ML20216J857
| ML20216J857 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/13/1998 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20216J846 | List: |
| References | |
| NUDOCS 9803240109 | |
| Download: ML20216J857 (700) | |
Text
{{#Wiki_filter:, t - i-ENCLOSURE 1 f l BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2' DOCKETNOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENTS CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS l l l PROPOSED UNIT 2 LICENSE CONDITION As part of the conversion of the Brunswick Steam Electric Plant (BSEP), Unit Nos. I and 2 from f the Current Technical Specifications (CTS) to the Improved Technical Specifications (ITS), as contained in Revision 1 ofNUREG-1433, " Standard Technical Specifications General Electric 1
- Plants, BWR/4," CTS 33.6.2, "End-Of-Cycle Recirculation Pump Trip System l
- Instrumentation," is being deleted for Unit 2. This system is only installed on Unit 2 and it is l l ' abandoned in place. In support of deletion of CTS 3.3.6.2, Carolina Power & Light (CP&L)
Company requests that the following Unit 2 license condition be established: The End-Of-Cycle Recirculation Pump Trip system instrumentation shall be maintained l inoperable (i.e., manually bypassed) during Mode 1, when thermal power is greater than i or equal to 30% rated thermal power. t i The wording of the proposed license condition is consistent with a previous footnote to l CTS 3.3.6.2, included in Amendment 123, dated April 30,1986, to the Unit 2 Technical . i Specifications,which stated: During the current cycle operation, the end-of-cycle recirculation pump trip (EOC-RPT) system will be inoperable (manually bypassed); therefore, Specification 3.3.6.2 above does not apply. The provisions of Specification 3.0.4 are not applicable. , 1 The proposed Unit 2 license condition will ensure that the EOC-RPT system will not be used without prior NRC approval. l El-1 9903240109 980313 POR ADOCK 05000324 P PDR
e I ENCLOSURE 2 4 ! BRUNSECK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 l DOCKET NO3. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 p SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENTS l CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS ADDITIONAL INFORMATION REGARDING DOC L.15 FOR ITS 3.8.1 AND L.2 FOR ITS 3.8.7 Revision A to the proposed license amendments revised ITS 3.8.1, "AC Sources - Operating," and ITS 3.8.7, " Distribution Systems-Operating," to allow continued operation of a unit for seven days with, either: (1) one of the opposite, shutdown, unit's balance of plant buses and the associated circuit path to the downstream 4.16 kV emergency bus, including the diesel generator associated with the 4.16 kV emergency bus inoperable, or (2) one of the opposite, shutdown, l unit's AC electrical distribution subsystem load groups inoperable. This change provides the required operational flexibility to perform any necessary electrical distribution system ! .. maintenance. A conference call was held on March 4,1998 to discuss the proposed changes. The following information, is being provided as a result of this call. i Additional Information Regarding Discussion of Change (DOC) L.15 for ITS 3.8.1 When a 4.16 kV emergency bus is inoperable (i.e., de-energized), the auxiliaries of the associated l ' diesel generator (DG) are also inoperable. These auxiliaries, required for DG OPERABILITY,- include the DG filter and prelube oil pump, the DG jacket water recirculation pump, the DG lube oil filter heater, the DG jacket water heater, a DG starting air compressor, the DG fuel oil transfer pump, the DG service waterjacket water valves, ventilation supply and exhaust fans, and DG control panel instrumentation. Restoration ofpower to the associated bus restores these DG auxiliaries to OPERABLE status. However, the DG cannot be declared OPERABLE until the L DG lube oil temperature and jacket water temperature are heated to within required limits and the DG starting air receiver is recharged. Upon restoratien of power to the bus powering the DG auxiliaries, it is expected, under worst case conditions, to take approximately four hours to restore DG lube oil temperature and jacket water temperature to within required limits and I hour to recharge the DG starting air receiver. Proposed ITS 3.8.1 Required Action B.3 requires both the affected offsite circuits and the DG to be restored to OPERABLE status within the seven day Completion Time. Therefore, in order to satisfy this requirement under worst case conditions, the associated bus must be restored at least four hours prior to the expiration of the seven day ,, Completion Time. If the affected DG is not restored within the seven day Completion Time, then a plant shutdown is required in accordance with ITS 3.8.1 Required Actions H.1 and H.2. j Additional Information Regarding DOC L.2 for ITS 3.8.7 l In the event of an inoperable AC electrical power distribution bus (e.g., substation bus E7 or vital bus 1E7) .when one or both of the units are in MODE 1,2 or 3, the Action of Current Technical i E2-1 j
l-i Specification (CTS) 3.8.2.1 provides an Allowed Outage Time (AOT) of eight hours for restoration of the bus. The eight hour AOT of CTS 3.8.2.1 is based on NUREG-0123, " Standard ! Technical Specifications for General Electric Boiling Water Reactors." The basis for the eight hour AOT in NUREG-0123 is that in this condition, in a standard design plant which has only L two AC electrical power distribution' divisions, one division is without AC power (i.e., no offsite l power to the division and the associated DG inoperable). In this condition, the standard design l-plant is more vulnerable to a complete loss of AC power. As a result, it is considered to be imperative that the unit operators' attention be focused on minimizing the potential for loss of power to the remaining division by stabilizing the unit and restoring power to the affected l division rather than entering and taking the actions for each of the supported Technical Specification components powered from the inoperable AC electrical power distribution division. The eight hour time limit, before requiring a unit shutdown in this condition, is ! considered to be acceptable because of the potential for decreased safety if the unit operators' attention is diverted from the evaluations and actions necessary to restore power to the affected division to the actions associated with taking the unit to shutdown within this time limit.
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However, the Brunswick Steam Electric Plant (BSEP) design of the AC Electrical Power , Distribution system includes four separate load groups (i.e., El, E2, E3, and E4), with two load E groups in each division. Therefore, an inoperable load group, for which the seven day AOT is , proposed, does not result in the loss of a complete AC Electrical Power Distribution system L division and the units are not as vulnerable as the standard design plant to a complete loss of l offsite power in this condition. In addition, the loads required for mitigation of accidents and l transients are not equally distributed across all four load groups (i.e., the El and E2 load groups l primarily serve Unit I loads and the E3 and E4 load groups primarily serve Unit 2 loads). The
- proposed seven day AOT for the operating unit would only apply to buses primarily associated with the shutdown unit and would only allow de-energization of one AC Electrical Power Distribution System load group at a time.
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l ENCLOSURE 3 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENTS CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS
SUMMARY
OF REVISION C CHANGES l The following discussions are provided for the attached Revision C Technical Specification changes, Bases changes, associated new or revised Discussion of Changes (DOCS), Current
- Technical Specification (CTS) markup pages, new or revised Justifications for Deviations =
(JFDs), and Operating License change.
- 1. The Improved Technical Specification (ITS) Surveillance Requirement (SR) 3.1.4.1 requirement to perform control rod scram time testing for all control rods following each I
refueling and the ITS SR 3.1.4.4 requirement for control rod scram time testing of only l affected control rods following any fuel movement within the reactor pressure vessel are revised, to require scram time testing of all control rods following fuel movement in the
- mactor pressure vessel, to be consistent with NUREG-1433, " Standard Technical Specifications General Electric Plants, BWR/4." The proposed change in the initial ITS submittal was based on an industry generic change to NUREG-1433. This generic change will not be approved in sufficient time to support ITS implementation. Therefore, this proposed change is withdrawn. Corresponding changes to the associated Bases are made. DOC L.1 for ITS 3.1.4 and the associated No Significant Hazards Evaluation l (NSHE) are also revised. Corresponding changes are also made to the NUREG-1433
!- markup pages and JFDs. JFD 5 for Section 3.1 is deleted and new JFD 10 for Bases l. Section 3.1 is added.
- 2. ITS SR 3.3.1.2.2 is revised to require verifying that "an OPERABLE SRM detector is located in: (a) the fueled region; (b) the core quadrant where CORE ALTERATIONS l are being performed, when the associated SRM is included in the fueled region; and (c) a core quadrant adjacent to where CORE ALTERATIONS are being performed, when the l associated SRM is included in the fueled region" instead of only requiring verification that the detector of"an OPERABLE SRM detector is located in the core quadrant where l
CORE ALTERATIONS are being performed and one is !ocated in the adjacent quadrant." As a result of providing the additional criteria on where the OPERABLE Source Range Monitors (SRMs) must be relocated, Note 2 to ITS SR 3.3.1.2.2 is also added to clarify that more than one of the three requirements ofITS SR 3.3.1.2.2 can be satisfied by the same SRM since only two SRMs are required to be OPERABLE. Corresponding changes are made to the associated Bases, CTS markup pages, and
.-NUREG-1433 markup pages. In addition, new DOC M.4 for ITS 3.3.1.2 is provided for this change. This change addresses NRC comment 3.3.1.2-3.
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- 3. ITS SR 3.3.1.2.4, which requires verification that SRM count rate is 2 3 counts per second (cps), is revised to require performance of SRM count rate verification once per 24 hours (1) in MODE 2, when Intermediate Range Monitors (IRMs) are on Range 2 or below, and (2) MODES 3 and 4, and MODE 5 when CORE ALTERATIONS are not in progress. Corresponding changes are made to the associated Bases, CTS markup pages, and NUREG-1433 markup pages. In addition, DOC M.1 for ITS 3.3.1.2 is revised and l new DOC M.5 for ITS 3.3.1.2 is provided for this change. This change addresses NRC comment 3.3.1.2-4. j
- 4. ITS SR 3.3.1.2.6 is revised to require performance of a CHANNEL FUNCTIONAL !
TEST of the SRMs once per 31 days (1) in MODE 2, with IRMs on Range 2 or below, and (2)in MODES 3 and 4. In addition, a Note is added to ITS SR 3.3.1.2.6 which allows the Surveillance to be delayed until 12 hours after the IRMs are on Range 2 or below. Corresponding changes are made to the associated Bases, CTS markup pages, and NUREG 1433 markup pages. In addition, new DOC M.6 for ITS 3.3.1.2 and new DOC L.12 for ITS 3.3.1.2, and an associated NSHE, are provided for this chan3e. T his change addresses NRC comment 3.3.1.2-6.
- 5. The Bases for ITS 3.3.1.2 Required Actions C.1, C.2.1.1, C.2.1.2, and C.2.2 are revised to clarify that if the Rod Worth Minimizer (RWM) is inoperable due to bypassing up to eight control rods in the RWM, the requirements ofITS 3.3.1.2 Required Action C.2.1.2 do not restrict reactor startup. Corresponding changes are also made to the NUREG-1433 !
markup pages. This change addresses an NRC comment.
- 6. The Completion Time provided in ITS 3.3.3.1 Required Action A.1 for restoration of post-accident monitoring instruments is reduced from 31 days to 30 days. Corresponding changes are made to the associated Bases, CTS markup pages, and NUREG '433 markup pages. In addition, new DOC M.6 for ITS 3.3.3.1 is provided for this change and DOC L.1 for ITS 3.3.3.1, and the associated NSHE, are revised. This change addresses NRC comments 3.3.3.1-9 and 3.3.3.1-10.
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- 7. The Completion Time provided in ITS 3.3.3.2 Required Action A.1 for restoration of remote shutdown monitoring instruments is reduced from 31 days to 30 days.
Corresponding changes are made to the associated Bases, CTS markup pages, and NUREG-1433 markup pages. In addition, new DOC M.1 for ITS 3.3.3.2 is provided for this change. This change addresses NRC comment 3.3.3.2-9.
- 8. The Applicable Safety Analyses, Limiting Condition for Operation (LCO), and Applicability Section of the Bases ofITS 3.3.4.1, Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation, is revised to more completely describe the plant specific safety analyses. The first sentence, "ATWS-RPT is not assumed in the safety analysis," is revised to "ATWS-RPT is not assumed to mitigate any accident or transient in the safety analysis." A corresponding change is also made to the NUREG-1433 markup page. ;
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- 9. A separate CHANNEL FUNCTIONAL TEST Surveillance Requirement (i.e, l ITS SR 3.3.5.1.6) is added for the Core Spray Pump Start-Time Delay Relay (i.e., ITS Table 3.3.5.1-1 Function 1.d), the Residual Heat Removal (RHR) Pump Start-Time Delay Relay (i.e., ITS Table 3.3.5.1-1 Function 2.f), and the Automatic Depressurization System Timers (i.e., ITS Table 3.3.5.1-1 Functions 4.b and 5.b). Corresponding changes are made to the associated Bases, CTS markup pages, and NUREG-1433 markup pages. l In addition, DOC A.10 for ITS 3.3.5.1 is deleted and DOC LE.1 for ITS 3.3.5.1 is i revised. This change addresses the same concem as identified in NRC comment 3.3.6.1-2.
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- 10. ITS 3.3.7.2, " Condenser Vacuum Pump Isolation Instrumentation," is revised m ;
Condition A from "One or more required channels inoperable" to "One or more channels inoperable." A corresponding change is also made to the NUREG-1433 markup page. This change is made to be consistent with the ITS Writers Guide since all channels of the l Condenser Vacuum Pump Isolation Instrumentation are required to be OPERABLE to satisfy the LCO. This change also addresses an NRC comment. I 1. The Applicability ofITS 3.3.8.2, Reactor Protection System (RPS) Electric Power Monitoring, is revised to include MODES 3 and 4 with any control rod withdrawn from a core cell containing one or more fuel assemblies. Corresponding changes are also made to Condition D ofITS 3.3.8.2. The RPS Electric Power Monitoring requirements for MODES 3 and 4 with any control rod withdrawn from a core cell containing one or more fuel assemblies were previously addressed in ITS 3.10.3, Single Control Rod Withdrawal - Hot Shutdown, and ITS 3.10.4, Single Control Rod Withdrawal - Cold Shutdown. As a result of this change,ITS 3.10.3 and ITS 3.10.4 are revised to delete the i redundant requirements for the RPS Electric Power Monitoring to be applicable when any I control rod is withdrawn from a core cell containing one or more fuel assemblies. Corresponding changes are also made to the associated Bases and NUREG-1433 markup pages. DOCS M.1 and L.1 for ITS 3.3.8.2, DOC L.2 for ITS 3.10.4, and JFD 29 for Section 3.3 are also revised and JFD 4 for Section 3.10 is deleted. These changes address an NRC comment.
- 12. DOC M.1 for ITS 3.3.6.1 is revised to more accurately reference the associated CTS instrumentation Function number that is affected by the change. " CTS Table 3.3.2-1 Functions 1.a and 5.a" is revised to " CTS Table 3.3.2-1 Functions 1.a.1 (valve group 8) and 5.a." This change addresses an NRC comment.
- 13. DOC LA.5 for ITS 3.3.6.1 is revised to incorporate thejustification (i.e., for the l relocation of the reactor steam pressure value associated with the bypassing of the
( isolation associated with the Condenser Vacuum - Low Function) that was provided in the response to NRC comment 3.3.6.1-14. This change addresses an NRC comment.
- 14. DOC L.12 for ITS 3.3.6.1 and associated NSHE are added to address the deletion of the requirement to deactivate the reactor vessel head spray valves closed when the Residual Heat Removal (RHR) Shutdown Cooling isolation instrumentation is inoperable and not l
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i l l restored within the required time period. This requirement,in CTS Table 3.3.2-1 Action 27, is deleted since the reactor vessel head spray lines are isolated with welded f I pipe caps and the reactor vessel head spray mode of the RHR system has been ! deactivated. Therefore, the isolation function of this instrumentation for the reactor vessel head spray valves is not necessary to ensure the associated lines are isolated. CTS i markup pages 8 of 42 and 29 of 42 for Specification 3.3.6.1 are also revised to reflect this change. This change addresses an NRC comment.
- 15. DOC A.3 for ITS 3.3.6.2 is revised to provide additional discussion justifying the administrative nature of the change to CTS Table 3.3.2-1 Action 23 for inoperable secondary containment isolation instrumentation. This change addresses an NRC comment.
- 16. DOC A.7 for ITS 3.3.7.1 is added to address the deletion of a Note that provides a one time exemption from the OPERABILITY requirements of CTS 3.3.5.5, " Control Room Emergency Ventilation System Instrumentation." This Note was added in recently issued j Amendments 191 and 222 for Units 1 and 2, respectively. The allowance of the Note will i expire prior to the implementation of the ITS. CTS markup pages 1 of 10 and 6 of 10 for j Specification 3.3.7.1 are also revised to reflect this change.
- 17. New JFD 32 for Section 3.3 is provided to justify maintaining the CTS allowance, in ITS 3.3.1.2 Required Action C.2.1.2, for bypassing up to eight control rods in the RWM without having to consider the RWM inoperable for the purpose of restricting reactor startup. This change addresses an NRC comment.
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- 18. The Background Section of the Bases for ITS 3.5.1, "ECCS - Operating," is revised to ;
more accurately describe the status of the RHR system cross tie valve (i.e., the cross tie valve is locked closed). A corresponding change is also made to the associated NUREG-1433 markup page.
- 19. DOC L 8 for ITS 3.6.1.3 and the associated NSHE are revised to more completely discuss closed system isolation valves, consistent with the Updated Final Safety Analysis Report (UFSAR) discussion, which are addressed by this change. '
- 20. The Bases ofITS 3.7.2, " Service Water (SW) System and Ultimate Heat Sink (UHS)," is revised in SR 3.7.2.4 to simplify the discussion related to the components (i.e., the diesel generators) that would be considered inoperable in the event of failure of this Surveillance. A corresponding change is also made to the associated NUREG-1433 markup page.
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- 21. DOC A.4 for ITS 3.7.3 is added to address the deletion of a Note that provides a one time exemption from the OPERABILITY requirements of CTS 3.7.2, " Control Room Emergency Ventilation System." This Note was added in recently issued Amendments 191 and 222 for Units 1 and 2, respectively. The allowance of the Note will E3-4
9 expire prior to the implementation of the ITS. CTS markup pages 1 of 8,2 of 8,5 of 8, j and 6 of 8 for Specification 3.7.3 are also revised to reflect this change.
- 22. ITS 3.8.1 Condition B and ITS 3.8.7 Condition A provide a 7 day time period for l restoration ofinoperabilities associated with a balance of plant bus, the associated i I
emergency buses, and associated diesel generator. The intent of these ACTIO'NS was to provide time to perform required maintenance. Therefore,ITS 3.8.1 and ITS 3.8.7 are
. revised to restrict use of these Conditions, Required Actions and Completion Times for planned maintenance only. Corresponding changes are made to the associated Bases and the NUREG-1433 markup pages. These changes address an NRC comment.
L 23. DOC L.15 for ITS 3.8.1 and DOC L.2 for ITS 3.8.7 are revised to provide additional - l_ discussion regarding the risk associated with the proposed allowed outage time changes l for offsite circuit inoperabilities, associated with a balance of plant bus, and for an emergency bus inoperability. The Tables in DOC L.15 for ITS 3.8.1 and DOC L.2 for ITS 3.8.7 are also revised to provide additional information regarding which buses support Technical Specification required equipment. These changes address NRC comments. j
' 24. Note I to SR 3.8.1.8 is modified to not restrict performance of the verification of the manual transier of the unit power supply from the preferred offsite circuit to the attemate
. offsite circuit when the unit is in MODE 1 or 2. The plant electrical system design includes breaker interlocks such that this manual transfer is performed without causing perturbations of the electrical distributions systems. Therefore, MODE restrictions on the
, performance of this Surveillance are not required and are deleted. Corresponding i changes are made to the associated Bases and CTS markup pages. In addition, new DOC . L.16 for ITS 3.8.1 and an associated NSHE are provided for this change. This change addresses an NRC comment.
- 25. For Unit 2, an Operating License Condition is proposed to be added to ensure that End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation is not credited for modifying applicable power distribution limits since the EOC-RPT Instrumentation requirements are to be deleted from the Unit 2 Technical Specifications.
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a l-ENCLOSURE 4 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 l SUPPLEMENT TO REQUEST FOR LICENSE AMENDMENTS l CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS ( 1 i l REVISION C - PAGE CHANGE INSTRUCTIONS i l J
ENCLOSURE 5 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 l S'UPPLEMENT TO REQUEST FOR LICENSE AMENDMENTS l CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS l l. l-l l l REVISION C l l l l l l l i l l-
l TSC 96TSB02 REVISION C INSERT AND REMOVAL INSTRUCTIGNS The following instructions are provided for use in updating the BNP ITS ! submittal (TSC 96TSB02). To facilitate the update process, the instructions have been divided by BNP ITS submittal volume number and associated divider tabs. To facilitate incorporation of the Revision C package into the ITS submittal, colored pages have been inserted between the material for each tab section. After incorporation of the Revision C pages into the BNP ITS , submittal, the colored pages should be discarded. VOLUME 3 - SECTION 3.1, SECTION 3.2 TAB - SECTION 3.1 l TAB - U1 ITS l REMOVE PAGE(S) INSERT PAGE(S) 3.1-12 thru 3.1-27 3.1-12 thru 3.1-26
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TAB - U1 ITS BASES REMOVE PAGE(S) INSERT PAGE(S) B 3.1-24 thru B 3.1-27 B 3.2-24 thru B 3.1-27 TAB - U2 ITS REMOVE PAGE(S) INSERT PAGE(S) 3.1-12 thru 3.1-27 3.1-12 thru 3.1-26 TAB - U2 ITS BASES REMOVE PAGE(S) INSERT PAGE(S)
- B 3.1-24 thru B 3.1-27 B 3.2-24 thru B 3.1-27 l Page 1 of 13 i
l l l I TSC 96TS802 REVISION C INSERT AS REMOVAL INSTRUCTIONS
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VOLUME 3 - SECTION 3.1, SECTION 3.2 TA8 - SECTION 3.1 (continued) TA8 - CTS / DOCS ! I REMOVE PAGEfS) INSERT PAGEfS) l Spec. 3.1.4 Markup page 1 of 6 Spec. 3.1.4 Markup page 1 of 6 Spec. 3.1.4 Markup page 4 of 6 Spec. 3.1.4 Markup page 4 of 6 ITS 3.1.4 DOC page 4 ITS 3.1.4 000 pages 4 and 5 l TA8 - ISTS/JFDs i I REMOVE PAGEfS) INSERT PAGEfS) Markup pages 3.1-12 and 3.1-13 Markup pages 3.1-12 and 3.1-13 l JFD Section 3.1 page 1 JFD Section 3.1 page 1 l TA8 - ISTS BASES /JFDs l REMOVE PAGE(S) INSERT PAGEfS) Markup page B 3.1-25 Markup page B 3.1-25 Insert B 3.1.4-1 (behind Insert B 3.1.4-1 (behind L Markup page B 3.1-25) Markup page B 3.1-25) i Markup page B 3.1-27 Markup page B 3.1-27 JFD Section 3.1 page 1 JFD Section 3.1 page 1 I TAB - NSHEs REMOVE PAGE(S) INSERT PAGE(S) , ITS 3.1.4 L.1 CHANGE pages 1 ITS 3.1.4 L.1 CHANGE pages 1 and 2 and 2 1 Page 2 of 13 l
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i I TSC 96TSB02 REVISION C INSERT AND REMOVAL INSTRUCTIONS I VOLUME 4 - SECTION 3.3 TAB - SECTION 3.3 TA8 - U1 ITS REMOVE PAGE(S) INSERT PAGE(S) 3.3-13 thru 3.3-74 3.3-13 thru 3.3-73 i TAB - U1 ITS BASES ,
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l REMOVE PAGE(S) INSERT PAGE(S) B 3.3-42 thru B 3.3-211 B 3.3-42 thru 8 3.3-212 l l l l TAB - U2 ITS l l REMOVE PAGE(S) INSERT PAGE(S) 3.3-13 thru 3.3-74 3.3-13 thru 3.3-73 l TAB - U2 ITS BASES l REMOVE PAGE(S) INSERT PAGE(S) B 3.3-42 thru B 3.3-212 B 3.3-42 thru 8 3.3-213 1 l l Page 3 of 13
l l TSC 96TS802 REVISION C INSERT AND REMOVAL INSTRUCTIONS VOLUME 5 - SECTION 3.3 TA8 - SECTION 3.3 TA8 - CTS / DOCS REMOVE PAGE(S) INSERT PAGE(S) Spec. 3.3.1.2 Markup page 1 of 6 Spec. 3.3.1.2 Markup page 1 of 6 Spec. 3.3.1.2 Markup pages 3 of 6 Spec. 3.3.1.2 Markup pages 3 of 6 and 4 of 6 and 4 of 6 Spec. 3.3.1.2 Markup page 6 of 6 Spec. 3.3.1.2 Markup page 6 of 6 ITS 3.3.1.2 DOC pages 2 thru 8 ITS 3.3.1.2 DOC pages 2 thru 10 Spec. 3.3.3.1 Markup page 3 of 10 Spec. 3.3.3.1 Markup page 3_of 10 Spec. 3.3.3.1 Markup page 5 of 10 Spec. 3.3.3.1 Markup page 5 of 10 Spec. 3.3.3.1 Markup page 8 of 10 Spec. 3.3.3.1 Markup page 8 of 10 Spec. 3.3.3.1 Markup page 10 of 10 Spec. 3.3.3.1 Markup page 10 of 10 ITS 3.3.3.1 DOC pages 3 thru 9 ITS 3.3.3.1 DOC pages 3 thru 9 Spec. 3.3.3.2 Markup page 1 of 6 Spec. 3.3.3.2 Markup page 1 of 6 Spec. 3.3.3.2 Markup page 4 of 6 Spec. 3.3.3.2 Markup page 4 of 6 ITS 3.3.3.2 DOC pages 1 thru 3 ITS 3.3.3.2 DOC pages I thru 3 Spec. 3.3.5.1 Markup pages 9 Spec. 3.3.5.1 Markup pages 9 i of 24, 10 of 24, and 11 of 24 of 24,10 of 24, and 11 of 24 Spec. 3.3.5.1 Markup pages 21 Spec. 3.3.5.1 Markup pages 21 of 24, 22 of 24, and 23 of 24 of 24, 22 of 24, and 23 of 24 ITS 3.3.5.1 DOC page 3 ITS 3.3.5.1 DOC page 3 ITS 3.3.5.1 DOC page 9 ITS 3.3.5.1 DOC page 9 Spec. 3.3.6.1 Markup page 8 of 42 Spec. 3.3.6.1 Markup page 8 of 42 Spec. 3.3.6.1 Markup page 29 of 42 Spec. 3.3.6.1 Markup page 29 of 42 ITS 3.3.6.1 DOC page 5 ITS 3.3.6.1 DOC page 5 ITS 3.3.6.1 DOC page 9 ITS 3.3.6.1 DOC page 9 ITS 3.3.6.1 DOC pages 29 thru 32 ITS 3.3.6.1 DOC pages 29 thru 32 l Page 4 of 13
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- TSC 96TS802 REVISION C l INSERT AND REMOVAL INSTRUCTIONS l VOLUME 5 - SECTION 3.3 TAB - SECTION 3.3 TA8 - CTS / DOCS (continued)
REMOVE PAGE(S) INSERT PAGE(S) ITS 3.3.6.2 DOC pages 1 thru 9 ITS 3.3.6.2 DOC pages 1 thru 9 Spec. 3.3.7.1 Markup page 1 Spec. 3.3.7.1 Markup page 1 of 10 of 10 Spec. 3.3.7.1 Markup page 6 Spec. 3.3.7.1 Markup page 6 of 10 of 10 , ITS 3.3.7.1 DOC pages 2 thru 9 ITS 3.3.7.1 DOC pages 2 thru 10 ITS 3.3.8.2 DOC pages 1 thru 6 ITS 3.3.8.2 DOC pages 1 thru 6 CTS 3/4.3.6.2 DOC page 1 CTS 3/4.3.6.2 DOC page 1 VOLUNE 6 - SECTION 3.3 TAB - SECTION 3.3 TAB - ISTS/JFDs RENOVE PAGE(S) INSERT PAGE(S) Markup pages 3.3-12 and 3.3-13 Markup pages 3.3-12 and 3.3-13 Markup page 3.3-16 Markup page 3.3-16 Markup page 3.3-23 Markup page 3.3-23 Markup page 3.3-27 Markup page 3.3-27 Markup pages 3.3-41 thru 3.3-43 Markup pages 3.3-41 thru 3.3-43 Markup pages 3.3-45 and 3.3-46 Markup pages 3.3-45 and 3.3-46 Markup page 3.3-A (after 3.3-74) Markup page 3.3-A (after 3.3-74) Markup pages 3.3-78 and 3.3-79 Markup pages 3.3-78 and 3.3-79 JFD Section 3.3 pages 5 and 6 JFD Section 3.3 pages 5 and 6 Page 5 of 13
f I ! J TSC 96TSB02 REVISION C INSERT AND REMOVAL INSTRUCTIONS 1 VOLUME 6 - SECTION 3.3 TA8 - SECTION 3.3 (continued) TA8 - ISTS BASES /JFDs REMOVE PAGE(S) INSERT PAGEfS) Markup pages B 3.3-40 and Markup pages B 3.3-40 and B 3.3-41 B 3.3-41 i Markup page B 3.3-42 Markup page B 3.3-42 , Markup page B 3.3-50 Markup page B 3.3-50 ' Markup page B 3.3-70 Markup page B 3.3-70 Markup page B 3.3-77 Markup page B 3.3-77 Markup page B 3.3-92 . Markup page B 3'.3-92 Markup page B 3.3-136 Markup page B 3.3-136 Markup page B 3.3-229 Markup page B 3.3-229 i Markup page B 3.3-231 Markup page B 3.3-231 TA8 - NSHEs REMOVE PAGEfS) INSERT PAGE(S) New page: ITS 3.3.1.2 L.12 CHANGE ! page 14 (behind ITS 3.3.1.2 L.ll CHANGE page 13) l ITS 3.3.3.1 L.1 CHANGE page 1 ITS 3.3.3.1 L.1 CHANGE page 1 New pages: ITS 3.3.6.1 L.12 CHANGE pages 13 and 14 (behind ITS 3.3.6.1 L.11 CHANGE page 12) ! Page 6 of 13
TSC 96TSB02 REVISION C INSERT AND RENOVAL INSTRUCTIONS . VOLUNE 8 - SECTION 3.5 TAB - SECTION 3.5 TAB - U1 ITS BASES REMOVE PAGEfS) INSERT PAGE(S) B 3.5-2 and B 3.5-3 8 3.5-2 and B 3.5-3 TAB - U2 ITS BASES , BEN _0_YI..fAGIIS1 INSERT PAGE(S) B 3.5-2 and B 3.5-3 8 3.5-2 and B 3.5-3 TAB - ISTS BASES /JFDs REMOVE PAGEfS) INSERT PAGE(S) Markup page B 3.5 2 Markup page B 3.5-2 I i l l 1 l Page 7 of 13
I l I TSC 96TS802 REVISION C INSERT AND REMOVAL INSTRUCTIONS . i VOLUNE 9 - SECTION 3.6 j TA8 - SECTION 3.6 i l TA8 - CTS / DOCS REMOVE PAGE(S) INSERT PAGE(S) l ITS 3.6.1.3 DOC page 9 ITS 3.6.1.3 DOC page 9 I l i VOLUNE 10 - SECTION 3.6 ' TA8 - SECTION 3.6 TAB - NSHEs i REMOVE PAGE(S) INSERT PAGEfS) ITS 3.6.1.3 L.8 CHANGE page 8 ITS 3.6.1.3 L.8 CHANGE page 8 l VOLUNE 11 - SECTION 3.7 TA8 - SECTION 3.7 TAB - U1 ITS BASES REMOVE PAGE(S) INSERT PAGEfS) B 3.7-19 B 3.7-19 TA8 - U2 ITS BASES REMOVE PAGE(S) INSERT PAGE(S) B 3.7-19 8 3.7-19 Page 8 of 13
TSC 96TSB02 REVISION C INSERT AND REMOVAL INSTRUCTIONS VOLUME 11 - SECTION 3.7 TA8 - SECTION 3.7 (continued) TA8 - CTS / DOCS REMOVE PAGE(S) INSERT PAGE(S) Spec. 3.7.3 Markup pages 1 of 8 Spec. 3.7.3 Markup pages 1 of 8 and 2 of 8 and 2 of 8 Spec. 3.7.3 Markup pages 5 of 8 Spec. 3.7.3 Markup pages 5 of 8 and 6 of 8 and 6 of 8 ITS 3.7.3 DOC pages 1 thru 3 ITS 3.7.3 DOC pages 1 thru 3 TAB - ISTS BASES /JFDs REMOVE PAGE(S) INSERT PAGE(S) Insert B 3.7.2-8 (1 page) (behind Insert B 3.7.2-8 (1 page) (behind Markup page 8 3.7-12) Markup page B.3.7-12) VOLUME 12 - SECTION 3.8 TA8 - SECTION 3.8 TAB - U1 ITS REMOVE PAGE(S) N LN.Sf_RT PAGE(S) 3.8-2 3.8-2 3.8-10 3.8-10 3.8-34 3.8-34 Page 9 of 13
TSC 96TS802 REVISION C INSERT AND REN0 VAL INSTRUCTIONS VOLUME 12 - SECTION 3.8 TA8 - SECTION 3.8 TAB - UI ITS BASES REMOVE PAGE(S) INSERT Pt.' (11 B 3.3-6 thru B 3.8-33 B 3.8-6 t m u B 3.8-33 8 3.8-75 and B 3.8-76 B 3.8-75 and B 3.8-76 TA8 - U2 ITS . REMOVE PAGE(S) INSERT PAGE(S) 3.8-2 3.8-2 l 3.8-10 3.8-10 3.8-34 3.8-34 TAB - U2 ITS BASES REMOVE PAGE(S) INSERT PAGE(S) B 3.8-6 thru 8 3.8-33 B 3.8-6 thru B 3.8-33 B 3.8-75 and B 3.8-76 B 3.8-75 and B 3.8-76 l l TAB - CTS / DOCS REMOVE PAGE(S) INSERT PAGE(S) Spec. 3.8.1 Markup page 5 of 16 Spoc. 3.8.1 Markup page 5 of 16 Spec. 3.8.1 Markup page 13 of 16 Spec. 3.8.1 Markup page 13 of 16 l ITS 3.8.1 DOC pages 16 thru 40 ITS 3.8.1 DOC pages 16 thru 41 l ITS 3.8.7 DOC pages 5 thru 29 ITS 3.8.7 DOC pages 5 thru 31 Page 10 of 13 l
l + TSC 96TSB02 REVISION C INSERT Ale REMOVAL INSTRUCTIONS VOLUME 13 - SECTION 3.8 l TAB - SECTION 3.8 ! TA8 - ISTS/JFDs itQ!QVE PAGEfS) _ INSERT PAGEfS) Insert 3.8.1-1A (behind Insert 3.8.1-1A (behind Markup page 3.8-1) Markup page 3.8-1) Markup page 3.8-8 Markup page 3.8-8 Insert 3.8.7-A (behind Markup Insert 3.8.7-A (behind Markup page3.8-38) page 3.8-38) JFD Section 3.8 pages 13 and 14 JFD Section 3.8'pages 13 and 14 TA8 - ISTS BASES /JFDs i REMOVE PAGE(S) INSERT PAGEfS) Insert B 3.8.1-5A (behind Insert Insert B 3.8.1-5A (behind Insert B 3.8.1-5) (behind Markup B 3.8.1-5) (behind Markup page B 3.8-5) page B 3.8-5) Insert B 3.8.1-5A (continued) Insert B 3.8.1-5A (continued) 8.3 (behind B 3.8.1-5A B.3 (behind B 3.8.1-5A (continued) B.2) (continued) B.2) Markup page B 3.8-20 Markup page B 3.8-20 Insert B 3.8.7-ACT A (behind Insert B 3.8.7-ACT A (behind Insert B 3.8.7-3) (behind Insert B 3.8.7-3) (behind
' Markup page B 3.8-81) Markup page B 3.8-81)
Insert B 3.8.7-ACT A (continued) Insert B 3.8.7-ACT A (continued) (behind B 3.8.7-ACT A) (behind B 3.8.7-ACT A) TAB - NSHEs REMOVE PAGEfS) INSERT PAGE(S) New Pages: ITS 3.8.1 L.16 CHANGE pages 21 and 22 (behind ITS 3.8.1 L.15 CHANGE page 20) Page 11 of 13
I I TSC 96TSB02 REVISION C INSERT AND REMOVAL INSTRUCTIONS V0 LUNE 14 - SECTIONS 3.9, 3.10 TAB - SECTION 3.10 TAB - U1 ITS l REMOVE PAGE(S) INSERT PAGE(S) 3.10-6 3.10-6 3.10-9 3.10-9 TAB - U1 ITS BASES . REMOVE PAGE(S) INSERT PAGE(S) B 3.10-13 B 3.10-13 B 3.10-18 B 3.10-18 i i TAB - U2 ITS REMOVE PAGE(S) INSERT PAGE(11 3.10-6 3.10-6 3.10-9 3.10-9 TAB - U2 ITS BASES REMOVE PAGE(S) INSERT PAGE(S) B 3.10-13 B 3.10-13 8 3.10-18 8 3.10-18 TAB - CTS / DOCS REMOVE PAGE(S) INSERT PAGE(S) ITS 3.10.4 DOC page 4 ITS 3.10.4 DOC page 4 Page 12 of 13
l TSC 96TSB02 REVISION C INSERT AND REMOVAL INSTRUCTIONS VOLUME 14 - SECTIONS 3.9, 3.10 TAB - SECTION 3.10 (continued) TAB - ISTS/JFDs E MOVE PAGE(S) INSERT PAGE(S) Markup page 3.10-6 Markup page 3.10-6 Markup page 3.10-9 Markup page 3.10-9 JFD Section 3.10 page 1 JFD Section 3.10 page 1 TAB - ISTS BASES /JFDs REMOVE PAGE(S) INSERT PAGE(S) Markup page B 3.10-13 Markup page B 3.10-13 Markup page B 3.10-18 Markup page B 3.10-18 l l Page 13 of 13 [ ] l i
i I Centrol Rod Scram Times 3.1.4 i j /"% l- (j 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control' Rod Scram Times LCO 3.1.4 a. No more than 10 OPERABLE control rods shall be " slow," in accordance with Table 3.1.4-1; and
- b. No more than 2 OPERABLE control rods that are " slow" shall occupy adjacent locations.
