ML20151R845
| ML20151R845 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/05/1988 |
| From: | Shelton D TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML20151R843 | List: |
| References | |
| NUDOCS 8808120320 | |
| Download: ML20151R845 (13) | |
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--Lig:nsa No..NPP-3 Ssrial Na. 1492' Enclosure Page 1 APPLICATION FOR AMENDMENT TO FACILITY OPF2ATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT'NO. 1 Attached are the requested changes to the Davis-Besse Nuclear Power Station, Unit No. 1 Facility Operating License No. NPF-3. Also included are the Safety Evaluation and Significant Hazards Consideration.
The proposed changes (submitted under cover letter Serial No. 1492) concern Section 3/4.3.3, Instrumentation, Remote Shutdown Instrumentation, Specification 3.3.3.5, Table 3.3-9, Remote Shutdown Monitoring Instrumentation, Instrument 7; Section 3/4.3.3, Instrumentation, Remote Shutdown Instrumentation, ;
Specification 4.3.3.5,-Table 4.3-6, Remote Shutdown j tionitoring Instrumentation Surveillance Requirements, j Instrument 5 and 7; 1
Section 3/4.3.3,. Instrumentation, Remote Shutdown Instrumentation, j Specification 4.3.3.5, Table 4.3-6, Remote Shutdown j Monitoring Instrumentation Surveillance Requirements, Footnote "*".
By D. C. Shelton, Vice President Nuclear l
Suorn to and subscribed before me this 5th day of August, 1988 M ,
No ary Public, State of Ohio JUDITH HIRSCH Notary Pelic Stete of Ohio lfy Commusion Ensires June 30.1992 gDR808120320 880805 -
p ADOCK 05000346 FDC j ,
Dockst Nc. 50-346 Lietnas No. NPF-3 S2 rial No. 1492 Enclosure Page 2 The following information is provided to aupport issuance of the requested changes (Attachment 3) to the Davis-Besse Nuclear Power Station, Unit No. 1 Operating License No.~ NPF-3, Appendix A, Technical Specifications, Specification 3.3.3.5 Table 3.3-9 and Specification 4.3.3.5 Table 4.3-6.
A.
Time Required to Implement: These changes will be implemented by the licensee within 45 days following NRC issuance of the License Amendment.
B.
Reason for Change (Facility Change Request (FCR) 87-0001): Revise the Technical Specifications to properly reference the Control Rod Position Switches for remote shutdovn instrumentation and delete the requirement perform o to
'hannel calibration of these switches every eighteen months.
C. Safety Evaluation:
See attached Safety Evaluation (Attachment No. 1).
D. Significant Hazards Consideration: See attached Significant Hazards Consideration (Attachment No. 2).
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,' , Dockat No. 50-346 LicInss No. NPF-3
. S; rial No. 1492 Attachment 1 Page 1 SAFETY EVALUATION Description of Proposed Activity General Design Criterion 19 of 10 CFR Part 50 Appendix A requires, in part, that "Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdovn of the reactor through the use of suitable procedures." Technical Specification 3.3.3.5 "Remote Shutdown Instrumentation" specifies remote shutdovn instrumentat!.on required in order to bring the station to a safe shutdown condition if control room habitability is lost. In order to ensure compliance with this Technical Specification, Surveillance Requirement 4.3.3.5 requires, in part, a channel check (monthly) and channel calibration (18 months) of the Control Rod Position Limit Switches. The Control Rod Position Limit Switches are stated to have a readout at System Logic Cabinet #4 of the Control Rod Drive Control System (CRDCS) in Tables 3.3-9 and 4.3-6 of the Technical Specifications. The indication at this cabinet, however, is actually provided by the Control Rod Position Switches, rather than the Position Limit Svitches.
