoyah autillary building crane will conform with NUREG-0612 guidelines, in that L the probability of a drop of a storage rack is extremely small. The crano has a tasted capacity of 80 tons. Tho maximum weight of any existing or replacemant storage rack and its associated handling tool is less than 15 tons. Therefore, there is ample safety f actor margin for movemer.tn of the storage racks by the auxiliary building crane. Special l[

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lifting devices, which have redundancy or a reted capacity sufficient te maintain edequat e nafety f actors, will also be utilized in the movements

[i of the stcrege racks. In accordence with NUREC-0612, Appendix D, the safety margir, ensuren that the probability of a load dt, p is ext; moly low.

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. Load travel over fuel stored in the cask loading area of the cask pit will be minimized and, in any case, will be prohibited unless an impact shield, uhich has been specifically designed for this purpose, is covering the area. Loads that are permitted when the shleid is in place must meet analytically determined weight, travel height, and cross-sectional area criteria that preclude penetration of the shleid. A TS has been proposed that incorporates the previously mentioned load criteria.

A nvel movement and rack changeout sequence has teen developed that '

illustrates that it will not be necessary to carry exiating er new racks e over fuel in the cask loading area or any region of the pool containirj fuel. A lateral-free zone clearance Irom stored fuel shall be maintained. Accordingly, it is concluded that the proposed installation '

activities will not significantly increase the probability of a load-handling accident. The consequences of a load-handling accident are unaffected by the proposed instal)ation activities.

! The consequences si a spent-fuel assembly drop were evaluaced, anu it was determined that Le racks will not be distorted such that they would not perform their safety function. The criticality acceptance criteriou, '

Kegg 1 0.95, is not violated, and the calculated doses are well within 10 CFR Part 100 guidelines. Thus, the consequences of this type of accident are not changed from previously evaluated spent-fuel assembly drops that have been found acceptable by NRC.

The existing TSs permit the transfer-canal gate and the divider gate in the spent-fuel pool to travel ovur fuel essemblies in the spent-fuel pool. Analysis showed that this drop causei less damage to the new racks than the fuel-assembly drop when it impacts the top of the rack. Rack damage is restricted to an area above the active fuel region.

The consequences of 3 seismic event have been evaluated. The new racks are designed and will be fabricated to meet the requirements of applicable portions of the NRC regulatory guides and published standards. Design margins have been provided for rack tilting, deflection, and movement such that the-rack

• do not impact each other or

, the spent-fuel-pit walls in-the active fuel region during the postulated ,

seismic events. The new free-standing racks are designed to maintain their integrity during and after a seismic event. The fuel assemblics also remain Antact and therefore no criticality concerns exist.

The spent-fuel pool system is a passive system with-the exception of the i fuel pool coc11ng train and heating, ventilating, end air-conditioning (HVAC) equipment. Redundancies in the cooling train and HVAC hardware art not reduced by the planned fuel storage densification. The potential increased heat load resulting from any additional storage of-spent fuel is well within the existing system cooling capacity. Therefore, the probability of occurrence or malfunction of safety equipment leading to the_ loss of cooling flow in the spent-fuel pool is not significantly-affected. Furthermore, the consequences of this type incident are not significantly increased from previously evaluated cooling system loss of flow malfunctions. Thermal-hydraulic scenarios assume the reracked pool ,

is appro:timately 85 nercent full with spent fuel assemh13es. From this

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I starting point, the remaining storage capaelty is utillroo by analyzing i both normal back-to-back and unplanned full core offloads using conservative assumptions and previously established methods, Calculated values include maximum poc1 water bulk temperaturo, coincident maximum pool water local temperature, the maximum froj claddiag temperature, time-to-boll after loss of cooling paths, and the effect of flow blockage in a storage coll.