APPLICABILITY: MODES I and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Be in MODE 3. 12 hours LCO not met. O SURVEILLANCE REQUIREMENTS
..................................... NOTE-----------------------------
During single control rod scram time Surve111ances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator. SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify each control rod scram time is Prior to within the limits of Table 3.1.4-1 with exceeding reactor steam dome pressure a: 800 psig. 40% RTP after fuel movement within the reactor C pressure vessel AND (continued) 'O Brunswick Unit 1 3.1-12 Amendment No.
Centrol Rod Scran Times 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 (continued) Prior to exceeding 40% RTP after ! each reactor i shutdown ! 2 120 days 1 SR 3.1.4.2 Verify, for a representative sample, each 120 days tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in dome pressure a 800 psig. MODE 1 i SR 3.1.4.3 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with declaring any reactor steam dome pressure. control rod 1 OPERABLE after work on control (~' \ rod or CRD System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram time Price to , is within the limits of Table 3.1.4-1 with exceeding l reactor steam dome pressure 2 800 psig. 40% RTP after ! work on control rod.or CRD System that could affect scram time , Brunswick Unit 1 3.1-13 Amendment No.
'l.
1 Control Rod Scras Times 3.1.4 Table 3.1.4-1 (page -1 of 1). Control Rod Scram Times
-------------------------------------NOTES------------------------------------ l
- 1. OPERABLE' control rods with scram times not within the limits of this Table '
are considered " slow."
- 2. Enter applicable conditions and Required Actions of LCO 3.1.3, " Control Rod OPERABILITY," for control: rods with scram times > 7 seconds to notch position 06. These control. rods are inoperable, in accordance with SR 3.1.3.4, and are not considered " slow."
SCRAM TIMES WHEN REACTOR STEAM E NOTCH POSITION PRESSURE 2: 800 psig a)(b) (seconds) 46 0.44 36 1.08 26 1.83 06 3.35 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero. (b) When reactor steam dome pressure is < 800 psig, established scram time limits apply, i I l l O Brunswick Unit 1 3.1-14 ' Amendment No. i i
_. = centrol Rod Scram Accumulators l 3.1.5 (*) . V 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Control Rod Scram Accumulators LCO 3.1.5 Each control rod scram accumulator shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS
..............................-------NOTE-------------------------......-...-.
l Separate Condition entry is allowed for each control rod scram accumulator. l l CONDITION REQUIRED ACTION COMPLETION TIME i A. One control rod scram A.1 --------NOTE--------- l accumulator inoperable Only applicable if with reactor stena the associated l f- g dome pressure control rod scram '(') 2: 950 psig. time was within the limits of Table 3.1.4-1 during the last scrtm time i Surveillance. Declare the 8 hours i associated control rod scram time
" sl ow. "
08 A.2 Declare the 8 hours ! associated control l
~
l rod inoperable. a- (continued) ; I 1 Brunswick Unit 1 3.1-15 Amendment No. i i
e Centrol R:d Scram Accumulators 3.1.5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME l B. Two or more control B.1 Restore charging 20 minutes from
- rod scram accumulators water header pressure discovery of inoperable with to a 940 psig. Condition B l reactor steam dome concurrent with pressure 2 950 psig. charging water header pressure
< 940 psig
! AND B.2.1 -------.N0TE--------- Only applicable if the associated control rod scram time was within the limits of l Table 3.1.4-1 during the last scram time l Surveillance. f3 d Declare the associated control 1 hour rod scram time
" sl ow. "
l OB B.2.2 Declare the I hour associated control j rod inoperable. (continued) ( Brunswick Unit 1 3.1-16 Amendment No. ( l
( Centrol Rod Scraa Accumulators 3.1.5 ACTIONS (continued)- l CONDITION REQUIRED ACTION COMPLETION TIME l l C. One or more control C.1 Verify all control Immediately upon rod scram accumulators rods associated with discovery of inoperable with inoperable charging water reactor steam dome accumulators are header pressure pressure < 950 psig. fully inserted. < 940 psig AND C.2 Declare the I hour associated control rod inoperable. D. Required Action B.1 0.1 --------NOTE--------- or C.1 and associated Not applicable if all Completion Time not inoperable control met. rod scram accu ='ilators are associated with fully inserted control O. rods. Manually scram the Immediately reactor. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
'SR 3.1.5.1 Verify each control rod scram accumulator 7 days pressure is 2: 940 psig.
1O l Brunswick Unit 1 3.1-17 Amendment No.
i ) Rod Pattern Centrol 3.1.6 Q 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control LCO 3.1.6 OPERABLE control rods shall comply with the requirements of l the banked position withdrawal sequence (BPWS). l l APPLICABILITY: MODES I and 2 with THERMAL POWER s 10% RTP. ACTIONS I CONDITION REQUIRED ACTION COMPLETION TIME i A. One or more OPERABLE A.1 --------NOTE--------- l control rods not in Control rod may be i i compliance with BPWS. bypassed in the rod ! l worth minimizer (RWM) l l or RWM may be j bypassed as allowed 1 by LC0 3.3.2.1,
" Control Rod Block "I$$$$$ $"*_.___
Move associated 8 hours , l control rod (s) to l l correct position. l 9.8 t l l A.2 Declare associated 8 hours i i control rod (s) j inoperable. (continued)- O Brunswick Unit 1 3.1-18 Amendment No. f. t
Rod Pattarn Centrol 3.1.6 ( ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Nine or more OPERABLE B.1 --------NOTE--------- control rods not in Control rod may be compliance with BPWS. bypassed in the RWM or RWM may be bypassed as allowed by LCO 3.3.2.1. Suspend withdrawal of Immediately control rods. NLD B.2 Manually scram the 1 hour reactor. SURVEILLANCE REQUIREMENTS (O.) SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply 24 hours with BPWS. O Brunswick Unit 1 3.1-19 Amendment No.
.)
SLC Systcm 3.1.7 I 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LC0 3.1.7 Two SLC subsystems shall be OPERABLE. l APPLICABILITY: MODES I and 2. ACTIONS i CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable. to OPERABLE status. B. Two SLC subsystems B.1 Restore one SLC 8 hours inoperable. subsystem to OPERABLE status. lO C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. l l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l SR 3.1.7.1 Verify available volume of sodium 24 hours l pentaborate solution is within the limits of Figure 3.1.7-1. J l (continued) O- Brunswick Unit 1 3.1-20 Amendment No. I l
l SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours solution is within the limits of Figure 3.1.7-2. SR 3.1.7.3 Verify temperature of pump suction piping 24 hours is within the limits of Figure 3.1.7-2. SR 3.1.7.4 Verify continuity of explosive charge. 31 days SR 3.1.7.5 Verify the concentration of boron in 31 days solution is within the limits of Figure 3.1.7-1. AND Once within O 24 hours after water or boron is added to solution i AND Once within 24 hours after solution temperature is restored within the limits of Figure 3.1.7-2 SR 3.1.7.6 Verify each pump develops a flow rate In accordance l 2 41.2 gpm at a discharge pressure with the ! d 1190 psig. Inservice ' Testing Program (continued) Brunswick Unit 1 3.1-21 Amendment No.
)
I SLC System 3.1.7 i SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify flow through one SLC subsystem from 24 months on a pump into reactor pressure vessel. STAGGERED TEST BASIS l' O ' i
)
l Brunswick Unit 1 3.1-22 Amendment No.
i SLC System 3.1.7 p d 22.0 21.0 ..- .-- , . _ y _ 3 ;. 20.0 h,e1 l g j, jdppd$ts N,sfddMs M.,:@:,@.:,0 1.a x c ,n,~ . , + :c .-.
.~ . , .
,o~ R.1
- z. . ; .: e ~ ; .
, ~_ ,,. . ..
$e.c 19.0 h. ~
Nk( v e,x^Nid,:,':@
~
i
$w$$hkEd+c ne "r _$yd I.
cn - I .:<w 3::- :. ew,.m::
.Tn. 2
.p. , lj *m u * '
i . -- g) 18.0 y"p; p I..N" s E M S g:j N l u)p - y [hthfhf l ACCEPTABLE REGION ; fyOp.g
}uat 17.0 . m, SEEpwee es .2, s . m. , A **7%-
16'0 8 15.0 35$$$ $@ M IWTWM 1 20 14.0 ;- CC M M REGION T MEl k, .ii_ $ [13 [ $.NCI! j;
. j.
~^ gg
( @ 13.0
~ ~ ' " " ~ "' ~' '
12.0 11.0 l 10.0 2000 2500 3000 3500 4000 4500 5000 NETVOLUME OF SOLUTION IN TANK (gals) Figure 3.1.7-1 (page 1 of 1) Sodium Pentaborate Solution Volume Versus Concentration Requirements f [e*%b Brunswick Unit 1 3.1-23 Amendment No.
SLC System 3.1.7 V 150 140 , . .m ,. . 130 [$ ACCEPTABLE . 120
& fj $ $ 1l 8 UNACCEPTABLE $. [ l 110 REGION sg & y'^ W-{p' k'g' j --
. ?
i . lR '" ; a H B V 90 w y p u pr k $ $h h ?h l w 80 ~ C' S $ \ W 70 '
/ UNACCEPTABLE 60 50 '
/
0 .5 10 13 15 17 19 21 25 14 16 18 20 SODIUM PENTABORATE SOLUTION CONCENTRATION (% by weight) Figure 3.1.7-2 (page 1 of 1) . Sodium Pentaborate Solution Temperature Versus Concentration Requirements r% \ Brunswick Unit 1 3.1-24 Amendment No.
SDV V:nt and Drain Valves 3.1.8 .f3 .V 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves LCO 3.1.8 Each SDV vent and drain valve shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS
-------------------------------------NOTE----------------__--_---__...-..-_-_.
Separate Condition entry is allowed for each SDV vent and drain line. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SDV vent A.1 Restore valve to 7 days or drain lines with OPERABLE status. one valve inoperable. O - B. One or more SDV vent 8.1 --------NOTE--------- or drain lines with An isolated line may both valves be unisolated under inoperable. administrative control to allow , draining and venting ! of the SDV. Isolate the 8 hours associated line. I C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. Brunswick Unit 1 3.1-25 Amendment No.
i' SDV Vent and Drain Valves 3.1.8 i i SURVEILLANCE REQUIREMENTS p SURVEILLANCE FREQUENCY-SR 3.1.8.1 -------------------NOTE-------------------- Not required.to be met on vent and drain valves closed during performance of SR 3.1.8.2. l Verify each SDV_ vent and drain valve is 31 days l open.- I l SR 3.1.8.2 Cycle each SOV vent and drain valve to the 31 days l fully closed and fully open position. SR 3.1.8.3 Verify each SDV vent and drain valve: 24 months l l a. Closes in s 30 seconds after receipt l of an actual or simulated scram l signal; and
- b. Opens When the actual or simulated scram signal is' reset.
l l g j Brunswick Unit ~l 3.1-26 Amendment No. !~ L
I l Centrol Rod Scram Times B 3.1.4 IO 'Q BASES (continued) ! SURVEILLANCE The four SRs of this LCO are modified by a Note stating that REQUIREMENTS during a single control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram accumulator. With the CR0 pump isolated, (i.e., charging valve closed) the influence of the CRD pump head does not i affect the single control rod scram times. During a full l core scram, the CRD pump head would be seen by all control l rods and would have a negligible effect on the scram l insertion times. SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is l based on an assumed control rod scram time. Measurement of l the scram times with reactor steam dome pressure 2 800 psig demonstrates acceptable scram times for the transients analyzed in Reference 4. Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing l effects of reactor steam dome pressure and stored l( ,\ accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure 2 800 psig ensures that the measured scram times will be within the specified limits at higher pressures. This test is performed for each control rod from its fully withdrawn position. Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed. To ensure that scram time testing is performed within a reasonable time following fuel movement within the reactor pressure vessel or following a shutdown 2120 days, all b control rods are required to be tested before exceeding 40% RTP following the shutdown. The specified frequencies are acceptable considering the additional surveillances lb performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by work on control d rods or the CR0 System. Q (continued) t Brunswick Unit 1 B 3.1-24 Revision No. ! I
Centrol Rod Scram Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.2 REQUIREMENTS (continued) Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods. The sample remains representative if no more than 20% of the control rods in the sample tested are determined to be " slow." With more than 20% of the sample declared to be " slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 20% criterion (i.e., 20% of the entire sample size) is satisfied, or until the-total number of " slow" control rods (throughout the core, from all surveillances) exceeds the i LC0 limit. For planned testing, the control rods selected for the sample should be different for each test. This test is performed for each control rod in the sample from its , fully withdrawn >osition. Data from inadvertent scrams ! should be used wienever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample. The 120 day
' l Frequency is based on operating experience that has shown control rod scram times do not significantly change over an l o)erating cycle. This Frequency is also reasonable based on :
q tie additional Surveillances done on the CRDs at more frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, " Control Rod Scram Accumulators." SR 3.1.4.3 . 1 When work that could affect the scram insertion time is ! performed on a control rod or the CR0 System, testing must l be done to demonstrate that each affected control rod i retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The ! required scram time testing must demonstrate the affected i control rod is still within acceptable limits. This test is performed for each affected control rod from its fully withdrawn position. In lieu of actually initiating a scram for each affected control rod, testing that adequately demonstrates the scram times are within acceptable limits is allowed to satisfy this SR. The test may include any series of sequential, overlapping, or total steps so the entire (continued) O Brunswick Unit 1 B 3.1-25 Revision No.
- i. , 1
[ l l Centrol Rod Scru Times B 3.1.4 BASES , l
- l. SURVEILLANCE SR 3.1.4.3 (continued)
L REQUIREMENTS f scram time sequence is verified. The limits for reactor .I L pressures < 800 psig are established based on a high probability of meeting the acceptance criteria at reactor
. pressures 2 800 psig. Limits for 2 800 psig are found in l Table 3.1.4-1 and do not apply for testing performed at
< 800 psig. If testing demonstrates the affected control rod does not meet these limits, but is within the 7-second l limit of Note 2 to Table 3.1.4-1, the control rod can be considered OPERABLE and " slow."
Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any - CRD for maintenance or modification; replacement of a ! control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, ~ accumulator, isolation - valve or check valve in the piping required for scram. The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. l SR 3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System, testing must be performed to demonstrate each affected control rod is still b within the scram time limits of Table 3.1.4-1 with the reactor steam dome pressure 2 800 psig. Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test. 4 For a control rod affected by work )erformed while shut down, however, a zero pressure and iigh pressure test may be required. This testing ensures that, prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions. This test is performed for each-affected control rod from its fully withdrawn position. Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria. g, (continued) 'O Brunswick Unit 1 B 3.1-26 Revision No.
Ccntrol Rod Scra: Times B 3.1.4 BASES SURVEILLANCE SR 3.1.4.4 (continued) REQUIREMENTS The Frequency of prior to exceeding 40% RTP is acceptable because of the capability tc test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. REFERENCES 1. USFAR, Section 3.1.2.2.1.
- 2. UFSAR, Section 4.2.1.1.8.
- 3. UFSAR, Section 4.3.2.
- 4. UFSAR, Chapter 15.
- 5. Letter from R.F. Janecek (BWROG) to R.W. Starostecki (NRC), BWR Owners Group Revised Reactivity Control System Technical Specifications, BWROG-8754, September 17, 1987.
- 6. 10 CFR 50.36(c)(2)(ii).
O O Brunswick Unit 1 B 3.1-27 Revision No.
Centrol Rod Scran Times 3.1.4 i 3.1 REACTIVITY CONTROL SYSTEMS (m] I 3.1.4 Control Rod Scram Times i LCO 3.1.4 a. No more than 10 OPERABLE control rods shall be " slow," ; in accordance with Table 3.1.4-1; and l l
- b. No more than 2 OPERABLE control rods that are " slow" shall occupy adjacent locations.
APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Be in MODE 3. 12 hours LCO not met. O ! SURVEILLANCE REQUIREMENTS l 1 ____...__.___.._________________.----NOTE------------------------------------- I During single control rod scram time Surveillances, the control rod drive l (CRD) pumps shall be isolated from the associated scram accumulator. SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify each control rod scram time is Prior to within the limits of Table 3.1.4-1 with exceeding reactor steam dome pressure 2 800 psig. 40% RTP after fuel movement within the reactor g, pressure vessel AND (continued)
\ Brunswick Unit 2 3.1-12 Amendment No.
Centrol Rod ScralTimes 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1-(continued) Prior to exceeding 40% RTP after each reactor. shutdown 2 120 days SR 3.1.4.2 Verify, for a representative sample, each 120 days tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in dome pressure 2 800 psig. MODE 1 SR 3.1.4.3 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with declaring any reactor steam dome pressure. control rod OPERABLE after work on control O rod or CR0 System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with exceeding reactor steam dome pressure 2 800 psig. 40% RTP after work on control rod or CR0 System that could affect i scram time ' b' 1 O Brunswick Unit 2 3.1-13 Amendment No. i
Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1) Control Rod Scram Times ....................................-NOTES-------------------------..--.--....
- 1. OPERABLE control rods with scram times not within the limits of this Table are considered " slow."
- 2. Enter applicable Conditions and Required Actions of LCO 3.1.3, " Control Rod OPERABILITY," for control rods with scram times > 7 seconds to notch position 06. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered " slow."
SCRAM TIMES WHEN E NOTCH POSITION REACTOR PRESSURE STEAM 2: 800 psigt D0f(a)(b)
-(seconds) 46 0.44 36
~1.08 26 1.83 06 3.35 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero. l 1
(b). When reactor steam dome pressure is < 800 psig, established scram time ! limits apply. j i l Brunswick Unit 2 3.1-14 Amendment No.
i l Centrol R:d Scram Accumulators 3.1.5 l l 3.1 REACTIVITY CONTROL SYSTEMS l 3.1.5 Control Rod Scram Accumulators LC0 3.1.5 Each control rod scram accumulator shall be OPERABLE. l l l APPLICABILITY: MODES I and 2. ACTIONS
._----NOTE-------------------------------...- -
Separate Condition entry is allowed for each control rod scram accumulator. CONDITION REQUIRED ACTION COMPLETION TIME A. One control rod scram A.1 --------NOTE--------- accumulator inoperable Only applicable if with reactor steam the associated dome pressure control rod scram O 2: 950 psig. time was within the limits of Table 3.1.4-1 during the last scram time Surveillance. Declare the 8 hours associated control rod scram time
" slow."
08 A.2 Declare the 8 hours associated control rod inoperable. (continued) O Brunswick Unit 2 3.1-15 Amendment No.
Centrol Rod Scraa Accumulators 3.1.5 ACTIONS (continued) j CONDITION REQUIRED ACTION COMPLETION TIME l .B.-.Two or more control B.1 Restore charging 20 minutes from discovery of rod scram accumulators water header pressure inoperable with ' to a 940 psig. Condition B reactor steam dome concurrent with pressure a 950 psig. charging water header pressure
< 940 psig 8N!Q B.2.1 --------NOTE---------
Only applicable if the associated control rod scram time was within the limits of Table 3.1.4-1 during the last scram time Surveillance. O Declare the associated control rod scram time 1 hour
" slow."
OE B.2.2 Declare the I hour associated control rod inoperable. (continued) O Brunswick Unit 2 3.1-16 Amendment No.
Central Rod Scram Accumulaters 3.1.5 (g ,j ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control C.1 Verify all control Immediately upon ! rod scram accumulators rods associated with discovery of inoperable with inoperable charging water i reactor steam dome accumulators are header pressure i pressure < 950 psig. fully inserted. < 940 psig l AND C.2 Declare the 1 hour l associated control i rod inoperable. D. Required Action B.1 D.1 --------NOTE--------- or C.1 and associated Not applicable if all Completion Time not inoperable control met. rod scram accumulators are : associated with fully l inserted control g rods. s ..................... l Manually scram the Immediately reactor. i SURVEILLANCE REQUIREMENTS . SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator 7 days pressure is 2: 940 psig. Brunswick Unit 2 3.1-17 Amendment No.
)
Rod Pattern Control 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control LCO 3.1.6 OPERABLE control rods shall comply with the requirements of the banked position withdrawal sequence (BOWS). APPLICABILITY: MODES I and 2 with THERMAL POWER s 10% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more OPERABLE A.1 --------NOTE--------- control rods not in Control rod may be compliance with BPWS. bypassed in the rod worth minimizer (RWM) or RWM may be bypassed as allowed by LCO-3.3.2.1,
" Control Rod Block O Instrumentation."
Move associated 8 hours control rod (s) to correct position. DE A.2 Declare associated 8 hours : c,ontrol rod (s) ! inoperable. i (continued) Brunswick Unit 2 3.1-18 Amendment No.
Rod Pattern Centrol 3.1.6 /~~N (j ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Nine or more OPERABLE B.1 --------NOTE--------- control rods not in Control rod may be compliance with BPWS. bypassed in the RWM or RWM may be bypassed as allowed by LCO 3.3.2.1. Suspend withdrawal of Immediately control rods. AND B.2 Manually scram the 1 hour reactor. / SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply 24 hours with BPWS. l I Brunswick Unit 2 3.1-19 Amendment No.
SLC Systea 3.1.7 ~Q,a 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE. APPLICABILITY: MODES I and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable. to OPERABLE status. B. Two SLC subsystems B.1 Restore one SLC 8 hours inoperable. subsystem to OPERABLE status. O C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. l l l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium 24 hours pentaborate solution is within the limits of Figure 3.1.7-1. (continued) O Brunswick Unit 2 3.1-20 Amendment No.
SLC Systea 1 3.1.7 l l
/~'N Q SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours solution is within the limits of Figure 3.1.7-2. SR 3.1.7.3 Verify temperature of pump suction piping 24 hours is within the limits of Figure 3.1.7-2. SR 3.1.7.4 Verify continuity of explosive charge. 31 days SR 3.1.7.5 Verify the concentration of boron in 31 days solution is within the limits of Figure 3.1.7-1. AND Once within T 24 hours after
.. water or boron is added to solution AND Once within 24 hours after solution temperature is restored within the limits of Figure 3.1.7-2 SR 3.1.7.6 Verify each pump develops a flow rate In accordance 2 41.2 gpm at a discharge pressure with the a 1190 psig. Inservice Testing Program (continued)
O Brunswick Unit 2 3.1-21 Amendment No.
1 SLC System 3.1.7 SURVEILLANCE REQUIRENENTS (continued) SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify flow through one SLC subsystem from 24 months on a pump into reactor pressure vessel. STAGGERED TEST BASIS O l O Brunswick Unit 2 3.1-22 Amendment No.
SLC System 3.1.7 O 22.0 21.0 y 20.0 t: bg 19.0 - g}. g 18.0 ------ - l MA - ACCEPTABLE REGION Wg 17.0 - Iz - l QO 16.0 ------- ' 15.0
~
UNACCEPTABLE
, @W@ 14.0 REGION 13.0 ;
12.0 11.0 1 I I 10.0 2000 2500 3000 3500 4000 4500 C000 NETVOLUME OF SOLUTION IN TANK (gals) Figure 3.1.7-1 (page 1 of 1) Sodium Pentaborate Solution Volume Versus Concentration Requirements O Brunswick Unit 2 3.1-23 Amendment No.
l l SLC System 3.1.7 i i 150 140 a (# 8 * *
. . =
l - l l.k bj 5 c - 4 130 8 * ' I h3CCEPTABLE i
% N (REGION] R ; '
~
- UNACCEPTABLE REGION h kl lN T% .' ' 't l1
^ .~ ',~ ~! '
110
.1:
i E oo ; a s -- m e p - m !p /r
@ 00
- E h ~
w 80 I ff ! (3 w/ W 70 " f UNACCEPTABLE 60 50 I , l 0 5 10 13 15 17 19 21 25 14 16 18 20 SODIUM PENTABORATE SOLUTION CONCENTRATION (% by weight) l I l Figure 3.1.7-2 (page 1 of 1) I Sodium Pentaborate Solution Temperature Versus Concentration Requirements ! (G
'j Brunswick Unit 2 3.1-24 Amendment No.
l (
I SDV Vent and Drain Valves i 3.1.8 I L 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves LCO 3.1.8 Each SDV vent and drain valve shall be OPERABLE. i APPLICABILITY: MODES I and 2. ACTIONS l
............................N0TE------------------------------------- i Separate Condition entry is allowed for each SDV vent and drain line.
]
l I CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SDV vent A.1 Restore valve to 7 days or drain lines with OPERABLE status. I one valve inoperable. ' B. One or more SDV vent B.1 --------NOTE--------- or drain lines with An isolated line may both valves be unisolated under inoperable, administrative control to allow i draining and venting of the SDV. l Isolate the 8 hours ; associated line. l C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. fVO l Brunswick Unit 2 3.1-25 Amendment No. 1
i-I SDV VInt and Drain Valves 3,1.8 l'i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 -------------------NOTE-------------------- Not required to be met on vent and drain valves closed during performance of l- SR 3.1.8.2. Verify each SOV vent and drain valve is 31 days open. SR 3.1.8.2 Cycle each SDV vent and drain valve to the 31 days fully closed and fully open position. L SR 3.1.8.3 Verify each SDV vent and drain valve: 24 months
- a. Closes in s 30 seconds after receipt of an actual or simulated scram i signal; and
- b. Opens when the actual or simulated scram signal is reset.
