ML20072A393

Proposed Tech Specs Re Spent Fuel Storage Rack Mod

ML20072A393
Person / Time
Site:	Fort Calhoun
Issue date:	06/06/1983
From:	OMAHA PUBLIC POWER DISTRICT
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ML20072A387	List: ML20072A386 ML20072A393
References
NUDOCS 8306100106
Download: ML20072A393 (15)

Text

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2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueling Operations (Continued)

(7) Direct communication between personnel in the con-trol room and at the refueling machine shall be available whenever changes in core geometry are taking place.

_(8) When irradiated fuel is being handled in the auxi-liary building, the exhaust ventilation from the spent-fuel pool area will be diverted through the charcoal filter.

(9) Prior to initial' core loading and prior to refuel-ing operations, a complete check out, including a load test, shall be conducted on fuel handling cranes that will be required during the refueling operation to handle spent fuel assemblies.

(10) A minimum of 23 feet of water above the top of-the core shall be maintained whenever irradiated fuel is being handled.

(11) Storage in Region 1-and Region 2 of the spent fuel racks shall be restricted to fuel assemblies having initial enrichment less than or equal to 4.0 weight percent of U-235.

(12) Storage in Region 2 of the spent fuel racks shall be restricted to those assemblies whose parameters fall within the " acceptable" region of Figure 2-10.

If any of the above conditions are not met, all refueling operations shall cease immediately, work shall be initiated to satisfy the required conditions, and no operations that may change the reactivity of the core shall be made. How-ever, refueling operations may commence and continue with less than 5 containment atmosphere and plant ventilation duct radiation monitors provided that gross, particulate and iodine monitors are monitoring the stack ef fluent.

These three plant ventilation duct radiation monitors will initiate closure of the containment pressure relief, air

, sample and purge system valves and shall employ a one-out-of-three logic for the initiation of VIAS.

e Irradiated fuel movement shall not be initiated before the

$@f reactor core has decayed for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the wo reactor has been operated at power levels in excess of 2%

80 eO rated power.

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@ The equipment and general procedures to be utilized during

- - refueling operations are discussed in the USAR. De tailed g instructions, the above specifications, and the design of

' est the fuel handling equipment incorporating built-in inter-locks and safety features provide assurance that no Amendment No. 5, H , 25, 43 2-38 ATTACHMENT A y , , - -y v-r--v,- , - - y , ,- , 3- . - . - - -,, e

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Ssent Fuel Pool Region 2 Storage Criterie Minimum Required Fuel Assembly Exposure es e Function of Inittel Enrichment to Permit Storage in Region 2 40000 Fuel ss neceptable for storage in Regson 2 for points above the line.

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2. 5 3. 0 3. 5 4. O Assembly Insttal Enrichment, w/o U-235 SPENT FUEL POOL REGION 2 OMAHA PUBLIC POWER DISTRICT FIGURE STORAGE CRITERIA FORT CALHOUN STATION-UNIT No.1 2-10

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l 2.0 LIMITING CONDITIONS FOR OPERATION i 2.8 Ref ueling Operations (Continued) incident could occur during the refueling operations that would result in a hazard to public health and safety.(1) When-ever changes are not being made in core geometry one flux moni-tor is sufficient. This permits maintenance of the instrument-ation. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition.

The shutdown cooling pump is used to maintain a uniform boron concentration.

The shutdown margin as indicated will keep the core subcriti-cal even if all CEA's were withdrawn from the core. During refueling operations, the reactor refueling cavity is filled with approximately 250,000 gallons of borated water. The boron concentration of this water (at least 1700 ppm boron) is sufficient to maintain the reactor subcritical by more than 5%, including allowance for uncertainties, in the cold condi-tion with all rods withdrawn.(2) Periodic checks of refueling water boron concentration ensure the proper shutdown margin.

Communication requirements allow the control room operator to inform the refueling machine operator of any impending unsafe condition detected from the main control board indicators during fuel. movement.