Although the proposed modification increases the pool heat load, results from the above analyses yield a maximum bulk temperature of approzinately 180 degrees Fahrer.helt which is below the bulk boiling temperaturn. Also tne maximum local water temperature is below nuclente balling condition values. Associated results from cotresponding loss of coolJng ,

evaluations give minimums of 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before boiling begins and 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> before the pool water level drops to the minimum regulred for shielding spent fuel. This is sufficient time to begin utilization of available alternate sources of makeup cooling water. Also, the offect of the increased thermal loading on the pool structure was evaluated and determined to be acceptable.

(2) Create the possibility of a new or different kind of accident from any accident previously analyzed.

The proposed modification has been evaluated in accordance with the guidance of the NRC position paper entitled, "0T Position for Review and Acceptance of Spent-Fuel Storage and Handling Applications"; eppropriate NRC regulatory guides; appropriate NRC standard review plans; and appropriate 1 Justry codes and standards. Proven analytical technology was used in designing the planned fuel storage expansion and will be utilized in the installation process. Basic reracking technology has been developed and demonstrated in over 80 applications for fuel pool capacity increases that have already received NRC staff approval. >

The TSs for the existing spent-fuel storage racks use burnup credit and

.f'-1 assembly administrative placement restrictions for criticality  :

cuatrol. The change to three-zono storage in the spent-fuel pool is J described in the. proposed change to.the design features section of the TSs. Additional evaltations were required to ensure that the criticality critorion is maintained.- These include the evaluation for the limiting criticality condition, i.e., the abnormal placement of an'unirradiated (fresh) fuel assembly of 4.95 weight percent enrichment into a storage cell location for irradiated fuel meeting the highest rack design burnup criterion. The evaluation for this case shows that the reactivity would exceed the limit in the absence of soluble boron. Soluble boron, for which credit is permittra under these abnormal conditions, ensures that reactivity is maintained substantially less than the design requirement.

Calculations indicate that a soluble poison coacentration of 685 parts per million (ppm) boron would be required to limit the maximum reactivity to a ke gg of 0.95, including uncertainties. This is less than the existing and proposed TS requirements of 2000 ppm..

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~4-It is not physically possible to inst all a fuel assembly outside and adjacent to a storage module in the spent-fuel storage pool. However, for a storage module i r;s t a l l ed in the cask loading area of the cask pit, there would be sufficient room ior such an extraneous assembly. The rmodule in this area is administratively limited by the preposed TS change to spent fuel only, and calculations show that t he raaximum kogg remains well below the 0.95 limit under this postulated accident condition, even in the absence of solubic boroa. To provide reactivity control assurance for the abnormal placement of a fresh assembly in the cask loading area module, a mo31fication to the existing TS has been proposod that requires boron concentration measurements while handling fuel in that area.

Although these changes required addressing edditional aspects of a previously analyzed accident, the possibility of a previously unanalyzed accident is not created. It is therefore concluded that the proposed reracking does not create the possibility of a new or different kind of accident fram any pic* lou:1y analyzed. [

(3) Involve a significant reduction in a margin of safety.

The design and technical review process applied to the reracking modification included addressing the following areas:

1. Nuclear criticality considerations.

2. Thermal-hydraulic considerations.

3. Mechanical, material, and structural considerations.

The established acceptance criterion for criticality is that the neutron multiplication factor shall be less than or equal to 0.95, including all uncertainties. The results of the criticality analysis for the new rack design demonstrate that this criterion is satisfied. The methods used in the criticality analysis conform to the applicable portions of NRC quidance and industry codes, standards, and specifications. In meeting '

3 the acceptance criteria for criticality in the spent-fuel pool and the cask loading area, such that kert is always 1,ss than 0.95 at a 95/95 percent probability tolerance level, the proposed amendment does not involve a significant reduction in the margin of safety for nuclear criticality.

Conservative methods and assuroptions were used to calculate the maximum fuel temperature and the increase in temperature of the w.ter in the spent-fuel-pit area. The thermal-hydraulic evaluation used methods previously employed. The proposed storage modification will increase the heat load in the spent-fuel pool, but the evaluation shows that the existing spent-fuel cooling system wi31 maintain the bulk pool water temperature at or below 180 degrees Fahrenheit. Thus it is demonstrated that the worst-case peak value of the pool bulk temperature is considerably lower than the hulk boiling temperaturo. Evaluation also shows that maximum local water temperatures along the hottest iuel assembly are below the nucleate boiling condition value. Thus there is no significant reduction in the margin of safety for thermal hydraulic or spent-fuel cooling considerations.

