ML20033C865
| ML20033C865 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 11/25/1981 |
| From: | TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML20033C855 | List: |
| References | |
| TASK-2.K.2.10, TASK-TM GL-82-16, GL-82-17, GL-82-23, GL-82-76, TAC-45184, TAC-49481, TAC-49482, TAC-49730, NUDOCS 8112040432 | |
| Download: ML20033C865 (9) | |
Text
{{#Wiki_filter:- e ADMINISTRATIVE CONTROLS 6.8_3 Temporary changes to procedures of 6'.8.1 above may be made ~ provided: The intent of the original procedure is not altered. a. The change is approved by two members of the plant management b. staff, at least one of whom holds a Senior Reacter Operator's License on the unit affected. The change is documented, reviewed by the SRB and approved c. by the Station Superintendent within 14 days of implementation? $!YoY Sd7 ATn9Calp w 6.9 rep 0RTING REOUIREMENTS ROUTINE rep 0RTS AND rep 0RTABLE OCCURRENCES. In addition to the applicable reporting requirements of Title 10, 6.9.1 Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted. STARTUp rep 0RT_ A summary report of plant startup and power escalation testing 6.9.1.1 shall be submitted.following (1) receipt of an operating license (2) amendment to.the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, themal, or hydraulic perfor-mance of the plant. The report shall address each of the tests identified in the 6.9.1.2 FSAR and shall include a description of the measured values of the i operating conditiens or characteristics obtained during the test program and a comparison of these values with design predictions and specifica-Any corrective actions that were required to obtain satisfactory Any additional specific details tions. operation shall also be described. required in license conditio cluded in this report. Startup, reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumptio 6.9.1.3 or comencement of commerical power operation, or (3) 9 months folicwing initial criticality, whichever is earliest.' If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or cemencement of cor=ercial 8112040432 811125 PDR ADOCK 05000346 P ryg 6-14 AmendmentNo./,12j unti 1 llUMts,etutt,
4 6.8.4 The following programs shall be established, implemented, and maintained: a. Primary Coolant Sources Outside Centainment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include makeup, letdown, seal injection, seal return, low pressure injection, containment spray, high pressure injection, aaste gas, primary sampling and reactor-coolant drain systems. The program shall include the following: (i) Preventive maintenance and/or periodic visual inspection re-quirements, and (ii) Integrated leak test requiremente for each system at refuel-ing cycle intervals or less. b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions, This program shall include the following: (i) Training of personnel, (ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analysis equipment. DAVIS-BESSE, UNIT 1
~ Docket No. 50-346 License No. NPF-3 Serial No. 757 November 25, 1981 Attachment II I. Change to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications Section 3.3.2.3, 4.3.2.3, Table 3.3-15, 4.3-15 and bases A. Time required to Implement This change is to be effective upon NRC approval and installation-of the ARTS System B. Reason for Change (Facility Change Request 79-283 Rev. A & B) To comply with NUREG 0737 Item II K.2.10 Safety-Grade Trip (Final ARTS) C. Safety Evaluation See Attached l i
Docket No. $@;M5@I License No. NPF-3 Snrial No.757 November 25, 1981 SAFETY EVALUATION This License Amendment Request (LAR) provides new Technical Specifications for " Limiting Condition for Operation" and surveillance requirements for the safety grade Anticipatory Reactor Trip System (ARTS). This LAR also calls for withdrawing the original submittal (July 13, 1979 Serial 527) of Technical Specification for the control grade interim ARTS. The previous submittal has not been acted upon by the NRC and since the control grade ARTS will be superceded by the safety grade ARTS following the 1982 refueling outage, further action on the previous submittal is not required and this withdrawal does not involve an unreviewed safety question. Without the institution of ARTS in the facility, when the main turbine trips at reactor power greater than 25% of full power, the.ICS will initiate a reactor power runback and the unit should stabilize when the reactor power reaches a level less than 25% full power. However, if the ICS is not properly tuned, high RCS pressure may occur causing a RPS reactor trip. The institution of ARTS,_in this case, is to prevent the ICS from challenging the RPS. In the event that both of the main feed pump turbines are lost, the ARTS will anticipate the problem and trip the reactor before the RCS reaches the RPS high pressure trip setpoint. ARTS is not required for the case when the main turbine trips at reactor power less than 25% full power since the turbine bypass valves are designed to fully handle this condition without challenging the RPS for tripping the reactor or without changing the reactor power or turbine header pressure. The installation of Arts will not degrade any safety functf a of the Reactor Protection System or SFRCS since ARTS is not intenden to replace RPS or SFRCS. The safety grades ARTS will replace all of the functions of the control grade ARTS. The new Technical Specifications are considered to be adequate to demonstrate and ensure the continued operability of ARTS. The surveillance requirement specified for this system ensure that the overall system functional capability is maintained and the periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. Pursuant to the above, it is concluded that the proposed Technical Specifications changes do not involve an unreviewed safety question.
