ML20128P365

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Application for Amend to License DPR-35,changing TS to Increase Allowed Fuel Assembly Storage Cells from 2,320 to 3,859 & Changing Max Loads Allowed to Travel Over Spent Fuel Assemblies to 2,000 Lbs.Holtec Rept HI092925 Encl
ML20128P365
Person / Time
Site: Pilgrim
Issue date: 02/11/1993
From: Kraft E
BOSTON EDISON CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20128P367 List:
References
BECO-93-016, BECO-93-16, NUDOCS 9302240256
Download: ML20128P365 (10)
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E. S.~ Kraft. Jr.

Vice President Nuclear Operations ,

7 nd staten orector February 11.-1993 BEco 93o16 U. S. Nuclear Regulatory Commission

- Document Control Desk Washington, DC 20555 License DPR-35 ,

Docket 50-293 PROPOSED-TECHNICAL SPECIFICATION CHANGE PILGRIM NUCLEAR POWER STATION SPENT FUEL STORAGE CAPACITY EXPANSION

References:

1. HOLTEC International Report HI-92925, " Pilgrim Nuclear Power Station Spent Fuel Storage Capacity Expansion", _ dated- ,

December,-1992. (Attached) '

2. Amendment No. 33. to Pilgrim Nuclear Power Station -(PNPS)-

Operating-License No.-DPR-35, dated Augus.t 17, 1978 Boston Edison Company (BEco) proposes the attached modification. to the -

L Technical : Specifications, Appendix A of Operating License DPR 35 for' the-Pilgrim' Nuclear Power Station (PNPS) in accordance with 10CFR50.90.

.The proposed modification to.the Technical Specifications does the following:

  • Increases the allowed- fuel assembly storage cells from 2320 to 3859 -

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  • Changes the maximum loads allowed to travel over the spent fuel-assemblies from 1000 lbs. to 2000 lbs' .
  • Changes the limiting characteristics'of assemblies.to be~ stored in the-ing

' spent fuel fromlattice and a maximum a' maximumaverage K-in, nitps 1.35 to a maximum yKranium weight. enrichment ,

Thetincrease' in the fuel' assembly storage cells is accomplished by installing

.six-additional stainless steel storage _ racks with Boral as the neutron-absorbing material. . - The new storage- racks will be designed and installed-in -

the spent fuel- pool such --that the ;K,,, of the' pool with spent. _ fuel ' assemblies' will . be~ less : than 'or equal to!0.95. The proposed change -in' the maximum-allowable loads traveling over the spent fuel assemblies is - required - to accommodate the installation of storage platforms over the new storage racks.

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  • To: U.S. Nuclear Regulatory Commission om: E. S. Kraft Page: 2 e: February 11, 1993 The proposed limitations on the K 3rf 3ntt and Uranium enrichment of fuel assemblies to be loaded in the spent fuel, pool ensures the K err of the spent fuel pool remains less than 0.95. TheattachedHOLTECReporE,on,HI-92925, provides the design basis for the proposed spent fuel storage racks and an analysis supporting the proposed modification of the Technical Specifications.

This report has been reviewed and accepted by our Nuclear Eii9i neering Department.

By Amendment No. 33 to the PNPS Operating License, the Nuclear Regulatory Commission authorized Boston Edison Company to increase the spent fuel storage capacity to 2320 cells. The NRC Safety Evaluation supporting the amendment concluded there is no impact on the final Environmental Statement previously issued by the Commission in granting Operating License DPR-35 to PNPS. Our review of the NRC's Safety Evaluation indicates the environmental findings and conclusio... reached by the NRC in granting Amendment No. 33 at e applicable to this proposed amendment.

The Boston Edison Company requests the NRC's approval of the proposed modification to the Technical Specifications by January, 1994, to allow our preparation for installation of new racks. To maintain full core off-load capability during and beyond Refueling Outage #10 (planned for April,1995),

we intend to install the first two new racks in the Fall of 1994, b0 Urk l E.S. Kraft)*' uclear Operations Vice Presrdent and Station Director Commonwealth of Massachusetts)

County of Plymouth )

Then personally appeared before me, E. S. Kraft, who being duly sworn, did state that he is Vice President Nuclear Operations and Station Director of Boston Edison Company and that he is duly authorized to- execute and file the submittal contained herein in the name and on behalf of Boston Edison Company and that the statements in said submittal are true to the best of his-knowledge and belief.

g lea My commission expires: dDS /IDATE

% [ }thNOTARY PUBLit !w  ;

3 . ;. y Attachments: A. DescriptionofProposedModificationtotheTechni@ g T ,i Specifications e J e.*g;-

B. Replacement Technical Specification Pages 4...-

C. Marked-Up Technical Specification Pages ,

  • '* *g..j D. HOLTEC Report, HI-92925, " Pilgrim Nuclear Power Station Spent Fuel Storage Capacity Expansion."

