ML19260A162

Tech Spec Change Request 56 Supporting Licensee Request to Change App a to License DPR-50 Re Spec of Time Limits for Completing Corrective Action for Abnormal Control Rod Driveline Sys Conditions.Certificate of Svc Encl

ML19260A162
Person / Time
Site:	Three Mile Island
Issue date:	05/24/1977
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19260A159	List: ML19260A158 ML19260A162
References
NUDOCS 7910290771
Download: ML19260A162 (8)

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!T"RCPCLITAIT EDISON COMPANY JERSEY CENTRAL PCWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC CCMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Stecification Chance Request No. 56 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. EPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request ,

proposed replacement pages for Appendix A are also included.

METROPOLITAN EDISON CCMPANY By Vice President Sworn and subscribed to me this M day of , 1977.

e Notary Pu kic g

(*'I'

$910 290 7 1480 254

4 U'iITED STATES OF AMERICA IIUCLEAR REGULATORY CCIGIISSIO:I IN THE GTTER OF DOCKET NO. 50-289 LICE:!SE NO. DPR-50 METROPOLITA'i EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No. 56 to Appendix A of the Operating License for "'hree Mile Island Nuclear Station Unit 1, has, on the date given 'celow, been filed with the U. S. Nuclear Regulatory Conmission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addresced as follows: .

Mr. '4eldon B. Arehart Mr. Harr/' B. Reese, Jr.

Board of Supervisors of Board of County Conmissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY By / b iice President Dated: May 2h,1977 1480 255

ihr<e Mile Island nuclear Station Unit 1 (TMI-1)

Operating License Uo. DPR-50 Docket Ko. 50-289, ,

Technical Specification Change Recuest No. 56 The licensee requests that the attached changed pages replace pages 3-33, 34, 3ha, 35 of Appendix A of the existing Technical Specification.

Reason for Proposed Chsnge The reason for this change is to provide uniformity in responding to abnormal CRD System conditions by specifying time limits for completing corrective action. This change vill also eliminate the need for a licensee event report when prompt corrective action is taken.

Safety Analysis Justifying Change The time limits stated in this proposed change are conservative with respect to B&W Standard Technical Specification and they provide assurance that the operating limits of the fuel vill not be exceeded. This proposed change does not involve an unreviewed safety questien in that (i) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased, (ii) the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created; and (iii) the margin of safety defined in the basis for any technical specification is not reduced.

1480 256

3 5.2 co:: TROL 'sCD 0?ct? AliD Po'ER DISTRIBUTIoII LI: CTS Anoli cability This specification applies to power distribution and operation of control rods during power operation.

Objective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion fron a hypothetical control rod ejection, and to assure core suberiticality after a reactor trip.

Specification 3 5.2.1 The available shutdown margin shall not be less than one percent Ak/k with the highest worth control rod fully withdrawn.

3 5 2.2 Operatie- vith inoperable rods:

a. Operation with more than one inoperable rod as defined in Specification h.7.1 and h.7.2.3 in the safety or regulating rod banks shall not be permitted. If more than one rod becomes inoperable, the reactor shall be placed in hot shutdown within four hours. Operation may continue i~f the system is restored to no more than one inoperable rod and operability of the remaining rods is demonstrated within 2h hours by a 2" exercise.

b. If a control rod in the regulating and/or safety rod banks is declared inoperable in the withdrawn position as defined in Specification Paragraph 4.7 1.1 and h.7 1.3, an evaluation shall be initiated immediately to verify the existence of one percent Ak/k hot shutdown margin. Boration may be initiated to increase the available rod worth either to compensate for the vorth of the inoperable rod or until the regulating Tahki~are fully withdrawn, whichever occurs first. Simultaneously, a program of exercising the remaining regulating and safety rods shall be initiated to verify operability.

c. If within one hour of determination of an inoperable rod as defined in Specification h.7.1, it is not determined that a one percent Ak/k hot shutdown cargin exists cisbining the vorth of the inoperable rod with each of the othe rods, the reactor shall be brought to the hot standby condition until this cargin is established.

1480 ?57 3-23

d. Following the determination of an inoperable rod as defined in Specification h.7.1, all rods shall be exercised within 2h hourc and exercised weekly until the rod problem is solved.

e. If a control rod in the regulating or cafety rod groups is declared inoperable per k.7 1.2, pover chall be reduced to 607, of the thermal power allowable for the reactor coolant pump combination within one hour.

p. =

1480 258 3-33a

f. If s control rod in the regulating or axial power shaping groups in deelsred inoperable per Specification h.T.l.2.,

operation may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Specification 4.7.1.2.

g. If the inoperable rod in Paragraph "e" above is in groups 5, o, 7, or 8, the other rods in the group may be trimmed to the same positicn. Normal operation of 100 percent of the thermal power allowable for the reactor coolant pump combination may then continue provided tnat the rod that was declared inoperable is maintained witPla allowable group average position limits in 3.5 2 5 3.5 2.3 The worth of single inserted control rods during criticality are limited by the restrictions of Specification 3.13.5 and the control Rod Position Limits defined in Specification 3.5.2 5 3 5 2.h Quadrant tilt:

a. Except for physics tests the quadrant tilt shall not exceed

+2.665 as datermined using the full incore detector system.

