ML20127B840
| ML20127B840 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/08/1993 |
| From: | Robert Stransky Office of Nuclear Reactor Regulation |
| To: | Dovach T COMMONWEALTH EDISON CO. |
| References | |
| TAC-M84992, TAC-M84993, NUDOCS 9301130153 | |
| Download: ML20127B840 (5) | |
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January 8, 1993 Ills 181tNLIM:
Docket files JRoc Docket flos. 50-373 f4RC & Local PDRs IKing and 50-374 PDill-2 r/f JDyer LMoore RStransky fir. Ihomas J. Kovath OGC Dilagan finclear Licensing Manager Gilill (8)
WJones Lommonwealth Edison Company-Suite 300 LGrimes ALRS (10)
OPUS West til DPA OL/llDLD 1400 OPUS Plate PDill-2 p/f Bllayton, Rll!
Downers Grove, Illinois 60515
Dear Mr. Kovath:
SUBJill:
l ASAl t ! LOUtilY $1 Alluti, UtillS 1 AllD 2 - LHAf1GE 10 llClifill Al SPltifICA110fl DAhlS $10110f1 3/4.1.3 (LAC fl05. M84992 AllD M84993) 11y let t er dated heptember 16, 1992, the Lommonwealth (dison Company (L100) pt oposed to thanqn l a5alle Lounty st ation, Units 1 and 2, lechnical
$pocification (15) Hases 3/4.1.3.
The proposed change would correct the bases to reflett a modift(ation to the control rod drive (CRD) support structures of both units.
In particular, you plan to lower the CRD support structure by approximately 0.5 inches, in order to increase the clearance between the LRD housings and the CRD support structure.
This modification would improve access by personnel working under the reactor vessel, thus decreasing the dose to these workers.
The st af f has reviewed your proposed change to the 15 Bases, as well as the associated 10 LIR 50.59 evaluation concerning the modif ication to the CRD support structure, and agrees that the proposed change to 15 Bases 3/4.1.3 is appropriate.
intlosed is a topy of revised Bases page li 3/41-3 f or both the Unit I and 2 lechnical Specifications.
If you have any questions regarding this change, please contact me at (301) 504-1346.
5interely, Omo :.1 ?%
Robert J. Stransky, Project Manager Pro.iect Directorate 111-2 Division of Reactor Projects lil/IV/V Offire of fluclear Reactor Regulation l oc l osures.
Unit 1 Bases page 11 3/4 1-3 Unit 2 Bases page B 3/4 1-3 cc w/ enclosure:
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1 January B, 1993 DISTRIBUTION:
Docket files JRoe Docket Nos. 50-373 NRC & Local PDRs TKing and 50-374 PDill-2 r/f JDyer CHoore RStransky Mr. Thomas J. Kovach OGC DHagan Nuclear Licensing Manager GHill (8)
WJones Commonwealth Edison Company-Suite 300 CGrimes ACRS (10)
OPUS West 111 OPA OC/LFDCB 1400 OPUS Place PDill-2 p/f BClayton, Rill Downers Grove, Illinois 60515
Dear Mr. Kovach:
SUBJECT:
LASALLE COUNTY STATION, UNITS 1 AND 2 - CHANGE TO TECHNICAL SPEclflCATION BASES SECTION 3/4.1.3 (TAC N05. M84992 AND M84993)
By letter dated September 16, 1992, the Commonwealth Edison Campany (CECO) proposed to change LaSalle County Station, Units 1 and 2, R.
ical Specification (IS) Bases 3/4,1.3.
The proposed change woulo correct the bases to reflect a modification to the control rod drive (CRD) support structures of both units.
In particular you plan to lower the CRD support structure by approximately 0.5 inches, in order to increase the clearance between the CRD housings and the CRD support structure.
This modification would improve access by personnel working under the reactor vessel, thus decreasing the dose to these workers.
