ML20062B616

Safety Evaluation Supporting Amend 133 to License DPR-40

ML20062B616
Person / Time
Site:	Fort Calhoun
Issue date:	10/12/1990
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20062B607	List: ML20062B604 ML20062B610 ML20062B616
References
NUDOCS 9010250259
Download: ML20062B616 (5)

Text

-

f *lovy

'[R71Og k UNITED STATES NUCLEAR REGULATORY COMMISSION

e' .' j r

AsMiwovow.o. c. aosss

%....+/

SAFETY EVALUATION BY THE OFFICE 0F NUCLEAR REACTOR REGULATION-RI. LATED TO AMENDMENT NO. 133 TO FACILITY OPERATING LICENSE NO. - QPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION. UNIT NO. 1 l i

DOCKET NO. 50-285 L

1.0 INTRODUCTION

By letters dated January 26, May 10, June 18, and August 2,1990 (References 1, 4, 5 and 6), Omaha Public Power District (OPPD) requested changes to the Fort Calhoun Unit 1 Technical Specifications.

The following changes are requested: (1) Increase the refueling-j boron concentration from 1800 to 1900 ppm; (2) Suspend certain sampling during periods when the fuel has been removed from the reactor vessel; (3) Revise the storage requirements of the spent fuel pool Region 2;.(4) Allow discharge of fuel assemblies from the core i

' directly to Region 2 of the spent fuel pool; and-(5) Delete the i requirements to provide a fuel performance report at the end-of each cycle.

As a result of the staff's review, the licensee supplemented its initial application with supplemental submittals which the staff i included in its original findings of no significant hazards consideration except for the August 2, 1990 submittal. This submittal corrected Technical Specification pages to include an omitted Greek letter p in its units. The August 2nd change was'within the scope of the notice' '

published in the Federal Reaister on July 25, 1990 (55 FR 30305), and-did not affect the initial determination.

2.0 DISCUSSION 2.1 Refuelina Boron Concentration The proposed amendment would increase the reactor vessel refueling boron concentration from 1800 to 1900 ppm in Technical Specification Sections Definitions, 2.3, 2.8, 2.14 and 4.4.- The limit on the boron i concentration during refueling ensures that the reactor will remain subcritical by at least 5% delta k/k-even if all control element assemblies (CEAs) are fully withdrawn from the core, as required by the plant Technical Specifications. In addition, the limit also provides 33.3 minutes before shutdown margin would be' lost in=the event of an inadvertent boron dilution assuming the maximum credible influx of unborated water. This meets the Fort Calhoun licensing-basis requirement of a dilution time to critical'of not less than 30 minutes during refueling. The shutdown margin and boron dilution events were calculated using approved methods. The proposed changes.

are, therefore, acceptable.

9010250259 901012 2 PDR ADOCK 05000285 a P PDC J

. . 2.2 Suspend Certain Samplina When All Fuel is Removed 1

2.2.1 Suspension of Boron Samplina When All Fuel'is Removed The proposed amendment would allow the suspension of boron sampling in  !

the reactor vessel-when all fuel has been removed.- Technical Specification  :

Sections 2.8 and 3.2 are affected. The reactor vessel coolant boron &

concentration requirement is based on the need for adequate shutdown-margin during refueling or fuel handling. When there is no fuel in the  ;

vessel, shutdown margin is not required and the need for. boron is  ;

eliminated. Therefore, the need for sampling the boron concentration is '

eliminated. The elimination of boron sampling and analyses when the core has been off-loaded is adequately covered by Footnote (3) to  ;

Technical Specification Table 3-4, Minimum Frequencies for Sampling Test, which requires reinitiation of sampling on shift prior to reintroduction of fuel into the reactor cavity. In addition, the proposed deletion of Section (6) to Technical Specification 2.8 is acceptable as it is consistent with the frequency now required by Table 3-4. The proposed changes are, therefore, acceptable.