L: l l i l O Brunswick Unit 2 3.1-26 Amendment No. , i 1
( i Centrol Rod Scraa Times B 3.1.4 I i Dr (Q BASES (continued) SURVEILLANCE The four SRs of this LC0 are modified by a Note stating that i REQUIREMENTS during a single control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated. (i.e., charging j valve closed) the influence of the CRD pump head does not l affect the single control rod scram times. During a full j core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times. i 1 SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on an assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure 2 800 psig demonstrates acceptable scram times for the transients analyzed in Reference 4. I Maximum scram insertion times occur et a reactor steam dome pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored
- accumulator energy. Therefore, demonstration of adequate
!( w scram times at reactor steam dome pressure 2 800 psig ensures that the measured scram times will be within the specified limits at higher pressures. This test is performed for each control rod from its fully withdrawn l l position. Limits are specified as a function of reactor l pressure to account for the sensitivity of the scram ' insertion times with pressure and to allow a range of l pressures over which scram time testing can be performed. To ensure that scram time testing is performed within a reasonable time following fuel movement within the reactor pressure vessel or following a shutdown 2120 days, all d control rods are required to be tested before exceeding 40% RTP following the shutdown. The specified Frequencies are acceptable considering the additional serveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by work on control b rods or the CRD System. g (continued) r Ov Brunswick Unit 2 B 3.1-24 Revision No.
t ! C:ntrol Rod Scraa Times l B 3.1.4 i p d BASES i SURVEILLANCE SR 3.1.4.2 REQUIREMENTS (continued) Additional testing of a sample of control rods is required i to verify the continued performance of the scram function ' during the cycle. A representative sample contains at least 10% of the control rods. The sample remains representative if no more than 20% of the control rods in the sample tested are determined to be " slow." With more than 20% of the rample declared to be " slow" per the criteria in lable 3.1.4-1, additional control rods are tested until this 20% criterion (i.e., 20% of the entire sample size) is satisfied, or until the total number of slow" control rods (throughout the core, from all surveillances) exceeds the LC0 limit. For planned testing, the control rods selected for the sample should be different for each test. This test is performed for each control rod in the sample from its fully withdrawn position. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample. The 120 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle. This Frequency is also reasonable based on the additional Surveillances done on the CRDs at more k frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, " Control Rod Scram Accumulators." i SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The required scram time testing must demonstrate the affected control rod is still within acceptable limits. This test is performed for each affected control rod from its fully withdrawn position. In lieu of actually initiating a scram for each affected control rod, testing that adequately demonstrates the scram times are within acceptable limits is allowed to satisfy this SR. The test may include any series of sequential, overlapping, or total steps so the entire (continued 1 A Brunswick Unit 2 8 3.1-25 Revision No.
l [. Centrol Rod Scram Times I 8 3.1.4 BASES l !~ SURVEILLANCE SR 3.1.4.3 (continued) i l REQUIREMENTS scram time sequence is verified. The limits for reactor pressures < 800 psig are established based on a high , probability of meeting the acceptance criteria at reactor pressures 2 800 psig. Limits for 2 800 psig are found in Table 3.1.4-1 and do not apply for testing performed at
< 800 psig. If testing demonstrates the affected control rod does not meet these limits, but is within the 7-second limit of Note 2 to Table 3.1.4-1, the control rod can be considered OPERABLE and " slow."
Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a l control rod; and maintenance or modification of a scram ! solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram. The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of o)erating conditions and the more frequent surveillances on ot1er aspects of control rod OPERABILITY. SR 3.1.4.4 When work that could affect the scram insertion time is i performed on a control rod or CRD System, testing must be performed to demonstrate each affected control rod is still d l within the scram time limits of Table 3.1.4-1 with the reactor steam dome pressure 2 800 psig. Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test. For a control rod affected by work performed while shut down, however, a zero pressure and high pressure test may be ; i required. This testing ensures that, prior to withdrawing l the control rod for continued operation, the control rod l scram performance is acceptable for operating reactor pressure conditions. This test is performed for each affected control rod from its fully withdrawn position. Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria. A.__ l-(continued) Brunswick Unit 2 B 3.1-26 Revision No.
Centrol Rod Scram Times B 3.1.4 n d BASES SURVEILLANCE SR 3.1.4.4 (continued) REQUIREMENTS The Frequency of prior to exceeding 40% RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. l REFERENCES 1. USFAR, Section 3.1.2.2.1.
- 2. UFSAR, Section 4.2.1.1.8.
- 3. UFSAR, Section 4.3.2. )
! 4. UFSAR, Chapter 15. i ! 5. Letter from R.F. Janecek (BWROG) to R.W. Starostecki (NRC), BWR Owners Group Revised Reactivity Control System Technical Specifications, BWROG-8754, September 17, 1987. , ! 6. 10 CFR 50.36(c)(2)(ii). r U l
- . I l
1 t I O Brunswick Unit 2 B 3.1-27 Revision No. l t
< p... ..... -t ; % ., +. . ~. :~ps 's u . .. . ,,...v.c ,_3 ,
o u. .m: 1
- 3. f. 3 . .
L v Q I __, i REACTIVITY CONTROL SYSTEMS - I. CONTROL ROD MAXIMUM SCRAM INSERTION TIMES f. LIMITINC CONDITION FOR OPERATION 3.1.3.2 The easimum scram insertion time of each control rod from tha futy
* * *! ~ withdrawn position to notch position 6,Choed on de-eneraisation of tne scram
(' fod,,lg,.(y ' .r pilot valve solenoids as time _ sero) shall not escoed 7.0 seconds. 7' l I APPLICARILITY: CONDITIONS 1 and 2. l ACTION: With the nazimum scram insertion time of one or more control rods escoeding ' 7.0 seconds, operation may continue and the provisions of Specification 3.0.4 are not applicable provided thats
- a. The control rod with the slou insertion time is declared inoperable, 4
- b. The requirements of Specification 3.1.3.1 are satisfied', and t
- c. If within the preset power level of the,RWM, the requirements of l.
# Specification 3.1.4.1.d are also satisfied, and The Surveillance Requirements of Specification 4.1.3.2.c are l j d.
performed at least once per 92 days when operation is continued with j' )
*/ [\ ' three or more control rods with sicu scram insertion times; Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
l- \ h Asle 'lo b4%* f*P-A SURvrltt.ANCE REQUIREMENTS _u _ The maximungeram ins _ertion time et tne control rods shall be
' demonstrated (thseeth messagemen9_ [WA mbr-s4e., c6,e p,e:, hee, g,toopif For sit control rods prio_r to THERMAL POWER exceedint 40% of RATED i.A.1 THERMA 1. POWER f ollowing(CORE-m.is.uAT4C@ after reactor shutdown
, g g ,, -
than 10 days 4.k
-w -% e% g**._
,
- d,:
q r @ that isbreste A ees For fspepeir.atH)af acte ind ual controt room sottowans - anEt on or'9581Ticatawaithe control rod or rod drive system
. ')
f g 3.l.dl.d/ h *^ 1 wFich could affect the scram (ipet'tishytime of those specific control ]' , ! r ., and 44q y, l t: , 41 radr:'~on a M 2 na. bet' ll at least once per
, 1d C For QUI oer free ^[
fg 1,l,al.2, Q120 days of operatioy. j.
$M5lJY".'I.4 fA.g
'(T5 se n." .
~
k M. i 3/4 1-5 Amendment No. 144 SRUNSWICK - UNIT 1 l' lok h Ti;
~
[
. u .. m . .
OCMCdko/\ 3.l. I ( bec S *
'3 I 5 .
i ( REACTIVITY CONTROL SYSTEMS CONTROL ROD MAXIMUM SCF AM INSERTION TIMES
- i l tTHITINC CONDITION FOR OP.ERATION
.(g 74 3.1.3.2 The maximum scram insertion time of each control rod from the full thdrawn position to notch position,6,(based on de energssat6on of the scram p Tot valve solenoids as time xerb shall not exceed Ja0 seconds.
[shok(a} APPLICABILITY 8 OPERATIONAL CONDITIONS 1 and 2. l ACTIONI With the maximum scram insertion time of one or more control rods exceeding 7.0 seconds, speration may continue and the provisions of Specification 3.0.4 i are not applicable provided thatt
- 4. The control rod with the slow insertion time is declared inoperable, 4
- b. The requirements of Specification 3.1.3.1 are satisfied, and ]
l c. If within the preset power level of the RWM, the requirements of l Specification 3.1.4.1.d are also satisfied, and
- d. The Surveillance Requirements of Specification 4.1.3.2.c are l performed at least once per 92 days when operation is continued with
.) h,\ three or more control rods with slow scram insertion timest l / I ' Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
f t gAw w.4 f 3.cghq SURVEILLANCE REQUIREMENTS i N The maximumMinpertion time of the control rods shall be G4 emonstrated(thpirth spurepefftMt.3;& cta,3er- dem do,.4 p.ess-M. :n 800 Peg h-i @ For all control rods prior to THERMAL POWER exceeding 40% of RATED eactor shutdown jf.3l.d.l -- THERMAL POWER fo11owinguCfC9Nr yTrtRAJ10N
" " a " W m s e ._ , ~ ~ _ ,.. _.9 4 h y.&j h,, 4. h ForfpeificsMg af fected individual control rods f ollowing b I L *g.
DJ mattre<oanh on %dimiert-4o3 the control rod or rod drive system which could affect the scram (Sed'rtJ43 time of those specific cor trol [.3 NO/s Sfy *^ rods e and gg
@ For(10Wheprok-fods efr~s reeStingArs' sis) at least once per !
. 120 days et operatic .
in MCI l 9"fr}sesty _ u s.i.v.4 -
@ SR.s.t.4.s
" iRuMSWrCi: UNIT 2
^ 2- 3/4 1-5 Amendment No. 175
.)
~
ppsk O v I l [ .
DISCUSSION OF CHANGES
~
ITS: 3.1.4 - CONTROL ROD SCRAM TIMES TECHNICAL CHANGES - LESS RESTRICTIVE LA.1 Surveillances are not required to be in the Technical ) (cont'd) Specifications to provide adequate protection of the public health and safety. Changes to the Bases will be controlled by the provisions of the Bases Control Program described in Chapter 5 of the ITS. LA.2 ITS SR 3.1.4.2 requires a " representative sample" of control rods be tested each 120 days of power operation in MODE 1 instead of l the currently required "10% of the control rods, on a rotating basis," (CTS 4.1.3.2.c). The details of what constitutes a j representative sample are to be relocated to the Bases. The requirements of ITS 3.1.4 and associated Surveillance Requirements are adequate to ensure scram time testing is performed. Therefore, the relocated details of what constitutes a representative sample are not required to be in the Technical Specifications to provide adequate protection of the public health and safety. Changes to the Bases will be controlled by the l provisions of the proposed Bases Control Program described in Chapter 5 of the ITS.
" Specific" L.1 CTS 4.1.3.2.a requires control rod scram time testing for all control rods prior to exceeding 40% RTP following CORE jG ALTERATIONS. The term " CORE ALTERATIONS" in CTS 4.1.3.2.a is lV replaced with " fuel movement within the reactor pressure vessel" in ITS SR 3.1.4.1. The intent of CTS 4.1.3.2.a is to verify that i'
- the negative reactivity insertion rate assumed in the safety analysis is still maintained after in vessel operations which could have significantly altered the negative reactivity insertion !
l rate. Although normal control rod movement is considered a CORE : ALTERATION with the reactor vessel head removed, this activity l does not have an impact on the negative reactivity insertion rate. Since normal control rod movement does not require verification l that the negative reactivity insertion rate assumed in the safety l analysis is still maintained during normal operations with the l vessel head on, it should also be allowed without a requirement to A i
! verify that the negative reactivity insertion rate assumed in the /_C_\ i safety analysis is still maintained with the reactor vessel head removed (i.e., defined as a CORE ALTERATION). The wording of ITS !
SR 3.1.4.1 provides a specific list of those CORE ALTERATIONS i which constitute in vessel operations which could significantly ' alter the negative reactivity insertion rate; specifically excluding control rod movement. In addition, scram time testing of control rods after in vessel operations related to control rod movement that could significantly alter the negative reactivity ! insertion rate (after work on a control rod that could affect ! scram times, e.g., CR0 replacement or CRDM overhaul) is still
. required to be performed by ITS SR 3.1.4.3 and ITS SR 3.1.4.4.
l l l
i DISCUSSION OF CHANGES ITS: 3.1.4 - CONTROL R00 SCRAN TINES TECHNICAL CHANGES - LESS RESTRICTIVE L.1 Therefore, this change is acceptable since the Surve111ances of (cont'd) ITS 3.1.4 are kdequate to ensure that the negative reactivity insertion rate assumed in the safety analysis is maintained. b Additionally, the reliability of the control rods is increased , since this change eliminates unnecessary testing for the control rods. RELOCATED SPECIFICATIONS { None I l 1 l O > l I
r l b(3 ./ Control Rod Scram Times 3.1.4 C,15 /boc.
- 3. ,1 REACTIVITY CONTROL SYSTEMS ~
l 3.1.4 Control Rod Scram Times LCO 3.1.4 a. No more than1107 0PERABLE control rods shall be " slow,' 3,g,g,3 in accordance with Table 3.1.4-1; and 5 si .s.4 b. No more than 2 OPERABLE control rods that are " slow"
/8 A shall occupy adjacent locations.
. APPLICABILITY: MODES 1 and 2.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l u,,a A. Requirements of the A.1 Be in MODE 3. 12 hours Q pcyog LCO not met, s.t.id mA , es) lf *Y t - SURVEILLANCE REQUIREMENTS __ .____- ____ .. NOTE---- During single control rod scram time Surve111ances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.
/M.t l
SURVEILLANCE FREQUENCY
~
SR 3.1.4.1 Verify each control rod scram time is Prior to - within the limits of Table 3.1.4-1 with exceeding
# 31*4 reactor steam dome pressure 218 psig. 40% RTP after
/AyLA.; fuel movement within the 4.t 31L reactor pressure vessel g
4.t d i
/u.) AN,Q (continued) l l
BWR/4-ST 3.1-12 Rev i, 04 U/795
/ /
j ,
~
., , .u r sra l
1 I Control Rod Scram Times . 3.1.4 cis/poc Stf2VEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 (continued) Prior to y t.S.2.a. . exceeding 40%.RTP after
/#'1 "' 8 each reactor D shutdown h 120 days SR 3.1.4.2 Verify, for a representative sample, each 120 days
- 4. p,1, c, tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in
/A.7., La.I, dome pressure 2 J800Kpsi . MODE 1 LA,1 4,i.t.a i
)$7i SR 3.1.4.3 Verify each affected control rod scram time Prior to l
' /M 2. is within the limits of Table 3.1.4-1 with declaring 4 I
any reactor steam dome pressure. control rod OPERABLE after (
.N ')
work on control rod or CRD System that could affect scram time
]
SR 3.1.4.4 Verify each affected control rod scram time Prior to ) 4 l.5.1.b is within the limits of Table 3.1.4-1 with exceeding ' reactorsteamdomepressure2f800},psig. 40% RTP after j
.z' 4,D work on control 08 1 rod or CRD 1 4, g,y ; System that i could affect .
[ scram time A ' d wn/g sis 3.1-13 -Ecv i, 04/G7/i N
. *e+. . ,e' .
-JUSTIFICATION FOR DEVIATIONS FROM NUREG-1433, REVISION 1 SECTION 3.1 - REACTIVITY CONTROL SYSTEMS O 1. The brackets are removed and the proper plant specific information/value is been provided.
- 2. The Frequency of NUREG-1433 SR 3.1.2.1 (BNP SR 3.1.2.1) . is changed in the BNP ITS from 1000 MWD /T to 1100 MWD /T since the average core exposure at BNP is tracked in metric tons. 1100 MWD / metric ton is approximately 1000 MWD /T. The BNP Bases clarifies that the units are in metric tons.
- 3. NUREG-1433 Specification 3.1.3, " Control Rod OPERABILITY," Required Actions associated with condition A and Required Action C.1 are modified by Notes that' allow bypassing the RWM. Additionally, NUREG-1433 Specification 3.1.6, " Rod Pattern Control," Required Actions A.1 and B.1 are also modified by Notes that allow bypassing the RWM. The corresponding Notes in BNP ITS 3.1.3 and ITS 3.1.6 are modified to include the allowance to bypass individual inoperable control rods in 4 the RWM. This change is consistent with current BNP licensing basis as reflected in CTS 3.1.4.1 Actions.
- 4. This bracketed requirement and associated information or information that references the bracketed information is deleted because the requirements are not applicable to BNP. The following requirements are renumbered or revised, where applicable, to reflect this deletion.
- 5. Not used.
a
- 6. The Control Rod Scram Time Table (Table 3.1.4-1) in NUREG-1433 Specification 3.1.4 is revised to more completely reflect the deletion of the 0 psig scram time acceptance criteria from the Table. The deletion for the o psip scram time acceptance criteria was approved per i Generic Change BWR-13, C6, and Revision 3 to BWR-13, C.6. Note (b) is revised to state "When reactor steam dome pressure is < 800 psig, established scram time limits apply."' In addition editorial changes are i made to the heading of the scram time column of the Table due to the '
deletion of the 0 psig scram time acceptance criteria.
- BNP UNITS 1 & 2 1 Revision 0
- ~
L ,.) l- Control Rod Scram Times o B 3.1.4 BASES (continued) SURVEILLANCE The four SRs of this LC0 are modified by a Note stating that REQUIR&.ENTS during a single control rod scram time surveillance,. the CRD pumps shall be isolated from the associated scram l accumulator. With the CRD. pump isolated, (i.e., charging valve closed) the influence of the CRD pump head does not affect the single control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times. SR 3.1.4.1 ~' The scram reactivity used in DBA and transient analyses is p ' based on an assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure 2 800 psig
, , demonstrates acceptable scram times for the transients analyzed in Reference @
Maximum scram insertion times occur at a reactor steam done - pressure of approximately 800 psig because of the competing I
. effects of reactor steam done pressure and stored i
i' i accumulator energy. Therefore, demonstration of adequate [7h4e#kpJu-ch scram times at reactor steam dose pressure 1800 psig .p, g 4 ,a ensures that the measured scram times will be within the k "".%"" specified limits at higher pressures.# Limits are specified as a function of reactor pressure to account for the. D@ D ' W ID .[' yg sensitivity of.the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performedm To ensure that scram time testing is performed within reasonable time following, fuel movement *" within the reactor ssure vessel 63Eh.a-shutdown 2 120 days control rods are required to be testedfollo% ' I beforeuelyovement reven exceeding 405 RTP.following (11mitef to m i othe cor shutdown. ells, fin is th[e b t intje t'of thi R that fily CRDs ociated4ith 4% I
'the j; ore cells,a facte the ts.are e'quiref l sto, tie sera ime test Howe ve /r,1 move if th'e react f shutd 20 day all control rede'are ran ed ta' , remain's Q h) s cram td tested the accitional survei ances requenc$wa.acceptuote puisv - v Tor control rod consicering n g ,Ig'g OPERABILITY, the frequent verification of adequate e accumulator pressure, and the required testing of control rods affected by work on control rods or the CAD System. ,
e i
~
(continued)
.i
.l I --8Wa/a STSP B 3.1-25 Rev%-04#12/95A I
g]v p .
1 l l l Insert B 3.1.4-1 140t used. i i l l l l l l i i
p
)
l i. O1 Control Rad Scram Times 8 3.1.4 BASES SURVEILLANCE SR 3.1.4.3 (continued) REQUIREMENTS Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram. The Frequency of once prior to declaring the affected
. control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. ,
SR 3.1.4.4 , , 4 When work that could affect the scram insertion time is performed on a control rod or CRD System, testing must be - to demonstrate each affected control rod is still within the Jimits of Table 3.1.4-1 with the reactor steam dome pressure 2 800 psig. Where work has been performed at
$cm- high reactor pressure, the requirements of SR 3.1.4.3 and O )s h SR 3.1.4.4 can be satisfied with one test. For a control rod affected by work performed while shut down, however, a zero pressure and high pressure test may be required.
testing ensures that, prior to withdrawing the control rodThis fN S,N ;, N for continued operation, the control rod scram performance c k is acceptable for operatina reactor oressure conditions. taf'%cM cah! redy VAlternatively, a control rod scram test during hydrostatic l km '.4 Cl% pressure testing could also satisfy both criteria. g W/pc... s s.n. The Frequency of @ prior to exceeding 40% RTP is acceptable because of the capability to test the control rod 3% p over a range of operating conditions and the more frequent surveillances on other aspects of control rod,0PERABILITY. REFERENCES 1. (10 E R lio. Appefidix A, # 3**E*M ib 1
- 2. QSAR, Section (9}.T.22:4 . 4 2. I. I. F
.3 J SAR, Section Q f I d (continued)
;- -MIR/A_STS B 3.1-27 1tev-17-04f07/95%
h ; lO } I. l
; .s . .>a . , ~ .
JUSTIFICATION FOR DEVIATIONS FROM NUREG-1433 REVISION 1 BASES SECTION 3.1 - REACTIVITY CONTROL SYSTEMS O 1. Changes are made (additions, deletions, and/or changes) to NUREG-1433 Bases which reflect the plant specific methodology, nomenclature, number, reference, system description, design analysis, or licensing basis description.
- 2. In the BNP safety evaluations, the SDM and CRDA analyses are mutually independent. The consideration of SDM is to assure that the reactor is shutdown and remains shutdown with the highest reactivity control rod withdrawn (and all other control rods inserted). Consequently, the consideration of SDM is no more appropriate for CRDA than it is for other accidents and transients. The CRDA assumes that the highest enthalpy control rod (it is highly probable that this will be different than the highest worth control rod determined for SDM) suddenly disengages from the stuck position and falls to the drive position.
Doppler reactivity tends to mitigate the event consequences with scram reactivity terminating the event. As a result, the NUREG-1433 Bases are modified to reflect the independence of the SDM and CRDA safety evaluations.
- 3. The criteria of the NRC Final Policy Statement on Technical Specifications Improvements have been included in 10 CFR 50.36(c)(2)(ii). Therefore, references in the NUREG-1433 Bases to the NRC Final Policy Statement are revised in the BNP ITS Bases to reference 10 CFR 50.36.
- 4. Editorial change made to NUREG-1433 Bases for enhanced clarity and readability of the BNP ITS Bases or to be consistent with similar i statements in other places in the NUREG-1433 Bases. l
- 5. The brackets are removed and the proper plant specific information/value )
is provided. l
- 6. Typographical / grammatical error corrected.
7.- The NUREG-1433 Bases is modified in the BNP ITS Bases to reflect those changes made to the associated BNP Specification. The following requirements are renumbered, where applicable, to reflect the changes.
- 8. The NUREG-1433 Bases is modified in the BNP ITS Bases to be consistent with the ITS Writer's Guide.
- 9. The Inservice Testing Program for BNP is not required to provide information for trend purposes. As a result, the wording associated with trending is not included in the BNP ITS Bases.
- 10. The NUREG-1433 Bases is modified in the BNP ITS Bases to more accurately /
reflect the requirements of the associated BNP Specification. ac\, O BNP UNITS 1 & 2 1 Revision 0
NO SIGNIFICANT HAZARDS EVALUATION I ITS: 3.1.4 - CONTROL R00 SCRAM TIMES ( L.1 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, Carolina Power & Light Company has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
CTS 4.1.3.2.a requires control rod scram time testing for all control ro'is prior to exceeding 40% RTP following CORE ALTERATIONS. The term
" CORE ALTERATIONS" in CTS 4.1.3.2.a is replaced with " fuel movement within the reactor pressure vessel" in ITS SR 3.1.4.1. The proposed change eliminates the performance of a Surveillance during a refueling activity that does not impact negative scram reactivity rate. Although normal control rod movement is considered a CORE ALTERATION with the reactor vessel head removed, this activity does not have an impact on the negative reactivity insertion rate. The Frequency of scram time testing is not considered to be an initiator of any analyzed event.
Therefore, this change does not significantly increase the probability I of a previously analyzed accident. The intent of CTS 4.1.3.2.a is to verify that the negative reactivity insertion rate assumed in the safety analysis is still maintained after in vessel operations which could have significantly altered the negative reactivity insertion rate. Since g o normal control rod movament is a refueling activity that does not impact ; ( negative scram reactivity rate and this activity is allowed to occur during routine operations without requiring a special performance of scram time testing, the consequences of an event occurring during normal control rod movement with the reactor vessel head installed are the same as the consequences of an event occurring during normal control rod movement with the reactor vessel head removed. In addition, scram time testing of control rods after in vessel operations related to control rod movement that could significantly alter the negative reactivity insertion rate (after work on a control rod that could affect scram times, e.g., CRD replacement or CRDM overhaul) is still required to be performed by ITS SR 3.1.4.3 and ITS SR 3.1.4.4. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical modification to the plant and does not introduce a new mode of operation. The Surveillances of ITS 3.1.4 continue to ensure that the negative reactivity insertion d rate assumed in the safety analys!s is maintained. Therefore, this change does not create the possibi'iity of a new or different kind of accident from any accident previously evaluated. O BNP UNITS 1 & 2 1 Revision 0
1 NO SIGNIFICANT HAZARDS EVALUATION ITS: 3.1.4 - CONTROL R0D SCRAM TIMES L.1 CHANGE (continued)
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change continues to verify that the negative reactivity rate assumed in the safety analysis is maintained after in vessel operations which could have significantly altered the negative reactivity insertion rate. As such, there is no impact on safety since d there is adequate assurance that the negative reactivity insertion rate assumed in the safety analysis is still maintained. Therefore, the proposed change does not involve a significant reduction in a margin of safety. O O l BNP UNITS 1 & 2 2 Revision 0
l SRM Instrumentaticn 3.3.1.2 i
/
l SURVEILLANCE REQUIREMENTS l
................................-----NOTE---------------------------------...--
i Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified condition. 1 1 SURVEILLANCE FREQUENCY l l SR 3.3.1.2.1 Perform CHANNEL CHECK. 12 hours l l l SR 3.3.1.2.2 ------------------NOTES------------------ A
- 1. Only required to be met during CORE 6 ALTERATIONS.
- 2. One SRM may be used to satisfy more A than one of the following, m Verify an OPERABLE SRM detector is 12 hours located in:
O (j a. The fueled region;
- b. The core quadrant where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region; and
- d. l l
1 L c. A core quadrant adjacent to where ! CORE ALTERATIONS are being performed, when the associated SRM is included j in the fuel region. i SR 3.3.1.2.3 Perform CHANNEL CHECK. 24 hours (continued) Brunswick Unit 1 3.3-13 Amendment No. l
SAM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l SR 3.3.1.2.4 ------------------NOTES-------------.---- l
- 1. Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
- 2. Not required to be met during a core spiral offload.
Verify count rate is 2 3.0 cps. 12 hours during CORE ALTERATIONS ANQ 24 hours g SR 3.3.1.2.5 Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.2.6 -------------------NOTE------------------ Not required to be performed until 12 hours after IRMs on Range 2 or below. d Perform CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.1.2.7 ------------------NOTES------------------
- 1. Neutron detectors are excluded.
- 2. Not required to be performed until 12 hours after IRMs on Range 2 or below.
Perform CHANNEL CALIBRATION. 24 months i Brunswick Unit 1 3.3-14 Amendment No. !
SRM Instrumentaticn , 3.3.1.2 [ Table 3.3.1.2-1 (pose 1 of 1) ( source Range Monitor Instrumentation APPLICABLE MODES OR OTHER REQUIRED SURVEILLANCE FUNCTical SPECIFIED CONDITIONS CHANNELS REQUIREMENTS
- 1. Source Renee Moniton 2(*) 3 SR 3.3.1.2.1 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.T 3,4 '2 st 3.3.1.2.3 st 3.3.1.2.4 st 3.3.1.2.6 SR 3.3.1.2.7 5 2(b) st 3.3.1.2.1 SR 3.3.1.2.2 st 3.3.1.2.4 SR 3.3.1.7.5 SR 3.3.1.2.7 (a) With IRMs on Range 2 or below.
(b) speclet novable detectors may be used in place of SPMs if connected to normat SRM circuits. ( Brunswick Unit 1 3.3-15 Amendment No.
PBDS 3.3.1.3 3.3 INSTRUMENTATION 3.3.1.3 Period Based Detection System (PBDS) LCO 3.3.1.3 One channel of PBDS instrumentation shall be OPERABLE. AND Each OPERABLE channel of PBDS instrumentation shall not indicate High-High Alarm. APPLICABILITY: THERMAL POWER and core flow in the Restricted Region specified in the COLR, THERMAL POWER and core flow in the Monitored Region specified in the COLR. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any OPERABLE PBDS A.1 Manually scram the Immediately O channel indicating High-High Alarm. reactor. B. Required PBDS channel B.1 --------NOTE-------- inoperable while in Only applicable if the Restricted Region. RPS Function 2.b, APRM Flow Biased Simulated Thermal Power-High, Allowable Value is
" Setup".
Initiate action to Immediately exit the Restricted Region. 0.B (continued) Brunswick Unit 1 3.3-16 Amendment No. 4
- PBDS 3.3.1.3 m
(,) ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Manually scram the Immediately reactor. C. Required PBDS channel C.1 Initiate action to 15 minutes inoperable while in exit the Monitored the Monitored Region. Region. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.3.1 Verify each OPERABLE channel of PBDS 12 hours (- instrumentation not in High-High Alarm. SR 3.3.1.3.2 Perform CHANNEL CHECK. 12 hours SR 3.3.1.3.3 Perform CHANNEL FUNCTIONAL TEST. 24 months Brunswick Unit 1 3.3-17 Amendment No.
Control Rrd B1cck Instrumentaticn 3.3.2.1 3.3 INSTRUMENTATION-3.3.2.1 Control Rod B;ock Instrumentation LCO 3.3.2.1 The control rod block instrumentation for each Function in Table 3.3.2.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.2.1-1. ACTIONS CONDITION REQUIRED ACTION COMPLETION TINE A. One rod block monitor A.1 Restore RBM channel 24 hours (RBM) channel to OPERABLE status, inoperable. B. Required Action and B.1 Place one RBM channel I hour associated Completion in trip. O- Time of Condition A not met. E Two RBM channels inoperable. C. Rod worth minimizer C.1 Suspend control rod Immediately ) (RWM) inoperable movement except by l during reactor scram. ) startup. E (continued) l O Brunswick Unit 1 3.3-18 Amendment No.
1 I Centrol Rod Block Instrumentaticn 3.3.2.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2= 1.1 Verify 2: 12 rods Immediately withdrawn. E C.2.1.2 Verify by Immediately administrative methods that startup with RWM inoperable, for reasons other than bypassed control rod (s), has not been performed in the last calendar year. AND 1 C.2.2 Verify movement of During control bypassed control rod movement rod (s) is in compliance with (-) \- banked position withdrawal sequence (BPWS) by a second licensed operator or other qualified member of the technical staff. D. RWM inoperable during D.1 Verify movement of During control reactor shutdown, bypassed control rod movement rod (s) is in accordance with BPWS by a second licensed operator or other qualified member of the technical staff. (continued) O' V Brunswick Unit 1 3.3-19 Amendment No.
I Control Rod Bleck Instrumentaticn 3.3.2.1 ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME l l E. One or more Reactor E.1 Suspend control rod Immediately l Mode Switch-Shutdown withdrawal. Position channels inoperable. AN_D E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies. SURVEILLANCE REQUIREMENTS
............N0TES------------------------------------
- 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod O 2.
Block Function. When an RBM channel is placed in an inoperable status solely for performance of required Surve111ances, entry into associated Conditions and hequired Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability. l SURVEILLANCE FREQUENCY l i SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. 92 days l (continued) l l Brunswick Unit 1 3.3-20 Amendment No.
I Centrol Rod Block Instrumentation 3.3.2.1 t ( l V SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY i SR 3.3.2.1.2 ------------------NOTE------------------- ! Not required to be performed until I hour i after any control rod is withdrawn at l s 10% RTP in MODE 2. Perform CHANNEL FUNCTIONAL TEST. 92 days l l l SR 3.3.2.1.3 ------------------NOTE------------------- Not required to be performed until I hour after THERMAL POWER is s 10% RTP in MODE 1. Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.4 ------------------NOTE------------------- s Neutron detectors are excluded.
........... _ . m ......................
Verify the RBM: 24 months l
- a. Low Power Range-Upscale Function is not bypassed when THERMAL POWER is 2: 29% RTP and s Intermediate Power Range Setpoint specified in the COLR.
g
- b. Intermediate Power Range-Upscale Function is not bypassed when THERMAL POWER is > Intermediate Power Range
- . Setpoint specified in the COLR and i
s High Power Range Setpoint specified e in the COLR. l c. High Power Range-Upscale Function is I not bypassed when THERMAL POWER is
> High Power Range Setpoint specified A .
in the COLR. U8 ! (continued) O Brunswick Unit 1 3.3-21 Amendment No.
l Control Rod Block Instrumentaticn 3.3.2.1 l [~
's SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.2.1.5 Verify the RWM is not bypassed when 24 months j THERMAL POWER is s 10% RTP. SR 3.3.2.1.6 ------------------NOTE------------------- Not required to be performed until I hour after reactor mode switch is in the shutdown position. l Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3 3.2.1.7 ------------------NOTE------------------- Neutron detectors are excluded. Perform CHANNEL CALIBRATION. 24 months k SR 3.3.2.1.8 Verify control rod sequences input to the Prior to RWM are in conformance with BPWS. declaring RWM OPERABLE following loading of sequence into RWH O Brunswick Unit 1 3.3-22 Amendment No.