I signed original and 37 copies cc: See next page

.-l To:. U.S. Nuclear Regulatory Commission From: E.S. Kraft Page: 3 Date: . February 11, 1993 cc: Mr. R. Eaton, Project Manager Division of Reactor Projects - I/II ,

Office of Nuclear Reactor Regulation Mail Stop: .14D1 U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Senior NRC Resident inspector Pilgrim Nuclear Power Station Mr. Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 305 South Street-Jamaica Plain, MA 02130

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q TATTACHMENT A TO BEC0 93415 DESCRIPTION OF PROPOSED MODIF! CATION TO THE TECHNICAL SPECIFICATIONE ,

PROPOSED MODIFICATION TO THE TECHNICAL SPECIFICATIONS Boston Edison Company proposes to modify the' Pilgrim -Nuclear Power 1 Station--

(PNPS)-Technical Specification sections- 5.5.C, D and E as follows:

Specification 5.5.C is revised to change the K innnit factor from 1.35 to 1.32-forU-235enrichmentupto4.6%averagedovertheaxfalplanarzoneofhighest

-average enrichment. Specification 5.5.D is revised to increase the Spent Fuel-Pool (SFP) storage capacity from 2320 to 3859 storage cells. Specification 5.5.E is revised to change the maximum loads allowed to' travel over the fuel assemblies from 1000 lbs. to 2000 lbs, to allow for the installation = of an.-

overhead platform upon which equipment may be stored.

This proposed Technical Specification change will support the installation of-six additional spent fuel storage racks in the spent fuel pool.-increasing =the cumulative storage capacity of the pool to 3859 fuel assemblies. 'The new storage racks are stainless steel with Boral as the! neutron absorbing =

material. At present, the spent fuel pool has 2333 storage cells. PNPS is

-currently licensed to store up to 2320 spent fuel assemblies by Specification-5.5.0. The additional six racks plus the existing 2333 storage cells will provide a cumulative storage capacity of 3859. This will extend the. reserve-full core off-lcad capability through the year 2012. The current storage cell capacity provides storage for spent fuel;' however, reserve for full core off -

load capability will be lost during and t.eyond Refueling.0utage #1u (1995).

The rack installation and the associated overhead platforms will: be accomplished in accordance with our commitments - to NUREG-0612, . " Control" of -

Heavy Loads to Nuclear Power Plants". An overhead platform will-be placed on

each of the new racks after storage cells are filled with- spent fuel assemblies. Thus the existing specification l_imit of 1000 lbs. must-increase-to 2000 lbs. to allow for the- platform installation. The new platform together with the lifting device weighs approximately 1850 lbs.

The maximum K r limit of fuel to be stored _ in- the ' spent: fuel pool Lis -

changedfrom1.3binit"to 1.32.

Additionally, to reflect the fact the reactivity -

of the fuel in the spent fuel pool is a function of U-235 enrichment as- well as K innnit - of the fuel in the standard core geometry, . a limit on U-235 enrichmentof4.6%wtisimposed. Together, these limits ensure K o f-the pool will remain ' O.95 -for all anticipated pool conditions,:,EncIuling' bosh abnormal and acc- 'it conditions.

While the proposed cha%e allows the addition of six storage racks containing a total of 1526 cells, only two racks will _be installed in the _ fuel pool _ upon receipt _of NRC's approval of this proposed Technical Specification amendment.

Two racks provide-an immediate 558 cell increase in storage' capacity .

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This interim increase will enable us to operate PNPS with full core of f-load

, capability until cycle #13 (now calculated to be the year 2003). Additional storage racks will be installed in the spent fuel pool as necessary.

BASIS FOR THE PROPOSED MODIFICATION The attached HOLTEC Report HI-92925, " Pilgrim Nuclear Power Station Spent fuel Storage Capacity Expansion", dated December 1992, provides the design basis and safety analysis performed to demonstrate the new spent fuel racks and the existing racks comply with applicable industry codes and standards. The report also provides information requested by the NRC position contained in "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", published in 1978 with a 1979 Addendum thereto.