The Heasurement Range attributed to the Position Limit Switches (0%, 25%, 50%,
75%, and 100%) actually applies to the Position Switches. In addition, the design of the Position Switches, as well as the Position Limit Switches, is such that calibration is not required following initial installation. Therefore, FCR 87-0001 proposes Tables 3.3-9 and 4.3-6 of Technical Specification 3/4.3.3.5 be revised, in order to correctly refer to the Control Rod Position Switches, rather than the Position Limit Svitches, and to delete the requirement for a channel calibration every eighteen months. In addition, a note allowing the extension of the due date for channel calibration of the steam generator outlet steam pressure indication at the Aux. Shutdown panel vill be deleted. The channel calibration was due on May 17, 1983 and the extension note allowed delay until September 17, 1983. This extension has expired and, thus, is no longer applicable. This Safety Evaluation is being prepared to evaluate the proposed amendment to Technical Specification 3/4.3.3.5 "Remote Shutdown Monitoring Instrumentation."
Systems Affected This FCR proposes a revision to Technical Specification 3.3.3.5. No physical modification to the station vill be made as a result of this change. The portions of the Technical Specification to be revised apply to the ContrF ' 1 Drive System (S.U.S. 55.01) and the Main Steam System (S.U.S. 83.01).
Applicable Documents Technical Specifications for Davis-Besse Nuclear Power Station Unit No. 1, Section 3.1.3.1 "Movable Control Assemblies-Group Height-Safety and Regulating
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Dockst No. 50-346 Licznsa No. NPF-3 S2 rial No. 1492 Attachment 1 Page 2 Groups," Section 3.1.3.2 "Group Height-Axial Power Shaping Rod Group,"
Section 3.1.3.4 "Rod Drop Time," Section 3.1.3.5 "Safety Rod Inse:: tion Limit,"
Section 3.1.3.6 "Regulating Rod Insertien Limits," Section 3.1.3.9 "Axial Power Shaping Rod Insertion Limits," Section 3.3.3.5 "Remote Shutdown Instrumentation."
Updated Safety Analysis Report for Davis-Besse Nuclear Power Station Unit No. 1, Appendix 3D "Conformance with NRC Generai Design Criteria, Safety Guides, and Information Guides," Section 7.4.1.1 "Systems Required for Safe Shutdovn-Control Rod Drive Control System (CRDCS)-Trip Portion," Section 7.7.1.3 ' Control Systems "RDCS-Vithout Trip Portion."
10 CFR 50 Appendix A-General Design Criterion for Nuclear Power Plants, Criterion 19 "Control Room."
Regulatory Guide 1.68.2, 1978 "Initial Startup Test Program to Damonstrate Remote Shutdown Capability for Vater-Cooled Nuclear Pover Plants."
B&V Specification 1086/NSS-14, "System Logic for Roller Nut Control Rod Drives."
Diamond Power Vendor Manual, "Volume II-Iastruction Manual for Control Rod Drive Control System."
Safety Function of Systems Affected The safety function of the trip portion of the Control Rod Drive Control System (CRDCS) is to trip the shim-safety Control Rod Drive Mechanisms (CRDHs) whenever it receives an automatic trip command from the RPS or Anticipatory Reactor Trip system (ARTS) or a manual trip command from the operator. The trip portion of the CRDCS is the only aspect of the CRDCS which affects public safety. The CEDCS trip logic is designed such that the removal of power to the CRDMs results in a free-fall gravity insertion of the control rods. Two independent trip methods are provided, in series, for removal of power to the trip mechanisms.
The primary method of CRDCS trip interrupts the three-phase AC power to the CRDM motor power supplies. Two, three-pole, metal clad power circuit breakers, equipped with instantaneous undervoltage coils and shunt trip devices, are provided in series in each of the parallel power circuits that feed the CRDH motors. Deenergization of the undervoltage coil or the shunt trip relay results in the breaker opening. The second trip method interrupts the gate control signals to the Silicone Controlled Rectifiers (SCRs) in each of the nine CRDH motor power supplies, and the motor return power supply. Ten relays, connected with their coils in parallel, are provided in each of the redundant halves of the power supplies. Contacts of these relays interrupt the gate control signals to the SCRs in each power supply which remover all power at the outputs of the motor power supplies. All CRDCS components influencing trip action are designed such that trip action is neither prevented nor delayed during or following a safe shutdown earthquake. No trip bypasses or interlocks are provided in the trip circuits.