In addition to the above engineered safety features, inter-locks are utilized during refueling operations to ensure safe handling. An excess weight interlock is provided on the lift-ing hoist to prevent movement of more than one fuel assembly at a time. In addition, interlocks on the auxiliary building crane will prevent the trolley from being moved over storage racks containing irradiated fuel, except as necessary for the handling of fuel. The restriction of not moving fuel in the reactor for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the power has been re-moved from the core takes advantage of the decay of the short half-life fission products and allows any failed fuel to purge

! itself of fission gases, thus reducing the consequences of l fuel handling accident.

l The ventilation air for both the containment and the spent l fuel pool area flows through absolute particulate filters and j radiation monitors before discharge at the ventilation dis-charge duct. In the event the stack discharge should indicate a release in excess of the limits in th*e technical specifi-cations, the containment ventilation flow paths will be closed automatically and the auxiliary building ventilation flow

( paths will be closed manually. In addition, the exhaust venti-l lation ductwork from the spent fuel storage area is equipped with a charcoal filter which will be manually put into oper-ation whenever irradiated fuel is being handled.(1)

References (1) FSAR, Section 9.5 (2) FSAR, Section 9.5.1.2 l Amendment No. 24 2-39

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... . -_- - - .. . _-~

TABLE 3-5 y (Continued)

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'! FSAR Section

+ Test Frequency Reference i

10c. (Continued) 4. Automatic and/or manual initi- At least once per-plant-

O ation of the system shall be operating cycle.

demonstrated.

m b

11. Containment Cool- 1. Demonstrate damper action. 1 year, 2 years, 5 years, 9.10 g ing and Iodine and every 5 years there-Removal Fuseable 2. Test a spare fuseable link. after.

Linked Dampers

12. Fuel Elements Visually inspect fuel elements re- During each refueling 3 ,

moved from the reactor. outage.

$ 13. Diesel Generator Calibrate. During each refueling 8.4.3

? Under-Voltage outage.

Relays

14. Motor Operated Verify the contactor pickup value During each refueling Safety Injection at < 85% of 460 V. outage.

Loop Valve Motor Starters ( IICV-311, 314, 317, 320, 327, 329, 331, 333, 312, 315, 318, 321)

, 15 . Pressurizer lieaters Verify control circuits operation for During each refueling post-accident heater use. outage.

}

3 j 16. Spent Fuel Pool Test neutron poison samples for Intervals of 1, 2, 4, j Region 1 Racks dimensional change, hardness change, 7, 11, 15, 20, and 25 i

and neutron attenuation change. years after installation.

4.0 DESIGN FEATURES 4.4 Fuel Storage 4.4.1 New Fuel Storage The new unirradiated fuel bundles will normally be stored in the dry new fuel storage. rack with an effective multi-plication factor of less than 0.9. The open grating floor below the rack and the covers above the racks, along with generous provision for drainage, precludes flooding of the new fuel storage rack.

New fuel may also be stored in shipping containers or in the spent fuel pool racks which have a maximum effective multiplication factor of 0.95 with Fort Calhoun Type C fuel and unborated water.

The new fuel storage racks are designed as a Class I structure.

4.4.2 Spent Fuel Storage Irradiated fuel bundles will be stored prior to off-site shipment in the stainless steel lined spent fuel pool.

The spent fuel pool is normally filled with borated water with a concentration of at least 1700 ppm.

The spent fuel racks are designed as a Class I structure.

Normally the spent fuel pool cooling system will maintain the bulk water temperature of the pool below 1200F. .

Under other conditions of fuel discharge, the fuel pool water temperature is maintained below 1400F.

The spent fuel racks are designed and will be maintained such that the calculated effective multiplication factor is no greater than 0.95 (including all known uncertain-ties) assuming the pool is flooded with unborated water.

The racks are divided into 2 regions. Region 1 racks are surrounded by Boraflex; Region 2 racks have no poison.

Acceptance criteria for fuel storage in Regions 1 and 2 are delineated in Section 2.8 of these Technical Speci-fications.

l Amendment No. , 43 4-4

5.0 ADMINISTRATIVE CONTROLS 5.8.3 a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant supervisory staff, at least one of whom holds a Senior Reactor Operator's License on the unit af-fected.

c. The change is documented, reviewed by the Plant Re-view Committee and approved by the Manager - Fort Calhoun Station within 14 days of implementation.

5.8.4 Written procedures approved per 5.8.2 above shall be im-plemented which govern the selection of fuel assemblies to be placed in Region 2 of the spent fuel racks (Techni-cal Specification 2.8(12). These procedures shall re-quire an independent verification of initial enrichment requirements and fuel burnup calculations for a fuel bundle to assure the " acceptance" criteria for placement in Region 2 are met. This independent verification shall be performed by individuals or groups other than those who performed the initial acceptance criteria assessment, but who may be from the same organization.