EFFECTS ON SAFETY FUNCTIONS

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Spent Fuel Storage Rack Design Review:

lhe additional new free-standing high density spent fuel storage racks will store fuel in discrete modules in the Spent Fuel Pool (SFP) with one module available for installation in the cask pit. Each cell is designed for storage of BWR fuel assemblies with K yfin,on in the standard core geometry of up to 1.32 and Uranium-235 enrichments up to 4.6% wt (with credit for burnable poison) while maintaining the required subcriticality (K,f f < 0.95) . While the HOLTEC Report discusses K mfinity for U-235 enrichment up to 4.9% wt., the propased Technical Specification limits the U-235 enrichment to 4.6% wt. to maintain additional conservatism.

The high density spent fuel storage rack cells are fabricated from 0.090-inch thick type 304L stainless steel (S/S) sheet material. The basic building block of the racks is a six-inch (nominal) box of 304L S/S fabricated from precision sheared channels. The Boral panels are located between the checkerboard boxes and the sheathing without a water gap. The cells are welded

  • together in a specified manner to become a free-standing structure that is seismically qualified without depending on neighboring modules or fuel pool _

walls for support. The nominal center-to-center spacing of the cells is 6.05 inches.

Safety Assessment:

The safety assessment of the proposed rack modules demonstrates their thermal-hydraulic, criticality and structural compliance with established requirements. Thermal-hydraulic compliance requires fuel cladding will not fail due to excessive thermal stress, and the steady' state pool bulk temperature will remain within the limits prescribed for the spent fuel pool.

Structural compliance is demonstrated by analysis showing the free-standing modules will not affect the spent fuel assemblies under all postulated seismic events, and the primary stresses in the module structure will remain below the ASME Code allowables. Analyses of the new and existing racks are presented in the attached HOLTEC Report. The structural qualification includes analytical demonstration that the subcriticality of the stored fuel will be maintained under accident scenarios such as fuel assembly drop or accidental misplacement of fuel outside a rack.

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h The criticality safety analyses demonstrates the neutron multiplication factor-for- the- stored fuel _ array is bounded by- the Technical Specification limit of:-

0.95 under-assumptions of 95% probability and 95%.confidenceLfor both new and-existing racks. The criticality analysis also sets the requirements on the length and density of the B-10 rod.

Since spent fuel is presently stored in the PNPS spent fuel pool, special -

administrative controls and procedures will be used --to minimize radiation exposure during the installation of the new spent' _ fuel racks. The evaluation ,

of postulated accidents with respect to nuclear criticality and radioactivity  !

release has shown acceptable results.

uncertainties, and postulated releases K[iofrnot does not exceed exceed 10CFR100 0.95,--including-

' acceptance criteria as described in HOLTEC Report HI-92925, Sections 4 and 9.

The analyses presented in the report clearly demonstrate- the - rack' module arrays possess wide margins of safety from all key - (thermal -hyd raul i c ,'

criticality, structural, and radiological) vantage points. The following "No Significant Hazards Consideration" evaluation is based on the analyses-4 described in the report.

NO SIGjJFICANT HAZARDS CONSIDERATION DETERMINATION The following evaluation demonstrates the proposed amendment does not exceed any of the three significant hazards considerations criteria of 10CFR50.92(c).

The analysis of this proposed modification has been, accomplished using currently accepted codes and standards. The three criteria are discussed below:

(ll. The procosed amendment does not involve a sionificant increase in- the probability or consecuences of an accident previously evaluated.

The analyses performed by HOLTEC demonstrate the acceptability of the proposed-Spent Fuel Storage expansion from a variety of perspectives. The analyses demonstrate Kgf will remain within acceptable limits even if an abnormal event, such as a fuel assembly misloading or assembly drop, should occur. It also has been demonstrated the spent fuel pool cooling system will continue to provide acceptable cooling of the stored assemblies,- and therei is sufficient time to take appropriate corrective action should: all cooling be inadvertently lost. The racks are designed to seismic' Class I requirements.

An assembly inadvertently dropped on the racks would not prohibit =the racks from performing their design function. The radiological consequences of a fuel handling accident remains within previously-established _ limits.

M9vement of fuel assemblies and racks necessary for rack installation will be performed in accordance with our commitments to NUREG 0612, entitled: " Control of Heavy Loads at Nuclear Power Plants." Thus, the probability of an accident involving assembly damage will not be significantly increased. Based on these considerations, the probability or consequences of a previously evaluated accident is not significantly increased by installation activities.

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To support the above cohclusion, BEco has considered the following _ potential scenarios:

  • A spent fuel assembly drop in the spent fuel pool.
  • A loss of spent fuel pool cooling system flow.

. A seismic event.

. A construction accident.