The function of the non-trip portion of the CRDCS is to provide for vithdrawal
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Dockst No. 50-346 LicInsa No. NPF-3 Sarisl No. 1492 Attachment 1 Page 3 and insertion of groups of Control Rod Assemblies (CRAs) to produce the desired reactor output. The CRAs are arranged in seven groups: four rod groups function as safety groups and three rod groups function as regulating rod groups. The safety groups are normally fully withdraun when the reactor is operating at power, with the regulating rods serving as the principal reactivity control medium. During startup, safety rods i through 4 are withdrawn to their "outlimit", which enables withdraval of regulating group 5. Once group 5 reaches 75% vithdrawal, as indicated by the Rod Position (or zone reference)
Svitches, group 6 vill be enabled and can be withdrawn. Group 7 can be withdrawn when group 5 is at its "outlimit", as indicated by the Rod Position Limit Svitches, and group 6 reaches 7N.' withdrawal. Upon insertion of regulating group 7 to a 25% vithdrawn position, group 6 can be inserted.
Similarly, group 5 can be inserted when group 6 has been inserted to 25%
vithdravn. Eight axial power shaping rod assemblies are provided to accommodate local reactivity perturbations. The CRDCS may be operated automatically by the Integrated Control System (ICS) or manually by the operator. A requirement for continuous boron control by the ICS is the withdraval of the Safety Groups to their "outlimit" and 25% or greater withdrawal of regulating group 5.
The non-trip portion of the CRDCS consists of two basic components: the motor control system and the system logic. This portion of the CRDCS is not Nuclear Safety Related. The Chapter 15 Accident Analysis demonstrates the adequacy of protection systems to cope with CRDCS malfuncticns. The motor control system comprises eight group power supplies and one auxiliary power supply. In order to initiate rod movement, the power supply sequentially energizes first two, then three, then two of the six CRDH motor-stator vindings in a stepping fashion to produce a rotating magnetic field for the CRDH motor to position the CRA.
Switching must is achieved by gating SCRs on for the period of time that each vinding be energized. Gating signals for the SCRs are generated by a motor-driven programmer consisting of 60 Hz synchronous mo'n- driving a multichannel photo optic encoder. Command signals to posiso.n the CRAs are introduced at the Programmer Motor. The system logic of the CRDCS contains those functions that control rod motion and functions which monitor system operation. Failures which could result in improper system operation are continuously monitored by fault detection circuits. Fault detectors are provided to indicate sequence faults, safety rods not fully withdrawn, and a programmer lamp fault. In the CRDCS, two methods of position indication are provided: relative and absolute. The relative position transducer consists of a small pulse-stepping motor, driven from the rod drive motor power supply, coupled to a potentiometer. The absolute position transducer consists of a series of 48 equally spaced reed switches, enclosed in a fiberglass housing and mounted on the outside of the upper motor tube of each CRDM. A magnet is mounted on the CRDM torque taker, and moves parallel to the reed switches as the CRA is moved. This magnetic field actuates the reed switches when the magnet passes in the vicinity of each switch. The reed switches are connected to a resistive network, which provides an analog voltage output proportional to the position of the CRDM leadscrev due to the switches sequentially opening and closing. Either this absolute position indication or the relative position indication may be displayed on the 61 single rod position meters located in the control room. Certain reed switches in the absolute position transducer provide specific signals utilized in the CRDCS logic. The Control Rod Position limit Svitches provide "inlimit" and "outlimit"
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Dockat No. 50-346 Licsnsa No. NPP-3 S, rial Ns. 1492 Attachment 1 Page 4
- signals to the CRDCS.