5.9 Reporting Requirements In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforce-ment unless otherwise noted.

5.9.1 Routine Reports

a. Startup Report. A summary report of plant startup and power escalation testing shall be submitted foi-lowing (1) receipt of an operating license, (2) amendment to the license involving a planned in-crease in power level, (3) installation of fuel that has a different design or has been manu-factured by a different fuel supplier, and (4) modi-fications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of the tests identified in the USAR and shall in general include a description of the measured values of the operat-ing conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license conditions based on other commitments shall be in-cluded in this report.

Amendment No. 9, 19, 35 5-10

m ._ _ , . . . _ - . _ _ _ _ .

5.0 ADMINISTRATIVE CONTROLS 5.9.1- Continued Startup reports shall be submitted within (1) 90 days following completion of the startup test pro-gram, (2) 90 days following resumption or commence-ment of commercial power operation, or (3) 9 months following. initial criticality, whichever is earli-est. If ._the Startup Report does not cover all three events (i.e.,. initial criticality, completion of startup test program, and resumption or commence-ment of commercial power operation), supplementary reports shall be submitted.at least every three months until-all three events have been completed.

b. Annual' Occupational Exposure Report. An annual occupational exposure report should be submitted

. prior to March 1 of each year. The report shall consist of a tabulation on an annual' basis of the number of station, utility and other personnel (in-cluding contractors) receiving exposures greater than 100 mrem /yr and their associated man rem ex-posure according to work and job functions,37 e.g.,

reactor operations _and surveillance, inservtce in-spection, routine' maintenance, special maintenance (describe maintenance),. waste processing, and re-fueling outages. . The dose assignment to various duty. functions may be estimates based on pocket

, dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual j

total dose need not be accounted for. In the aggre-gate, at least 80% of the total whole body dose re-l ceived from external sources shall be assigned to specific major work functions.

c. Monthly Operating Report. Routine reports of oper-ating statistics and shutdown experience shall be p submitted on a monthly basis to the Director, Of fice of Management . Information and Program Con-trol, U. S. Nuclear Regulatory Commission, Washing-ton, D.C. 20555, with a copy to the appropriate Re-gional Office, to arrive no later than the fif-teenth of each month following the calendar month covered by the report.

' 5.9.2 Reportable Occurrences Reportable occurrences,-including corrective actions and measures to prevent reoccurrence, shall be reported to the NRC. Supplemental reports may be required to fully i

3/ This tabulation supplements the requirements of 5 20.407 of 10 CFR Part 20.

i Amendment No. 9', 35 5-11 1,

5.0 ADMINISTRATIVE CONTROLS 5.9.2 Continued describe final resolution of occurrence. In case of cor-rected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date,

a. Prompt Notification With Written Followup. The types of events listed below shall be reported as expeditiously as possible, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the appro-priate Region Office, or his designate no later than the_first working day following the event, with a written followup report within two weeks.

The written followup report shall include, as a minimum, a completed copy of a licensee event re-port form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide com-plete explanation of the circumstances surrounding the event.

(1) Failure of the reactor protection system or other systems subject to limiting safety set-tings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limit-ing safety system setting in the technical specifications or failure to complete the required protective function.

Note: Instrument drif t discovered as a result of testing need not be reported under this item but may be reportable under items 2.a(5),

2.a(6), or 2.b(1) below.

(2) Operation of th'e unit or affected systems when any parameter or operation subject to a limiting condition is less conservative than the least conservative aspect of the limit-ing condition for operation established in the technical specifications.

Note: If specified action is taken when a system is found to be operating between the most conservative and the least conservative as-pects of a limiting condition for operation listed in the technical specifications, the limiting condition for operation is not con-sidered to have been violated and need not be reported under this item, but it may be reportable under item 2.b(2) below.

5-12

5.0 ADMINISTRATIVE CONTROLS 5.9.2 Continued (3) Abnormal degradation discovered in fuel clad-ding, reactor coolant pressure boundary, or primary containment.

Note: Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item.

(4) Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions greater.than or equal to 1.0% Ak/k; a calcu-lated reactivity balance indicating a shut-down margin less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity in-sertion of more than 0.5% ak/k; or oc-currence of any unplanned criticality.