2.2.2 Suspension of Chloride Samplina When All Fuel is Removed The proposed amendment would allow the suspension of chloride sampling in the reactor vessel when fuel is removed. Table 3-4 of Technical Specification 3.2 is affected. The chloride ion concentration limit is i

established to prevent any potential degradation of fuel mechanical design properties or Reactor Coolant System (RCS) piping. When fuel is not present, the mechanical design properties of the RCS are not subject I

! to potential degradation due to intergranular stress corrosion cracking induced by a high chloride concentration. The licensee's chemical.

control procedure provides adequate assurance that chloride contamination is not likely. The sampling requirements of Technical Specification Table 3-4, Minimum Frequencies for Sampling Test, require reinitiation of sampling one shift prior to reintroduction of fuel into the reactor cavity. This provides for detection of chloride contamination and subsequent corrective action prior to reintroduction of fuel into the reactor. The proposed changes are, therefore, acceptable.

2.3 Storaae Requirements for Spent Fuel Pool Recion 2-The proposed amendment would modify Technical Specification Figure 2-10 to allow spent fuel assemblies stored in Region 1, which currently do not meet the minimum burnup requirements for storage-in Region 2, to be moved to Region 2 if a' full length CEA is inserted into the fuel assembly.

The spent fuel storage racks consist of two distinct regions. Region 1 can accept either new or irradiated fuel. Region 2, however, can accept only spent fuel meeting the minimum exposure requirements currently specified in Figure 2-10.

. 3 ..

l Analyses were performed for the licensee by Pickard, Lowe and Garrick l (PLG) to determine the reactivity of the spent fuel racks (k-eff) as a j function of initial U-235 enrichment and exposure. PLG has demonstrated ,

their ability to calculate the reactivity of. spent fuel' pool. Packs in I the past. The models and codes which have been used in the ahalyse's have been thoroughly qualified by comparison with measured critical experiments and other published data. -Therefore, the' staff considers PLG and the methods used adequately qualified.. The effects of calculational biases and uncertainties, manufacturing and mechanical tolerances, and postulated accidents have been adequately accounted for in the determination l of the minimum allowable exposure, as'a function of initial enrichment, '

for an assembly with a CEA in place to be moved to Region 2. In addition,. ,

the determination of the minimum amount of boron-10 (B-10) remaining'in  ;

the CEAs assumed full CEA insertion during their residence in the core.. '

This is conservative since the CEAs had only limited insertion.during their exposure in-the core.

The results of the analyses show that a k-eff of 0.9450 can be maintained in the Region 2 rack with CEAs inserted in fuel assemblies with an initial- '

enrichment of 3.25 weight percent-(w/o) U-235 after achieving an exposure of 4,900 MWD /MTU. Equivalently, for an initial enrichment of 4.00 w/o ,

U-235, the required exposure is 12,800 MWD /MTU. This meets the;NRC fuel ,

a

' storage criterion which specifies that the worst case k-eff for the storage rack, including all biases and uncertainties, must be no greater than 0.95.

The 45 CEAs currently residing in the spent fuel pool have been shown to have maintained their structural integrity by a combination of inspections and calculations. However, because of the recent CEA failures in older CE designs attributed to irradiation assisted' stress corrosion cracking

! (IASCC), the licensee was asked to evaluate the continued acceptable use of these 45 CEAs for suberiticality during their lifetime in the spent fuel pool. The-licensee stated that the relatively benign environment of the spent fuel pool, as compared to the reactor core would not cause future IASCC. The staff, therefore, concludes that these CEAs are acceptable-for use as neutron absorbers. Administrative controls.will-require the CEAs to remain in the fuel assemblies while they are resident in Region

2. In addition, a clip will be attached to tie the CEA and fuel assembly together in Region 1 prior to transfer to Region 2. The clip will not -

be able to be removed by the grapple on the fuel handling machine.~ This physical restraint system will prevent the inadvertent removal of~a CEA from the fuel assembly after it is inserted. The staff concludes that the administrative controls and the physical restraints are' sufficient to preclude the inadvertent removal of a CEA from a fuel assembly and- .;

i thus maintain the required amount of subcriticality in the spent fuel l pool The proposed changes are, W refore, acceptable.

