ML17254A685
| ML17254A685 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/16/1985 |
| From: | Lear G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17254A686 | List: |
| References | |
| NUDOCS 8512230221 | |
| Download: ML17254A685 (10) | |
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO. 50-244 R.
E.
GINNA NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.12 License No.
DPR-18 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Rochester Gas and Electric Corporation (the licensee) dated February 27,
- 1985, as supple-mented June 10, June 26 and July 11, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
8512230221 851216 PDR ADOCK 05000244 P
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No.
DPR-18 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.12
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 16, 1985.
George E. Lear, Director PWR Project Directorate Pl Division of PWR Licensing-A, NRR
ATTACHMENT TO LICENSE AMENDMENT NO. 12 FACILITY OPERATING L'ICENSE NO.
DPR-18 DOCKET NO. 50-244 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE 1-8 3 ~ 1 1 2
3.11-4 5.4-2 to 5.4-4 INSERT 1-8 3.11-2 3.11-4 5.4-2 to 5.4-4 5.4-4a
Dose E uivalent I-131 The dose equivalent I-131 shall be that concentration of I-131 which alone would produce the same thyroid dose as the quantity and i sotopi c mixture of I-1 31, I-132, I-133, I-134 and I-135 actually present.
The dose conversion factors used for this calculation shall be those for the adult thyroid dose via inhalation, contained in NRC Regulatory Guide
- 1. 109 Rev.
1 October 1977.
Re ortable Event A Reportable Event shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.
Canisters Containin Consolidated Fuel Rods Canisters containing consolidated fuel rods are stainless steel canisters containing the fuel rods of no more than two fuel assemblies which have decayed at least five years and are capable of being stored in a storage cell of the spent fuel pool.
1-8 Amendment No. 8,12
e.
Charcoal dsorbers shall be installed in e
3.11.2 ventilation system.exhaust from the spent fuel storage pit'area and shall oe operable.
Radiation levels in the spent fuel storage area shall be monitored continuously.
3.11.3 A load in excess of one fuel assemb'ly and its handling tool shall never be stationed or permitted to pass over storage racks containsng spent fuel.
3.11.4 The spent tuel pool temperature shall be limited to 150 F.
The restriction of 3.11.3 above shall not apply to the movement of cannisters containing consolidated fuel rods sf the spent fuel rack beneath the transported canister contain only spent fuel that has decayed at least 60 days since reactor shutdown.
Basis Charcoal adsorbers will reduce significantly the consequences of a retueling accident which.considers the clad failure of a single irradiated fuel assembly.'herefore,'harcoal adsorbers should be employed whenever irradiated fuel is being handled.
This p
requires that the ventilation system should be operating ana drawing air through the adsorbers.
The desired air flow path, when handling irradiated fue'I, is from tne outside of the building into tbh operating floor area, toward the spent fuel storage pit, into the area exhaust ducts, through the adsorbers, and out through the ventilation system exhaust to the facility vent.
Operation of a 3.11-2 Amendment No. 5, gP, 12
The spent fuel pool temperature is limited to 150'F because if the spent fuel pool cooling system is not at that, temperature, sufficient time (approximately 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />) is available to provide backup cooling, assuming the maximum anticipated heat load (full core discharge 6 previously stored fuel), until a temperature of 180'F is reached, the temperature at which the structural integrity of the pool was analyzed and found acceptable.
The requirement of 3.1L5 insures that should a handling accident occur during the movement of a consolidated fuel cannister (as described in 5.4.) the dose at the exclusion area boundary would satisfy the requirements of 10CFR100.
References (1),
PSAR - Section 9.3-1 (2)
ANS-5.1 (N 18.6), October 1973
- 3. 11-4 Amendment N0.12
5.4.4 Cannisters containing consolidated fuel rods may be stored in either Region 1 or 2 provided that:
a.
the average burnup and initial enrichment of the tuel assemblies from which the rods were removed b.
satisfy the requirements of 5.4.2 and 5.4.3
- above, and the average decay heat of the fuel assembly from 5.4. 5 5.4.6 which the rods were removed is less than 2150 BTU/hr The requirements of 5.4.4a may be excepted for those consolidated fuel assemblies of Region RGAF2.
The spent fuel storage pit is filled with borated water at a concentration to match that used in the reactor cavity and refueling canal during refueling operations whenever there is fuel in the pit.
Basis The center to center spacing of Region 1 insures that Keff <
0.95 for the enrichment limitations specified in 5.4.2 and for a postulated missile impact the resulting dose at the EAB would be within the guidelines of 10CFR100 In Region 2, Keff <
0.95 is insured by the addition of fixed neutron poison (boraflex) in each of the Region 2 storage locations, and a minimum burnup requirement as a function of initial enrichment for each fuel assembly design.
The 60 day cooling time requirement insures that for a postulated missile impact the resulting dose at the EAB would be within the guide-lines of 10CFR100.
5.4-2 Amertdaent Bo, 12
~ f The two curves of Figure 5.4-2 divide the fuel assembly designs into two groups.
The first group is all fuel delivered prior to January 1, 1984.
This incorporates all Exxon and Westing-house HIPAR designs used at Ginna.
The second curve is for the Westinghouse Optimized Fuel Assembly design delivered to Ginna beginning in February 1984.
The assembly average burnup is calculated using INCORE generated power sharing data and the actual plant operating history.
The calculated assembly average burnup should be reduced S
by 10% to account for uncertainties.
An uncertainty of 4% is associated with the measurement of power sharing.
The additional 6% provides additional margin to bound the burnup uncertainty associated with the time between measurements and updates of core burnup.
The curves of Figure 5.4-2 incorporate the uncertainties of the calculation. of assembly reactivity.
The calculations of fuel assembly burnup for comparison to the curves of Figure 5.4-2 to determine the acceptability for storage in Region 2 shall be independently checked.
'The record
"'of these calculations shall be kept for as long as fuel assemblies
= remain in the pool.
The fuel storage cannisters are designed so that, normally, they can contain the equivalent number of fuel rods from two fuel assemblies in a close packed array, and can be stored in either Region 1 or Region 2, rack locations.
The close packed array will insure the K~ of the rack configuration containing any number of cannisters will be less than that for stored fuel assemblies at the same burnup and initial enrichment.
The exception 5.4-3 Amendment ND. 12
of paragraph 5.4.5 is possible because the consolidated configuration is substantially less reactive than that of a fuel assembly.
The maximum decay heat requirement will insure that local and film boiling will not occur between the close packed fuel rods if the C
pool temperature is maintained at or below 150'F.
The decay heat of the assembly will he determined using ANS 5.1, ASB 9-2 or other acceptable substitute standards.
With the addition of the storage of consolidated fuel cannisters, the theoretical storage capacity of the pool would he increased to 2032 fuel assemblies (2x1016).
- However, due to limitation on the heat removal capability of the spent fuel pool cooling system, the storage capacity is limited to 1016.fuel assemblies.
References
- 1. letter, J.E. Maier to H.R. benton, january 18, 1984.
2.
Letter J.E. Haier to H.R. X)enton, january 18, 1984.
3.
Criticality.Analysis of Region 2 of the Ginna MDR Spent Fuel Storage Rack, Pickard, Lowe and Garrick, Inc.
March 8, 1984.
4.
Letter, T.R. Robbins, Pickard, Lowe and Garrick, Inc. to J.D. Cook, RGB March 15, 1984.
5.
Letter, D.M. Crutchfield to J.E. Maier, November 5, 1981.
5.4-4 Amendment No. 1Ã
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