As detailed in Section 4 of HOLTEC Report H1-92925, BECo evaluated the consequences of a spent fuel assembly drop in the spent fuel pool and found the criticality acceptance criterion, Kg 5 0.95, is not violated. Also there is no significant change in the radiclogical consequences of a fuel assembly drop from the previous analyses m the calculated doses are well within 10CFR100 guidelines. Analysis sc . .; th:tt dropping a spent fuel assembly on the racks will not prohibit the racks from performing their safety.

function. Thus, the consequences of this type of accident are_ not significantly changed from the previously evaluated spent fuel assembly drops-.

Certain racks in the pool will be equipped with overhead storage platforms.

These platforms are flat plate structures. They serve to store miscellaneous items and protect the fuel assemblies stored underneath from damage._ Dropping the platform from a height of 4 inches above the rack-(a possible situation if the platform is ever moved in the pool) was analyzed. It was determined that dropping spent fuel from 4" above the racks is a more severe event than a 4" drop of the platform with an assumed dry weight of 2000-lbs. Therefore, the fuel drop scenarios bound the platform drop condition.

During refueling activities, when the heat load in the pool is greatest,- an intertie is available between the fuel pool cooling system and either loop of the Residual Heat Removal (RHR) system. The RHR pump and heat exchanger-configuration provides greater cooling capacity for full core off-loads and as a backup to the normal fuel pool cooling system. This system will function during a loss of offsite power by utilizing e.mergency diesel generator AC-power. The-analysis in Section 5 of HOLTEC Report HI-92925 determined cooling capacities and maximum temperatures as well as the - time-to-boil without cooling. The calculations show that if cooling is lost at the instant when the pool water reaches its maximum value during a full core off-load, there is a minimum of 6.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before bulk boiling can occur.

- During reactor power operation, the normal fuel pool cooling system is used with either of the two pumps and heat exchangers capable of maintaining the fuel pool well below boiling. In the event of a -loss of offsite power, a temporary AC power interconnection is used to operate-one or both pumps. Due to lower spent fuel pool heat loads during plant operation, more than 16_ hours are available before bulk boiling can occur. Thus, the consequences of_ this -

event type are not significantly increased from previously- evaluated loss of cooling system flow events.

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The consequences ~ of a seismic- event- have - been evaluated. The = additional new

^ racks will meet design and fabrication requirements of applicable NRC-

- Regulatory Guides : and industry standards. Seismic ana_ lyses on : the new and ,

existing racks were performed- using both - single rack 3-D -(opposed- phase motion) and Whole Pool . Multi-Rack (WPMR) models. The results- of these analyses indicate a large margin of kinematic and stress safety. The-kinematic margin against rack-to-rack impact is at least I na inches or rack-- ,

to-wall .mpact is at least 2 me inches for all racks in the poel. Likewise, the maximum rack primary stresses imder the Safe Shutdown Earthquake (SSE) condition are less than 50% of the allowable ASME Code value. Finally, the maximum bending moments and through-thickness- shear in the supporting pool structure under factored load conditions are less than 80% of the respective allowables. The new free-standing racks are Asigned, as are the existing free-standing racks, so that the integrity of the racks and the pool structure is maintained during and af ter a seismic event. Thus, the consequences of 'a -

seismic event are not increased from previously evaluated events.

The consequences of a construction accident have been considered. A heavy load will not be carried in the spent fuel pool area until all fuel in the l pool has decayed for a minimum of three months. Per NUREG 0612 this provides i sufficient time for the decay of gaseous radionuclides in - the fuel (gap

! activity) such that an assumed accidental releast of gasses from damage to all stored fuel assemblies results in a potential offsite dose less than 10CFR100 l limits. In addition, there is no equipment essential to the safe shutdown of the reactor or employed to mitigate the consequences of an accident beneath, adjacent to, or otherwise within the area of influence of any loads to be handled during th!s expansion modification. Therefore, the consequences of a j construction accident are not significantly increased from previously evaluated events.

NUREG-0554, entitled: " Single-Failure-Proof Cranes for Nuclear Power Plants", -

provides guidance for the design, fabrication, installation and testing of new cranes that are of a high reliability design. NUREG-0612, Appendix C, entitled: " Modification of Existing Cranes," provides guidelines -on the implementation of NUREG-0554 at operating plants. An evaluation of storage -

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rack movements to be performed by the PNPS Reactor Building-crane demonstrated the probability of a drop of a storage rack is= extremely small. The -Reactor Building crane has a rated capacity of 100- tons ard incorporates a design safety factor of five. The maximum weight of any existing or replacement storage rack and its associated handling tool is 15 tons. Therefore, there is an ample safety factor margin for movements of the storage racks by the Reactor Building crane.