1 l Centrol Rod Bleck Instrumentaticn 3.3.2.1 1 n Table 3.3.2.1 1 (page 1 of 1) ! Control Rod Stock Instrtmentation APPLICA8LE l MODES OR l OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWA8LE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE
- 1. Rod Btock Monitor
- a. Low Power Range --Upscale (a) 2 sa 3.3.2.1.1 (h)
SR 3.3.2.1.4 SR 3.3.2.1.7
- b. Intermediate Power (b) 2 sR 3.3.2.1.1 (h)
Range -4Jpscale SR 3.3.2.1.4 SR 3.3.2.1.7
- c. High Power Range -Upscale (c),(d) 2 Sk 3.3.2.1.1 (h)
SR 3.3.2.1.4 SR 3.3.2.1.7
- d. Inop (d),(e) 2 SR 3.3.2.1.1 NA
- e. Downscate (d),(e) 2 SR 3.3.2.1.1 NA SR 3.3.2.1.7
- 2. Rod Worth Minialzer 1(I) 2(I) 1 SR 3.3.2.1.2 NA
[m} l , sR 3.3.2.1.3 (/ SR 3.3.2.1.5 SR 3.3.2.1.8
- 3. Reactor Mode switch -shutdown (g) 2 SR 3.3.2.1.6 NA Position (a) THERMAL POWER t 29% RTP and 5 Intermediate Power Range setpoint specified in the COLR and MCPR < 1.70.
(b) THERMAL POWER > Intermediate Power Range Setpoint specified in the COLR and 5 High Power Range Setpoint specified in the COLR and MCPR < 1.70. (c) THERMAL POWER > High Power Range setpoint specified in the COLR and < 90% RTP and MCPR < 1.70. (d) THERMAL POWER t 90% RTP and MCPR < 1.40. (e) THERMAL POWER t 29% and < 90% RTP and MCPR < 1.70. ; (f) With THERMAL POWER 5 10% RTP. (g) Reactor mode sultrh in the shutdown position. (h) Attowable Value specified in the COLR.
/~
f
\'
Brunswick Unit 1 3.3-23 Amendment No.
l l Feedwater and Main Turbine High Water Level Trip Instrumentaticn 3.3.2.2 ,V (~) 3.3 INSTRUMENTATION 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation LCO 3.3.2.2 Three channels of feedwater and main turbine high water level trip instrumentation shall be OPERABLE. APPLICABILITY: THERMAL POWER 2: 25% RTP. l ACTIONS l
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One feedwater and main A.1 Place channel in 7 days turbine high water trip. level trip channel l((~')s inoperable. B. Two or more feedwater B.1 Restore feedwater and 4 hours l and main turbine high main turbine high l water level trip water level trip channels inoperable. capability. I i l C. Required Action and C.1 Reduce THERMAL POWER 4 hours , associated Completion to < 25% RTP. Time not met. l ! I l i I I l l I O Brunswick Unit 1 3.3-24 Amendment No. i
Feedwater and Main Turbine High Water Level Trip Instrumentaticn 3.3.2.2 n , U SURVEILLANCE REQUIREMENTS 1 ( -------------------------------------NOTE------------------------------------- l When a channel is placed in an inoperable status solely for performance of l required Surve111ances, entry into associated Conditions and Required Actions ! may be delayed for up to 6 hours provided feedwater and main turbine high water level trip capability is maintained. i l SURVEILLANCE FREQUENCY l SR 3.3.2.2.1 Perform CHANNEL CHECK. 24 hours 4 SR 3.3.2.2.2 Perform CHANNEL CALIBRATION. The 24 months l Allowable Value shall be s 207 inches. SR 3.3.2.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST, 24 months including valve actuation, u l l Brunswick Unit 1 3.3-25 Amendment No. i 1 l
PAM Instrumentatien 3.3.3.1 !6 h 3.3 INSTRUMENTATION 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation j LCO 3.3.3.1 The PAM instrumentation for each Function in Table 3.3.3.1-1 shall be OPERABLE. l l APPLICABILITY: MODES I and 2. i l ACTIONS l
....__........_______................N0TES------------------------------------
- 1. LCO 3.0.4 is not applicable.
- 2. Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME l A. One or more Functions with one required A.1 Restore required channel to OPERABLE 30 days d l channel inoperable. status. l
- B. Required Action and B.1 Initiate action in Immediately l associated Completion accordance with Time of Condition A Specification 5.6.6.
not met. i L C. One or more Functions C.1 Restore one required 7 days /\ with two required channel to OPERABLE O channels inoperable. status. l (continued) l Brunswick Unit 1 3.3-26 Amendment No. l
PAM Instrumentatien 3.3.3.1 ACTIONS (contir.aed) CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C Table 3.3.3.1-1 for not met. the channel. E. As required by E.1 Be in H0DE 3. 12 hours Required Action 0.1 and referenced in Table 3.3.3.1-1. F. As required by F.1 Initiate action in Immediately Required Action D.1 accordance with and referenced in Specification 5.6.6. Table 3.3.3.1-1. O SURVEILLANCE REQUIREMENTS
....................................-NOTE-------------------------------------
These SRs apply to each Function in Table 3.3.3.1-1. SURVEILLANCE FREQUENCY 4 SR 3.3.3.1.1 Perform CHANNEL CHECK. 31 days d SR 3.3.3.1.2 Perform CHANNEL CALIBRATION of the 92 days Drywell and Suppression Chamber H, and 02 Analyzers. (continued) O Brunswick Unit 1 3.3-27 Amendment No.
PAM Instrumentation 3.3.3.1 l SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY ! 1 ! SR 3.3.3.1.3 Perform CHANNEL CALIBRATION for each 24 months required PAM Instrumentation channel , except for Drywell and Suppression l Chamber H, and 0, Analyzers. i l O i l i 'O Brunswick Unit 1 3.3-28 Amendment No.
r l PAM Instrumentatien I 3.3.3.1
^j (s /
j Table 3.3.3.1 1 (page 1 of 1) i Post Accident Monitoring Instruaantation j CONDITIONS 1 REFERENCED REGUIRED FROM REQUIRED FUNCTION CMANNELS ACit0N D.1
- 1. Reactor Vesset Pressure 2 E
- 2. Reactor Vessel Water Level
- a. -150 inches to +150 inches 2 E
- b. O inches to +210 inches 2 E
- c. +150 inches to +550 inches 2 E
- 3. Suppression Chamber Water Level 2 E
- 4. Suppression Chamber Water Temperature 2 E
- 5. Suppression Chamber Pressure 2 E
- 6. Drywell Pressure 2 E T. Drywett Temperature 2 E
- 8. PCIV Position 2perpenetgtg E flow path
- 9. Drywell and suppression Chamber H,10, Analyzer 2 E
- 10. Drywett Area Radiation 2 F (a) Not required for isolation valves whose assoc {ated penetration flow path la isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
(b) only one position indication channet is required for penetration flow paths with only one Instatted control room indication channel. l l'\
\'} Brunswick Unit 1 3.3-29 Amendment No.
[
Remote Shutdown Monitoring Instrumentation 3.3.3.2 3.3 INSTRUMENTATION 3.3.3.2 Remote Shutdown Monitoring Instrumentation LCO 3.3.3.2 The Remote Shutdown Monitoring Instrumentation Functions ) shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS k bb 35b ks not $ppkkc$bke
- 2. Separate Condition entry is allowed for each Function.
I CONDITION REQUIRED ACTION COMPLETION TIME g- A. One or more required Functions inoperable. A.1 Restore required Function to OPERABLE-30 days d status. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized. (continued) Brunswick Unit 1 3.3-30 Amendment No.
1 l Remote Shutdown Monitoring Instrumentatien 3.3.3.2 3.3 'INSTRUMENTATIO;- 3.3.3.2 Remote Shutdown Monitoring Instrumentation l ! l L L LCO 3.3.3.2 The Remote Shutdown Monitoring Instrumentation Functions shall be OPERABLE. l APPLICABILITY: MODES I and 2. l
- ACTIONS l
..................................... NOTES---------------------------------.-.
- 1. LCO 3.0.4 is not applicable.
- 2. Separate Condition entry is allowed for each function.
CONDITION REQUIRED ACTION CONPLETION TINE n A. One or more required Functions inoperable. A.1 Restore required Function to OPERABLE 30 days b status. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion l Time not met. d i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required 31 days l instrumentation channel that is normally energized. l (continued) O Brunswick Unit 1 3.3-30 Amendment No. I
l Remote Shutdown M:nitoring Instrumentatien 3.3.3.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l SR 3.3.3.2.2 Perform CHANNEL CALIBRATION for each 24 months required instrumentation channel. 'O l I i O Brunswick Unit 1 3.3-31 Amendment No.
ATWS-RPT Instrumentatien 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation LCO 3.3.4.1 Two channels per trip system for each ATWS-RPT instrumentation Function listed below shall be OPERABLE:
- a. Reactor Vessel Water Level-Low Level 2; and
- b. Reactor Vessel Pressure-High.
APPLICABILITY: MODE 1. ACTIONS
...____.....................-NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION COMPLETION TIME m REQUIRED ACTION i A. One or more channels A.1 Restore channel to 14 days inoperable. OPERABLE status, i 1 0.8 A.2 --------NOTE--------- Not applicable if inoperable channel is the result of an inoperable breaker. P1 ace channel in 14 days trip. (continued) Brunswick Unit 1 3.3-32 Amendment No.
- ATWS-RPT Instrumentaticn l 3.3.4.1 l
ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME l B. One Function with B.1 Restore ATWS-RPT trip 72 hours ATWS-RPT trip capability. capability not maintained. l I C. Both Functions with C.1 Restore ATWS-RPT trip I hour ATWS-RPT trip capability for one 4 capability not Function. I maintained. D. Required Action and D.1 Remove the associated 6 hours associated Completion recirculation pump (s) Time not met. from service. 0_ R N D.2 Be in MODE 2. 6 hours
]
SURVEILLANCE REQUIREMENTS _____..__...___........______ .. __.-h0TE------------------------------------- When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions ; may be delayed for up to 6 hours provided the associated Function maintains l ATWS-RPT trip capability. j
..........__.................................___.________..............__..... 1 SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL CHECK. 24 hours (continued) l I
O Brunswick Unit 1 3.3-33 Amendment No.
i ATWS-RPT Instrumentatien i 3.3.4.1 l SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE ! l FREQUENCY 4 SR 3.3.4.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.1.3 Calibrate. the triy units. 92 days i l SR 3.3.4.1.4 Perform CHANNEL CALIBRATION. The 24 months Allowable Values shall be:
- a. Reactor Vessel Water Level-Low Level 2: 2: 101 inches; and
- b. Reactor Vessel Pressure-High:
s 1147 psig. SR 3.3.4.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST 24 months including breaker actuation. i i
. O
ECCS Instrumentation 3.3.5.1 l 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation l l LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.5.1-1. ACTIONS
-------------------------------------NOTE--------------------------...-.------
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.1-1 for the channel. B. As required by B.1 --------NOTES-------- Required Action A.1 1. Only applicable and referenced in in MODES 1, 2, Table 3.3.5.1-1. and 3.
- 2. Only applicable for Functions 1.a, 1.b, 2.a, and 2.b.
Declare supported I hour from feature (s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature (s) in both divisions AND l (continued) i
%J Brunswick Unit 1 3.3-35 Amendment No.
l ECCS Instrumentatien 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2_ --------NOTE--------- Only applicable for Functions 3.a and 3.b. Declare High Pressure 1. hour from Coolant Injection discovery of (HPCI) System loss of HPCI inoperable. initiation capability j AND , B.3 Place channel in 24 hours trip. C. As required by C.1 --------NOTES-------- Required Action A.1 1. Only applicable and referenced in in MODES 1, 2, O Table 3.3.5.1-1. and 3.
- 2. Only applicable for Functions 1.c, 1.d, 2.c, 2.d, and 2.f.
Declare supported I hour from feature (s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature (s)in both divisions AND ; C.2 Restore channel to 24 hours OPERABLE status. (continued) l s
ECCS Instrumentation 3.3.5.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. As required by D.1 --------NOTE--------- Required Action A.1 Only applicable if and referenced in HPCI pump suction is Table 3.3.5.I-1. not aligned to the suppression pool. Declare HPCI System I hour from inoperable. discovery of loss of HPCI initiation capability AND D.2.I Place channel in 24 hours trip. OB D.2.2 Align the HPCI pump 24 hours suction to the suppression pool. (continued) l l l O Brunswick Unit 1 3.3-37 Amendment No.
ECCS Instrumentatien 3.3.5.1 I ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. As required by E.1 Declare Automatic 1 hour from Recuired Action A.1 Depressurization discovery of > anc referenced in System (ADS) valves loss of ADS Table 3.3.5.1-1. inoperable. initiation capability in both trip systems AND E.2 Place channel in 96 hours from ) trip. discovery of ) inoperable channel concurrent with HPCI or reactor core isolation cooling (RCIC) inoperable l AND 8 days (continued) O Brunswick Unit 1 3.3-38 Amendment No.
ECCS Instrumentation 3.3.5.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TINE F. As required by F.1 Declare ADS valves I hour from Required Action A.1 inoperable. discovery of and referenced in loss of ADS - Table 3.3.5.1-1. initiation capability in both trip systems-AND F.2 Restore channel to 96 hours from OPERABLE status. discovery of inoperable channel concurrent with HPCI or RCIC inoperable AND 8 days G. Required Action and G.1 Declare associated Immediately associated Completion supported feature (s) Time of Condition B, inoperable. C, D, E, or F not met. l l O Brunswick Unit 1 3.3-39 Amendment No.
ECCS Instrumentatien-3.3.5.1 SURVEILLANCE REQUIREMENTS
-------------------------------------NOTES----------------------------.--.....
- 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.
- 2. When a channel is placed in an inoperable status solely for performance of i required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows:. (a) for up to 6 hours for Function 3.c; l and.(b) for up to 6 hours for Functions other than 3.c provided the associated Function or the redundant Function maintains ECCS initiation capability.
SURVEILLANCE FREQUENCY l SR 3.3.5.1.1 Perform CHANNEL CHECK. 24 hours SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days O SR 3.3.5.1.3 Calibrate the trip unit. 92 days SR 3.3.5.1.4 Perform CHANNEL CALIBRATION. 24 months SR 3.3.5.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.5.1.6 Perform CHANNEL FUNCTIONAL TEST. 24 months $s ! l l l O Brunswick Unit 1 3.3-40 Amendment No.
ECCS Instrumentation 3.3.5.1 Table 3.3.5.1 1 (page 1 of 4)
% Emercancy Core Cooling system Instruesntation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM .
SPECIFIED PER REQUIRED SURVEILLANCE All.0WASLE FU6.CTION CONDITIONS FUNCTION ACTION A.1 REeulREMENTS VALUE
- 1. Core spray system
- a. Reactor vesset Water 1,2,3, 4 8 SR 3.3.5.1.1 t 13 inches Levet dow Level 3 SR 3.3.5.1.2 4(*), 5(a) g, 3,3,5,g,3 SR 3.3.5.1.4 SR 3.3.5.1.5
- b. Drywell 1,2,3 4 8 SR 3.3.5.1.1 s 1.8 psis Pressure --High SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
- c. Reactor steam Dome 1,2,3 4 C st 3.3.5.1.1 t 402 psig Pressure --Low SR 3.3.5.1.2 and SR 3.3.5.1.3 5 425 psig sa 3.3.5.1.4 SR 3.3.5.1.5 4(*), 5(*) 4 8 SR 3.3.5.1.1 a 402 psis SR 3.3.5.1.2 and SR 3.3.5.1.3 5 425 polg sR 3.3.5.1.4 SR 3.3.5.1.5
- d. Core spray Ptap 1,2,3, 2 C sa 3.3.5.1.4 t 14 seconds start -Time Delay 1 per ptmp SR 3.3.5.1.5 and Relay 4(a), 5(a) SR 3.3.5.1.6 s 16 seconds
- 2. Low Pressure Coolant Injection (LPCI) system
- a. Reactor vessel Water Level --Low Level 3 1,2,3, 4 8 SR 3.3.5.1.1 t 13 inches SR 3.3.5.1.2 4(*), 5(*) st 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
- b. DrywetL 1,2,3 4 8 SR 3.3.5.1.1 s 1.8 pelg Pressure --High SR 3.3.5.1.2 SR 3.3.5.1.3 sa 3.3.5.1.4 sa 3.3.5.1.5 (continued)
(a) Wen associated subsystem (s) are required to be OPERABLE. O V Brunswick Unit 1 3.3-41 Amendment No.
ECCS Instrumentatien 3.3.5.1 Table 3.3.5.1 1 (page 2 of 4) Emergency Core Cooling system Instrsamentation APPLICA8LE CONDITIONS NODES REQUIRED REFERENCED OR OTNER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
- 2. LPCI systam (continued)
- c. Reactor steam Dome Pressure --Low 1,2,3 4 C SR 3.3.5.1.1 t 402 psis SR 3.3.5.1.2 and SR 3.3.5.1.3 s 425 psis SR 3.3.5.1.4 st 3.3.5.1.5 4(a),$(a) 4 5 sR 3.3.5.1.1 t 402 psig sa 3.3.5.1.2 and sa 3.3.5.1.3 5 425 psIg SR 3.3.5.1.4
$R 3.3.5.1.5
- d. Reactor steam Dome 1(b) 2(b)
, , 4 C st 3.3.5.1.1 2 302 psis Pressure -Low SR 3.3.5.1.2 (Recirculation Ptsup 3(b) sR 3.3.5.1.3 Discharge Valve SR 3.3.5.1.4 Permissive) SR 3.3.5.1.5
- e. Reactor Vessel shroud 1,2,3 2 8 SR 3.3.5.1.1 t 50 inches Levet SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 st 3.3.5.1.5
- f. RHR Ptsup start -Time 1,2,3, 4 C SR 3.3.5.1.4 t 9 seconda Delay Relay 1 per ptmp SR 3.3.5.1.5 and 4(*), 5(*) SR 3.3.5.1.6 s 11 seconds
- 3. High Pressure Coolant Injection (NPCI) system
- a. Reactor Vesset Water 1, 4 S SR 3.3.5.1.1 t 101 inches Level -Low Level 2 SR 3.3.5.1.2 2(*), 3(c) st 3.3.5.1.3 .
sR 3.3.5.1.4 l SR 3.3.5.1.5 '
- b. Dryweti 1, 4 e SR 3.3.5.1.1 s 1.8 psIe Pressure -411gh SR 3.3.5.1.2 2(*),3(C) SR 3.3.5.1.3 st 3.3.5.1.4 l sR 3.3.5.1.5 (continued)
(a) h associated st& system (s) are required to be OPERABLE. (b) With associated recirculation ptmp discharge valve or recirculation ptmp discharge bypass valve open. (c) With reactor steam done pressure > 150 psig. \ Brunswick Unit 1 3.3-42 Amendment No.
ECCS Instrumentatien 3.3.5.1 p) s V Table 3.3.5.1 1 (page 3 of 4) Emergency Core cooling system Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FRON SPECIFIED PER REQUIRED gURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMEN1s VALUE
- 3. NPCI system (continued)
- c. Reactor vessel Water Level -Nigh 1, 2 C st 3.3.5.1.1 s 207 inches SR 3.3.5.1.2 2(*3, 3(CI SR 3.3.5.1.3 st 3.3.5.1.4 SR 3.3.5.1.5
- d. Condensate storage 1, 2 D st 3.3.5.1.2 t 23 feet Tank Level -Low SR 3.3.5.1.4 4 inches 2(*3, 3(CI SR 3.3.5.1.5
- e. suppression Chan6er 1, 2 D SR 3.3.5.1.2 s 2 feet Water Level -4tish SR 3.3.5.1.4 2(83, 3(*) st 3.3.5.1.5
- 4. Automatic Depressurization system (ADS) Trip system A
- a. Reactor vessel Water 1, 2 E st 3.3.5.1.1 1 13 inches Level -Low Level 3 SR 3.3.5.1.2 2(C), 3(*) SR 3.3.5.1.3, SR 3.3.5.1.4 SR 3.3.5.1.5 O* b. ADS Timer 1, 2(CI, 3(*)
1 F SR 3.3.5.1.4 SR 3.3.5.1.5 sR 3.3.5.1.6 s 108 seconds c
- c. Reactor vessel Water 1, 1 E st 3.3.5.1.1 t 153 Inches Level -Low Level 1 st 3.3.5.1.2 2(C), 3(*) st 3.3.5.1.3 st 3.3.5.1.4 SR 3.3.5.1.5
- d. Core spray Pwp 1, 2 F SR 3.3.5.1.2 t 102 psis Discharge SR 3.3.5.1.4 and Pressure -High 2(83, 3(*) st 3.3.5.1.5 s 130 psig
- e. RNR (LPCI Mode) Po p 1, 4 F SR 3.3.5.1.2 t 102 psis Discharge 2 per pmp st 3.3.5.1.4 and Pressure ~411gh 2(*I, 3(C) st 3.3.5.1.5 s 130 psis (centinued)
(c) With reactor steam done pressure > 150 psig. _ l 1 1 I C]' %~ Brunswick Unit 1 3.3-43 Amendment No.
ECCS Instrumentatien 3.3.5.1 ( Table 3.3.5.1 1 (page 4 of 4) Emergency Core Cooling system Instrumentotton APPLICA8LE . CONDITIONS 8MBEs OR REQUIRED REFERENCED OTHER CHANNELS FROM sPECIFIED PER REQUIRED SURVEILLANCE ALLOW 48LE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE J
- 5. ads Trip system B
- e. Reactec Vessel Water 1, 2 E sR 3.3.5.1.1 t 13 inches Level -Low Level 3 st 3.3.5.1.2 2(*I, 3ICI sa 3.3.5.1.3 st 3.3.5.1.4 st 3.3.5.1.5
- b. ads Timer 1, 1 F SR 3.3.5.1.4 5 108 socorais SR 3.3.5.1.5 2I *3, 3(*) SR 3.3.5.1.6 1
- c. Reactor vessel Water 1, 1 E sa 3.3.5.1.1 t 153 inches Level -Low Level 1 sa 3.3.5.1.2 2ICI, 3(*) sR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
- d. Core sprey Pump 1, 2 F SR 3.3.5.1.2 a 102 psis Discharge SR 3.3.5.1.4 and Pressure -elish 2(*), 3I *) sR 3.3.5.1.5 5 130 psig
- e. RNA (LPCI Mode) Puup 1, 4 F $R 3.3.5.1.2 1 102 psig Discherpe 2 per ptmp SR 3.3.5.1.4 and Pressure -High 2(CI, 3(c) SR 3.3.5.1.5 s 130 psis
.. . ~ , _ .
(c) With reactor steam dame pressure > 150 psis. 1 l l l 1 l l lt l Brunswick Unit 1 3.3-44 Amendment No.
,a RCIC System Instrumentation 4 3.3.5.2 hs_./ 3.3 INSTRUMENTATION 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3.3.5.2 The RCIC System instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE. APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig. ACTIONS
..............................-NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION- REQUIRED ACTION CONPLETION TINE A. One or.more channels A.1 Enter the Condition Immediately n V inoperable. referenced in Table 3.3.5.2-1 for the channel. B. As required by B.1 Declare RCIC System I hour from Required Action A.1 inoperable. discovery of and referenced in loss of RCIC Table 3.3.5.2-1. initiation capability 8!S l B.2 Place channel in 24 hours i trip. I C. As required by C.1 Restore channel to 24 hours Required Action A.1 OPERABLE status, and referenced in Table 3.3.5.2-1. 1 (continued) O Brunswick Unit 1 3.3-45 Amendment No. I
RCIC System Instrumentatien 3.3.5.2 ACTIONS (continued)' CONDITION REQUIRED ACTION COMPLETION TIME l 1 l D. As required by D.1 ------ NOTE--------- Required Action A.1 Only applicable if l and referenced in RCIC pump suction is Table 3.3.5.2-1. not aligned to the l suppression pool. 1 ..................... Declare RCIC System I hour from inoperable. discovery of loss of RCIC initiation capability MD D.2.1 Place channel in 24 hours trip. 03 D.2.2 Align RCIC pump 24 hours suction to the g suppression pool. E. Required Action and E.1 Declare RCIC System Immediately associated Completion inoperable. Time of Condition B, C, or D not met. O Brunswick Unit 1 3.3-46 Amendment No.
l RCIC System Instrumentation 3.3.5.2 l ,. '( SURVEILLANCE REQUIREMENTS
...............................------NOTES------------------------------------
- 1. Refer to Table 3.3.5.2-1 to determine which SRs. apply for each RCIC Function.
r
- 2. When a channel is placed in an inoperable status solely for performance of required Swteillances, entry into associated Conditions and Required Actions may 1,e delayed as follows: (a) for up to 6 hours for Function 2; and (b) for up to 6 hours for Functions 1 and 3 provided the associated Function maintains RCIC initiation capability.
SURVEILLANCE FREQUENCY SR 3.3.5.2.1 Perform CHANNEL CHECK. 24 hours SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days
~.
SR 3.3.5.2.3 Calibrate thE r y eaits. 92 days SR 3.3.5.2.4 Perform CilANNEL CALIBRATION. 24 months SR 3.3.5.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months ( C Brunswick Unit 1 3.3-47 Amendment No. 1
RCIC Systeni Instrumentation i 3.3.5.2 /^) 6 '(/ Table 3.3.5.2 1 (page 1 of 1) Reactor Core Isolation Cooling system Instrumentation CONOIT10Ns REQUIRED REFERENCED CHANNELS FROM REQUIRED SURVEILLANCE ALLOWASLE ' FUNCTION PER FUNCTION ACTION A.1 REQUIREMENTS VALUE
- 1. Reactor vessel Water 4 5 SR 3.3.5.2.1 t 101 trches Levet dow Level 2 st 3.3.5.2.2 sa 3.3.5.2.3 SR 3.3.5.2.4 SR 3.3.5.2.5
- 2. Reactor Vessel Water 2 C SR 3.3.5.2.1 5 207 inches j Level -4tish SR 3.3.5.2.2 SR 3.3.5.2.3 SR 3.3.5.2.4
- SR 3.3.5.2.5 i
- 3. Condensate storage Tank 2 D st 3.3.5.2.2 e 23 feet Levet d ow st 3.3.5.2.4 '
SR 3.3.5.2.5 Brunswick Unit 1 3.3-48 Amendment No.
Pri ary Containment Isolation Instrumentaticn l- 3.3.6.1 i O V 3.3 INSTRUMENTATION 3.3.6.1 Primary Containment Isolation Instrumentation l l LCO 3.3.6.1 The primary containment isolation instrumentation for each l Function in Table 3.3.6.1-1 shall be OPERABLE. i l ! APPLICABILITY: According to Table 3.3.6.1-1. l ACTIONS l .....................................N0TE-------------------...-..-.--.....--. Separate Condition entry is allowed for each channel. l CONDITION REQUIRED ACTION COMPLETION TIME l A. One or more required A.1 Place channel in 12 hours for l channels inoperable. trip. Functions 2.a, ! 2.b, and 6.b b AN.D 24 hours for Functions other than Functions
, 2.b, and g l
B. One or more functions B.1 Restore isolation I hour with isolation capability. capability not maintained. 1 l C. Required Action and C.1 Enter the Condition Immediately associated Completion referenced in l Time of Condition A Table 3.3.6.1-1 for i , or B not met. the channel. (continued) ; O Brunswick Unit 1 3.3-49 Amendment No.
I
)
Primary Containment Isolation Instrumentatien 3.3.6.1 ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME l D. As required by D.1 Isolate associated 12 hours Required Action C.1 main steam line and referenced in (MSL). l Table 3.3.6.1-1. l M i D.2.1 Be in MODE 3. 12 hours ! AND D.2.2 Be in MODE 4. 36 hours l l E. As required by E.1 Be in MODE 2. 6 hours I Required Action C.1 and referenced in Tabl e 3.3.6.1-1. l l l i F. As required by F.1 Isolate the affected I hour Required Action C.I penetration flow and referenced in path (s). i Table 3.3.6.1-1.
]
G. Required Action and G.1 Be in MODE 3. 12 hours i associated Completion Time for Condition F AND not met. G2 Be in MODE 4. 36 hours E i As required by Required Action C.1 , and referenced in l Table 3.3.6.1-1. l (continued) O Brunswick Unit 1 3.3-50 Amendment No.
r - 1 L Primary Centainment Isolatien Instrumentatien l 3.3.6.1 4 ACTIONS (continued) I CONDITION REQUIRED ACTION CONPLETION TINE l I H. As required by H.1 Declare associated I hour Required Action C.1 standby liquid ! and referenced in control subsystem Table 3.3.6.1-1. (SLC) inoperable.- E
- H.2 Isolate the Reactor I hour Water Cleanup (RWCU)
System. I. As required by I.1 Initiate action to Immediately Required Action C.1 restore channel to end referenced in OPERABLE status. ! Table 3.3.6.1-1. ' B I.2 Initiate action to Immediately isolate the Residual l( Heat Removal (RHR)
-Shutdown Cooling (SDC) System.
l l 1 m Brunswick Unit 1 3.3-51 Amendment No.
l l Prinary Containment Isolatien Instrumentation 3.3.6.1
't i
U SURVEILLANCE REQUIREMENTS l
....................................-NOTES------------------..---.-..----...--
- 1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary 1 Containment Isolation Function.
- 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required i
Actions may be delayed as follows: (a) for up to 2 hours for Functions 2.c, 2.d, 3.a 3.b, 3.e, 3.f, 3.g, 3.h, 4.a. 4.b, 4.e, 4.f, 4.g, / 4.h, 4.1, 4.k. 5.a 5.b, 5.e, 5.f, and 6.a; and (b) for up to 6 hours for U"\ all other Functions provided the associated Function maintains isolation capability.
.............................................................................. i I
SURVEILLANCE FREQUENCY 1 SR 3.3.6.1.1 Perform CHANNEL CHECK. 24 hours I l SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days tQ () SR 3.3.6.1.3 Calibrate the trip unit. 92 days SR 3.3.6.1.4 Perform CHANNEL CALIBRATION. 92 days l i l l SR 3.3.6.1.5 Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.6.1.6 Perform CHANNEL CALIBRATION. 24 months l SR 3.3.6.1.7 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months l l (continued) O v Brunswick Unit 1 3.3-52 Amendment No. l
I Primary Containment Isolaticn Instrumentatien 3.3.6.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l SR 3.3.6.1.8 ------------------NOTES-----------------
-1. Radiation detectors are excluded. ,
g i 2. The sensor response time for l Functions _1.a and 1.c may be assumed ! to be the design sensor response i time. i Verify the ISOLATION INSTRUMENTATION 24 months on a l RESPONSE TIME is within limits. STAGGERED TEST BASIS SR 3.3.6.1.9 Perform CHANNEL FUNCTIONAL TEST. 24 months bl I l l l i l
'O Brunswick Unit 1 3.3-53 Amendment No.
l.