These signals are utilized to ensure regulating rods are not withdrawn until the safety group rods are fully withdrawn, prohibit further invard motion of a rod group that is fully inserted, allov automatic control by the ICS, and are indicated on the Diamond (Rod Control) panel. The Control Rod Position (ur zone reference) Switches provide signals corresponding to 0%, 25%,
50%, 75%, and 100% rod withdrawal. These signals are utilized in sequencing rod insertion or withdraval, and are indicated at System Logic Cabinet #4 of the CRDCS.
No adjustment, other than raising or lovering the complete absolute position transducer, $s provided for the Control Rod Position Switches or the Control Rod Position Limit Switches.
Effects on Safety This Safety Evaluation proposes that Table 3.3-9 of Technical Specification 3.3.3.5 and Table 4.3-6 of Surveillance Requirement 4.3.3.5 be revised to refer to the Control Rod Position Switches, rather than the Control Rod Position Limit Switches, months.
and to delete the requirement for channel calibration every eighteen Changing the reference to the Position Switches vill correctly identify the instrteents and vill make the Technical Specifications consistent with the Vendor manual and other site documentation. The Position Iimit Svitches have an indication on the Diamond Panel, but no remote readout is provided. The Control Rod Position Limit Switches are stated to have a readout at System Logic Cabinet #4 of the Control Rod Drive Control System (CRDCS) in Tables 3.3-9 and 4.3-6 of the Technical Specifications. The indication at this cabinet, however, is actually provided by the Control Rod Position Switches, rather than the Position Limit Svitches. The Measurement Range attributed to the Position Limit Switches (0%, 25%, 50%, 75%, and 100%) actually applies to the Position Switches.
Therefore the tables should be revised to refer to the Control Rod Position Svitches. Surveillance Requirement 4.3.3.5 requires a calibration of the Control Rod Position Limit Switches every eighteen months. The design of the Position Limit Svitches, as well as the Position Svitches, is such that calibration is not required following ini.tial installation. The only adjustment possible the CRDM.
is lovering or raising the fiberglass tube containing the switches at verifying that Operability of the Position Switches is presently established by the 0% position lights indicate when the control rods are fully inserted, and the remaining position lights (25%, 50%, 75%, and 100%) indicate as the CRAs are withdrawn (ST 5013.02). The reed switches are permanently affixed within the fiberglass tube and their relative position cannot be changed. Therefore, verification of the 0% position switches satisfies the channel check and channel calibration required by Surveillance Requirement 4.3.3.5.
The calibration requirement is not necessary, however, as no adjustment is actually required. Based upon these considerations, changing the instrument reference from Control Rod Position Limit SwitcLes to Control Rod Position Switches in Tables 3.3-9 and 4.3-6 of the Technical Specifications vill not affect the safety function of any system or component.
The note in Table 4.3-6 which allows an extension of the due date for channel calibration of the Steam Generator Outlet Steam Pressure indication at the Aux Shutdovn panel is obsolete, as the evtension date has elapsed. Therefore, deleting the nov obsolete extension note vill have no effect on the safety
Dockst No. 50-346 Lictna No. NPF-3 Serial No. 1492 Attachment 1 Page 5 function of any system or component since the calibration is performed in compliance with surveillance requirements per 4.3.3.5.
Unreviewed Safety Question Evaluation l
Per 10CFR50.59, the following considerations are addressed to determine whether an unreviewed safety question exists as a result of the proposed change:
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j The proposed change does not increase the probability of occurrence of an l accident as previously evaluated in the USAR. The proposed changes to the i Technical Specifications vill correct the instrument descriptions to correspond to the Vendor Manual and other site documentation. The Channel Calibration requirement for the Control Rod Position Switches may be deleted, as no adjustment is necessary following initial installation. These changes vill not l degrade safety. (10CFR50.59(a)(2)(i))
The proposed change does not increase the consequences of an accident as previously evaluated in the USAR. The proposed changes to the Technical
, Specifications Function are clerical of any system in nature and vill have no affect on the Safety or component.
(10CFR50.59(a)(2)(1))
The proposed change does not increase the probability of occurrence of a 1 malfunction of equipment important USAR. to safety as previously. evaluated in the The proposed changes to the Technical Specifications vill not degrade the trip portion of the Control Rod Drive System or any other safety system.