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The mechanical, material, and structural design of the new spent-fuel racks is in accordance with app 11 cable portions of "NRC OT poaltion for Review and Acceptance of Spent-Fuel Storage and Handling Applications,"  !

dated April 14, 1978 (as modified January 18, 1979), as well as other applicable NRC guidance and Industry codes. The primary safety function of the spent-fuel racks is to maintain the fuel assemblies in a sate configuration through all normal and abnormal loading conditions.

Abnormal loadings that have been evaluated with acceptable results and discussed previously include the effect of an earthquake and the impact because of the drop of a fuel assembly. The rack materials used are compatible with the fuel assemblies and the environment in the spent-fuel pool. The structural design for the new racks provides tilting, deflection, and movement margins ruch that the racks do not impact each other or_the spent-fuel-pit walls in the active fuel region during the postulated seismic events. Also the spent-fuel assemblies th6mselves remain intact and no criticality concerns exist. In addition, finite element _ analysis methods were used to evaluate the continued structura) acceptability of the spent-fuel pit. The analysis was performed in accordance with " Building Code Requitements for Reinforced Concrete" (ACI 318-63, 77). Therefore, with rtspect to mechanical, material, and structural considerations, there is no sicnificant reduction in a margin of safety.

In summary, the proposed spent-fuel storage modifications do not

1. Involve a significant increase in the probability or consequences of an accident previously eviluated; or

2. Yreate the possibility of a new or different kind of accident from any accident previously evaluateds er

3. Involve a significant reduction in a margin of safety.

Therefore, TVA has determined that the proposed amendments as described do not involve significant hazard considerations and that the criteria of 10 CPR 50.91 have accordingly been met.

TVA has also reviewed the NRC examples of licensing amendments considered not likely to involve significant hazards considerations as provided in the final adoption of 10 CFR 50.92 published on page 7751 of the Federal Reoister, Volume 51, No. 44, March 6, 1986. Ex ample (X) provides four criteria that, if satisfied by a reracking request, indicate that it is i likely no significant hazards considerations are involved. The criteria and how TVA's amendment request for Sequoyah complies are indicated

below.

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[ Criterion (1):

1 The storage expansion method consists of either replacing existing racks with a design that allows closer spacing between stored spent-fuel assemblies or placing additional racks of the original design on the pool floor if space permits.

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The Sequoyah Nuclear Plant teracking invalves replacing the existing racks with a design that allows c30ser spacing betwoon stored fuel assemblies and also provides additional rack storage on the pool floor where space permits.

CLiittipn (2):

The storage expansion method does not involve rod consolidation or double tiering.

i Proposed Amendment:

The Sequoyah racks are not double tiered, and all racks will sit on the floor of the spent-fuel pool. Additionally, the amendment application does not involve consolidation of spent fuel.

Criterion Cal  ;

The k egg of the pool is maintained less than or equal to 0.95.

hoposed Amendment:

The design of the now spent-fuel racks contains a neutron absorber, Boral, to allow close storage of spent-fuel assemblies while ensuring ,

that the k gg o remains less than 0.95 under all normal operating conditions with unborated water in the pool and less than 0.95 under abnormal conditionr with soluble boron in the pool.

CriterioD (4):

No new technology or unproven technology is utilized in either the construction process or the analytical techniques necessary to justify the expansion.

Proposed Amendment:

The construction processes and analytical techniques used in the fabrication and design are substantially the same as those of aumerous other rack installations. Thus, no new or unproven technology is utilized in the construction ur analysis of the high-density, spent-fuel racks at Sequoyah. TVA's Contractor, Holtec International, has

-previously supplied licensable racks of very similar design for about 10 other_reracking projects.

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