Primary Centainment Isolatien Instrumentaticn l 3.3.6.1 Table 3.3.6.1 1 (page 1 of 5) Primary contalrument Isolation Instrumentation APPLICA8LE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS TROM l sPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE l FUNCTION CONDITIONS SYsfEM ACTION C.1 REQUIREMENTS VALUE
- 1. Main steam Line Isolation l
- a. Reactor Vesset Water 1,2,3 2 D sn 3.3.6.1.1 t 13 inches Level --Low Level 3 st 3.3.6.1.2 st 3.3.6.1.3 st 3.3.6.1.6 st 3.3.6.1.7 SR 3.3.6.1.8
- b. Main steam Line 1 2 E st 3.3.6.1.1 t 825 psis Pressure d ow SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 l I
- c. Main steam Line 1,2,3 2 per D SR 3.3.6.1.1 5 138% rated i Flow --High MSL sR 3.3.6.1.2- steam flow I SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 SR 3.3.6.1.8
- d. Condenser Vacuun dow 1, 2 D st 3.3.6.1.1 2 7.5 Inches SR 3.3.6.1.2 He vacuun p 2("I, 3(a) st 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- e. Main steam Isotation 1,2,3 2 D sR 3.3.6.1.2 5 197'F Valve Pit SR 3.3.6.1.6 Tenperature --High SR 3.3.6.1.7 l
- 2. Primary Containment '
isolation
- a. Reactor Vessel Water Level d ow Levet 1 1,2,3 2 c st 3.3.6.1.1 t 153 inches SR 3.3.6.1.2 SR 3.3.6.1.3 st 3.3.6.1.6 SR 3.3.6.1.7
- b. Drywell Pressure .-41gh 1,2,3 2 C SR 3.3.6.1.1 5 1.8 psig SR 3.3.6.1.2 sR 3.3.6.1.3 l sR 3.3.6.1.6 SR 3.3.6.1.7 (continued)
(a) With any turbine stop valve not closed, s Brunswick Unit 1 3.3-54 Amendment No.
l i l Pricary Containment Isolation Instrumentation 4 3.3.6.1 Table 3.3.6.1 1 (page 2 of 5) I Primary Contafrument Isolation Instrumentation i l APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM ' SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE )
- 2. Primary Contaltunent Isolation (continued)
- c. Main Stack 1,2,3 1 F SR 3.3.6.1.2 (b) i Radiation -Hfgh SR 3.3.6.1.6 ]
SR 3.3.6.1.7 i st 3.3.6.1.8 g l
- d. Reactor Building 1,2,3 1 G sa 3.3.6.1.1 s 16 mR/hr Exhaust st 3.3.6.1.2 Radiation -High sR 3.3.6.1.6 SR 3.3.6.1.7
- 3. High Pressure Coolant Injection (NPCI) system Isolation
- a. HPCI fleam Line 1,2,3 1 F SR 3.3.6.1.1 5 275% rated Flow -High SR 3.3.6.1.2 steam flow SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- b. HPCI steam Line 1,2,3 1 F SR 3.3.6.1.6 t 4 seconds and Flow --High Time Delay
~ SR 3.3.6.1.7 s 12 seconds Relay SR 3.3.6.1.9 V c. HPCI steam supply Line 1,2,3 2 F SR 3.3.6.1.2 t 104 psig -
Pressure -tow sa 3.3.6.1.4 SR 3.3.6.1.7
- d. HPCI Turbine Exhaust 1,2,3 2 F SR 3.3.6.1.2 5 9 psis Diaphragm SR 3.3.6.1.6 Pressure -High SR 3.3.6.1.7
- e. Drywell Pressure -High 1,2,3 i F SR 3.3.6.1.1 s 1.8 psis SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- f. HPCI steam Line Area 1,2,3 1 F sa 3.3.6.1.5 s 200*F L.-
Tenperature -High SR 3.3.6.1.6 SR 3.3.6.1.7
- g. HPCI steam Line Tunnel 1,2,3 1 F st 3.3.6.1.5 5 200*F Ambient SR 3.3.6.1.6 i Teaperature -High SR 3.3.6.1.7 1
- h. HPCI steam Line Tunnet 1,2,3 1 F SR 3.3.6.1.5 5 50'F Differentist SR 3.3.6.1.6 ,
Temperature --High SR 3.3.6.1.7 ) (continued) j l (b) Allowable Value established in accordance with the methodology in the Offsite Dose Calculation Manual. l l 1 l l O l ( ' Brunswick Unit 1 3.3-55 Amendment No. l i
)
l
.)
V l 1 Primary Centainment Isolatten Instrumentation j l 2.3.6.1 :
,~
1
\ \
Table 3.3.6.1 1 (page 3 of 5) ; Primary contaltunent Isolation Instrumentation i APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWASLE FUNCTION CONDITIONS SYSTEM AC1XON C.1 REQUIREMENTS VALUE. 3 NFCI System Italetion (continued) I. HPCI Equipment Area Tenperature -Nish 1,2,3 2 F SR 3.3.6.1.5 SR 3.3.6.1.6 5 175'T h SR 3.3.6.1.7
- 4. Reactor Core Isolation Cooling (RCIC) System Isolation
- a. RCIC Steam Line Flow -Migh 1,2,3 1 F SR 3.3.6.1.1 5 275% rated SR 3.3.6.1.2 steam flow SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- b. RCIC Steam Line 1,2,3 1 F SR 3.3.6.1.6 t 4 seconds !
Flow --High Time Delay SR 3.3.6.1.7 and l Relay SR 3.3.6.1.9 s 12 seconds l
- c. RCIC Steam Supply Line 1,2,3 2 F SR 3.3.6.1.2 5 53 psig Pressure -Low SR 3.3.6.1.4 SR 3.3.6.1.7 O
i
- d. RCic Turbine Exhaust Diaphragm Pressure -411gh 1,2,3 2 F SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7 5 6 pelo
- e. Drywell Pressure -High 1,2,3 1 F SR 3.3.6.1.1 s 1.8 psig SR 3.3.6.1.2 j SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- f. RCIC Steam Line Area 1,2,3 1 F SR 3.3.6.1.5 5 175'F '
Tenperature --High SR 3.3.6.1.6 ~ SR 3.3.6.1.7
- g. RCIC Steam Line Tunnel 1,2,3 1 F SR 3.3.6.1.5 s'200*F Anblent SR 3.3.6.1.6 Temperature -High SR 3.3.6.1.7 I
- h. RCIC Steam Line Tunnel 1,2,3 1 F SR 3.3.6.1.5 s 30 minutes and Area SR 3.3.6.1.6 Temperature -High Time SR 3.3.6.1.7 Delay
- 1. RCIC Steam Line Tunnel 1,2,3 i F SR 3.3.6.1.5 s 50*F Differentlet SR 3.3.6.1.6 Temperature High SR 3.3.6.1.7 (continued) l l
l Brunswick Unit 1 3.3-56 Amendment No.
Primary Containment Isolation Instrumentation 1 3.3.6.1 l b Tabte 3.3.6.1 1 (page 4 of 5) Primary Cantainment f ootation Instrumentation i APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE
- 4. RCIC System isolation (continued)
J. RCIC Equipment Area 1,2,3 2 F SR 3.3.6.1.5 5 175'F Temperature -High SR 3.3.6.1.6 l SR 3.3.6.1.7
- k. RCIC Equipment Area 1,2,3 1 F sa 3.3.6.1.5 s 50*F Differentist sa 3.3.6.1.6 l Temperature ~41(gh SR 3.3.6.1.7 '
- 5. Reactor Water cleanup !
(RWCU) system Isolation
]
- a. Differentist 1,2,3 1 F SR 3.3.6.1.5 s 73 spm Flow -High SR 3.3.6.1.6 i SR 3.3.6.1.7 i
- b. Differentlet 1,2,3 1 F SR 3.3.6.1.5 5 30 minutes j Flow -Hlgh Time Delay SR 3.3.6.1.6 l CR 3.3.6.1.7 l
- c. Area 1,2,3 3 F st 3.3.6.1.5 5 150*F
[mT Temperature -High 1 per room SR 3.3.6.1.6 SR 3.3.6.1.7 \w e/ ;
- d. Area ventitation 1,2,3 2 F SR 3.3.6.1.5 5 50*F '
Offferential SR 3.3.6.1.6 i Temperature .-4tigh SR 3.3.6.1.7 l
- e. Piping Outside RWCu 1,2,3 1 F SR 3.3.6.1.5 s 120*F Rooms Area SR 3.3.6.1.6 Tenperature -High SR 3.3.6.1.7
- f. SLC system Initiation 1,2 1(CI H SR 3.3.6.1.7 NA
- g. Reactor vessel Water 1,2,3 2 F SR 3.3.6.1.1 t 101 inches Level -Low Level 2 SR 3.3.6.1.2 SR 3.3.6.1.3 sR 3.3.6.1.6 SR 3.3.6.1.7 (continued)
(c) SLC system initiation only frputs into one trip system. ! l i j 1 Brunswick Unit 1 3.3-57 Amendment No. i l i
[ Pri::ary Centainment Isolation Instrumentation l i 3.3.6.1 Table 3.3.6.1-1 (page 5 of 5) Primary Contalrument Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOW 48LE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE
- 6. RER shutdown Cooling system Isolation
- a. Reactor steam Dome Pressure -4tish 1,2,3 1 F SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.7 5 137 psig h
- b. Reactor Vesset Water 3,4,5 2(d) I sa 3.3.6.1.1 t 153 inches Level -Low Levet 1 sa 3.3.6.1.2 st 3.3.6.1.3 st 3.3.6.1.6 SR 3.3.6.1.7 (d) In MODES 4 and 5, provided RMR shutdown Cooling system Integrity maintained, only one channet per trip system with an isolation signet avaltable to one RHR shutdown cooling pump suction isolation valve is required.
O i I l 1 O ! V Brunswick Unit 1 3.3-58 Amendment No.
Secondary Centainment Isolation Instrumentatien 3.3.6.2 , d(S 3.3 INSTRUMENTATION 3.3.6.2 Secondary Containment Isolation Instrumentation LCO 3.3.6.2 The secondary containment isolation instrumentation for each Function in Table 3.3.6.2-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.6.2-1. i ACTIONS ,
)
--------------------------.----------NOTE------------------------------.---.--
Separate Condition entry is allowed for each channel. l CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in 12 hours for i inoperable. trip. Function 2 l f3 y AND 24 hours for Functions other than Function 2 i B. One or more Functions B.1 Restore isolation 1 hour < with isolation capability. ' capability not maintained. 1 C. Required Action and C.1.1 Isolate the 1 hour associated Completion associated Time not met, penetration flow paths, l (continued) A V Brunswiek Unit-1 l 3.3-59 Amendment No. l
I I Seccndary Containment Isolaticn Instrumentatien ! 3.3.6.2 i ' V(~h ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l j C. (continued) C.I.2 Declare associated I hour secondary containment ! isolation dampers inoperable. AND . C.2.1 Place the associated I hour standby gas treatment (SGT) subsystem (s) in operation. E , C.2.2 Declare associated I hour SGT subsystem (s) inoperable. 3 (v SURVEILLANCE REQUIREMENTS
-------------------------------------NOTES------------------------------------ ;
- 1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary '
Containment Isolation Function. ; l
- 2. When a channel is placed in an inoperable status solely for performance of l required Surveillances, entry into associated Conditions and Required j Actions may be delayed as follows: (a) for up to 2 hours for Function 3 '
and (b) for up to 6 hours for Functions 1 and 2 provided the associated Function maintains isolation capability. SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. 24 hours (continued) l bO) Brunswick Unit 1 3.3-60 Amendment No.
i Secondary Containment Isolation Instrumentation l 3.3.6.2 i SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days l SR 3.3.6.2.3 Calibrate the trip unit. 92 days : I l SR 3.3.6.2.4 Perform CHANNEL CALIBRATION. 24 months : SR 3.3.6.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months !
. O
Secondary Containment Isolaticn Instrumentation l 3.3.6.2 I 1
\g Table 3.3.6.2 1 (page 1 of 1) secondary Contalement Isolation Instrumentation APPLICA8LE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWASLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE
- 1. Reactor vessel Water 1,2,3, 2 SR 3.3.6.2.1 t 101 inches Level -Low Level 2 SR 3.3.6.2.2
$R 3.3.6.2.3 SR 3.3.6.2.4 sa 3.3.6.2.5
- 2. 0"ywell Pressure -High 1,2,3 2 sa 3.3.6.2.1 s 1.8 psig SR 3.3.6.2.2 SR 3.3.6.2.3 l SR 3.3.6.2.4 SR 3.3.6.2.5
- 3. Reactor Building Exhaust 1,2,3, 1 SR 3.3.6.2.1 s 16 sR/hr Radiation -4tish (a),(b) sR 3.3.6.2.2 sR 3.3.6.2.4 SR 3.3.6.2.5 I
(a) During operations with a potential for draining the reactor vessel. (b) During CORE ALTERATIONS and during movement of irradiated fuel assembtles in secondary contelnment. I v i l I l
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Brunswick Unit 1 3.3-62 Amendment No.
f. l' f CREV System Instrumentaticn 3.3.7.1 ('T V 3.3 INSTRUMENTATION 3.3.7.1 Control Room Emergency Ventilation (CREV) Systeru Instrumentation LCO 3.3.7.1 Two channels per trip system of the Control Building Air Intake ' Radiation-High Function shall be OPERABLE. b APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ACTIONS
......................___.-----------NOTE---------------------.-...........---
Separate Condition entry is allowed for each channel.
,q CONDITION REQUIRED ACTION COMPLETION TIME b
A. One or more channels A.1 Place one CREV 7 days inoperable. subsystem in the radiation / smoke protection mode of operation. A B. CREV System initiation B.1 Place one CREV 1 hour capability not subsystem in the maintained. radiation / smoke protection mode of operation. O Brunswick Unit 1 3.3-63 Amendment No. l
I-I CREV System Instrumentation 3.3.7.1 O SURVEILLANCE REQUIREMENTS 1 l
..................................... NOTE---------------------...----..---..--
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains CREV initiation capability. 1 I SURVEILLANCE- FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK. 2. hours SR 3.3.7.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.7.1.3 Perform CHANNEL CALIBRATION. The 24 months Allowable Value shall be s 27 mR/hr. CD SR 3.3.,.,.. ,e,f,,m too,c S,S1E, ,,,C1,oNAL 1ES1. ,. mont,s l O Brunswick Unit 1 3.3-64 Amendment No.
Ccndenser Vacuum Pump Isolatien Instrumentation 3.3.7.2 O V 3.3 INSTRUMENTATION 3.3.7.2 Condenser Vacuum Pump Isolation Instrumentation LCO 3.3.7.2 Four channels of the Main Steam Line Radiation-High Function for condenser vacuum pump isolation shall be OPERABLE. APPLICABILITY: MODES I and 2 with a condenser vacuum pump in service. b ACTIONS
-------------------------------------NOTE-------------------.--.-..---..__---.
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels inoperable. A.1 Restore channel to OPERABLE status. 12 hours b E b A.2 --------NOTE--------- Not applicable if inoperable channel is the result of an-inoperable condenser vacuum pump trip breaker or isolation valve. Place channel or 12 hours associated trip system in trip. (continued) -p G Brunswick Unit 1 3.3-65 Amendment No.
Cendtnser Vacuum Pump Isslaticn Instrumentatien . i 3.3.7.2 ACTIONS (continued) CONDITION. REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Isolate condenser 12 hours
- associated Completion vacuum pumps.
Time of Condition A not met. E EL B.2 Isolate main steam 12 hours lines. Condenser vacuum pump isolation capability- E not maintained. B.3 Be in MODE 3. 12 hours SURVEILLANCE REQUIRENENTS
.................................-----NOTE-------------------------------------
When a channel is placed in an inoperable status solely for performance of d O required Surve111ances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains condenser. vacuum pump isolation capability. SURVEILLANCE FREQUENCY SR 3.3.7.2.1 Perform CHANNEL CHECK. 24 hours i SR 3.3.7.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.7.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Value shall be s 6 x background. (continued) 1 l O Brunswick Unit 1 3.3-66 Amendment No.
Condenser Vacuun Purp Isolation Instrumentation 3.3.7.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY ! l l SR 3.3.7.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 24 months i including condenser vacuum pump trip l breaker and isolation valve actuation. I l l
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I l l O a: O Brunswick Unit 1 3.3-67 Amendment No.
LOP Instrumentation 3.3.8.1 13 V 3.3 INSTRUMENTATION 3.3.8.1 Loss of Power (LOP) Instrumentation LC0 3.3.8.1 The LOP instrumentation for each function in Table 3.3.8.1-1 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, When the associated diesel generator is required to be OPERABLE by LCO 3.8.2, "AC Sources-Shutdown." ACTIONS
.....................................N0TE----------------------------.-.-----.
Separate C u dition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in I hour O inoperable. trip. B. Required Action and B.1 Declare associated Immediately associated Completion diesel generator (DG) Time not met. inoperable. O Brunswick Unit 1 3.3-68 Amendment No.
LOP Instrumentation 3.3.8.1 f3 V SURVEILLANCE REQUIREMENTS
....................................-NOTES-----------------.-..-.------------.
- 1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.
- 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required-Actions may be delayed for up to 2 hours provided: (a) for Function 1, the associated Functions maintains initiation capability for three DGs; and (b) for Function 2, the associated Function maintains DG initiation capability.
SURVEILLANCE FREQUENCY SR 3.3.8.1.1 Perfe'1 CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.8.1.2 Perform CHANNEL CALIBRATION. 18 months O SR 3.3.8.1.3 Perform CHANNEL CALIBRATION. 24 months SR 3.3.8.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months 1 l O Brunswick Unit 1 3.3-69 Amendment No.
LOP Instrumentatien 3.3.8.1 1 I Table 3.3.8.1-1 (page 1 of 1) Loss of Power Instrumentation REQUIRED CHANNELS SURVEILLANCE ALLOWASLE FUNCTION PER BUS REQUIREMENTS VALUE
- 1. 4.16 kV Emergency Bus Undervoltage (Loss of Voltage)
- a. Bus undervoltage 1 SR 3.3.8.1.2 2 3115 y and 5 3400 V SR 3.3.8.1.4
- b. Time Delay 1 SR 3.3.8.1.2 1 0.5 seconds and SR 3.3.8.1.4 s 2.0 seconds
- 2. 4.16 kV Emergency sus Undervoltage (Degraded Voltage)
- a. Bus Undervoltage 3 SR 3.3.8.1.1 2 3706 V and 5 3748 V SR 3.3.8.1.3 SR 3.3.8.1.4
- b. Time Delay 3 SR 3.3.8.1.1 t 9.0 seconds and SR 3.3.8.1.3 s 11.0 seconds SR 3.3.8.1.4 L
. Brunswick Unit 1 3.3-70 Amendment No.
! i ! I I RPS Electric Power Monitoring 3.3.8.2 A
- Q 3.3 INSTRUMENTATION 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring LCO 3.3.8.2 Two RPS electric power monitoring assemblies shall be OPERABLE for each inservice RPS motor generator set or alternate power supply.
APPLICABILITY: MODES 1 and 2, MODES 3, 4, and 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. d ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or both inservice A.1 Remove associated 72 hours power supplies 'vith inservice power one electric power supply (s) from monitoring assembly service, inoperable. B. One or both inservice B.1 Remove associated I hour power supplies with inservice power both electric power supply (s) from monitoring assemblies service, inoperable. C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or B not met in MODE 1 or 2. (continued) t Brunswick Unit 1 3.3-71 Amendment No.
e RPS Elcctric Power Monitcring 3.3.8.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 _ Initiate action to Immediately associated Completion fully insert all Time of Condition A insertable control or B not met in rods in core cells MODE 3, 4, or 5 with any control rod containing one or more fuel assemblies. b withdrawn from a core cell containing one or more fuel assemblies. l l 1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY s SR 3.3.8.2.1 ------------------NOTE------------------- l Only required to be performed prior to ' entering MODE 2 from MODE 3 or 4, when in MODE 4 for a 24 hours. Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.8.2.2 Perform CHANNEL CALIBRATION for each RPS 24 months motor generator set electric power monitoring assembly. The Allowable Values shall be:
- a. Overvoltage s 129 V.
- b. Undervoltage 2 105 V.
- c. Underfrequency n 57.2 Hz. l (continued)
Brunswick Unit 1 3.3-72 Amendment No.
RPS Electric Power Monitoring 3.3.8.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.8.2.3 Perform CHANNEL CALIBRATION for each RPS 24 months alternate power supply electric power monitoring assembly. The Allowable Values shall be:
- a. Overvoltage s 132 V.
- b. Undervoltage 2 108 V.
- c. Underfrequency 2 57.2 Hz.
SR 3.3.8.2.4 Perform a system functional test. 24 months l l I Brunswick Unit 1 3.3-73 Amendment No.
i SRM Instrumentation l B 3.3.1.2 i LO BASES i SURVEILLANCE SR 3.3.1.2.1 and SR 3.3.1.2.3 (continued) i J 1- REQUIREMENTS l 'The Frequency of once every 12 hours for SR 3.3.1.2.1 is based on operating experience that demonstrates channel failure is rare. While in MODES 3 and 4, reactivity changes are not expected; therefore, the 12 hour Frequency-is relaxed to 24 hours for SR 3.2.1.2.3. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.1.2.2 To provide adequate coverage of potential reactivity changes in the core, one SRM is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed, and.the other OPERABLE SRM must be in an adjacent quadrant containing fuel. Note 1 states that the SR is required to be met only during CORE ALTERATIONS. It is not required to b be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a , review of plant logs to ensure that SRMs required to be
- OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE.
Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRM. The g 12 hour Frequency is based upon operating experience and supplements operational controls over refueling activities that include steps to ensure that the SRMs required by the LCO are in the proper quadrant, SR 3.3.1.2.4 This Surveillance consists of a verification of the SRM instrument readout.to ensure that the SRM reading is greater than a specified minimum count rate with the detector inserted to the normal operating level, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient " source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate. (continued) (O Brunswick Unit 1 B 3.3-42 Revision No. l l
1 SRM Instrumentatien B 3.3.1.2 C BASES SURVEILLANCE SR 3.3.1.2.4 (continued) REQUIREMENTS To accomplish this, the SR is modified by Note I that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant. itith four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical. In addition, Note 2 states that this requirement does not have to be met during a core spiral offload. A core spiral offload encompasses offloading 1. cell on the edge of a continuous fueled region (the core cell can be offloaded in any sequence). If the core is beltg unloaded in this manner, the various core configurations encountered will not be critical. The Frequency is based upon channel redundancy and other information available in the control room, and ensures that the required channels are frequently monitored while core reactivity changes are occurring. When no reactivity chanaes are in )rogress, the Frequency is relaxed from 12 hou:s to 24 iours. d SR 3.3.1.2.5 and SR 3.3.1.2.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly. SR 3.3.1.2.5 is required in MODE 5, and the 7 day frequency en wres that the channels are OPERABLE while core reactivity changes could be ' in progress. This Frequency is reasonable, based on operating experience and on other Surveillances (such as a CHANNEL CHECK), that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. SR 3.3.1.2.6 is required to be met in MODE 2 with .IRMs on Range 2 or below, ud in MODES 3 and 4. Since core ' reactivity changes do not normally take place in MODES 3 and 4 and core reactivity changes are due only to control rod movement in MODE 2, the frequency is extended from 7 days to 31 days. The 31 day Frequency is based on c ' operating experience and on other Surveillances (such as CHANNEL CHECK) that ensure proper functioning between A CHANNEL FUNCTIONAL TESTS. (2 , (continued) 4 O Brunswick Unit 1 B 3.3-43 Revision No.
1 SRM Instrumentaticn i B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.5 and SR 3.3.1.2.6 (continued) i REQUIREMENTS l The Note to the Surveillance allows the Surveillance to be l delayed until entry into the specified condition of the Applicability (THERMAL POWER decreased to IRM Range 2 or below). The SR must be performed within 12 hours after IRMs are on Range 2 or below. The allowance to enter the Applicability with the 31 day Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. 6 l Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances. SR 3.3.1.2.7 Performance of a CHANNEL CALIBRATION at a Frequency of O' 24 months verifies the performance of the SRM detectors and associated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status. The neutrca detectors are excluded from the CHANNEL CALIBRATION /, Note 1) because they cannot readily be adjusted. The deter tors are fission chambers that are designed to have a .?latively constant sensitivity over the range and with an aceracy specified for a fixed useful life. Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability. The SR sust be performed in MODE 2 within 12 hours of entering MODE 2 with IRMs on Range 2 or below. ! The allowance to enter the Applicability with the 24 month frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while , on IRM Range 3, the plant would not be expected to maintain i steady state operation at this power level. In this event, j (continued) O -Brunswick Unit 1 B 3.3-44 Revision No.
L 1 i f' SRM Instrumentation B 3.3.1.2 l BASES l SURVEILLANCE SR 3.3.1.2.7 (continued) REQUIREMENTS the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to i perform the Surveillances. REFERENCES None. 1 O l l LO Brunswick Unit 1 B 3.3-45 Revision No. 1
a I PBDS ! B 3.3.1.3 B 3.3 INSTRUMENTATION l B 3.3.1.3 Period Based Detection System (PBDS) l l BASES l BACKGROUND General Design Criteria 12 requires protection of fuel thermal safety limits from conditions caused by neutronic/ thermal hydraulic instability. Neutronic/ thermal l hydraulic instabilities can result in power oscillations which could result in exceeding the MCPR Safety Limit (SL). The MCPR SL ensures that at least 99.9% of the fuel rods avoid boiling transition during normal operation and during an anticipated operational occurrence (A00) (refer to the Bases for SL 2.1.1.2). The PBDS provides the operator with an indication that conditions consistent with a significant degradation in the stability performance of the reactor core has occurred and the potential for imminent onset of neutronic/ thermal hydraulic instability may exist. Indication of such degradation is cause for the operator to initiate an immedicte reactor scram if the reactor is being operated in either the Restricted Region or Monitored Region. The Restricted Region and Monitored Region are defined in the COLR. The PBDS instrumentation of the Neutron Monitoring System (NMS) consists of two channels. PBDS channel A includes input from 13 local power range monitors (LPRMs) within the reactor core and PBDS channel B includes input from 11 LPRMs l within the reactor core. All LPRMs are utilized from each l of the axial levels except for the D level detectors. These inputs are continually monitored by the PBDS for variations in the neutron flux consistent with the onset of neutronic/ thermal hydraulic instability. Each channel ' includes separate local indication and separate control room l High-High Alarms. While, this LC0 specifies OPERABILITY requirements only for one monitoring and indication channel of the PBDS, if both are OPERABLE, a High-High Alarm from either channel results in the need for the operator to take actions, j The primary PBDS component is a card in the NHS with analog inputs and digital processing. The PBDS card has an automatic self-test feature to periodically test the L hardware circuit. The self-test functions are executed ; j_ during their allocated portion of the executive loop ' p (continued) I Brunswick Unit 1 B 3.3-46 Revision No. 1
PBDS B 3.3.1.3 BASES BACKGROUND . sequence. Any self-test failure indicating loss of critical (continued) function results in a common control room " Inoperative" alarm. The inoperable condition is also displayed by an indicating light on the card front panel. A manually initiated internal test sequence can be actuated via a recessed push button. This internal test consists of simulating alarm and inoperable conditions to verify card OPERABILITY. Further descriptions of the PBDS are provided in References 1 and 2. Actuation of the PBDS High-High Alarm is not postulated to occur due to neutronic/ thermal hydraulic instability during operation outside the Restricted Region and the Monitored Region. Periodic perturbations can be introduced into the thermal hydraulic behavior of the reactor core from external sources such as recirculation system components and the I pressure and feedwater control systems. These perturbations can potentially drive the neutron flux to oscillate within a frequency range expected for neutronic/ thermal hydraulic instability. The presence of such oscillations may be recognized by the period based algorithm of the PBDS and i could result in a High-High Alarm. Actuation of the PBDS High-High Alarm outside the Restricted Region and the O Monitored Region indicate the presence of a source external to the reactor core and are not indications of neutronic/ thermal hydraulic instability. APPLICABLE Analysis, as described in Section 4 of Reference 1, SAFETY ANALYSES confirms that A00s initiated from outside the Restricted Region without stability control and from within the Restricted Region with stability control are not expected to result in neutronic/ thermal hydraulic instability. The stability control applied in the Restricted Region (refer to
' LCO 3.2.3, " Fraction of Core Boiling Boundary (FCBB)") is established to prevent neutrcnic/ thermal hydraulic instability during operation in the Restricted Region.
Operation in the Monitored Region is only susceptible to instability under operating conditions beyond those analyzed in Reference 1. The types of transients specifically evaluated are loss of flow and coolant temperature decrease which are limiting for the onset of instability. The initial conditions assumed in the analysis are reasonably conservative and the immediate post-event reactor conditions are significantly stable. However, these assumed initial conditions do not bound each individual parameter (continued) O Brunswick Unit 1 B 3.3-47 Revision No. a
PBDS. B 3.3.1.3 I BASES APPLICABLE which impacts stability performance-(Ref.1). The PBDS l SAFETY ANALYSES instrumentation provides the operator with an indication
'(continued) that conditions consistent with a significant degradation in the stability performance of the reactor core has occurred and the potential for imminent onset of neutronic/ thermal hydraulic instability may exist. Such conditions are only postulated to result from events initiated from initial conditions beyond the conditions assumed in the safety.
analysis (refer to Section 4, Ref.1). The PBDS has no safety function and is not assumed to
-function during any UFSAR design basis accident or transient analysis. However, the PBDS provides the only indication of I the imminent onset of neutronic/ thermal hydraulic instability during operation in regions of the operating domain potentially susceptible to instability. Therefore, the PBDS is included in the Technical Specifications. ;
l LC0 One PBDS channel is required to be OPERABLE with a minimum of eight LPRM inputs to monitor reactor neutron flux for indications of imminent onset of neutronic/ thermal hydraulic instability. A PBDS channel may be considered OPERABLE with O six LPRM inputs when the distribution of OPERABLE LPRMs provides: a) at least one OPERABLE LPRM in each core quadrant or b) at least two OPERABLE LPRMs in the core ! quadrant opposite any core quadrant with no OPERABLE LPRMs. l The required distribution of the LPRMs when a PBDS channel is considered OPERABLE with as few as six OPERABLE LPRMs ensures a minimum of two OPERABLE LPRMs in opposite core quadrants. This distribution ensures that, for all l postulated orientations and modes of oscillation, there are I at least two OPERABLE LPRMs in the core quadrants in which
.the local neutron flux will oscillate with a frequency within the range monitored by the PBDS. OPERABILITY ;
requires the ability for the operator to be immediately i alerted to a High-High Alarm. This is accomplished by the ! I instrument channel control room alarm. The LCO also l requires reactor operation be such that the High-High Alarm l is not actuated by any OPERABLE PBDS instrumentation , channel. ; 1 l APPLICABILITY At least one of two PBDS instrumentation channels is required to be OPERABLE during operation in either the , Restricted Region or the Monitored Region specified in the l (continued) O Brunswick Unit 1 B 3.3-48 Revision No. l i
e PBDS t B 3.3.1.3 O V BASES e APPLICABILITY COLR. Similarly, operation with the PBDS High-High Alarm (continued) of any OPERABLE PBDS instrumentation channel is not allowed in the Restricted Region or the Monitored Region. Operation in these regions is susceptible to instability (refer to the Bases for LCO 3.2.3 and Section 4 of Ref.1). OPERABILITY of at least one PBDS instrumentation channel and operation with no indication of a PBDS High-High Alarm from any. OPERABLE PBDS instrumentation channel is therefore required , during operation in these regions. The boundary of the Restricted Region in the Applicability of this LCO is analytically established in terms of thermal j power and core flow. The Restricted Region is defined by the APRM Flow Biased Simulated Thermal Power-High Control l Rod Block setpoints, which are a function of reactor I recirculation drive flow. The Restricted Region Entry Alarm , (RREA) signal is generated by the Flow Control Trip l Reference (FCTR) card using the APRM Flow Biased Simulated Thermal Power-High Control Rod Block setpoints. As a result, the RREA is coincident with the Restricted Region boundary under all anticipated operating conditions when the setpoints are not " Setup," and provides the indication t regarding entry into the Restricted Region. However, APRM ! Flow Biased Simulated Thermal Power-High Control Rod Block l signals provided by the FCTR card, that are not coincident with the Restricted Region boundary, do not generate a valid RREA. The Restricted Region boundary for this LCO Applicability is specified in the COLR. When the APRM Flow Biased Simulated Thermal Power-High Control Rod Block setpoints are " Setup" the applicable ; setpoints used to generate the RREA are moved to the interior boundary of the Restricted Region to allow controlled operation within the Restricted Region. While the setpoints are " Setup" the Restricted Region boundary remains defined by the normal APRM Flow Biased Simulated j Thermal Power-High Control Rod Block setpoints. Parameters such as reactor power and core flow available at the reactor controls, may be used to provide immediate confirmation that entry into the Restricted Region could reasonably have occurred. The Monitored Region in the Applicability of this ! LCO is analytically established in terms of thermal power and core flow. However, unlike the Restricted Region ! boundary the Monitored Region is not specifically monitored l l by plant instrumentation to provide automatic indication of l L entry into the region. Therefore, the Monitored Region 1 (continued) O Brunswick Unit 1 B 3.3-49 Revision No. l l
PBDS L B 3.3.1.3 BASES-l APPLICABILITY boundary is defined solely in terms of thermal power and ! (continued) core flow.. The Monitored Region boundary for this LCO l Applicability is specified in the COLR.