(10CFR50.59(a)(2)(1))
The proposed change does not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the USAR. The proposed changes to the Technical Specifications vill correct an instrument reference and delete an unnecessary calibration requirement for the Control Rod Position Svitches. (10CFR50.59(a)(2)(i))
The proposed different typa change does not create the possibility of an accident of a than previously evaluate:d in the USAR. The proposed changes to thecomponent.
or Technical Specifications vill not affect the safety function of any system (10CFR50.59(a)(2)(ii)) I The proposed change does not create the possibility of a malfunction of a different type than previously evaluated in the USAR.
No new adverse environment vill be created by the clerical changes proposed to the Technical Specifications.
The for anyproposed change Technical does not reduce the margin of safety as defined in the bases Specification.
The basis for Technical Specification 3.3.3.5 ensures that the station can be brought to a safe shutdown condition from outside the control room.
Evaluation Technical vill not affect the margin of Specification. safety for this, or any other,The) clerica (10CFR50.59(a)(2)(iii))
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-Dockat-No. 50-346 Lic'nsa No. NPF-3 Striil Na. 1492 Attachment 1 Page 6 Conclusion Based upon t.he above considerations, it is. concluded that the proposed. change does not involve an unreviewed safety question.
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Dockst No. 50-346 Lic nsa No. NPF-3 Smrial No. 1492 Attachment 2 Page 1 SIGNIFICANT HAZARDS CONSIDERATION Description of Proposed Activity The purpose of this Significant Hazards Consideration is to review and evaluate proposed changes to Technical Specification 3/4.3.3.5 Tables 3.3-9 and 4.3-6.
These proposed changes properly reference the Control Rod Position Switches and delete the requirement to perform a channel calibration of these switches due to their non-adjustability. In addition, a proposed administrative change to delete a footnote for which the applicability of the time frame has expired is also included.
General Design Criterion 19 of 10 CFR Part 50 Appendix A requires, in part, that "equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdovn of the teactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures." Technical Specification 3.3.3.5 "Remote Shutdown Instrumentation" specifies remote shutdown instrumentation required in order to bring the station to a safe shutdown condition if control room habitability is lost. In order to ensure compliance with this Technical Specification, Surveillance Requriement 4.3.3.5 requires, in part, a channel check (monthly) and channel calibration (18 months) of the Control Rod Position Limit Svitches. The Control Rod Position Limit Svitches are stated to have a readout at System Logic Cabinet #4 of the Control Rod Drive Control System (CRDCS) in Tables 3.3-9 and 4.3-6 of the Technical Specifications. The indication at this cabinet, however, is actually provided by the Control Rod Position Switches, rather than the Position Limit Svitches.
The Measurement Range attributed to the Position Limit Svitches (0%, 25%, 50%,
75%, and 100%) actually applies to the Position Switches. In addition, the design of the Position Svitches, as well as the Position Limit Switches, is such that calibration is not required following initial installation. Therefore, FCR 87-0001 proroses Tables 3.3-9 and 4.3-6 of Technical Specification 3/4.3.3.5 <
be revised, in order to correctly refer to the Control Rod Position Switches, j rather than the Position Limit Svitches, and to delete the requirement for a '
channel calibration every eighteen months. In addition, a note allowing the extension of the due date for channel calibration of the steam generator outlet i steam pressure indication at the Auxiliary Shutdown panel vill be deleted. The l channel calibration was due on May 17, 1983 and the extension note allowed delay until September 17, 1983. The extension has expired and, thus, is no longer applicable. (This note deletion was previously proposed in Serial No. 1407, dated November 2, 1987.)