(5) Failure or malfunction of one or more com-ponents which prevents or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the SAR.

(6) Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional require-ments of systems required to cope with accidents analyzed in the SAR.

Note: For items 2.a(5) and 2.a( 6) reduced re-dundancy that does not result in a loss of system function need not be reported under this section but may be reportable under items 2.b(2) and 2.b(3) below.

(7) Conditions arising from natural or man-made events that, as a direct result of the event require plant shutdown, operation of safety systems, or other protective measures re-quired by technical specifications.

5-13

5.0 ADMINISTRATIVE CONTROLS 5.9.2 Continued (8) Errors discovered in the transient or acci-dent analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could-have permitted reactor operation in a manner less conservative than assumed in the analyses.

(9) Performance of structures, systems, or com-ponents that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report or technical specifications that require remedial action or corrective measures to prevent the existence or develop-ment of an unsafe condition.

Note: This item is intended to provide for report-ing of potentially generic problems.

b. Thirty Day Written Reports. The reportable oc-currences discussed below shall be the subject of written reports to the Director of the appropriate Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event re-port form. Information provided on the licensee event report form shall be supplemented, as needed,

! by additional narrative material to provide com-plete explanation of the circumstances surrounding the event.

( 1) Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.

(2) Conditions leading to operation in a de-graded mode permitted by a limiting con-dition for operation or plant shutdown required by a limiting condition for oper-ation.

5-14

5.0 ADMINISTRATIVE CONTROLS 5.9.2 Continued Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configurations as described in items 2.b(1) and 2.b(2) need not be reported except where test results themselves reveal a degraded mode as described above.

(3) Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems.

(4) Abnormal degradation of systems other than those specified in item 2.a(3) above de-signed to contain radioactivelmaterial re-sulting from the fission process.

Note: Sealed sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in techni-cal specifications need not be reported under this item.

t 5-14a

.' O 5.10 Record Retention 5.10.1 The following records shall be retained for at least five years:

a. Records and logs of facility operation covering time interval at each power level.

b. Records and logs of principal maintenance activi-ties, inspections, repair and replacement of princi-pal items of equipment related to nuclear safety.

c. REPORTABLE OCCURRENCE Reports.

d. Records of ~ surveillance activities, inspections and calibrations required by these Technical Specifi-cations.

e. Records of reactor tests and experiments.

f. Records of changes made to Operating Procedures,

g. Records of radioactive shipments.

h. Records of sealed source leak tests and results,

i. Records of annual physical inventory of all source material of record.

5.10.2 A complete record of the analysis employed in the selec-tion of any fuel assembly to be placed in Region 2 of the spent' fuel racks will be retained as long as.that bundle remains in Region 2 (reference Technical Specifications 2.8(12) and 5.8.4).

5-18 i

.. ,o DISCUSSION These Technical Specification changes are being made to adminis-tratively support an application by the District dated March 12, 1982, to install high density spent fuel storage racks at the Fort Calhoun Station. These racks will be divided into 2 regions; Region 1, which utilizes poisons, and Region 2, which takes credit for fuel burnup to limit reactivity below a specified value. The intent of these Technical Specification changes is to:

(1) Provide an administrative control over the selection of fuel assemblies to be placed in Region 2 to assure these as-semblies meet the necessary criteria for storage in that area. The acceptance criteria will be based on Figure 2-10,

" Spent Fuel Pool Region 2 Storage Criteria",.which will be added to the Technical Specifications by this change. Justi-fication for this figure was presented in the January 21, 1983 submittal for the spent fuel storage rack application.

The statement describing who may perform an independent re-view in Technical Specification 5.8.4 was based on 10 CFR 50, Appendix B, Part III ( Design Control) .

(2) Provide assurance that surveillance requirements for periodic testing of neutron poison samples in the Region 1 spent fuel racks are met.

The following hazards considerations have been made in accordance with 10 CFR 50.92:

(1) The changes make the Technical Specifications more conserva-tive and therefore do not involve an increase in the proba-bility or consequences of an accident previously evaluated.

( 2) The changes will support a safety analysis which is present-ly ongoing. It constitutes an administrative change which makes the Technical Specifications more conservative by ensuring adherence to the safety analysis and therefore will not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) The changes require additional administrative controls in-volving use of the above mentioned spent fuel storage racks.

They will tend to increase the margin of safety rather than reduce it.

ATTACHMENT B a