-Operation outside the Restricted Region and the Monitored Region is not susceptible to neutronic/ thermal hydraulic instability even under extreme postulated conditions.
l ~ ACTIONS 'Ad i. If_at any time while in the Restricted Region or Monitored Region, an OPERABLE PBDS instrumentation channel indicates a High-High Alarm, the operator is required to initiate an immediate reactor scram. Verification that the High-High Alarm is valid may be performed without delay against J another output from a PBDS card observable from the reactor controls in the control room prior to the manual reactor scram. This provides assurance that core conditions leading to neutronic/ thermal hydraulic instability will be mitigated. This Required Action and associated Completion Time does not allow for evaluation of circumstances leading to the High-High Alarm prior to manual initiation of O reactor scram. B.1 and B.2 Operation with the APRM Flow Biased Simulated Thermal L Power-High Function (refer to LC0 3.3.1.1, Table 3.3.1.1-1, Function 2.b) " Setup" requires the stability control applied ' in the Restricted Region (refer to LCO 3.2.3) to be met. Requirements for operation with the stability control met are established to prevent reactor thermal hydraulic instability during operation in the Restricted Region. When i the APRM Flow Biased Simulated Thermal Power-High Control i Rod Block setpoints are not " Setup" uncontrolled entry into : the Restricted Region is identified by receipt of a valid j RREA. Immediate confirmation that the RREA is valid and ; indicates an actual entry into the Restricted Region may be l performed without delay. Immediate confirmation constitutes i observation that plant parameters immediately available at the reactor controls (e.g., core power and core flow) are reasonably consistent with entry into the Restricted Region. This immediate confirmation may also constitute recognition that plant parameters are rapidly changing during a (continued) q lO l Brunswick Unit 1 B 3.3-50 Revision No. l i
PBDS l B 3.3.1.3 BASES l ACTIONS B.1 and B.2 *(continued) ! transient (e.g., a recirculation aump trip) which could reasonably result in entry into t1e Restricted Region. While the APRM Flow Biased Simulated Thermal Power-High Control Rod Block setpoints are " Setup," operation in the Restricted Region may be confirmed by use of plant i parameters such as reactor power and core flow available at l the reactor controls. With the required PDDS channel 4 inoperable while in the Restricted Region, the ability to monitor conditions indicating the potential for imminent i onset of neutronic/ thermal hydraulic instability as a result ! of unexpected transients is lost. Therefore, action must be immediately initiated to exit the Restricted Region. i Exit of the Restricted Region can be accomplished by control rod insertion and/or recirculation flow increases. Actions to restart an idle recirculation loop, withdraw control rods or reduce recirculation flow may result in unstable reactor conditions and are not allowed to be used to comply with this Required Action. , O The time required to exit the Restricted Region will depend V on existing plant conditions. Provided efforts are begun without delay and continued until the Restricted Region is exited, operation is acceptable based on the low probability i of a transient which degrades stability performance l occurring simultaneously with the required PBDS channel inoperable. Required Action B.1 is modified by a Note that specifies i that initiation of action to exit the Restricted Region only ; applies if the APRM Flow Biased Simulated Thermal l Power-High function is " Setup". Operation in the i Restricted Region without the APRM Flow Biased Simulated Thermal Power-High Function " Setup" indicates uncontrolled entry into the Restricted Region. Uncontrolled entry is consistent with the occurrence of unexpected transients, which, in combination with the absence of stability controls being met may result in significant degradation of stability performance. Under these conditions with the required PBDS instrumentation channel inoperable, the ability to monitor conditions indicating the potential for imminent onset of neutronic/ thermal hydraulic instability is lost and continued operation is not justified. Therefore, Required Action B.2 requires immediate reactor scram. (continued) O a Brunswick Unit 1 B 3.3-51 Revision No.
PBDS B 3.3.1.3
. BASES ACTIONS L1 (continued)
In the Monitored Region the PBDS High-High Alarm provides indication of degraded stability performance. Although not !. anticipated, operation in the Monitored Region is l susceptible to neutronic/ thermal hydraulic instability under l postulated conditions exceeding those previously assumed in the safety analysis. With the required PBDS channel l ' inoperable while in the Monitored Region, the ability to l monitor conditions indicating'the potential for imminent l onset of neutronic/ thermal hydraulic instability is lost. Therefore, action must be initiated to exit the Monitored Region. Actions to restart an idle recirculation loop, withdraw control rods or reduce recirculation flow may result in approaching unstable reactor conditions and are not allowed to be used to comply with this Required Action. Exit of the Monitored Region is accomplished by control rod insertion and/or recirculation flow increases. However, actions which reduce recirculation flow are allowed provided the FCBB is . recently (within 15 minutes) verified to be s 1.0. Recent !- verification of FCBB being met, provides assurance that with the PBDS inoperable, planned decreases in recirculation drive flow should not result in significant degradation of core stability performance. The Completion Time of 15 minutes ensures timely operator action to exit the region consistent with the low probability that reactor conditions exceed the initial conditions assumed in the safety analysis. The time required to exit the Monitored Region will depend on existing plant conditions. Provided efforts are begun within 15 minutes and continued until the Monitored Region is exited, operation is acceptable based on the low probability of a transient which degrades stability performance occurring simultaneously with the required PBDS channel inoperable. SURVEILLANCE SR 3.3.1.3.1 REQUIREMENTS
- During operation in the Restricted Region or the Monitored Region the PBDS High-High Alarm is relied upon to indicate conditions consistent with the onset of neutronic/ thermal hydraulic instability. Verification that each OPERABLE (continued)
O Brunswick Unit 1 8 3.3-52 Revision No. 1 I
PBDS B 3.3.1.3 BASES SURVEILLANCE SR 3.3.1.3.1 (continued) REQUIREMENTS channel of PBDS instrumentation is not in High-High Alarm every 12 hours provides assurance of the proper indication of the alarm during operation in the Restricted Region or the Monitored Region. The 12 hour Frequency supplements less formal, but more frequent, checks of alarm status during operation. SR 3.3.1.3.2 Performance of the CHANNEL CHECK every 12 hours ensures that a gross failure of instrumentation has not occurred. This CHANNEL CHECK is normally a comparison of the PBOS indication to the state of the annunciator, as well as comparison to the same parameter on the other channel if it is availaole. It is based on the assumption that the instrument channel indication agrees with the immediate indication available to the operator, and that instrument channels monitoring the same parart.eter should read similarly. Deviations between the instrument channels could n be an indication of instrument component failure. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication i and readability. . The 12 hour Frequency is based on the CHANNEL CHECK Frequency requirement of similar Neutron Monitoring System components. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. 3.3.1.3.3 SR A CHANNEL FUNCTIONAL TEST is performed for each required PBDS channel to ensure that the system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the PBDS includes manual initiation of an internal test sequence and verification of appropriate alarms and inop conditions being reported. (continued) l v Brunswick Unit 1 B 3.3-53 Revision No.
1 PBDS B 3.3.1.3 BASES SURVEILLANCE SR 3.3.1.3.3 (continued) REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST at a Frequency of 24 months verifies the performance of the PBDS and associated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status. The alarm circuit is designed to operate for over 24 months with sufficient accuracy on signal amplitude and signal timing considering environment, initial calibration, and accuracy drift (Ref. 2). REFERENCES 1. NEDO 32339-A, Reactor Stability Long Term Solution: Enhanced Option I-A, July 1994.
- 2. NEDC-32339, Supplement 2, Reactor Stability Long Term Solution: Enhanced Option I-A Solution Design.
April 1995. O V O Brunswick Unit 1 8 3.3-54 Revision No.
IL l' Control Rod Block Instrumentaticn B 3.3.2.1 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES l BACKGROUND Control rods provide the primary means for control of reactivity changes. Control rod block instrumentation i includes channel sensors, logic circuitry, switches, and l relays that are _ designed to ensure that specified fuel design limits are not exceeded for postulated transients and l accidents. During high power, operation, the rod block l monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch-Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities. l The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint
- during control rod manipulations. It is assumed to function
, to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint specified in the COLR. The RBM . has two channels,-either of which can initiate a control rod I block when the channel output exceeds the control rod block ' i setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. The RBM channel signal is generated . by averaging a set of local power range monitor (LPRM) ! signals at various core heights surrounding the control rod being withdrawn. A signal from one average power range i monitor (APRM) channel assigned to each Reactor Protection 1 System (RPS) trip system supplies a reference signal for the . RBM channel in the same trip system. This reference signal J l 1s used to determine which RBM range setpoint (low, intermediate, or high) is enabled. If the APRM is i indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically l bypassed if a peripheral control rod is selected (Ref.1). i A rod block signal is also generated if an RBM downscale i trip or an inoperable trip occurs, since this could indicate , a problem with the RBM channel. The downscale trip will (continued) O Brunswick Unit 1 B 3.3-55 Revision No. i i
i Centrol Rod Bicck Instrumentatien B 3.3.2.1 BASES u BACKGROUND occur if the RBM channel signal decreases below the (continued) downscale trip setpoint after the RBM channel has been
, normalized. The inoperable trip will occur during the
, nulling (normalization) sequence, if the RBM channel fails to null, too few LPRM inputs are available, if a module is '; not plugged in, or the function switch is moved to any position other than " Operate." The purpose of the RWM is to control rod patterns during startup and shutdown, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances _ from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed. The RWM is a single channel system i that provides input into the RMCS rod withdraw permissive ! O circuit. With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained.- This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the ' shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods. APPLICABLE 1. Rod Block Monitor SAFETY ANALYSES, LCO, and The RBM is designed to prevent violation of the MCPR APPLICABILITY SL and the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event.. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 2. A statistical analysis of RWE events was performed to (continued) O Brunswick Unit 1 B 3.3-56 Revision No.
i Control Rod B1cck Instrumentatien B 3.3.2.1 , I BASES l l APPLICABLE I. Rod Block Monitor (continued) l SAFETY ANALYSES, LCO, and determine the RBM response for both channels for each event. APPLICABILITY From these responses, the fuel thermal performance as a l function of RBM Allowable Value was determined. The Allowable Values are chosen as a function of power level. Based on the specified Allowable Values, operating limits are established. The RBM Function satisfies Criterion 3 of Reference 3. l l Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Value for the associated power range, to ensure that no single instrument failure can preclude a rod block from this Function. The actual setpoints are calibrated consistent ! with applicable setpoint methodology.
)
Trip setpoints are specified in the setpoint calculations. I The setpoints are selected to ensure that the trip settings do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its O Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process 1 p rameter exceeds the setpoint, the associated device (e.g., ' tiip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint value, by accounting for calibration based errors. These calibration i based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation i uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined
! by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.
I- (continued) Brunswick Unit 1 B 3.3-57 Revision No.
1 Control Rod 81cck Instrumentation B 3.3.2.1 o I 'd BASES l 4 APPLICABLE 1. Rod Block Monitor (continued) SAFETY ANALYSES, LCO, and The RBM is assumed to mitigate the consequences of an RWE i APPLICABILITY event when operating 2 29% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE (Ref. 2). When operating < 90% RTP, analyses (Ref. 2) have shown that with an initial MCPR 21.70, no RWE event will result in exceeding the MCPR SL. Also, the analyses demonstrate that when operating at 2 90% RTP with MCPR 21.40, no RWE event will result in exceeding the MCPR SL (Ref. 2). Therefore, under these conditions, the RBM is also not required to be OPERABLE. J
- 2. Rod Worth Minimizer The RWM enforces the banked position withdrawal sequence )
(BPWS) to ensure that the initial conditions of the CRDA analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in References 4, 5, and 6. The BPWS requires that control rods g+ be moved in groups, with all control rods assigned to a specific group required to be within specified banked >V positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, " Rod Pattern Control." The RWM Function satisfies Criterion 3 of Reference 3. The RWM is a microprocessor-based system with the principle task to reinforce procedural control to limit the reactivity worth of control rods under lower power conditions. Only one channel of the RWH is available and required to be OPERABLE. Special circumstances provided for in the Required Action of LCO 3.1.3, " Control Rod OPERABILITY," and LC0 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. As required by these conditions, one or more control rods may be bypassed in the RWM or the RWM may be bypassed. However, the RWM must be considered inoperable and the Required Actions of this LCO followed since the RWM can no longer enforce compliance with the BPWS. (continued) O Brunswick Unit 1 B 3.3-58 Revision No.
Control Rod Bicek Instrumentet Jcn B 3.?. 2.1 BASES APPLICABLE 2. Rod Worth Minimizer (continued) SAFETY ANALYSES, LCO, and Compliance with the BPWS, and therefore OPERABILITY of the l APPLICABILITY RWM, is required in MODES I and 2 when THERMAL POWER is s 10% RTP. When THERMAL POWER is > 10% RTP, there is no - posrible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a CRDA (Refs. 5 and 6). In MODES 3 and 4, all control rods are required to be inserted into the core; therefore, a CRDA cannot occur. In MODE 5, since only a i single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SOM ensures that the consequences of a CRDA are acceptable,- since the reactor will be subcritical.
- 3. Reactor Mode Switch-Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode e Switch-Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.
The Reactor Mode Switch-Shutdown Position Function satisfies Criterion 3 of Reference 3. Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on ! reactor mode switch position. Durin'g shutdown conditions (MODE 3, 4, or 5), no positive reactivity insertion events are analyzed because assumptions ; are that control rod withdrawal blocks are provided to ' prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod withdrawal block is required to be OPERABLE. During MODE 5 with the. reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LCO 3.9.2, " Refuel Position One-Rod-Out Interlock") provides the required control rod withdrawal blocks. (continued) O -Brunswick Unit 1 B 3.3-59 Revision No.
Control Rod Block Instrumentation B 3.3.2.1. I d BASES (continued) ACTIONS 8.d With one R8M channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function; however, overall reliability is reduced because a single failure in the remaining OPERABLE channel can result i in no control rod block capability for the R8M. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Completion Time of 24 hours is based on the low probability of an event occurring coincident with a failure in the remaining OPERABLE channel. B.1 If Required Action A.1 is not met and the associated Completion Time has expired, an RBM channel must be placed in the tripped condition within I hour. If both RBM channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip within 1 hour. This initiates a control rod withdrawal block, thereby ensuring that the RBM O function is met. ! i The-1 hour Completion Time is' intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or _ tripping of inoperable channels. C.1. C.2.1.1. C.2.1.2. and C.2.2 With the RWM Function inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence. However, the overall reliability is reduced because a single operator error can result in violating the control rod sequence. Therefore, control rod movement must be immediately suspended except by scram. Alternatively, startup may continue if at least 12 control l rods have already been withdrawn, or a reactor startup with j the RWM inoperable, for reasons other than one or more control rods bypassed in the RWM, was not performed in the last 12 months. Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant (continuel). i o Brunswick Unit 1 B 3.3-60 Revision No. l
Centrol Rod Blcck Instrumentaticn B 3.3.2.1 b d BASES ACTIONS C.1. C.2.1.1. C.2.1.2. and C.2.2 (continued) logs and control room indications. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2. Required Action C.2.2 allows for the RWM Function to be performed manually and requires double verification of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. One or more control rods may be bypassed in the RWM or the RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LC0 3.1.3 and LCO 3.1.6 may require bypassing one or more control rods in the RWM or bypassing the RWM, during which
'.ime the RWM must be considered inoperable with Condition C entered and its Required Actions taken. In the event one or more control rods are bypassed in the RWM (up to 8 control rods may be bypassed in accordance with the RWH design), hs Required Action C.2.1.2 does not restrict reactor startup.
l u 1 With the RWM Function inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence. Required Action 0.1 allows for the RWM Function to be performed manually and requires double verification of compliance with the prescribed rod sequence by a second licensed operator (Reactor 0)erator or Senior Reactor Operator) or other qualified mem)er of the technical staff. One or more control rods may be bypassed in the RWM or the RWM may be bypassed under these conditions to allow the reactor shutdown to continue. E.1 and E.2 With one Reactor Mode Switch-Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. However, since the Required Actions are (continued) l v Brunswick Unit 1 8 3.3-61 Revision No.
Centrol Rod B1cck Instrumentatien B 3.3.2.1 BASES. ACTIONS E.1 and E.2 (continued) consistent with the normal action of an OPERABLE Reactor Mode Switch-Shutdown Position Function (i.e., maintaining all control rods inserted), there is no distinction between having one or two channels inoperable. In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies will ensure that the core is subcritical with adequate SDM ensured by LCO 3.1.1. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Control Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1. The Surveillances are modified by a Note to indicate that when an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 7) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour { testing allowance does not significantly reduce the l probability that a control rod block will be initiated when necessary. SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel { to ensure that the channel will perform the intended , function. It includes the Reactor Manual Control System 1 (continued) i 4 O -Brunswick Unit 1 B 3.3-62 Revision No.
Centrol Rod Block Instrumentatien 8 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.1 (continued) REQUIREMENTS input. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on reliability analyses (Ref. 8). SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the RWM is performed by selecting a control rod not in compliance with the prescribed sequence and verifying proper annunciation of the selection error, and by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs. As noted in the SRs, SR 3.3.2.1.2 is not required to be perfomed until I hour after any control rod is withdrawn in MODE 2. As noted, A SR 3.3.2.1.3 is not required to be performed until I hour after THERMAL POWER is s 10% RTP in MODE 1. This allows entry into MODE 2 for SR 3.3.2.1.2, and entry into MODE 1 when THERMAL POWER is s 10% RTP for SR 3.3.2.1.3, to perform the required Surveillance if the 92 day Frequency is not met per SR 3.0.2. The I hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. Operating experience has demonstrated these components will usually pass the Surveillances when performed at the 92 day Frequency. Therefore, the Frequency is acceptable from a reliability standpoint. SR 3.3.2.1.4 J The RBM setpoints are automatically varied as a function of I power. Three Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values I automatically change are based on the APRM signal's input to l each RBM channel. Below the minimum power range setpoint, l the RBM is automatically bypassed. These power range j setpoints (low power range setpoint, intermediate power i range setpoint, and high power range setpoint) must be ] (continued) O Brunswick Unit 1 B 3.3-63 Revision No. I i
Control Rod Bleck Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.4 (continued) REQUIREMENTS verified periodically to be less than or equal to the specified Allowable Values in the COLR. If any power range setpoint is nonconservative, then the affected RBM channel is considered inoperable. Alternatively, the RBM power range channel can be placed in the conservative condition (i.e.,enablingtheproperRBMsetpoint). If placed in this condition, the SR is met and the RBM channel is not considered inoperable. As noted, neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.3 and SR 3.3.1.1.8. The 24 month Frequency is based on the actual trip setpoint methodology utilized for these channels. SR 3.3.2.1.5 The RWM is automatically bypassed when power is above a specified value. The power level is determined from steam flow signals. The automatic bypass setpoint must be O verified periodically to be > 10% RTP. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The Frequency is based on the trip setpoint methodology utilized for the low power setpoint channel. SR 3.3.2.1.6 i A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode l Switch--Shutdcwn Position Function to ensure that the 1 channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch-Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown : position and verifying a control rod block occurs. l As noted in the SR, the Surveillance is not required to be performed until I hour after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be (continued) O Brunswick Unit 1 B 3.3-64 Revision No. l
Centrol Rod Bleck Instrumentation B 3.3.2.1 BASES SURVEILLANCE :SR 3.3.2.1.6 (continued) REQUIREMENTS performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the 24 month Frequency is not met per SR 3.0.2. The'l hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. SR 3.3.2.1.7 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANi4EL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The CHANNEL CALIBRATION may be performed O electronically. As noted, neutron detectors are excluded from the CHANNEL , CALIBRATION because they are passive devices, with minimal l drift, and because of the difficulty of simulating a i meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.3 and SR 3.3.1.1.8. The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.2.1.8 The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible. (continued) O Brunswick Unit 1 8 3.3-65 Revision No. 4
l Centrol Rod Blcck Instrumentatica B 3.3.2.1
- (f~)
j BASES (continued) REFERENCES 1. UFSAR, Section 7.6.1.1.5.
- 2. NEDC-31654P, Maximum Extended Operating Domain Analysis For Brunswick Steam Electric Plant, February 1989.
- 3. 10CFR50.36(c)(2)(ii).
- 4. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Unit I and 2, September 1995.
- 5. UFSAR Section 15.4.
- 6. NRC SER, Acceptance for Referencing of Licensing Topical Report NEDE-24011-P-A; General Electric l Standard Application for Reactor Fuel, Revision 8, Amendment 17, December 27, 1987.
- 7. GENE-770-06-1-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, I December 1992.
- 8. NEDC-30851P-A, Supplement 1, Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation, October 1988.
O 'srunswick Unit 1 8 3.3-66 Revision No. I
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Feedwater and Main Turbine High Water Leval Trip Instrumentatien B 3.3.2.2 A U B 3.3 INSTRUMENTATION B 3.3.2.2 Feedwater and Nain Turbine High Water Level Trip Instrumentation BASES BACKGROUND The feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow. With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level setting causing the trip of the two feedwater pump turbines and the main turbine. High water levels signals are provided by three narrow range l sensors of the Digital Feedwater Control System. These . three level sensors sense the difference between the l pressure due to a constant column of water (reference leg) l and the pressure due to the actual water level in the reactor vessel (variable leg). The three level signals are input into a digital control computer. The digital control computer provides three output signals to the high water i level trip channels. Each high water level trip channel - consists of a relay whose contacts form the trip logic. The high water level trip logic is arranged as a two-out-of-three logic, that trips the two feedwater pump turbines and the main turbine. The digital control computer processes the reactor water level input signals and compares them to pre-established setpoints. When the setpoint is exceeded, the associated channel output relay actuates, which then 1 outputs to the main turbine and feedwater pump trip ) initiation logic. I A trip of the feedwater pump turbines limits further l increase in rea : tor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the ) main turbine and closure of the stop valves protects the )
-turbine from damage due to water entering the turbine. I i
I APPLICABLE The feedwater and main turbine high water level trip 1 SAFETY ANALYSES instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref. 1). (continued) O Brunswick Unit 1 B 3.3-67 Revision No.
Fcedwater and Main Turbine High Water Level Trip Instrumentatien B 3.3.2.2 BASES APPLICABLE The high water level trip indirectly initiates a reactor SAFETY ANALYSES scram from the main turbine trip.(above 30% RTP) and trips (continued) the feedwater pumps, thereby terminating the. event. The f reactor scram mitigates the reduction in MCPR. Feedwater. and main turbine high water level trip instrumentation satisfies Criterion 3 of Reference 2. LCO The LCO requires three channels of the reactor vessel high water level instrumentation to be OPERABLE to ensure that
-the feedwater pump turbines and main turbine trip on a valid high water level signal. Two of the three channels are needed to provide trip signals in order for the feedwater and main turbine trips to occur. Each channel must have its setpoint set within the s)ecified Allowable Value of SR 3.3.2.2.2. The A11owa)1e Value is set to ensure that the thermal limits are not exceeded during the event. The actual setpoint is calibrated to be consistent with the applicable setpoint methodology assumptions. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS.
O Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. j Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device I,e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection (continued) O Brunswick Unit 1 B 3.3-68 Revision No. l
l Feedwater and Main Turbine High Water Level Trip Instrumentatien B 3.3.2.2 l i BASES l LCO because instrumentation uncertainties, process effects, (continued) calibration tolerances, instrument drift, and severe i environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. APPLICABILITY The feedwater and main turbine high water level trip instrumentation is required to be OPERABLE at ;t: 25% RTP to ensure that the fuel cladding integrity Safety Limit and the cladding 1% plastic strain limit are not violated during the feedwater controller failure, maximum demand event. As discussed in the Bases for LCO 3.2.1, " Average Planar Linear Heat Generation Rate (APLHGR)," and LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," sufficient margin to these limits exists below 25% RTP; therefore, these requirements are only necessary when operating at or above this power level. ACTIONS A Note has been provided to modify the ACTIONS related to feedwater and main turbine high water level trip instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to ap)1y for each additional failure, with Completion Times )ased on initial entry into the Condition. However, the Required Actions for inoperable feedwater and main turbine high water level trip instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable feedwater and main turbine high water level trip instrumentation channel. A.1 With one channel inoperable, the remaining two OPERABLE channels can provide the required trip signal. However, overall instrumentation reliability is reduced because a single failure in one of the remaining channels concurrent (continued) Brunswick Unit 1 B 3.3-69 Revision No.
l
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Feedwater and Main Turbins High Water Level Trip Instrumentation B 3.3.2.2 BASES ACTIONS A.1 (continued) with feedwater controller failure, maximum demand event, may result in the instrumentation not being able to perform its intended function. Therefore, continued operation is only allowed for a limited time with one channel inoperable. If the inoperable channel cannot be restored to OPERABLE status within the Completion Time, the channel must be placed in the tripped condition per Required Action A.1. Placing .the inoperable channal in trip would conservatively compensate ; for the inoperability, restore capability to accommodate a l single failure, and allow operation to continue with no-further restrictions. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in a feedwater or main turbine trip), Condition C must be entered and its Required Action taken. l The Completion Time of 7 days is based on the low , probability of the event occurring coincident with a single l failure in a remaining OPERABLE channel. l l N With two or more channels inoperable, the feedwater and main turbine high water level trip instrumentation cannot perform its design function (feedwater and main turbine high water level trip capability is not maintained). Therefore, continued operation is only permitted for a 4 hour period, during which feedwater and main turbine high water level trip capability must be restored. The trip capability is considered maintained when sufficient channels are OPERABLE or in trip such that the feedwater and main turbine high water level trip logic will generate a trip signal on a valid signal. This requires two channels to each be OPERABLE or in trip. If the required channels cannot be restored to OPERABLE status or placed in trip, Condition C must be entered and its Required Action taken. The 4 hour Completion Time is sufficient for the operator to take corrective action, and takes into account the likelihood of an event requiring actuation of feedwater and main turbine high water level trip instrumentation occurring (continued) O Brunswick Unit 1 B 3.3-70 Revision No.
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'Feedwat:r and Main Turbine High Water Lovel Trip Instrumentation B 3.3.2.2 O
V. BASES ACTIONS Rd (continued) during this period. It is also consistent with the 4 hour Completion Time provided in LCO 3.2.2 for Required
- Action A.1, since this instrumentation's purpose is to preclude a MCPR violation.
l l C.d l ' With the required channels not restored to OPERABLE status
' or placed in trip, THERMAL POWER must be reduced to
< 25% RTP within 4 hours. As discussed in the Applicability section of the Bases, operation below 25% RTP results in sufficient margin to the required limits, and the feedwater and main turbine high water level trip instrumentation is not required to protect fuel integrity during the feedwater controller failure, maximum demand event. The allowed .}
Completion Time of 4 hours is based on operating experience to reduce THERMAL POWER to < 25% RTP from full power conditions in an orderly manner and without challenging plant systems. j ,p V SURVEILLANCE The Surve111ances are modified by a Note to indicate that i REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains feedwater
. and main turbine high water level trip capability. Upon L completion of the Surveillance, or expiration of the 6 hour
! allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 3) assumption that 6 hours is the average time l required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the feedwater pump turbines and main turbine will trip when necessary. SR 3.3.2.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures i' that a gross failure of instrumentation has not occurred. A l CHANNEL CHECK is normally a comparison of the parameter (continued) lO l Brunswick Unit 1 B 3.3-71 Revision No. I ! I l l
1 Fredwatcr and Main Turbine High Water Level Trip Instrumentatien B 3.3.2.2 BASES-SURVEILLANCE SR 3.3.2.2.1 (continued) REQUIREMENTS indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels, or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, l including indication and readability. If a channel is ' outside the criteria, it may be an indication that the instrument has drifted outside its limits. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of q channel status during normal operational use of the displays Q associated with the channels required by the LCO. SR 3.3.2.2.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel l responds to the measured parameter within the necessary ! range and accuracy. CHANNEL CALIBRATION leaves-the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude 1 of equipment drift in the setpoint analysis. SR 3.3.2.2.3 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the feedwater and (continued) O Brunswick Unit 1 B 3.3-72 Revision No.
Feedwater and Mais furbine High Watsr Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE SR 3.3.2.2.3 (continued) REQUIREMENTS main turbine valves is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a valve is incapable of operating, the associated instrumentation would also be inoperable. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. REFERENCES 1. UFSAR, Section 15.1.2.
- 2. 10 CFR 50.36(c)(2)(ii).
- 3. GENE-770-06-1-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation Technical Specifications, December 1992.
^
f3 V ( w/ Brunswick Unit I ~ B 3.3-73 Revision No.
i PAM Instrumentatien l B 3.3.3.1 I B 3.3 INSTRUMENTATION B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation BASES l l BACKGROUND The primary purpose of the PAM instrumentation is to display
- in the control room plant variables that provide information l
required by the control room operators during accident situations. This information provides the necessary support l for the operator to take the manual actions for which no L automatic coistrol is provided and that are required for safety systems to accomplish their safety functions for Design Basis Events. The instruments that monitor these variables are designated as Type A, Category I, and non-Type A, Category I, in accordance with Regulatory Guide 1.97 (Ref. 1). The OPERABILITY of the accident monitoring instrumentation
. ensures that there is sufficient information available on selected plant parameters to monitor and assess plant status and behavior following an accident. This capability is consistent with the recommendations of Reference 1.
APPLICABLE The PAM instrumentation LCO ensures the OPERABILITY of SAFETY ANALYSES Regulatory Guide 1.97, Type A variables so that the control . room operating staff can: )
- Perform the diagnosis specified in the Emergency I Operating Procedures (EOPs). These variables are restricted to preplanned actions for the primary
, success path of Design Basis Accidents (DBAs), (e.g., loss of coolant accident (LOCA)), and
- Take the specified, preplanned, manually controlled actions for which no automatic control is provided, which are required for safety systems to accomplish their safety function.
The PAM instrumentation LCO also ensures OPERABILITY of Category I, non-Type A, variables so that the control room
- operating staff can
- Determine whether systems important to safety are performing their intended functions; 1 (continued) l O Brunswick Unit 1 B 3.3-74 Revision No.
( PAM Instrumentaticn B 3.3.3.1 ! A IV 1 BASES l I ' APPLICABLE
- Determine the potential for causing a gross breach of l SAFETY ANALYSES the barriers to radioactivity release; l- (continued) l
- Determine whether a gross breach of a barrier has occurred;.and
- Initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.
The plant s)ecific Regulatory Guide 1.97 Analysis (Ref. 2)
- documents tie process that identified Type A and Category I, j non-Type A, variables.
Accident monitoring instrumentation that satisfies the definition of Type A in Regulatory Guide 1.97 meets Criterion 3 of Reference 3. Category I, non-Type A, instrumentation is _ retained in Technical Specifications (TS) because they are intended to' assist' operators in minimizing the consequences of accidents. Therefore, these Category I variables are important for reducing public risk. l LCO LCO 3.3.3.1 requires two OPERABLE channels for all but one Function to ensure that no single failure prevents the operators from being presented with the information ! necessary to determine the status of the plant and to bring the plant to, and maintain it in, a safe condition following that accident. Furthermore, providing two channels allows a CHANNEL' CHECK during the post accident phase to confirm the validity of displayed information, i The exception to the two channel requirement is primary containment isolation valve (PCIV) position. In this case, the important information is the status of the primary containment penetrations. The LCO requires one position indicator for each active (e.g., automatic) PCIV. This is sufficient to redundantly verify the isolation status of , each isolable penetration either via indicated status of the ' active valve and prior knowledge of passive valve or via ! system boundary status. If a normally active PCIV is _ known l to be closed and deactivated, position indication is not ; (~ needed to determine status. Therefore, the position l I indication for closed and deactivated valves is not required to be OPERABLE. , (continued) l O Brunswick Unit 1 B 3.3-75 Revision No.
.r.
I
PAM Instrumentaticn B 3.3.3.1 O v BASES LC0 The following list is a discussion of the specified l (continued) instrument Functions listed in Table 3.3.3.1-1 in the accompanying LCO.
- 1. Reactor Vessel Pressure Reactor vessel pressure is a Type A and Category I variable provided to support monitoring of Reactor Coolant System (RCS) integrity and to verify operation of the Emergency Core Cooling Systems (ECCS). Two independent-pressure transmitters with a range of 0 psig to 1500 psig monitor pressure and are indicated in the control room. Wide range j instruments are the primary indication used by the operator '
during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel. 2.a. 2.b. 2.c. Reactor Vessel Water Level i Reactor vessel water level is a Type A and Category I , variable provided to support monitoring of core cooling and ) r- to verify operatien of the ECCS. Channels from three (- different ranges of water level provide the PAM Reactor Vessel Water Level Function. The water level channels measure from -150 inches to +550 inches. Water level is measured by independent differential pressure transmitters for each required channel. The output from these channels is recorded on independent recorders or read on indicators, which are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel. l
- 3. Suppression Chamber Water Level Suppression chamber water level is a Type A and Category I variable provided to detect a breach in the reactor coolant pressure boundary (RCPB). This variable is also used to verify and provide long term surveillance of ECCS function.
The wide range suppression pool water level measurement provides the operator with sufficient information to assess the status of both the RCPB and the water supply to the ECCS. The wide range water level indicators are capable of monitoring the suppression pool water level from the bottom (continued) Brunswick Unit 1 B 3.3-76 Revision No.