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O.. :. t - a y INSTRUMENTATION ANTICIPATORY REACTOR TRIP SYSTDI INSTRUMENTATION ~ LIMITING CONDITION FOR OPERATION 3.3.2.3 The Anticipatory Reactor Trip System instrunentation channels of Table 3.3-15 shall be OPERABLE. APPLICABILITY: As shoOn in Table 3.3-15 A cTioM : As S H oto M 1Q T AS L. E 3 3-15" SURVEILLANCE REQUIREMEh'IS 4.3.2.3 The Anticipatory Reactor Trip System shall be deconstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAt TEST during the = odes and at the frequencies shown in Table 4.3-15. e b M S 4 e e 4 4 9
.r TABLE 3.3-15 ANTICIPATORY REACTOR TRIP SYSTEM INSTRUMENTATION FUNCTIONAL TOTAL NO. CHANNELS MINIMUM APPLIC-UNIT OF TO CHANNELS ABLE CHANNELS TRIP OPERABLE MODES ACTION.
- 1. Turbine Trip 4
2 3 1* 16
- 2. Main Fead Pump Turbine Trip 4
2 3 1,2 17
- 3. Output Logic 4
2 3 1,2 18 ACTION 16 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirements, restore the inoperable channel to OPERABLE status within 72 hours or reduce reactor power to less than 25% full power within the next 6 hours. ACTION 17 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLL requirements, restore the inoperable channel to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours. ACTION 18 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed provided bett. of the following conditions are satisfied: a) The control rod drive trip breaker associated with the inoperable channel is placed in the tripped condition within one hour. b) The Minimum Channels OPERABLE requirement is met; however, one additional control rod drive trip breaker associated.with another channel may be tripped for up to 2 hours for surveillance testing per Specification 4.3.2.3, after reclosing the control l rod drive trip breaker opened in a) above.
- Applicable only above 25% reactor power.
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a e a TABLE 4.3-15 ANTICIPATORY REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CilANNEL MODES *IN WilICH CIIANNEL CilANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT CllECK CALIBRATION TEST REQUIRED 1. Turbine Trip S N.A. M 1* l l 2. Main Feed Pump Turbine Trip S N.A. M 1,2 l 3. Output Logic N.A. N.A. M 1,2 l O Applicable only above 25% reactor power d l
.I i Bases 3/q.3. d 3 A NT1CtPATORY 'EEAcTOR Trip Sysreu (ARTS) De-OPERA Bt UTY of th-c Anh'cipatory Reacter Tytp Systein insivumentafion ensures that I) the. a.ssocictied yeach>y -fvip acfion tdt ll be. inikate_d k*en Yhe rvuin futbine -fvig (abooe 45[ nil power) oY hofh y fhe. n1ain y. d fumf }urbines 'lYiy
- 2) Se Sf<cified coincidence. logic i.s M4inYained 3) suffic'nent redundancy is ataintained 1~o perrnilen channci
' ?* be ou~$~ of.Sevoice, pr +estiny or maintenance.. he OPERAbtUTY cf +his system i.s repired 15 minim 1 e. -fin ivnpact of 3 ffallsjenfs in ihs secondary sysicms on the prirnary Sy'hm. Idi+houf* ihe AR.T5 When fhe rnaid lurbine fri s af reactor power greater lhor1 asp f full ower', lhe Ic5 will inif(ate ct runback of reoclor pou>cr. The unit p should stabili e when reacfor power reaches ca. level less han RS/ 3 full power. However, if res is not properly luned, hiyh Rcs fressure, rnay occur during se, potoer runback. causing ct reac/or hip. The instifulien of ARTS. in 4his ca.se., is to prevent rne. 2cs pm challenging ne RPs. AP-Ts is not repire.d pr the when the main furbine 6ips case at rencioY foWer fcGs hh an .2 6 {-ufi power S nce kh4 & Thlns h pass valves are de sig ned ib fully handle fhis condilion wi1houl-3 challenging Tne R.Ps or wilhout ct reac+or potuei change.or Wi%out a bbine h<ader pre.ssure c.hange-Ys.L eVenf of 0- loss of hofh MQin f<ed pu kQYbines, We hRTS n WiX onticipate, fhe frcblew1 and will ivs](ate. reacfor frip for fo hi h* pressure + rip seipolnb Res reoching +he. 9
Bases The syywjllaMCe. vcylvemenf5 specified fvfhis Syshm ensum ht k overall 9 stem pnchbeal capabiliG l.s maintalned cornfaraN i% OYI IMQ! dfSljn SfcndardS The pviod[C SUYNil{QMCc it.5f4 f performe.d at she minimum f19aencie.s an sufftelent to demonsfrafe. 4his cafabildy. e 9 i l l I _}}