ML20044B136
| ML20044B136 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 07/06/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044B134 | List: |
| References | |
| NUDOCS 9007170386 | |
| Download: ML20044B136 (6) | |
Text
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SAFETY. EVALUATION.BY.THE OFFICE OF NUCLEAR. REACTOR. REGULATION
.r REL ATED.T0. AMENDNENT.NO.144..T0 FACILITY.0PERATING. LICENSE NO..DPR.53 i
AND. AMENDMENT.NO.127. TO. FACILITY.0PERATING-LICENSE.NO..DPR-69 BALTIMORE. GAS.AND ELECTRIC COMPANY.
CALVERT-CLIFFS #JCLEAR. POWER. PLANT, UNITS.1.AND.2 e
DOCKET.NOS. 50-317.AND.50-318
1.0 INTRODUCTION
~
i By letter dated April 30, 1990, Baltimore Gas and Electric Company, the licensee for.Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2, i
proposed Technical 1 Specification-(TS) changes to Appendix A of Operating-
. License Nos. DPR-53 and DPR-69 for Calvert Cliffs Units 1 and 2,
respectively. The proposed TS changes were requested to allow one-time movements of two separate special spent fuel shipping casks to and from the cask pit passing over spent fuel assemblies stored in the spent fuel pool. The t
first cask (33 inches in diameter,16 feet long, ud approximately 18 tons in
. weight) wil1~be used to ship 13 selected spent fuel rods to Chalk River, Canada.' The second smaller cask will be used to ship a reactor vessel weld
- material surveillance capsule to Combustion Engineering, both in support of hot-cell work sponsored by the Electric Power Research Institute (EPRI). The
-staff's-review and evaluation-addresses the.offsite radiological consequences and cr.iticality concerns resulting from a postulated spent fuel shipping cask
- drop accident (the larger cask) over the spent-fuel assemblies stored in the spent fuel pool.
2.0 BACKGROUND
The current Section 3.9.13 of the Calvert Cliffs TS, " Spent Fuel Cask Handling.
m Crane," prohibits a spent fuel shipping cask from passing over any area within one shipping-cask length of any stored fuel assembly. This TS requirement was the= result of the. licensee's. response to NUREG-0612. " Control.of Heavy Loads at'-
- Nuclear Power. Plants." The proposed TS changes would allow the present fuel,
. stored as the result of the extended Unit 2 outage, to remain in an area which currently prohibits storage when a fuel shipping cask is being moved into and 7
away'from the cask pit passing over that area. The weight of the spent fuel shipping casks-described above are bounded by the weight of the heavy loads (cask drop accident) considered during the NUREG-0612 review.
9007170386 900706 PDR ADOCK 0500o381 P
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. r The specific change requested will add the following footnotes to the Calvert Cliffs-Units-1 and 2,-TS Sections 3/4.9.13:
"These conditions are modified to permit shipping cask travel to and from the _ cask pit in the presence of fuel within one cask length radius of the pathway provided the boric acid concentration in the spent fuel pool is greater than or equal to 1000 ppm AND the following criteria'are met by all assemblies within one cask length radius of the pathway:-
- 1) initial enrichment less than or equal to 4.1' w/o U-235, 2) Burnup greater than~or equal to 28,000 MWD /MTU, and 3) greater than 440 days-elapsed from the shutdown of the last operating cycle in which the assembly was present in the -core. Crane interlocks and physical stops which restrict a spent fuel shipping cask from passing over any area
.within one shipping cask length of any fuel assembly not satisfying the above criteria:shall be demonstrated OPERABLE within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to-using the crane for moving a cask within one cask length of fuel assemblies meeting the above criteria. These modifications are applicable only for shipment of fuel rods supporting the EPRI sponsored hot-cell work and for the shipment of a reactor vessel weld material surveillance capsule."
Calvert Cliffs Units 1 and 2 share a common spent fuel pool that is located outside the containment in the auxiliary building and is divided by a dam into two identical halves, the-north and south pools. The spent fuel shipping cask pit is-located on the Unit 1 side, north of the dividing dam in the pool. The licensee estimated that a maximum of approximately 500 spent fuel assemblies could be located in the pool within the radius of one cask length from the cask pathway at any time during the cask movement.
3.0 EVALUATION The staff evaluated postulated fuel handling cask drop accident using assumptions contained in Positions C.I.a through C.I.k of Regulatory Guide (RG) 1.25 and the procedures specified in Standard _ Review Plan (SRP) Section 15.7.5.(NUREG-0800).
In addition, the specified assumptions postulate that the dropped fuel cask would damage all 500 spent fuel assemblies by falling
.onto its side and' rolling in a direction in which it could break open all fuel
' rods in-the 500 spent fuel assemblies.
Instantaneous puff release of noble gasses and radioiodine from the gaps of the broken rods occur as gas bubbles pass up through the water covering the fuel. All radioactivity reaching the auxiliary building atmosphere is exhausted within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> through engineered safety feature filtered exhaust systems:to.the environment. As stipulated in the proposed TS change request, all radioactive material in the 500 spent fuel assemblies that were damaged by-W the dropped fuel cask is assumed to have decayed for a period of greater than
-440 days.