4 PAM Instrumentation 8 3.3.3.1 BASES LCO - 3. Suporession Chamber Water Level (continued) of the ECCS suction lines to 5 feet above the normal pool water level. Two wide range suppres tion pool water level signals are transmitted from separate differential pressure I
- transmitters for each channel. The out >ut of one of these channels is recorded on a recorder in tie control room. The outpM of the other channel is read on an indicator in the control room. These instruments are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
- 4. Suppression Chamber Water Temperature Suppression chamber water temperature is a Type A and Category I variable provided to detect a condition that could potentially-lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach. The suppression chamber water temperature instrumentation, which measures from 40'F to n 240*F, allows operators to detect trends in suppression pool water temperature in sufficient time to take action to V prevent steam quenching vibrations in the suppression pool.
Suppression pool temperature is monitored by-12 pairs of temperature sensors spaced around the suppression pool. A pair of sensors is located near each of the quenchers on the discharge lines of the 11 safety / relief valves. Each pair of sensors is located so as to sense the representative temperature of that sector of the suppression pool even with the associated safety / relief valve open. The outputs for the sensors are indicated on two microprocessors in the control room. The signals from the sensors are conditioned by the two microprocessors to provide an average water temperature. Average water temperature is recorded on two independent recorders in the control room. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channels.
- 5. SuDDression Chamber Pressure Suppression chamber pressure is a Type A and Category I variable provided to detect a condition that could potentially lead to containment breach and to verify the (continued)
Brunswick Unit 1 B 3.3-77 Revision No.
i l l~ PAM Instrumentatien I B 3.3.3.1 O. V BASES LCO 5. Suopression Chamber Pressure (continued) effectiveness of ECCS actions taken to prevent containment breach. Suppression chamber pressure is indicated in the control room from two separate pressure transmitters. The range of indication is O psig to 75 psig. These instruments are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
- 6. Drywell Pressure Drywell pressure is a Ty>e A and Category I variable provided to detect breac1 of the RCPB and to verify ECCS functions that operate to maintain RCS integrity. Two wide range drywell pressure signals are transmitted from separate pressure transmitters for each channel. The output of one of these channels is recorded on a recorder in a control room. The output of the other channel is read on an indicator in the control room. The pressure channels measure from -5 psig to 245 psig. These instruments are the primary indication used by the operator during an accident.
y Tt erefore, the PAM Specification deals specifically with this portion of the instrument channel.
- 7. Drywell Temperature !
Drywell temperature is a Type A and Category I variable provided to detect a breach of the RCPB and to verify the effectiveness of ECCS functions that operate to maintain RCS integrity. Sixteen temperature sensors are located in the drywell to monitor drywell temperature. The sensors are divided into two divisions for redundancy. The signals from these sensors are conditioned by two divisionalized l microprocessors. Drywell temperature is recorded by two pairs of divisionalized recorders in the control room. The range of the recorders is from 40*F to 440*F. These recorders are the primary indication used by the operator ' during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel. (continued) Brunswick Unit 1 B 3.3-78 Revision No.
PAM Instrumentaticn B 3.3.3.1 i L] BASES ' l LCO 8. Primary Containment Isolation Valve (PCIV) Position , (continued) 1
.PCIV position, a Category I variable, is provided for verification of containment integrity. In the case of PCIV position, the important information is the isolation status
, of the containment penetration. The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active PCIV in a containment penetration flow path, i.e., two ' total channels of PCIV l position _ indication for a penetration flow path with two l active valves. For containment penetrations with only one active PCIV having control room indication, Note (b)' requires a single channel of valve position indication to be
- OPERABLE. This is sufficitnt to redundantly verify the isolation status of each isolable penetration via indicated
! status of the active valve, as applicable, and prior knowledge of passive valve or system boundary status. If a l penetration flow path is isolated, position indication for the PCIV(s) in the associated penetration flow path is not ! needed.to determine status. Therefore, the position L -indication for valves in an isolated penetration flow path is not required to be OPERABLE. The PCIV position PAM instrumentation consists of position-l switches, associated wiring and control room indication for l active PCIVs (check valves and manual valves are not required to have position indication). Therefore, the PAM S)ecification deals specifically with these instrument , ciannels. 1
- 9. Drywell and Suporession Chamber Hydroaen and Oxvaen Analyzers i l
Drywell and suppression chamber hydrogen and oxygen analyzers are Type A and Category I instruments provided to detect high hydrogen or oxygen concentration conditions that represent a potential for containment breach. This variable is also important in verifying the adequacy of mitigating i actions. The drywell and suppression chamber hydrogen and l
- oxygen analyzers PAM instrumentation consists of two j i independent gas analyzer systems. Each gas analyzer system i consists of a hydrogen analyzer and an oxygen analyzer. The l analyzers are capable of determining hydrogen concentration ;
in the range of 0% to 30% and oxygen concentration in the= , range of 0% to 25%. Each gas analyzer system must be (continued) O Brunswick Unit 1 B 3.3-79 Revision No.
]
l PAM Instrumentation B 3.3.3.1 BASES l LCO 9. Drywell and SuDDression Chamber Hydroaen and Oxvaen Analyzers (continued) capable of sampling the drywell and the suppression chamber. There are two independent recorders in the control room to display the results. Therefore, the PAM Specification deals specifically with these portions of the analyzer channels. j l
- 10. Drywell Area Radiation l Drywell area radiation is a Category I variable provided to monitor the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans. Post accident drywell area radiation levels are monjtored by four instruments, each with a range of 1 R/hr to 10 R/hr. The outputs of these channels are indicated and recorded in the control room. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
,o APPLICABILITY The PAM instrumentation LCO is applicable in MODES I and 2.
[] These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1 and 2. In MODES 3, 4, and 5, plant conditions are such that the likelihood of an event that would require PAM instrumentation is extremely low; therefore, PAM instrumentation is not required to be i OPERABLE in these MODES. ACTIONS Note I has been added to the ACTIONS to exclude the MODE change restriction of LC0 3.0.4. This exception allows entry into the applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require plant , shutdown. This exception is acceptable due to the passive l function of the instruments, the operator's ability to ' diagnose an accident using alternative instruments and methods, and the low probability of an event requiring these instruments. Note 2 has been provided to modify the ACTIONS related to i PAM instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables (continued) ! Brunswick Unit 1 B 3.3-80 Revision No.
PAM Instrumentatien B 3.3.3.1 A U BASES ACTIONS expressed in the Condition discovered to be inoperable or (continued) not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Requi. red Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable PAM instrumentation channels provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable PAM Function. A.d When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account d the remaining OPERABLE channels, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAN instrumentation during this interval. O u l If a channel has not been restored to OPERABLE status in 30 days, this Required Action specifies initiation of action in accordance with Specification 5.6.6, which requires a b! ! written report to be submitted to the NRC. This report discusses the results of the root cause evaluation of the i inoperability and identifies proposed restorative actions. ; This Required Action is appropriate in lieu of a shutdown ; requirement, since another OPERABLE channel is monitoring - the Function, and given the likelihood of plant conditions that would require information provided by this instrumentation.
.G.d When or.e or more Functions have two required channels that b are inoperable (i.e., two channels inoperable in the same Function), one channel in the Function should be restored to "4 OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the (continued)
Brunswick Unit 1 B 3.3-81 Revision No.
PAM Instrumentationi B 3.3.3.1 BASES , ACTIONS [d (continued) availability of alternate means to obtain the required information. Continuous' operation with two required
-channels' inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM-instrumentation. Therefore, requiring restoration of one inoperable channel of the~ Function limits the~ risk that the PAM Function will be in a degraded condition should an accident occur.
D.d This Required Action directs entry into the appropriate Condition referenced in Table 3.3.3.1-1. The applicable Condition referenced in the Table is Function dependent. Each time an inoperable channel has not met the Required Action of Condition C and the associated Completion Time has expired, Condition D is entered for that channel and provides for transfer to the appropriate subsequent-Condition. O f.d - For the majority of Functions in Table 3.3.3.1-1, if any Required Action and associated Completion Time of Condition C is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. F.1 Since alternate means of monitoring primary containment area radiation are available, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.6. These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. The (continued) O Brunswick Unit 1 B 3.3-82 Revision No.
PAM Instrumentation B 3.3.3.1 O BASES ACTIONS Ed (continued) report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAN channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels. SURVEILLANCE As noted at the beginning of the .SRs, the following SRs REQUIREMENTS apply to each PAM instrumentation Function in Table 3.3.3.1-1. SR 3.3.3.1.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations O- between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. . A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar plant instruments located. throughout the plant. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its. limit. The Frequency of 31 days is based upon plant operating experience, with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is rare. .The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of those displays associated with the channels required by the LCO. (contihued) O Brunswick Unit 1 B 3.3-83 Revision No.
.i
1 PAM Instrumentatien B 3.3.3.1 O O BASES SURVEILLANCE SR 3.3.3.1.2 and SR 3.3.3.1.3 REQUIREMENTS (continued) These SRs require a CHANNEL CALIBRATION to be performed. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies the channel responds to measured parameter with the necessary range and accuracy. For Function 9, the CHANNEL CALIBRATION shall be performed using standard gas samples containing a nominal:
- a. Zero volume percent hydrogen, balance nitrogen;
- b. Seven to ten volume percent hydrogen, balance nitrogen;
- c. Twenty-five to thirty volume percent hydrogen, balance nitrogen;
- d. Zero volume percent oxygen, balance nitrogen;
- e. Seven to ten volume percent oxygen, balance' nitrogen; i and Twenty to twenty-five volume percent oxygen, balance O
f. nitrogen. For Function 10, the CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one point ~ calibration check of the detector below 10 R/hr with an installed or portable gamma source. The 92 day Frequency for CHANNEL CALIBRATION of the drywell and suppression chamber hydrogen and oxygen analyzers is based on operating experience. The 24 month Frequency for CHANNEL CALIBRATION of all other PAM Instrumentation of Table 3.3.3.1-1 is based on operating experience and consistency with the BNP refueling cycles. REFERENCES 1. Regulatory Guide 1.97, Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Revision 2, December 1980. (continued) O Brunswick Unit 1 B 3.3-84 Revision No.
PAM Instrumentation B 3.3.3.1 BASES REFERENCES 2. NRC Safety Evaluation Report, Conformance to (continued) Regulatory Guide 1.97, Rev. 2, Brunswick Steam Electric Plant, Units 1 and 2,' May 14,1985.
- 3. 10 CFR 50.36(c)(2)(ii).
l f' i l 1 i l w Brunswick Unit 1 8 3.3-85 Revision No.
i ) i I 1; Rimote Shutdrwn M:nitcring Instrumentaticn B 3.3.3.2 3 B 3.3 INSTRUMENTATION B 3.3.3.2 Remote Shutdown Monitoring Instrumentation l BASES i l BACKGROUND The remote shutdown monitoring instrumentation provides the control room operator with sufficient instrumentation to support placing and maintaining the plant in.a safe shutdown condition from a location other than the control room. This
- capability is necessary to protect against the possibility
- of the control room becoming inaccessible. A safe shutdown condition is defined as MODE 3. With the plant in MODE 3, the Reactor Core Isolation Cooling (RCIC) System, the safety / relief valves, and the Residual Heat Removal (RHR) l System can be used to remove core decay heat and meet all safety requirements. The long term supply of water for the
! RCIC System and the ability to operate shutdown cooling from outside the control room allow extended operation in MODE 3. In the event that the control room becomes inaccessible, the operators can monitor the status of the reactor and primary containment and the operation of the RCIC and RHR Systems at iO l the remote shutdown panel and place and maintain the plant in MODE 3. Controls and selector switches will have to be operated locally at the switchgear, motor control panels, or other local stations. The plant is in MODE 3 following a j plant shutdown and can be maintained safely in MODE 3 for an l extended period of time. ! The OPERABILITY of the remote shutdown monitoring instrumentation Functions ensures that there is sufficient information available on selected plant parameters to place and maintain the plant in MODE 3 should the control room become inaccessible. APPLICABLE The remote shutdown monitoring instrumentation is required SAFETY ANALYSES to provide equipment at appropriate locations outside the control room with a design capability to monitor the prompt shutdown of the reactor to MODE 3, including the necessary instrumentation to support maintaining the plant in a safe condition in MODE 3. The criteria governing the design and the specific system requirements of the remote shutdown monitoring instrumentation are located in the UFSAR (Ref.1). (continued) O Brunswick Unit 1 B 3.3-86 Revision No.-
I-l 1 Remote Shutdown Monitoring Instrumentation I L B 3.3.3.2
)
O BmS
- APPLICABLE The Remote Shutdown Monitoring Instrumentation is considered L SAFETY ANALYSES an important contributor to reducing the risk of accidents; I
(continued) as such, it meets Criterion 4 of Reference 2. LCO The Remote Shutdown Monitoring Instrumentation LCO provides the requirements for the OPERABILITY of the monitoring instrumentation necessary to support placing and maintaining the plant in MODE 3 from a location other than the control room. The monitoring instrumentation required are listed in ; Table B 3.3.3.2-1. The monitoring instrumentation are those required for:
)
- Reactor pressure vessel (RPV) pressure control;
- Decay heat removal; and
- RPV inventory control. bl The remote shutdown monitoring instrumentation is OPERABLE
'~ if all instrument channels needed to support the remote shutdown monitoring function are OPERABLE with readouts t . displayed external to the control room. The remote shutdown monitoring instruments covered by this j LC0 do not need to be energized to be considered OPERABLE. This LCO is intended to ensure that the instruments will be OPERABLE if plant conditions require that the remote shutdown monitoring instrumentation be placed in operation. APPLICABILITY The Remote Shutdown Monitoring Instrumentation LCO is applicable in MODES 1 and 2. This is required so that the plant can be placed and maintained in MODE 3 for an extended period of time from a location other than the control room. This LCO is not applicable in MODES 3, 4, and 5. In these MODES, the plant is already subcritical and in a condition of reduced Reactor Coolant System energy. Under these conditions, considerable time is available to restore (continued) ! 1 l l O Brunswick Unit 1 B 3.3-87 Revision No. l 1
Remote Shutdown Monitcring Instrumentaticn B 3.3.3.2 [ V BASES APPLICABILITY necessary instrument functions if control room instruments (continued) or control becomes unavailable. Consequently, the LCO does not require OPERABILITY in MODES 3, 4, and 5. ACTIONS A Note is included that excludes the MODE change restriction of LCO 3.0.4. This exception allows entry into an applicable MODE while relying on the ACTIONS even though the ; ACTIONS may eventually require a plant shutdown. This t exception is acceptable due to the low probability of an I event requiring this system. { Note 2 has been provided to modify the ACTIONS related to Remote Shutdown Monitoring Instrumentation Functions. Section 1.3, Completion Times, specifies that once a l Condition has been entered, subsequent divisions, l subsystems, components, or variables expressed in the ; Condition, discovered to be inoperable or not within limits, j will not result in separate entry into the Condition. l Section 1.3 also specifies that Required Actions of the l Condition continue to apply for each additional failure, with Completion Times based on initial entry into the O Condition. However, the Required Actions for inoperable () Remote Shutdown Monitoring Instrumentation Functions provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable Remote Shutdown , Monitoring Instrumentation Function. ' A.1 I Condition A addresses the situation where one or more required Functions of the remote shutdown monitoring instrumentation is inoperable. This includes any function , listed in Table B 3.3.3.2-1. The Required Action is to j restore the Function (all required channels) to OPERABLE status within 30 days. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room. (continued) O Brunswick Unit 1 B 3.3-88 Revision No.
l i Remote Shutdown Monitaring Instrumentaticn B 3.3.3.2 '(' ( BASES ACTIONS Bl (continued) If the Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without I challenging plant systems, j l SURVEILLANCE SR 3.3.3.2.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of n excessive instrument drift in one of the channels or Q something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. As specified in the Surveillance, a CHANNEL CHECK is only required for those channels that are normally energized. For Function 2 of Table B 3.3.3.2-1, the CHANNEL CHECK requirement does not apply to the N017 instrument loop since this instrument loop has no displayed indication. The CHANNEL CHECK requirement does apply to the remaining instruments of Function 2. The Frequency is based upon plant operating experience that demonstrates channel failure is rare. (continued}. Brunswick Unit 1 B 3.3-89 Revision No.
Remote Shutd:wn Monitcring Instrumentatien l B 3.3.3.2 i BASES SURVEILLANCE SR 3.3.3.2.2 REQUIREMENTS (continued) CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel responds to measured parameter values with the necessary range and accuracy. The 24 month Frequency is based upon operating experience l and consistency with the BNP refueling cycle. REFERENCES 1. UFSAR, Section 7.4.4. 1
- 2. 10CFR50.36(c)(2)(ii).
O Brunswick Unit 1 B 3.3-90 Revision No.
I Remote Shutd:wn Monitoring Instrumentation B 3.3.3.2 f '
\ Table B 3.3.3.2 1 (page 1 of 1) l Remote shutdown monitoring Instrumentation l l
REQUIRED READOUT NLSSER OF I FUNCTION LOCAT10N CHANNELS
- 1. Reactor vesset Pressure (a) 1
- 2. Reactor vessel Water Level (e) g
- 3. S p ession Chamber Water Level (a) 1
- 4. Suppression Chamber Water Temperature (a) g
- 5. Drywett Pressure (g) .j
- 6. Drywell Temperature (a) 9
- 7. Residual Heat Ranovat System Flow (a) 1 l
l (a) Remote Shutdown Panel, Reactor Building 20 ft. Elevation. O l l 1 l l l i
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l l l s Brunswick Unit 1 8 3.3-91 Revision No.
), ATWS-RPT Instrumentaticn B 3.3.4.1 O f
-B'3.3 -INSTRUNENTATION L B 3.3.4.1 Anticipated Transient Without Scram' Recirculation Pump Trip (ATWS-RPT) Instrumentation BASES BACKGROUND The ATWS-RPT System initiates an RPT, adding negative reactivity, following events in which a scram does not, but should occur, to lessen the effects of an ATWS event.
Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level-Low Level 2 or Reactor Vessel Pressure-High setpoint is reached, the recirculation pump drive motor _ breakers trip. The ATWS-RPT System (Ref.1) includes sensors, relays, and circuit breakers that are necessary to cause initiation of an RPT. The channels include electronic equipment (e.g., trip units) that compare measured input signals with . pre-estabitshed setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal tu the trip logic. The ATWS-RPT consists of two independent tri systems, with two channels of Reactor Vessel Pressure-Higk and two channels of Reactor Vessel Water Level-Low Level. 2 in each trip system. Each ATWS-RPT trip system is a two-out-of-two logic for each Function. Thus, either two Reactor Water Level-Low Level 2 or two Reactor Vessel Pressure-High signals are needed to trip a trip system. The outputs of the channels in a trip system are combined in a logic so that either trip system will trip both recirculation pumps (by tripping the respective drive motor breakers). There is one drive motor breaker provided for.each of the two recirculation pumps for a total of two breakers. The output of each trip system is provided to these recirculation pump breakers. APPLICABLE The ATWS-RPT is not assumed to mitigate any accident or SAFETY ANALYSES, transient in the safety analysis. The ATWS-RPT initiates an p LCO, and RPT to aid in preserving the integrity of the fuel cladding APPLICABILITY following events in which a scram does not, but should, occur. Based on its contribution to the reduction of overall plant risk, however, the instrumentation meets Criterion 4 of Reference 2. (continued) Brunswick Unit 1 B 3.3-92 Revision No.
s
'ATWS-RPT Instrumentatien B 3.3.4.1
-BASES lL l
. APPLICABLE The OPERABILITY of the ATWS-RPT is dependent on the l l SAFETY ANALYSES, OPERABILITY of the individual instrumentation channel l LCO, and Functions. Each Function must have a required number of l APPLICABILITY OPERABLE channels in each trip system, with their setpoints (continued) within the specified Allowable Value of SR 3.3.4.1.4. The actual setpoint is ' calibrated consistent with applicabic setpoint methodology assumptions. Channel OPERABILITY also includes the associated recirculation pump drive motor breakers.
' Allowable Values are specified for each ATWS-RPT Function specified in the LCO. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. -A channet is inoperable if its actual trip setting is not within its required Allowable Value. . Trip setpoints are those predetermined values of l output at which an action should take place. The setpoints
- are compared to the actual process parameter (e.g., reactor t
vessel water level), and when the measured output value of l the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the design analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection I because instrumentation uncertainties, process effects, I calibration tolerances, instrument drift,-and severe i environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. The individual Functions are required to be OPERABLE in MODE 1 to protect against common mode failures of the ! Reactor Protection System by providing a diverse trip to mitigate the consequences of a postulated ATWS event. The i
- Reactor Vessel Pressure-High and Reactor Vessel Water l
(continued) i Brunswick Unit 1 B 3.3-93 Revision No.
I ATWS-RPT Instrumentatien B 3.3.4.1 BASES APPLICABLE Level-Low Level 2 Functions are required to be OPERABLE in SAFETY' ANALYSES, MODE I, since the reactor is producing significant power and LCO, and the recirculation system could be at high flow. During this APPLICABILITY MODE, the potential exists for pressure increases or low (continued) water level, assuming an ATWS event. In MODE 2, the reactor is at low power and the recirculation system is at low flow; thus, the potential is low for a pressure increase or low water level, assuming an ATWS event. Therefore, the ATWS-RPT is not necessary. In MODES 3 and 4, the reactor is shut down with all control rods inserted; thus, an ATWS l event is not significant and the possibility of a I significant pressure increase or low water level is , negligible. In MODE 5, the one rod out interlock ensures that the reactor remains subcritical; thus, an ATWS event is not significant. In addition, the reactor pressure vessel (RPV) head is not fully tensioned and no pressure transient threat to the reactor coolant pressure boundary (RCPB) i exists. l Tne specific Applicable Safety Analyses and LCO discussions i are listed below on a Function by Function basis. I
- a. . Reactor Vessel Water Level-Low Level 2 Low RPV water level indicates the capability to cool 3 the fuel may be threatened. 'Should RPV water level i decrease too far, fuel damage could result. 1 Therefore, the ATWS-RPT System is initiated at Level 2 ,
to aid in maintaining level above the top of the l' active fuel. The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant bolloff. Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual l water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Level 2, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Vessel Water Level-Low Level 2 Allowable Value is (continued) O Brunswick Unit 1 B 3.3-94 Revision No.
[. ATWS-RPT Instrumentation B 3.3.4.1 BASES i APPLICABLE a. Beactor Vessel Water level-Low levell (continued) SAFETY ANALYSES, , LCO, and chosen so that the system will not be initiated after APPLICABILITY a Level I scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point.
- b. Reactor Vessel Pressure-Hiah Excessively high RPV pressure may rupture the RCPB.
An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron i flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Vessel Pressure-High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the A overpressurization event, the RPT aids in the Q termination of the ATWS event and, along with the safety / relief valves, limits the peak RPV pressure to less than the ASME Section III Code Service Level C limits (1500 psig). The Reactor Vessel Pressure-High signals are initiated from four pressure transmitters that monitor reactor vessel pressure. Four channels of Reactor Vessel Pressure-High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this function on a valid signal. The Reactor Vessel Pressure-High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code Service Level C allowable Reactor Coolant System pressure. I i ACTIONS A Note has been provided to modify the ACTIONS related to ATWS-RPT instrumentation channels. Section 1.3, Completion ! Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into (continued) O j b ! Brunswick Unit 1 B 3.3-95 Revision No. '
ATWS-RPT Instrumentatien B 3.3.4.1 BASES ACTIONS the Condition. Section 1.3 also specifies that Required (continued) Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable ATWS-RPT instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable ATWS-RPT instrumentation channel. A.1 and A.2 With one or more channels inoperable, but with ATWS-RPT capability for each Function maintained-(refer to. Required Actions B.1 and C.1 Bases), the ATWS-RPT System is capable of performing the intended function. However, the reliability and redundancy of the ATWS-RPT instrumentation ; is reduced, such that a single failure in the remaining trip I system could result in the inability of the ATWS-RPT System to perform the intended function. . Therefore, only a limited time is allowed to restore the inoperable channels to OPERABLE status. Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of ATWS-RPT, 14 days is provided to restore the inoperable channel (Required Action A.1). Alternately, the ; inoperable channel may be placed in trip (Required l Action A.2), since this would conservatively compensate for ; the inoperability, restore capability to accommodate a I single failure, and allow operation to continue. As noted, i placing the channel in trip with no further restrictions is ; not allowed if the inoperable channel is the result of an ' inoperable breaker, since this may not adequately compensate for the inoperable breaker (e.g., the breaker may be inoperable such that it will not open). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in an RPT), or if the inoperable channel is the result of an , inoperable breaker, Condition D must be entered and its 1 Required Actions taken. ' (continued) O Brunswick Unit 1 B 3.3-96 Revision No.
i l- ATWS-RPT Instrumentatien B 3.3.4.1 BASES ! ACTIONS M l (continued) ! Required Action 8.1 is intended to ensure that appropriate ! tctions are taken if multiple, inoperable, untripped i channels within the same Function result in the Function not maintaining ATWS-RPT trip capability. A Function is considered to be maintaining ATWS-RPT trip capability when sufficient channels are OPERABLE or in trip such that the
- ATWS-RPT System will generate a trip signal from the given Function on a valid signal, and both recirculation pumps can be tripped. This requires two channels of the Function in the same trip system to each be OPERABLE or in trip, and the recirculation pump drive motor breakers to be OPERABLE or in
.. trip. The 72 hour Completion Time is sufficient for the operator L to take corrective action (e.g., restoration or tripping of channels) and takes into account the likelihood of an event-requiring actuation of the ATWS-RPT instrumentation during this period and that one Function is still maintaining j' ATWS-RPT trip capability. O u Required Action C.1 is intended to ensure that appropriate Actions are taken if multiple, inoperable, untripped channels within both functions result in both Functions not maintaining ATWS-RPT trip capability. The description of a function maintaining ATWS-RPT trip capability is discussed j in the Bases for Required Action B.1 above. The I hour Completion Time is sufficient for the operator to take corrective action and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period. D 1 and 0.2 With any Required Action and associated Completion Time not met, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least M00E:2 within 6 hours (Required Action D.2). Alternately, the associated recirculation pump (s) may be removed from service since this (continued) I-O Brunswick Unit 1 B 3.3-97 Revision No.
f ATWS-RPT Instrumentatien B 3.3.4.1 BASES l ACTIONS D.1 and D.2 (continued) performs the intended function of the instrumentation (Required Action D.1). The allowed Completion Time of 6 hours is reasonable, based on operating experience, both to reach MODE 2 from full power conditions and to remove a recirculation pump from service in an orderly manner and without challenging plant systems. SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a ' channel is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for op in 6 hours provided the associated Function maintains A1WS-RPT trip capability. Upon completion of the survelliance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 3) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the ( probability that the recirculation pumps will trip when necessary. SR 3.3.4.1.1 Performance of the CHANNEL CHECK once every 24 hours ensures j that a gross failure of instrumentation has not occurred. A 1 CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument ; channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is (continued) O Brunswick Unit 1 B 3.3-98 Revision No.
l ATWS-RPT Instrumentation B 3.3.4.1 (~) (/ BASES
)
SURVEILLANCE SR 3.3.4.1.1 (continued) REQUIREMENTS outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel fr.ilure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO. ' SR 3.3.4.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required l channel to ensure that the channel will perform the intended ' function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. l The Frequency of 92 days is based on the reliability l analysis of Reference 3. I SR 3.3.4.1.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than , the Allowable Value specified in SR 3.3.4.1.4. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the i channel performance is still within the requirements of the plant design analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability ' analysis of Reference 3. SR 3.3.4.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel 4 responds to the measured parameter within the necessary (continued) O Brunswick Unit 1 B 3.3-99 Revision No.
r I t ATWS-RPT Instrumentatien B 3.3.4.1 BASES i SURVEILLANCE SR 3.3.4.1.4 (continued) REQUIREMENTS range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency is based upon the' assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.4.1.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the design function. Therefore, if a breaker is incapable of operating, the associated instrument channel (s) would be inoperable. 1 The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated these components will-usually pass the Surveillance when performed at the 24 month Frequency. REFERENCES 1. UFSAR Sections 5.4.1.2.4 and 7.6.1.3.1.
- 2. 10CFR50.36(c)(2)(ii).
- 3. GENE-770-06-1-A, Bases for Changes To Surveillance Test Intervals and Allowed Out-of-Service Times For Selected Instrumentation Technical Specifications, December 1992.
O Brunswick Unit i B 3.3-100 Revision No. 1 I
ECCS Instrumentation B 3.3.5.1 O (/ B 3.3 INSTRUMENTATION B 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation BASES BACKGROUND The purpose of the ECCS instrumentation is to initiate appropriate responses from the systems to ensure that the fuel is adequately cooled in the event of a design basis accident or transient. For most anticipated operational occurrences and Design Basis Accidents (DBAs), a wide range of dependent and independent parameters are monitored. The ECCS instrumentation actuates core spray (CS), the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System, high pressure coolant injection (HPCI), Automatic Depressurization System (ADS), and the diesel generators (DGs). The equipment involved with each of these systems is described in the Bases for LCO 3.5.1, "ECCS-Operating" or LCO 3.8.1, "AC Sources-0perating."
) Core Spray System The CS System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Level 3 or Drywell Pressure-High coincident with Reactor Steam Dome Pressure-Low. Each of these diverse variables is. monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the trip units are connected to relays whose cc. tacts are arranged in a one-out-of-two taken twice logic (i.e., two trip systems) for each Function.
The CS System initiation signal is a sealed in signal and must be manually reset. The CS System can be reset if reactor water level and high drywell pressure have been restored. Upon receipt of an initiation signal, the CS pumps are started approximately 15 seconds after power is available to limit the loading of the AC power sources. (continued) O Brunswick Unit 1 B 3.3-101 Revision No.
ECCS Instrumentation B 3.3.5.1
, BASES BACKGROUND Core Sorav System (continued)
The CS test line isolation valve, which is also a primary containment isolation valve'(PCIV), is closed on a CS initiation' signal to allow full system flow assumed in the accident analyses and maintain primary containment isolated in the event CS is not operating. The CS System also monitors the pressure in the reactor to ensure that, before the injection valves open, .the reactor pressure has fallen to a value below the CS System's maximum design pressure. The variable is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic.> Low Pressure Coolant In.iection System The LPCI is 'an operating mode of the Residual Heat Removal (RHR) System, with two LPCI subsystems. The LPCI subsystems may be initiated by automatic or manual means. Automatic O initiation occurs for conditions of Reactor Vessel Water Level-Low Level 3 or Drywell Pressure-High coincident with Reactor Steam Dome Pressure-Low. Each of these diverse variables is monitored by four redundant transmitters, which, in turn, are connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic (i.e., two trip systems) for each Function. Once an initiation signal is received by the LPCI control circuitry, 1 the signal is sealed in until manually reset. ; Upon receipt of an initiation signal, the LPCI pumps are started approximately 10 seconds after power is available to I limit the loading of the AC power sources. l The RHR test line suppression pool cooling isolation valve, suppression pool spray isolation valves, and containment spray isolation valves (which are also'PCIVs) are also closed on a LPCI initiation signal to allow the full system l flow assumed in the accident analyses and maintain primary containment isolated in the event LPCI is not operating. i (continued) i O Brunswick Unit 1 B 3.3-102 Revision No.
i ECCS Instrumentatien B 3.3.5.1 .O O BASES BACKGROUND Low Pressure Coolant Injection System (continued) The LPCI System monitors the pressure in the reactor to ensure that, before an injection valve opens, the reactor pressure has fallen to a value below the LPCI System's maximum design pressure. The variable is monitored by four i redundant transmitters, which are, in turn, connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic. Additionally, instruments are provided to close the recirculation loop pump discharge valves to ensure that LPCI flow does not bypass the core when it injects into the recirculation lines. The variable is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic. Low reactor water level in the shroud is detected by two additional instruments to automatically isolate other modes of RHR (e.g., suppression pool cooling) when LPCI is required. One in::trument closes LPCI loop A valves and the q y other instrument closes LPCI loop B valves. Manual overrides for these isolations are provided, t Hiah Pressure Coolant Injection System The HPCI System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Level 2 or Drywell Pressure-High. Each of these variables is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic for each Function. ; The HPCI test line isolation valve is closed upon receipt of a HPCI initiation signal to allow the full system flow assumed in the accident analysis. , The HPCI System also monitors the water levels in the condensate storage tank (CST) and the suppression pool because these are the two sources of water for HPCI operation. Reactor grade water in the CST is the normal (continued) O Brunswick Unit 1 B 3.3-103 Revision No.