l I

l 5

4

, y , - -

+

l' . .  !

l l, 4'.

l 2.4 Direct Transfer of Spent Fuel from Reactor Core to Region 2 1

The proposed amendment would modify Technical Specification Section 2.8 to allow spent fuel to be transferred directly from the react 6r core to Region 2 of the spent fuel pool. Originally, the staff required fuel to- 1 be moved directly from the core to Region 1 and then to Region 2 only i after proper verification had been made of sufficient burnup for acceptable.- '

storage in Region 2. An interlock on the fuel handling crane at Fort s Calhoun prevented fuel from being transferred from the core directly to Region 2. Subsequently, the staff modified its requirements-and allowed a by pass of the interlock such that sufficiently irradiated fuel could_ be ,

directly placed into Region 2 if two independent verifications of burnup adequacy were performed (Ref. 2). A fuel burnup determination is-performed at Fort Calhoun prior to fuel movement into Region 2 of the spent fuel-pool using Special Procedures SP-BURNUP-1. The staff concludes that this procedure adequately ensures that the _ fuel will meet the acceptance criteria.

for burnup prior to movement to Region 2. In addition, the potential risk  ;

of dropping a fuel assembly with the additional handling required in a two-step operation is minimized. The proposed change is, therefore, acceptable.

2.5 End of Cycle Fuel Performance Report The requested revision to Technical Specification Section 5.9.3.h would eliminate the requirement for a fuel performance report at the end of each cycle. This request is administrative because it has been previously approved in the Safety Evaluation Report for Amendment No. 77 (Ref 3)-

but was inadvertently retained in the Technical Specifications. The reportable events required by 10 CFR 50.73 would still require the licensee to report unique, widespread or unexpected fuel cladding failures in the reactor or spent fuel pool.

2.6 Findinas Based on the above evaluation, the staff finds that the requested changes to the Fort Calhoun Unit 1 Technical Specifications are acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

The amendment involves a change .in a requirement with respect to the installation or use of a facility component-located within the restricted area as defined in 10 CFR Part 20 and changes in surveillance requirements.

The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and.that there is no significant increase in individual or cumulative occupational radiation exposures. The Commission has previously issued a proposed finding that the amendment-involves-no significant hazards consideration and.there has been no public comment on-such finding. Accordingly, the amendnient meets the eligibility criteria I

for categorical exclusion set forth in 10 CFR Section 51.22(c)(9). Pursuant-  !

to 10 CFR 51.22(b), no environmental ~ impact statement or environmental assessment need be prepared in connection with the issuance of the amendment, .(

I

- .. - _._m __m._.- -

4.0 CONCLUSION

. i The staff has concluded, based on the considerations discusseh above, that:

(1) there is reasonable assurance that the health and safety of the public.

will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, l

and the issuance of the amendment will not be inimical to the common defense and security or to the health and-safety of the public.

5. 0 REFERENCES

1. Letter from K.J. Morris (OPPD) to USNRC, " Application for Amendment of Operating License, LIC-89-1171," January 26, 1990.

2. Memorandum from Frederick J. Hebdon (NRC PD-IV) to James L. Milhoan (NRC Reg. IV), "Ft. Calhoun Spent Fuel Storage Procedures," July 19, 1989.

3. Letter from E.G. Tourigny (NRC) to W.C. Jones (OPPD), " Amendment No.

77 to Facility Operating License No. DPR-40 for the Fort Calhoun

. Station, Unit 1," April 26, 1984.

4. Letter from W.G. Gates (OPPD) to USNRC, " Application for Amendment [

of Operating License, LIC-90-0378," May 10. 1990.-

5. Letter from W.G. Gates (OPPD) to USNRC, " Application for Amendment of Operating License, LIC-90-0500," June 18,1993.

6. Letter from W.G. Gates (OPPD) to USNRC, " Application for Amendment of Operating License, LIC-90-0645," August 2, 1990. '

Dated: Oc.t;ber 12, 1990 Principal Contributors:

L. Kopp, SRXB S. Koscielny, EMCB l

l 1

I r

_ - - - _ _ .-, ,,