Therefore, it is concluded that the proposed amendment supporting the addition-I of spent fuel racks in the spent fuel pool does not involve a' significant i increase in the probability or consequences -of any accident previously

-evaluated.

j (2). The proposed amendment does not create the possibility of - a new or different kind of accident from any accident previously evaluated.

j No unproven technology is involved either in the installation process or in i the analytical techniques necessary to justify the planned fuel storage expansion. The basic technology for fuel pool expansion has been developed t

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and demonstrated in over 80 applications for fuel pool capacityj increases-

_4 previously approved by the NRC.

HOLTEC has evaluated the proposed modification in accordance with the guidans of an NRC position paper entitled: "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," with-appropriate NRC Regulatory  ;

Guides, with NRC Standard Review Plans, and with industry codes and standards.

In addition, BEco has reviewed several- previous NRC- Safety Evaluation' Reports-for rack installation applications similar to this proposed modification.

Based upon the foregoing, the proposed rack installation does;not_ create. the-oossibility of a new or different of accident from any accident ~ previously evaluated.

(31. The proposed amendment does not involve a sianificant reduction-in_4 marain of safety.

The HOLTEC report demonstrates the acceptability of adding new racks from a variety of perspectives including ~ criticality, thermal-hydraulic, radiological, seismic and structural considerations. The results of these analyses provide the basis for our conclusion that the changes do not involve a significant reduction in a margin of safety.

The established acceptance criterion for criticality is that the effective neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions. This margin of safety has been adhered to in the criticality analysis methods in developing-the new rack design.

The methods used in the criticality analysis conform to the -applicable.

portions of the appropriate NRC guidance and industry codes, standards, and specifications. In meeting the acceptance criteria for criticality in the spent fuel peol such that Km is always less than or equal to 0.95, including uncertainties at a 95%/95% probability / confidence level, the proposed:

amendment does not involve a significant reduction in the margin of safety for.

nuclear criticality.

It is recognized that a one-to-one correspondence between the k-infinity of a bundle in the standard core geometry and the. keft in the fuel rack-does~ not -

exist. The effect o_f higher fuel enrichments on the neutron energy: spectrum is to reduce the reactivity worth of- the Boral absorber, resulting in a lower value of limiting kinfin:ty. _ In order to provide a complete specification of fuel that can be stored in the PNPS pool, the criteria for both k-infinny and fuel enrichment needs to be prescribed. Calculations have been performed to demonstrate that all fuel assemblies of up to 4.9% wt - planar-average U-235 enrichment with a kinrinity of 1.32 or less can be stored in- the-PNPS_ spent- fuel pool with K m less than or equal to 0.95.

Conservative methods were used to; calculate the maximum fuel temperature- and -

the -increase,in_ temperature of the water in the-spent fuel _ pool. The thermal-hydraulic evaluation used methods previously employed for evaluations of the present spent fuel racks to demonstrate the temperature margins of safety are maintained. The proposed modification will increase the heat load in_ the spent fuel pool. The evaluation shows the existing spent fuel cooling -system 6

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will maintain the' bulk 1 pool water temperature at or below 142*F during

, refueling.

The evaluation also shows that maximum local water temperatures along the hottest fuel assembiy are below the nucleate boiling condition value.- Thus, there is no significant. reduction in the margin of safety caused by -thermal-hydraulic or spent fuel cooling concerns.

The main safety function of the spent- fuel pool and the racks'is to maintain

'the spent fuel assemblies in a safe configuration through all normal or abnormal loadings. Abnormal loadings that have been considered are the_ effect of an earthquake, the drop o_f a spent fuel assembly, or the drop of any other heavy object. The mechanical, material, and structural _ design of the new spent fuel racks is in accordance with NRC guidance. The rack materials used are compatible with the spent fuel pool and the spent fuel assemblies. The structural considerations of the new racks -and existing racks address margins of safety to preclude tilting, deflection or movement, thereby ensuring the racks do not impact each other during postulated seismic events. In addition the spent- fuel assemblies remain intact. and no criticality concerns exist.

Thus, the margin of safety is not significantly reduced by the proposed rack additions.

Conclusion:

In view of the above, BEco has determined' the proposed amendment does not involve significant hazards consideration and the criteria of 10CFR50.92 have-been met.

This change has been reviewed and recommended for approval by the Ope 2 sons-Review Committee and reviewed by the Nuclear Safety Review -an' Audit Committee.

SCHEDULE OF CHANGE This change will become effective 30-days following BECa's receipt of the Commission's approval.

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