b. When the fu11 incore detector system is not available and except for physics tests quadrant tilt shall not exceed +1.hT5 as determined using the minimum incore detector system.

c. When neither incore detector system above is available and except for physics tests quadrant tilt shall not exceed +0.81%

as determined using the power range channels displayed on the console for each quadrant (out of core detector system).

d. Except for physics tests if quadrant tilt exceeds the tilt limit power shall be reduced immediately to below the power level cutoff (see Figures 3 5-2A, 3 5-2B and 3.5-2C)T "MdFe-over, the power level cutoff value shall be reduced 2 percent for each 1 percent tilt in excess of the tilt limit. For less than four pump operation, thermal power shall be reduced 2 percent of the thermal power allovable for the reactor coolant pump combination for each 1 percent tilt in excess of the tilt limit.

e. Within a period of h hours, the quadrant power tilt shall be reduced to less than the tilt limit except for physics tests, or the following adjustments in setpoints and limits shall te made:

1. The protection c/ stem reactor pover/ imbalance envelope trip setpoints shs11 te reduced 2 percent in pcVer for ea= rercen: t11:.

3-3' 1480 259

2. 7te control rod group withdrawal limits (Figures 3.5-2A, 3.5-2B, 3.5-2C, 3.5-23, 3 5-2E, 3 5-2F, 3.5-2K, 3.5-2L, and 3.5-2M) shall be reduced 2 percent in power for each 1 percent tilt in excesc of the tilt limit.

3. The operational imbalan:e limits (Figure 3.5-2G, 3.5-2H and 3.5-2I) shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.

f. Except for physics or diagnostic testing, i' quadrant tilt is in excess of +25 725 determined using the full incore detector system (FIT), or +2k.09% determined using the minimum incore '

detector system (MIT) if the FIT is not available, or +21 39%

determined using the out of core detector system (0"') when neither the FIT nor MIT are.available, the reactor will be placed in the hot shutdown condition. Diagnosite testing during power operation with c. quadrant tilt is permitted provided that the thermal power allowable is restricted as stated in 3.5.2.h.d above.

g. Quadrant tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15 percent of rated power.

3-3ha 1480 260

m O ~G j 3 5.2.5 Control Rod Fositions: g

[fiY'Ci)i/d

a. Operating rod group overlap shall not exceed 25% +5% between two sequential groups except for physics tests. If overlap exceeds 25% +5%, the reactor shall be placed in hot shutdown within four hours unless the overlap condition is corrected.

b. Position limits are specified for regulating and axial power shaping centrol rods. Except for physics tests or exercising control rods, the regulating control rod insertion /vithdrawal limits are specified on Figures 3 5-2A, 3.5-2B, and 3 5-2C for four pump operation and Figures 3 5-2D, 3.5-2E, and 3.5-2F for three or two pump operation. Also excepting physics tests or exercising control rods, the axial power shuping control rod insertion /vithdrawal limits are specified on Figures 3.5-2K, 3 5-2L, and 3 5-2M. If any of these control rod position limits are exceeded, corrective measures shall be taken immedi-ately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within four hours.

c. Except for physics tests, power shall not be increased above the power level cutoff (See Figures 3 5-2A, 3 5-2B and 3.5-2C) unless the xenon reactivity is within 10% of the equilibrium value for operation at rated power and asymptotically approach-ing stability.

d. Core imbalance shall be monitored on a minimum frequency of once every two hours during power operation above 40% of rated power. Except for physics tests, corrective measures (reduction of imbalance by APSR movements and/or reduction in reactor power) shell be taken to maintain operation within the envelope define'l by Figures 3 5-2G, 3.5-2H and 3 5-2I. If the imbalance is not within the envelope defined by Figures 3 5-2G, 3.5-2H and 3 5-21 corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within four hours, reactor power shall be reduced until imbalance limits are met. _ _ _ . - _

e. Safety rod limits are given in 3.1.3 5 3 5 2.6 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.

3.5 2 7 A power map shall be taken at periodic intervals of 10 to 30 full power days using the incore instrumentation detection system, to verify that the power distribution is within the limits shown in Figure 3. 5-2I.

ssses 1480 261 The power-imbalance envelope defined in Figures 3 5-2G, 3 5-2H, and 3.5-2I is buJad on IOCA ana]yses which have defined the taximum linear heat rate (see sicure 3.5-2J ) such that the taximum clad terperature vill not exceed the

.inal Acceptance Criteria (2200F). Operation outside of the power imbalance envelope alone does not constitute a situstion that would cause the Final Acceptance Criteria to be exceedeJ should s LOCA recur. The power impalance envelope represent; the boundary of opers'. en l'.mited by the Final Acceptance Criteria only 12 the con:rcl rods are at ;he withdrawal /incertion limits as defined by 2-1