Systems Affected
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'This license amendment request proposes a revision to Technical Specification 3.3.3.5. No physical modification to the station vill be made as {
j a result of this change. The portions of the Technical Specification to.Le i revised apply to the Control Rod Drive System and Main Steam System. '
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Docket No. 50-346 Lic nsa No. NPP-3 Sariel No. 1492 Attachment 2 Page 2 Applicable Documents Technical Specifications for Davis-Besse Nuclear Power Station Unit No. 1, Section 3.1.3.1 "Movable Control Assemblies-Group Height-Safety and Regulating Groups," Section 3.1.3.2 "Group Height-Axial Pover Shaping Rod Group,"
Section 3.1.3.4 "Rod Drop Time," Section 3.1.3.5 "Safety Rod Insertion Limit,"
Section 3.1.3.6 "Regulating Rod Insertion Limits," Section 3.1.3.9 "Axial Power Shaping Rod Insertion Limits," Section 3.3.3.5 "Remote Shutdown Instrumentation."
Updated Safety Analysis Report for Davis-Besse Nuclear Power Station Unit No. 1, Appendix 3D "Conformance with NRC General Design Criteria, Safety Guides, and Information Guides," Section 7.4.1.1 "Systems Required for Safe Shutdown-Control Rod Drive Control System (CRDCS)-Trip Portion," Section 7.7.1.3 "Control Systems-CRDCS-Vithout Trip Portion."
10 CFR 50 Appendix A-General Design Criteria for Nitclear Power Plants, criterion 19 "Control Room."
Regulatory Guide 1.68.2, 1978, "Initial Startup Test Program to Demonstrate Recote Shutdown Capability for Vater-Cooled Nuclear Pover Plants."
B&V Specification 1086/NSS-14, "System Logic for Roller Nut Control Rod Drives."
Diamond Power Vendor Manual, "Voltme II-Instruction Manual for Control Rod Drive Control System."
Safety Function of Systems Affected The sr.fety function of the trip portion of the control Rod Drive Control System (CRDCS) is to trip the shim-safety Control Rod Drive Mechanisms (CRDMs) whenever it receives an automatic trip command from the Reactor Protection System (RPS) or Anticipatory Reactor Trip System (ARTS) or a manual trip command from the operator. !
The trip portion of the CRDCS is the aspect of the CRDCS which affects safety. The CRDCS trip logic is designed such that the removal of power 1 to the CRDMs results in a free-fall gravity insertion of the control rods. Two independent trip methods are provided, in series, for removal of power to the trip mechanisms.
The primary method of CRDCS trip interrupts the three-phase AC power to the CRDM motor power supplies. Two, three-pole, metal clad power circuit breakers. equipped with instantaneous undervoltage coils and shunt trip !
devices, are provided in series in each of the parallel pover circuits that feed the CRDM motors. t Deenergization of the ur.dervoltage coil or the shunt trip relay results in the breaker opening. The second trip method interrupts the gate control signals to the Silicone Controlled Rectifiers (SCRs) in each of the l nine CRDM motor power supplies, and the motor return power supply. Ten relays, !
connected with their coils in parallel, are provided in each of the redundant halves of the power supplies. Contacts of these relays interrupt the gate control signals to the SCRs in each power supply which removes power at outputs of the motor power supplies. CRDCS components influencing trip the action are designed such that trip action is neither prevented nor delayed during or following a safe shutdown earthquake. No trip bypasses or interlocks are provided in the trip circuits.
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. S: rial Ns. 1492 Attachment 2 Page 3 The function of the non-trip portion of the CRDCS is to provide for withdrawal and insertion of groups of Control Rod Assemblies (CRAs) to produce the desired reactor output. The CRAs are arranged in seven groups: four rod groups function as safety groups and three rod groups function as regulating rod groups. The safety groups are normally fully withdrawn when the reactor is operating at power, with the regulating rods servinF as the principal reactivity control medium. During startup, safety rod groups 1 through 4 are withdrawn to their "outlimit", which enables withdraval of regulating group 5. Once group 5 reaches 75% vithdrawal, as indicated by the Control Rod Position (or zone reference) Svitches, group 6 vill be enabled and can be withdrawn. Group 7 can be withdrawn when group 5 is at its "outlimit", as indicated by the Rod Position Limit Svitches, and group 6 reaches 75% vithdrawal. Upon insertion of re:gulating group 7 to a 25% vithdrawn position, group 6 can be inserted.