4 3
The staff com uted the offsite doses fc; the Calvert Cliffs Exclusion Area 5
^ Boundary (EAB and Low Population Zone (LPZ) boundary using the assumptions described above, the assumptions contained in Regulatory Guide 1.25, and the procedures specified in SRP Section 15.7.5.
These computed offsite dotes are-
. ell within the acceptance criteri: given in Section 15.7.5 of the SRP and the w
' exposure guidelines of 10 CFR Par'; 200. The attached tables provide the results of the staff's calculations of the offsite Joses (Table 1) and :he assumptions used in calculating offsite doses (Tiole 2).
Criticality concerns were also considered because a cask drop would cause 0 geometrical distortion of the fuel / rack system.
Because the distortion.is difficult'to predict, assumptions were m*.de to bound the most reactive configuration.
For the calculations, the licensee assumed that the geometry of individual fuel assemblies was not deformed (this assumption maximized reactivity) and the storage racks were deformed to remove the inter-storage cell gap (neutron flux trap). Additionally, the poison material contained within the racks was ignored and replaced by pool water. Although NUREG-0612 states that criticality analysis for a dropped load may assume that the poison material integral to'the rack remains in place, the licensee conservatively chose to ignore this benefit. The licensee also assumed an initial fuel assembly enrichment limit of 4.1 w/o U-235, a minimum assembly burnup of 28,000 MWD /MTV, and a minimum boron concentration in the-spent fuel pool of 1000 ppm.
The staff finds the assumptions related to enrichment, burnup, and boron concentration, which are incorporated directly into the proposed TS change are consistent with NUREG-0612 and are, therefore, acceptable.
a
.The restrictions imposed by the footnote will be confirmed by procedures which-require that the serial numbers of each fuel assembly located within one cask length radius of the pathway be checked to assure that all the stored fuel within the are conforms to the restrictions. The procedures also require n
l verification that the boron concentration is greater than or equal to 1000 ppm
, prior to moving the cask into or out of the-spent fuel pool. ~
A two-dimensional analysis using the previously approved DOT-IV computer code r'
was performed for an infinite array of fuel assembly storage modules distorted as described above, The results yielded a Keff of 0.898.
Because L
~ uncertainties are less than 0.03, the Keff value with uncertainties will be no L
greater than 0.928. Consequently, the results are well within the criticality l
limit of a Keff equal to or less than 0.95 specified in Criterion II of p
NUREG-0612, Section 5.1.
The Keff of.095 is the currently approved limit in l
the TS for refueling operations as identified in the TS Bases 3/4.0, " Refueling l
Operations."
l3 The staff has determined that the proposed TS changes are acceptable based on its evaluation discussed above. -The offsite radiological consequences are 1
well within the acceptance criteria based on the staff's independent analysis using he assumptions in RG.I.25 and SRP Section 15.7.5.
The results of the i
licensee's criticality aralysis, using previously approved methodologies, are well Oithin the criticality limits of a Keff equal to or less than.095 specidied in NUREG-0612 and SRP Section 9.1.2.
I
.4 4 '. 0 ENVIRONNENTAL. CONSIDERATION These amendments involve a change in a requirement with respect to the
' installation or use of the facilities' components located within the restricted
[
areas as defined in 10 CFR Part 20. The staff has determined that these-amendments involve 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 exposure. The Comission has previously issued a proposed finding that these amendments involve no significant hazards consideration and there has been no public comment on such finding.
Accordingly, these amendments meet the eligibility criteria 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 these amendments.
5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that: (1)there r
is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed maniter, and (2) such activities will be conducted in compliance with the Comission's regulations and the issuance of
'these amendments will not' be inimical to the comon defense and security or to the health and safety cf the public.
Dated:
July 6, 1990 PRINCIPAL. CONTRIBUTORS:
-J.
Lee
- 10. Mcdonald P
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Table 1-Calculated Radiological Consequences following fuel cask drop accident (rem)
- Thyroid Dose 0.01 0.01.
75 Whola Body-Dose 0.05 0.01 6
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Table 2-
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Assumptions used for estimating the radiological consequences following a-postulated fuelt cask drop accident in fuel pool Parameter and Unit of Measure Quantity Power level, MWt -
2700 Number of fuel assemblies damaged
.500
= Shutdown time, days
'440 t
Inventory released from damaged rods, %
lodine 10 Noble Gases 30 Pool decontamination factors lodine 100 Nob.le gases 1
Iodine forms in atmosphere above pool, %
. Elemental 75
' Organic 25
' lodine 1 removal efficiencies for auxiliary building gas-treatment system (spent fuel pool area). %
Elemental
-No filters assumed-Organic No filters assumed.
Atmospheric dispersion factor, sec/m8 1.3 E-4
' Auxiliary building mixing efficiency, t 0
9 i