The staff has reviewed your proposed change to the TS Bases, as well as the associated 10 CFR 50.59 evaluation concerning the modification to the CRD support structure, and agrees that the proposed change to TS Bases 3/4.1.3 is appropriate.
Enclosed is a copy of revised Bases page B 3/4 1-3 for both the Unit 1 and 2 Technical Specifications, if you have any questions regarding this change, please contact me at (301) 504-1346.
Sincerely, Cripr.3: Eigr4x y Robert J. Stransky, Project Manager Project Directorate 111-2 Division of Rea: tor Projects lil/IV/V Office of Nuclear Reactor Regulation
Enclosures:
Unit 1 Bases page B 3/4 1-3 Unit 2 Bases page B 3/4 1-3 cc w/ enclosure:
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k Mr. Thomas J. Kovach LaSalle County Station Commonwealth Edison Company Unit Nos. I and 2 CC:
7 Phillip P. Steptoe, Esquire Robert Cushing Sidley and Austin Chief, Public Utilities Division One first National Plaza Illinois Attorney General's Office Chicago, Illinois 60603 100 West Randolph Street Chicago, Illinois 6060)
Assistant Attorney General 100 West Randolph Street Michael I. Miller, Esquire Suite 12 Sidley and Austin Chicago, Illinois 60601 One first National Plaza Chicago, Illinois 60690 Resident inspector /LaSalle, NPS U. S. Nuclear Regulatory Commission Mr. G. Diederich Rural Route No. 1 LaSalle Station Manager P. O. Box 224 LaSalle County Station Marseilles, Illinois 61341 Rural Route 1 P. O. Box 220 Chairman Marsellies, Illinois 61341 LaSalle County Board of Supervisors LaSalle County Courthouse Ottawa, Illinois 61350 Attorney General 500 South 2nd Street Springfield, Illinois 62701 Chairman Illinois Commerce Commission Leland Building 527 Cast Capitol Avenue Springfield, Illinois 62706 Illinois Department of Nuclear Safety Office of Nuclear facility Safety 1035 Outer Park Drive Springfield, Illinois 62704 Regional Administrator, Region _ Ill V. S. Nuclear Regulatory Commission 799 Roosevelt Road, Bldg. #4 Glen-Ellyn, Illinois 60137 Robert Neuman Office of Public Counsel State of Illinois Center 100 W. Randolph l
Suite 11-300 Chicago, Illinois 60601 i
l l
REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued)
In addition, the automatic CRD charging water header low pressure scram (see Table 2.2.1-1) initiates well before any accumulator loses its full capa-bility to insert the control rod.
With this added automatic scram feature, the surveillance of each individual accumulator check valve is no longer necessary to demonstrate adequate stored energy is available for normal scram action.
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR.
The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod drive coupling integrity.
The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and there-fore that other parsmeters are within their limits, the control rod position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a control rod to less than 3.65 inches in the event of a housing failure.
The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system.
The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components, 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident.
The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm.
Thus requiring the RhH to be OPERABLE when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER provides adequate control.
The Rhh provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.
-LA SALLE - UNIT 1 B 3/4 1-3 Letter Dated: January 8, 1993
4 REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued)
In addition, the automatic CRD charging water header low pressure scram (see Table 2.2.1-1) initiates well before any accumulator loses its full capa-bility to insert the control rod.
With this added automatic scram feature, r
the surveillance of each individual accumulator check valve is no longer necessary to demonstrate adequate stored energy is available for normal scram action.
Control rod coupling integrity is required to ensure compliance with the l
analysis of the rod drop accident in the FSAR.
The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod drive coupling integrity.
The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control-rod position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a control rod to less than 3.65 inches in the event of a housing failure.
The amount of l
rod reactivity which could be added by this small amount of rod-withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system.
The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 CONTROL R0D PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure i
that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident.
The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawel.
When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there-is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm.
Thus requiring the RWM to be OPERABLE when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER provides edequate control.
The RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.
LA SALLE - UNIT 2 B 3/4 1-3 Letter Dated: January 8, 1993
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