Similarly, group 5 can be inserted when group 6 has been inserted to 25%
vithdravn. Eight axial power shaping rod assemblies are provided to accommodate local reactivity perturbations.
ICS or manually by the operator. The CRDCS may be operated automatically by the A requirement for continuous boron control by the ICS is the vithdraval of the Sefety Groups to their "outlimit" and 25% or greater withdraval of regulating group 5.
The non-trip partion of the CRDCS consists of two basic components: the motor controlRelated.
Safety system and the system logic. This portion of the CRDCS is not Nuclear The Chapter 15 Accident Analysis demonstrates the adequacy of
' protection systems to cope with CRDCS malfunctions. The motor control system comprises eight group power supplies and one auxiliary power supply. In order to initiate then three,rod movement, the power supply sequentially energizes first two, then two of the six CRDM motor-stator vindings in a stepping fashion to produce a rotating magnetic field for the CRDM motor to position the CRA.
Switching must is achieved by gating SCRs on for Oe period of time that each vinding be energized.
Gating signals for the SCRs are generated by a motor-driven photo-optic encoder. of 60 Hz synchronous motor driving a multichannel programmer consisting Programmer Motor. Command signals to position the CRAs are introduced at the The system logic of the CRDCS contains those functions that control rod motion and functions which monitor system operation. Failures which could result in improper system operation are continuously monitored by fault detection circuits.
safety rods not fully withdrawn, and a programmer lamp fault. Fault detectors t In the CRDCS, two methods of position indication are provided relative and absolute. !
The relative position transducer consists of a small pulse-stepping motor, driven from the rod drive motor power supply, coupled to a potentiometer. ,
The absolute position transducer consists of a series of 48 equally spaced reed l switches, upper motorenclosed in aCRDM.
tube of each fiberglass housing and mounted on the outside of the !
and moves parallel to the reed switches as the CRA is moved.A magnet This magnetic is mounted on field switch. actuates the reed switches when the magnet passes in the vicinity of each The reed svitches are connected to a resistive network, which provides an analog voltage output proportional to the position of the CRDM leadscrev due !
to the switches sequentially opening and closing. Either this absolute position indication rod positionormaters the relative position located in the indication may be displayed on the 61 single control room.
Certain reed switches in the absolute position transducer provide specific signals utilized in the CRDCS 1
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Dockat No. 50-346-Licsnsa No. NPF-3
. ~Ssrial No. 1492 Attachment 2 ,
Page 4 logic. The signals to tSe Control CRDCS. Rod Position Limit Switches provide "inlimit" and "outlimit" These signals are utilized to ensure regulating rods are not withdrawn until the safety group rods are fully vithdrawn, prohib!t.further inward motion of a rod group that is fully inserted, allow automatic control by the ICS, and are indicated on the Diamond (Rod Control) panel. The Control Rod t Position 50%, 75%, (orand zone 100%reference) Svitches provide signals corresponding to 0%, 25%,
rod withdrawal.
. These signals are utilized in sequencing rod insertion CRDCS.
or withdrawal, and are indicated at System Logic Cabinet #4 of the No adjustment, other than raising or lowering the complete absolute position transducer, is provided for the control Rod Position Svitches or the Control Rod Position Limit Switches.
Effects on Safety This Significant Hazard Consideration proposes (nat Table 3.3-9 of Technical Specification 3.3.3.5 and Table 4.3-6 of Surveillance Requirement 4.3.3.5 be revised to refer to the Control Rod Position Svitches, rather than the Control' '
Rod Position Limit Switches, and to delete the requirement for channel calibration every eighteen months. Changing the reference to the Position ;
Switches vill correctly identify the instruments and vill make the Technical Specifications consistent with the vendor manual and other site documentation.
The remote Position readout Limit Switches have an indication on the Diamond Panel, but no is provided.
The Control Rod Position Limit Switches are stated to have a readout at System Logic Cabinet #4 of the Control Rod Drive Control System (CRDCS) in Tables 3.3-9 and 4.3-6 of the Technical Specifications. The
- indication at this cabinet, however, is'actually provided by.the Control Rod Position Switches, rather than the Position Limit Switches. The Measurement Range attributed to the Position Limit Switches (0%, 25%, 50%, 75%, and 100%)
actually applies tc the Pcsition Switches.
Therefore the tables should be revised to refer to the Control Rod Position Svitches.
Surveillance Requirement 4.3.3.5 requires a calibration of the Control Rod I Poaition Limit Switches every eighteen months. The design of the Position Limit Swit-hes, as well as the Position Switches, is such that calibration is not required following initial installation. The only adjustment possible is levering or raising the fiberglass tube containing the switches at the CRDM.
Operability of the Position Svitches is presently established by verifying that the 0% position lights indicate when the control rods are fully inserted, and the remaining position lights (25%, 50%, 75%, and 100%) indicate as the CRAs are withdrawn.
The reed switches are permanently affixed within the fiberglass tube
' and their relative position cannot be changea. Therefore, verification of the 0% position switches satisfies the channel check and channel calibration required by Surveillance Requirement 4.3.3.5. The calibration requirement is not necessary, however, as no adjustment is actually required. Based upon these t I
considerations, changing the instrument reference from Control Rod Position Limit Switches to Control Rod Position Svitches in Tables 3 3-9 and 4.3-6 of tie Technical Specifications and deleting the requirement for a channel calibration for the Control Rod Position Suitches in Table 4.3-6 of the Technical l Specification vill not affect the safety function of any system or component. !
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Footnote "*" in Table 4.3-6 which allows an extension of the due date for i channel calibration of the Steam Cenerator Outlet Steam Pre.rsure indication at
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Docket No. 50-346 L.icansa No. NPF-3
, S2 rial No. 1492 Attachment 2 Page 5 the Auxiliary Shutdown panel is obsolete, as the extension date has elapsed.
Therefore, deleting the nov obsolete extension note vill have no effect on the safety function of any system or component since the calibration is performed in compliance with surveillance requi.ements per 4.3.3.5.
SIGNIFICANT HAZARDS ANALYSIS The Commission has provided standards in 10CFR50.92(c) for determining whether a significant hazards consideratior, exiscs. A proposed amendment to an Operating License for a facility involves no significant Fazards consideration if operation of the facility in accordance with the proposed amendment vould nott (1) involve a significant increase in the probability or consequences of an accident previlusly evaluated, (2) create the possibility of a new or diff.erent kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. The Company has reviewed the proposed change and determined that:
The proposed amendment does not involve a significant increase in the probabilitychanges proposed or consequene-s are a/ of an accident previously evaluated because the between plant configu- : ,.i , testinginand istrative nature and provide for consistency license requirements. The Control Rod Positisn Switches vill be properly referenced, which is consistent with the current plant design, and the channel calibration requirements of the switches vill be deleted based on their non-adjustability after initial installation.
These proposed changes vill have no effect on the safety function of any system or component (10CFR50.92(c)(1)).
The proposed amendment does not create the possibility of a new or different kind of accident than previously evaluated because the proposed changes are administrative ation, in nature and provide for consistency between plant configur-testing and license requirements, and do not affect any of the assumptions used in previous accident evaluations.
All accidents continue to be bounded by previous analysis and the proposed administrative changes will not ,
introduce the possibility of any new !
(10CFR50.92(c)(2)). or different hind of accident The proposed amendment does not in<olve a significant reduction in a margin of safety because the changes are administrative in nature and provide for consistency between plant configuration, testing and license requirements. The changes are being made to properly reference existing instrumentation and to correct the channel calibration requirements to account for the non-adjustability of components of that instrumentation.
involve no reduction in a margin of safety (10CFR50.92(c)(3)).The proposed changes Based on the above reasoning, Toledo Edison has determined that the proposed amendment does not involve a significant hazards consideration.