-j ECCS Instrumentatien B 3.3.5.1
/* BASES-BACKGROUND Hiah Pressure Coolant In.iection System (continued) source. Upon receipt of a HPCI initiation signal, the CST suction valve is automatically signaled to open it is normally in the open position).unless both suppre(ssion pool suction valves are open. If the water level in the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes. Two level switches are used to detect low water level in the CST. Either switch can cause the suppression pool suction valves to open and the CST suction valve to close. Two level switches are also used to detect high water level in the suppression pool. Either switch can cause an automatic swap of the HPCI pump suction valves. The suppression pool suction valves also automatically open and the CST suction valve closes if high water level is detected in the suppression pool. To pravent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction path must be open before the CST suction path is automatically isolated. The HPCI System provides makeup water to the reactor until O the reactor vessel water level reaches the Reactor Vessel Water Level-High trip, at which time the HPCI turbine ! trips, which causes the turbine's stop valve and the ! injection valve to close. This variable is monitored by two l transmitters, which are, in turn, connected to two trip ; units. The outputs of the trip units are connected to i relays whose contacts are arranged in a two-out-of-two logic to provide high reliability of the HPCI System. The HPCI System automatically restarts if a Reactor Vessel Water Level-tow Level 2 signal is subsequently received. , Automatic Depressurization System-The ADS may be initiated by either automatic or manual means. Automatic initiation occurs when signals indicating Reactor Vessel Water Level-Low Level 3; and confirmed Reactor Vessel Water Level-Low Level 1; and CS or RHR (LPCI Mode) Pump Discharge Pressure-High are all present and the ADS Timer has timed out. There are two transmitters for Reactor Vessel Water Level-Low Level 3 and one transmitter for confirmed Reactor Vessel Water Level-Low Level 1 in (continued) O Brunswick Unit 1 B 3.3-104 Revision No.
L 1-p ECCS Instrumentatien B 3.3.5.1 , D C BASES 1 1 BACKGROUND Automatic Depressurization System (continued) each of the two ADS trip systems. Each of these transmitters connects to a trip unit, which then drives a l relay whose contacts form the initiation logic. Each ADS trip system includes a time delay between satisfying the initiation logic and the actuation of the ADS valves. The ADS Timer time delay setpoint chosen is long . enough that the HPCI System has sufficient operating time to i recover to a level above Reactor Vessel Water Level-Low l Level 3, yet not so long that the LPCI and CS Systems are unable to adequately cool the fuel if the HPCI System fails to maintain that level. An alarm in the control room is annunciated when either of the timers is timing. Resetting the ADS initiation signals resets the ADS Timers. The ADS also monitors the discharge pressures of the four LPCI pumps and the two CS pumps. Each ADS trip system includes two discharge pressure permissive switches from one CS pump and from each LPCI pump in a Division (i.e., Division II LPCI subsystems B and D input to ADS trip system A, and Division I LPCI subsystems A and C input to O ADS trip system B). The signals are used as a permissive for ADS actuation, indicating that there is a source of core coolant available once the ADS has depressurized the vessel. One CS pump or two RHR pumps in a LPCI loop are sufficient to permit automatic depressurization. The ADS logic in each trip system is arranged in two strings. Each string has a contact from Reactor Vessel Water Level-Low Level 3. One of the two strings in each trip system also has a confirmed Reactor Vessel Water Level-Low Level I contact and an ADS Timer. All contacts in both logic strings must close, the ADS timer must time out, and a CS or LPCI pump discharge pressure signal must be present to initiate an ADS trip system. Either the A or B trip system will cause all the ADS relief valves to open. Once the ADS Timer has timed out and the ADS initiation signal is present, the trip system is sealed in until manually reset. Manual inhibit switches are provided in the control room for the ADS; however, their function is not required for ADS OPERABILITY (provided ADS is not inhibited when required to be OPERABLE). (continued) Brunswick Unit 1 B 3.3-105 Revision No.
1 F . ECCS Instrumentatien
- B 3.3.5.1 BASES BACKGROUND Diesel Generators i-
' (continued) The DGs may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Level 3 or Drywell Pressure-High coincident with Reactor Steam Dome Pressure-Low. The DGs are also initiated upon loss of voltage signals. (Refer to the Bases for LCO 3.3.8.1, " Loss of Power (LOP) Instrumentation," for a discussion of these signals.) Each of these diverse variables is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the four trip units are connected to relays whose contacts are connected to a one-out-of-two taken twice logic to initiate all DGs. The DGs receive their initiation signals from the CS System initiation logic. The DGs can also be started manually from the control room and locally from the associated DG room. Upon receipt of a loss of coolant accident (LOCA) initiation ; signal, each DG is automatically started, is ready to load within 10 seconds, and will run in standby conditions (rated voltage and frequency, with the DG output breaker open). The DGs will only energize their respective 4.16 kV-emergency buses if a loss of offsite power occurs. (Refer O to Bases for LCO 3.3.8.1.) APPLICABLE The actions of the ECCS are explicitly assumed in the safety SAFETY ANALYSES, analyses of References 1, 2, and 3. The ECCS is initiated LCO, and to preserve the integrity of the fuel cladding by limiting APPLICABILITY the post LOCA peak cladding temperature to less than the 10 CFR 50.46 limits. ECCS instrumentation satisfies Criterion 3 of Reference 4. Certain instrumentation Functions are retained for other i' reasons and are described below in the individual Functions discussion. l The OPERABILITY of the ECCS instrumentation is dependent l upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.1-1. Each function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. The actual setpoint is calibrated . consistent with applicable setpoint methodology assumptions.- i (continued) O Brunswick Unit 1 8 3.3-106 Revision No. ,
l l ECCS Instrumentation , B 3.3.5.1 O \b BASES APPLICABLE Allowable Values are specified for each ECCS Function SAFETY ANALYSES, specified in the table. Trip setpoints are specified in the LCO, and setpoint calculations. The setpoints are selected to ensure APPLICABILITY that the trip settings do not exceed the Allowable Value (continued) between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable i Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of . the process parameter exceeds the setpoint, the associated i device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. pd The trip setpoints and Allowable Values determined in this j manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions that may require ECCS (or DG) initiation to mitigate the consequences of a design basis transient or accident. Table 3.3.5.1-1 footnotes (a), (b), and (c) specifically indicate other conditions when certain ECCS Instrumentation Functions are required to be OPERABLE. To ensure reliable ECCS and DG function, a combination of Functions is required to provide j primary and secondary initiation signals. The specific Applicable Safety Analyses, LCO, and l Applicability discussions are listed below on a Function by J Function basis. (continued) O O Brunswick Unit 1 B 3.3-107 Revision No.
ECCS Instrumentation B 3.3.5.1 V
- BASES APPLICABLE- Core Sorav and Low Pressure Coolant Injection Systems SAFETY ANALYSES, LCO, and 1.a. 2.a. Reactor Vessel Water level-Low Level 3 APPLICABILITY (continued) Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
The low pressure ECCS and associated DGs are initiated at Reactor Vessel Water Level-Low Level 3 to ensure that core spray and flooding functions are available to prevent or minimize fuel damage. The Reactor Vessel Water Level-Low
. Level 3 is one of the Functions assumed to be OPERABLE and capable of initiating the ECCS and associated DGs during the transients analyzed in References 1 and 3. In addition, the Reactor Vessel Water Level-Low Level 3 Function is directly assumed in the analysis of the recirculation line break (Ref. 5). The core cooling function of the ECCS, along with the scram action of the Reactor Protection System (RPS),
ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Reactor Vessel Water Level-Low Level 3 signals are O initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Reactor Vessel Water Level-Low Level 3 Allowable Value { is chosen to allow time for the low pressure core flooding j systems to activate and provide adequate cooling. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. Four channels of Reactor Vessel Water Level-Low Level 3 Function are only required to be OPERABLE when the ECCS or DG(s) are required to be OPERABLE to ensure that no single instrument failure can preclude ECCS and DG initiation. Refer to LCO 3.5.1 and LCO 3.5.2, "ECCS-Shutdown," for Applicability Bases for the low pressure ECCS subsystems; and LCO 3.8.1 and LC0 3.8.2, "AC Sources-Shutdown," for Applicability Bases for the DGs. 1.b. 2.b. Drywell Pressure-Hiah High pressure in the drywell-could indicate a break in the reactor coolant pressure boundary (RCPB). The low pressure i I (continued) 1 Brunswick Unit 1 B 3.3-108 Revision No. ! l
ECCS Instrumentatien B 3.3.5.1
'"ASES i
APPLICABLE . 1.b. 2.b. Drywell Pressure-Hiah (continued) SAFETY ANALYSES, LCO, and ECCS and associated DGs are initiated upon receipt of the APPLICABILITY Orywell Pressure-High Function coincident with Reactor l Steam Dome Pressure-Low Function in order to minimize the I possibility of fuel damage. The Drywe11 ~ Pressure-High l Function is directly assumed in the analysis of the recirculation line break (Ref. 5). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. ) High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible to be indicative of a LOCA inside primary containment. The Drywell Pressure-High Function is required to be OPERABLE when the ECCS or DG is required to be OPERABLE in conjunction with times when the primary containment is required to be OPERABLE. Thus, four channels of the CS and LPCI Drywell Pressure-High Functions are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single O. instrument failure can preclude ECCS and DG initiation. In MODES 4 and 5, the Drywell Pressure-High Function is not required, since there is insufficient energy in the reactor to pressurize the primary containment to Drywell Pressure-High setpoint. Refer to- LCO 3.5.1 for. Applicability Bases for the low pressure ECCS' subsystems and i to LC0 3.8.1 for Applicability Bases for the DGs. ! 1.c. 2.c. Reactor Steam Dome Pressure-tow i Low reactor steam dome pressure signals are used as l permissives for the low pressure ECCS subsystems. This l ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these' subsystems' maximum design pressure. The low reactor steam dome pressure signals are also used in the Drywell' Pressure-High logic circuits to distinguish high drywell pressure caused by a LOCA from that caused by loss of drywell cooling. The Reactor Steam Dome
~ Pressure-Low is one of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS and associated DGs during the transients analyzed in References 2 and 3. In addition, the Reactor Steam Dome (continued)
O . Brunswick Unit 1 B 3.3-109 Revision No.
y ECCS Instrumentaticn j B 3.3.5.1 {
\
. BASES APPLICABLE 1.c. 2.c. Reactor Steam Dome Pressure-Low (continued) i SAFETY ANALYSES,
! LCO, and Pressure-Low Function is directly assumed in the analysis j APPLICABILITY of the recirculation line break (Ref. 5). The core cooling l function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains l below the limits of 10 CFR 50.46. The Reactor Steam Dome Pressure-Low signals are initiated from four pressure transmitters that sense the reactor done pressure. The Allowable Value is low enough to prevent overpressuring the equipment in the low pressure ECCS, but high enough to ' ensure that the ECCS injection prevents the fuel peak l cladding temperature from exceeding the limits of 10 CFR 50.46. Four channels of Reactor Steam Dome Pressure-Low Function are only required to be OPERABLE when the ECCS or DG(s) are required to be OPERABLE to ensure that no single instrument failure can preclude ECCS and DG initiation. Refer to LC0 3.5.1 and LCO 3.5.2 for Applicability Bases for the low ,Cj\ pressure ECCS subsystems; and LCO 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs. 1.d. 2.f. Core SDray and RHR Pump Start-Time Delay Relays The purpose of these time delays is to stagger the start of the CS and RHR pumps that are in each of Divisions I and II, thus limiting the starting transients on the 4.16 kV emergency buses. These functions are necessary when power is being supplied from either the normal power sources (offsite power) or the standby power sources (DGs). The Core Spray Pump Start-Time Delay Relays and the RHR Pump. Start-Time Delay Relays are assumed to be OPERABLE in the accident and transient analyses requiring ECCS initiation. That is, the analyses assume that the pumps will initiate when required and excess loading will not cause failure of the power sources. There are eight RHR Pump Start-Time Delay Relays, two channels in each of the RHR pump start logic circuits. There are six CS pump start timers arranged such that there ; are four separate channels of the Core Spray Pump Start i Time-Delay Relay Function, two channels in each of the CS (continued) l ! B 3.3-110 Revision No. ! Brunswick Unit 1 I
ECCS Instrumentatien B 3.3.5.1 BASES APPLICABLE 1.d. 2.f. Core Sorav and RHR Pumo Start-Time Delav Relays SAFETY ANALYSES (continued) LCO, and APPLICABILITY pump start logic circuits. Each channel consists of an individual.10 second timer and a 5 second timer. The 5 second timer is common to both channels associated with a CS pump start logic circuit. Each 10 second timer associated with a'CS pump start logic channel is shared with an RHR pump start logic channel. While two time delay relay channels are dedicated to a single CS pump start logic, a single failure of a 5 second CS pump timer could result in the failure of the two low pressure ECCS pumps, powered from the same 4.16 kV emergency bus, to perform their intended function within the assumed ECCS RESPONSE TIME (e.g., as in the case where both ECCS pumps on one 4.16 kV emergency bus start simultaneously due to an inoperable time delay relay). This still leaves four of the six low pressure ECCS pumps OPERABLE. Additionally, a failure of both shared time delay relay channels in an RHR and CS pump start logic circuit would also leave four of the six low pressure ECCS pumps OPERABLE as described above. As a result, to satisfy the single failure criterion (i.e., O. loss of one instrument does not preclude ECCS initiation), only one channel per pump of the Core Spray and RHR Pump Start-Time Delay Relay functions are required to be OPERABLE when the associated ECCS subsystera is required to be OPERABLE. Refer to LCO 3.5.1 and LC0 3.5.2 for Applicability Bases for the ECCS subsystems. The Allowable Values for the Core Spray and RHR Pump Start-Time Delay Relays are chosen to be long enough so that most of the starting transient of the previously started pump is complete before starting a subsequent pump on the same 4.16 kV emergency bus and short enough so that ECCS operation is not degraded. 2.d. Reactor Steam Dome Pressure-Low (Recirculation Pump Discharoe Valve Permissive) Low reactor steam dome pressure signals are used as permissives for recirculation pump discharge valve closure and recirculation pump discharge bypass valve closure. This ensures that the LPCI subsystems inject into the proper RPV location assumed in the safety analysis. The Reactor Steam 'J (continued) i O Brunswick Unit 1 8 3.3-111 Revision No. l I l 1 l
)
l ECCS Instrumentation B 3.3.5.1 ( v BASES l APPLICABLE 2.d. Reactor Steam Dome Pressure-Low (Recirculation Pumo SAFETY ANALYSES Discharae Valve Permissive) (continued) i LCO, and l APPLICABILITY Dome Pressure-tow is one of the Functions assumed to be l OPERABLE and capable of closing the valve (s) during the i transients analyzed in References 2 and 3. The core cooling function of the ECCS, along with the scram action of the i RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Steam Dome Pressure-Low Function is directly assumed in the analysis of the recirculation line break (Ref. 5). The Reactor Steam Dome Pressure-Low signals are initiated i from four pressure transmitters that sense the reactor dome pressure. The Allowable Value is chosen to ensure that the valves close prior to commencement of LPCI injection flow into the , core, as assumed in the safety analysis. ' Four channels of the Reactor Steam Dome Pressure-Low Function are only required to be OPERABLE in MODES 1, 2, and 3 with the associated recirculation pump discharge valve { open or the associated recirculation pump discharge bypass valve open. With the valve (s) closed, the function of instrumentation has been performed; thus, the Function is not required. In MODES 4 and 5, the loop injection location is not critical since LPCI injection through the recirculation loop in either direction will still ensure that LPCI flow reaches the core (i.e., there is no significant reactor steam dome back pressure). 2.e. Reactor Vessel Shroud level The Reactor Vessel Shroud Level function is provided as a permissive to allow the RHR System to be manually aligned from the LPCI mode to the suppression pool cooling / spray or drywell spray modes. The permissive ensures that water in the vessel is at least two thirds core height before the manual transfer is allowed. This ensures that LPCI is available to prevent or minimize fuel damage. This function may be overridden during accident conditions as allowed by plant procedures. The Reactor Vessel Shroud Level Function is implicitly assumed in the analysis of the recirculation (continued) O Brunswick Unit 1 B 3.3-112 Revision No.
i l ECCS Instrumentation B 3.3.5.1
.m BASES APPLICABLE 2.e. Reactor Vessel Shroud level (continued)
SAFETY ANALYSES, LCO, and line break (Ref. 5) since the analysis assumes that no LPCI APPLICABILITY flow diversion occurs when reactor water level is below the Reactor Vessel Shroud Level. Reactor Vessel Shroud Level signals are initiated from two level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Reactor Vessel Shroud Level Allowable Value is chosen to l allow the low pressure core flooding systems to activate and provide adequate cooling before allowing a manual transfer. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero j point. Two channels of the Reactor Vessel Shroud Level Function are only required to be OPERABLE in MODES 1, 2, and 3. In MODES 4 and 5, the specified initiation time of the LPCI subsystems is not assumed, and other administrative controls t' are adequate to control the valves that this function ) isolates (since the systems that the valves are opened for are not required to be OPERABLE in MODES 4 and 5 and are normally not used). HPCI System 3.a. Reactor Vessel Water Level-Low level 2 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the HPCI System is initiated at Level 2 to maintain level above the top of the active fuel. The Reactor Vessel Water level-Low I Level 2 is one of the Functions assumed to be OPERABLE and I capable of initiating HPCI during the transients analyzed in References 2, 3, and 6. Reactor Vessel Water Level-Low Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. s n (continued) U Brunswick Unit 1 B 3.3-113 Revision No.
ECCS Instrumentation B 3.3.5.1 C' BASES APPLICABLE 3.a. Reactor Vessel Water Level-Low level 2 (continued) SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level-Low Level 2 Allowable Value APPLICABILITY is low enough to avoid a HPCI- System start from normal reactor level transients (e.g., a reactor scram without the loss of feedwater flow) and high enough to avoid initiation of low pressure ECCS at Reactor Vessel Water Level-Low Level 3 during a transient resulting from a complete loss of feedwater flow. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. Four channels of Reactor Vessel Water level-Low Level 2 Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI initiation. Refer to LCO 3.5.1 for HPCI Applicability Bases. 3.b. Drywell Pressure-Hiah High pressure in the drywell could indicate a break in the RCPB. The HPCI System is initiated upon receipt of the Drywell Pressure-High Function in order to minimize the b] possibility of fuel damage. The Drywell Pressure-High Function is not assumed in accident or transient analyses. It is retained since it is a potentially significant contributor to risk. High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible to be indicative of a LOCA inside primary containment. ! Four channels of the Drywell Pressure-High Function are required to be OPERABLE when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI initiation. Refer to LLO 3.5.1 for the Applicability Bases for the HPCI System. 3.c. Reactor Vessel Water level-Hiah High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the Reactor Vessel Water Level-High signal is used to trip the HPCI turbine to prevent overflow into the main steam lines (MSLs) which precludes an unanalyzed event. (continued) Brunswick Unit 1 8 3.3-114 Revision No.
l l ECCS Instrumentaticn l 8 3.3.5.1 i BASES 4 APPLICABLE 3.c. Reactor Vessel Water Level-Hiah (continued) SAFETY ANALYSES, LCO, and Reactor Vessel Water Level-High signals for HPCI are APPLICABILITY initiated from two level transmitters from the narrow range water level measurement instrumentation. Both Reactor l Vessel Water Level-High signals are required in order to ' close the HPCI turbine stop valve. This ensures that no single instrument failure can preclude HPCI initiation. The Reactor Vessel Water Level-High Allowable Value is high enough to avoid interfering with HPCI System operation i during reactor water level recovery resulting from low ' reactor water level events and low enough to prevent flow from the HPCI System from overflowing into the MSLs. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. Two channels of Reactor Vessel Water Level-High Function are required to be OPERABLE only when HPCI is required to be OPERABLE. Refer to LCO 3.5.1 for HPCI Applicability Bases. , 1 3.d Condensate Storace Tank level-Low 1 Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source. Normally the suction valves between HPCI and the CST are open and, upon receiving a HPCI initiation signal, water for HPCI injection would be taken from the CST. However, if the water level in the CST falls below a preselected level, , first the suppression pool suction valves automatically ! open, and then the CST suction valve automatically closes. This ensures that an adequate supply of makeup water is available to the HPCI pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes. The Function is implicitly assumed in the accident and transient analyses (which take credit for HPCI) since the analyses assume that the HPCI suction source is the suppression pool. The Condensate Storage Tank Level-Low signal is initiated from two level switches. The logic is arranged such that either level switch can cause the suppression pool suction valves to open and the CST suction valve to close. The l (continued) i 'G Brunswick Unit 1 8 3.3-115 Revision No.
II l ECCS Instrumentaticn B 3.3.5.1 BASES APPLICABLE-3.d Condensate Storaae Tank level-Low (continued) SAFETY ANALYSES LCO, and Condensate Storage Tank Level-Low function Allowable Value APPLICABILITY is high enough to ensure adequate pump suction head while water is being taken from the CST. Two channels of the Condensate Storage Tank Level-Low function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI swap to suppression pool source. Refer to LC0 3.5.1 for HPCI Applicability Bases. 3.e. Suppression Chamber Water Level-Hiah Excessively high suppression pool water could impact operation of the HPCI and Reactor Core Isolation Cooling (RCIC) exhaust vacuum breakers resulting in an inoperable HPCI or RCIC System. Therefore, signals indicating high suppression pool water level are used to transfer the suction source of HPCI from the CST to the suppression pool to eliminate the possibility of HPCI continuing to provide additional water from a source outside containment. To prevent losing suction to the pump, the suction valves are ( interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes. This function is implicitly assumed in the accident and transient analyses (which take credit for HPCI) since the analyses assume that the HPCI suction source is the suppression pool. The Suppression Chamber Water Level-High signal is initiated from two level switches. The logic is arranged i such that either switch can cause the suppression pool suction valves to open and the CST suction valve to close. The Allowable Value for the Suppression Chamber Water Level-High function is chosen to ensure that HPCI will be aligned for suction from the suppression pool before the , water level reaches the point at which the HPCI and RCIC 1 exhaust vacuum breakers become inoperable. The Allowable Value is referenced from the suppression chamber water level zero. Suppression chamber water level zero is one inch ; below the torus centerline. (continued) O Brunswick Unit 1 B 3.3-116 Revision No.
ECCS Instrumentatien i B 3.3.5.1 i BASES APPLICABLE 3.e. Suppression Chamber Water Level-Hioh (continued) l SAFETY ANALYSES, i LCO, and Two channels of Suppression Chamber Water Level-High APPLICABILITY Function are required to be OPERABLE only when HPCI is (continued) required to be OPERABLE to ensure that no single instrument failure can preclude HPCI swap to suppression pool source. Refer to LC0 3.5.1 for HPCI Applicability Bases. Automatic DeDressuriZation System (ADS 1 4.a. 5.a. Reactor Vessel Water level-tow level 3 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, ADS receives one of the signals necessary for initiation from this function. The Reactor Vessel Water Level-Low Level 3 is one of the functions assumed to be OPERABLE and capable of initiating the ADS during the accident analyzed in References 2 and 5. The core cooling function of the ECCS, along with the scram action.of the RPS, ensures that the fuel peak cladding temperature remains below the limits of ! ,O 10 CFR 50.46. l V Reactor Vessel Water Level-Low Level 3 signals are initiated from four icvel transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Level 3 Function are required to be OPERABLE only when ADS is required to be ; OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two channels input to ADS trip system A, while the other two channels input to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases. The Reactor Vessel Water Level-Low Level 3 Allowable Value is chosen to allow time for the low pressure core flooding i systems to initiate and provide adequate cooling. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point, , (continued) r k Brunswick Unit 1 B 3.3-117 Revision No.
ECCS Instrumentaticn B 3.3.5.1-O V - BASES APPLICABLE 4.b. 5.b. ADS Timer SAFETY ANALYSES, LCO, and The purpose of the ADS Timer is to delay depressurization of APPLICABILITY the reactor vessel to allow the HPCI System time to maintain (continued) reactor vessel water level. Since the rapid depressurization-caused by ADS operation is one of the most severe transients on the reactor vessel,.its occurrence should be limited. By delaying initiation of the ADS Function, the operator is given the chance to monitor the success or failure of the HPCI System to maintain water level, and then to decide whether or not to allow ADS to initiate, to delay initiation further by recycling the timer, or to inhibit initiation permanently. The ADS Timer Function is assumed to be OPERABLE for the' accident analyses of References 2 and 5 that require ECCS initiation and assume failure of the HPCI System. There are two ADS Timer relays, one in each of the two ADS trip systems. The Allowable Value for the ADS Timer is chosen to be long enough to allow HPCI to start and avoid an inadvertent blowdown yet short enough so that there is still time after depressurization for the low pressure ECCS subsystems to provide adequate core cooling. Two channels of the ADS Timer Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. One channel inputs to ADS trip system A, while the other channel inputs to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases. 4.c. 5.c. Reactor Vessel Water level-Low Level 1 The Reactor Vessel Water Level-Low Level 1 Function is used by the ADS only as a confirmatory low water level signal. ADS receives one of the signals necessary for initiation from Reactor Vessel Water Level-Low Level 3 signals. In order to prevent spurious initiation of the ADS due to spurious Level 3 signals, a Level 1 signal must also be received before ADS initiation commences.. Reactor Vessel Water Level-Low Level 1 signals are l initiated from two level transmitters that sense the 1 difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Allowable (continued) O Brunswick Unit 1 B 3.3-118 Revision No.
I ECCS Instrumentatien B 3.3.5.1 U('M BASES APPLICABLE 4.c. 5.c. Reactor Vessel Water Level-Low Level 1 SAFETY ANALYSES, (continued) LCO, and APPLICABILITY Value for Reactor Vessel Water Level--Low Level 1 is selected at the RPS Level 1 scram Allowable Value for convenience. Refer to LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation," for the Bases discussion of l this Function. The Allowable Value is referenced from l reference level zero. Reference level zero is 367 inches I above the vessel zero point. l Two channels of Reactor Vessel Water Level-Low Level 1 Function are only required to be OPERABLE when the ADS is , required to be OPERABLE to ensure that no single instrument j failure can preclude ADS initiation. One channel inputs to ADS trip system A, while the other channel inputs to ADS ! trip system B. Refer to LC0 3.5.1 for ADS Applicability l Bases, l 4.d. 4.e. 5 d. 5.e. Core Soray and RHR (LPCI Mode) Pump l Discharae Pressure-Hiah A The Pump Discharge Pressure-High signals from the CS and V RHR pumps are used as permissives for ADS initiation, indicating that there is a source of low pressure cooling l l water available once the ADS has depressurized the vessel. Pump Discharge Pressure-High is one of the Functions assumed to be OPERABLE and capable of permitting ADS l initiation during the events analyzed in References 2 and 5 , with an assumed HPCI failure. For these events the ADS l depressurizes the reactor vessel so that the low pressure ' ECCS can perform the core cooling functions. This core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature l remains below the limits of 10 CFR 50.46. Pump discharge pressure signals are initiated from twelve pressure switches, two on the discharge side of each of the six low pressure ECCS pumps. In order to generate an ADS permissive in one trip system, it is necessary that only one CS pump (both channels for the pump) indicate the high discharge pressure condition or two RHR pumps in one LPCI loop (one channel for each pump) indicate a high discharge pressure condition. The Pump Discharge Pressure-High Allowable Value is less than the pump discharge pressure when the pump is operating at all flow ranges and high (continued) O Brunswick Unit 1 B 3.3-119 Revision No.
L l ECCS Instrumentaticn i B 3.3.5.1 1 N BASES , APPLICABLE 4.d. 4.e. 5 d. 5.e. Core Spray and RHR (LPCI Mode) Pump l SAFETY ANALYSES, Discharae Pressure-Hiah (continued) LCO, and APPLICABILITY enough'to avoid any condition that results in a discharge pressure permissive when the CS and LPCI pumps are aligned i for injection and the pumps are not running. The actual operating point of this function is not assumed in any transient or accident analysis. Twelve channels of Core Spray and RHR (LPCI Mode) Pump l Discharge Pressure-High Functions are only required to be ! OPERABLE when the ADS is required to be OPERABLE to ensure ! that no single instrument failure can preclude ADS , initiation. Two CS channels associated with CS pump B and ! four LPCI channels associated with RHR pumps B and D are required for trip system A. Two CS channels associated with CS pump A and four LPCI channels associated with RHR pumps A and C are required for trip system B. Refer to LC0 3.5.1 for ADS Applicability Bases. ACTIONS A Note has been provided to modify the ACTIONS related to O ECCS instrumentation channels. Section 1.3, Completion ( Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable ECCS instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable ECCS instrumentation channel. j i 6d Required Action A.1 directs entry into the appropriate ' Condition referenced in Table 3.3.5.1-1. The applicable Condition referenced in the Table is function dependent. Each time a channel is discovered inoperable, Condition A is ' entered for that channel and provides for transfer to the i appropriate subsequent Condition. (continued) Brunswick Unit 1 B 3.3-120 Revision No.
ECCS Instrumentatien L. 8 3.3.5.1 13 V BASES ACTIONS B.1. B.2. and B.3 (continued) Required Actions B.1 and B.2 are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in redundant automatic initiation capability being lost for the feature (s). Required Action B.1 features would be-those that are initiated by functions 1.a. 1.b, 2.a, and 2.b i (e.g., low pressure ECCS). . The Required Action B.2 system would be HPCI. For Required Action B.1, redundant automatic initiation capability is lost if (a) two Function 1.a channels are inoperable and untripped in the same trip system, (b) two Function 2.a channels are inoperable and untripped in the same trip system, (c) two Function 1.b channels are inoperable and untripped in the same system, or (d) two Function 2.b channels are inoperable and untripped in the same trip system. For low pressure ECCS, since each inoperable channel would have Required Action B.1 applied separately (refer to ACTIONS Note), each inoperable channel L would only require the affected portion of the associated system of low pressure ECCS and DGs to be declared inoperable. However, since channels in both associated low pressure ECCS subsystems (e.g., both CS subsystems) are O inoperable and untripped, and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in the associated low pressure ECCS and cgs being concurrently declared inoperable. For Required Action B.2, redundant automatic initiation capability is icst if two Function 3.a or two Function 3.b channels are inoperable and untripped in the same trip system. In this situation (loss of redundant automatic initiation capability), the 24 hour allowance of Required Action B.3 is i not appropriate and the feature (s) associated with the l inoperable, untripped channels must be declared inopera'ae ! within I hour. As noted (Note 1-to Required Action B.1), i' Required Action B.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the low pressure ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of initiation capability for 1 24 hours (as allowed by Required Action B.3) is allowed during MODES 4 and 5. There is no similar Note provided for Required Action B.2 since HPCI instrumentatio'n is not required in MODES 4 and 5; thus, a Note is not necessary. (continued) O Brunswick Unit 1 B 3.3-121 Revision No. l L : L _
I
;i ECCS'Instrumentatien B 3.3.5.1 BASES ACTIONS B.1. B.2. and B.3- (continued)
Notes are also provided (Note 2 to Required Action B.1 and the Note to Required Action B.2) to delineate which Required-Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable. This ensures that the proper loss of initiation capability check is performed. Required Action B.1 (the Required l Action for certain inoperable channels in the low mressure ' ECCS subsystems) is not applicable to Function 2.e', since this Function provides backup to administrative controls , ensuring that operators do not divert LPCI flow from injecting into the core when needed. Thus, a total loss of Function 2.e capability for 24 hours is allowed, since the I LPCI subsystems remain capable of performing their intended i function. { i The Completion Time is intended to allow the operator time i to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal !
" time zero" for beginning the allowed outage time " clock."
For Required Action 8.1, the Completion Time only begins upon discovery that a redundant feature in the same system , (e.g., both CS subsystems) cannot be automatically initiated due to inoperable, untripped channels within the same Function as described in the paragraph above. For Required Action B.2, the Completion Time only begins upon discovery that the HPCI System cannot be automatically initiated due to two inoperable, untripped channels for the associated ' Function in the same trip system. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. , Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 7) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.3. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. (continued) O Brunswick Unit 1 B 3.3-122 Revision No.
i ECCS Instrumentation B 3.3.5.1 p) q BASES ACTIONS B.I. B.2 and B.3 (continued) J 1 Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition G must be entered and its Required Action taken. C.1 and C.2 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the same function result in redundant automatic initiation capability being lost for the feature (s). Required Action C.1 features would be those that are initiated by functions 1.c, l.d, 2.c, 2.d, and 2.f (i.e., low pressure ECCS). Redundant automatic initiation capability is lost if either (a) two Function 1.c channels are inoperable in the l same trip system, (b) two Function 2.c channels are inoperable in the same trip system, (c) two Function 2.d 1 channels are inoperable in the same trip system, or (d) two or more required Function 1.d and 2.f channels associated with low pressure ECCS pumps powered from separate 4.16 kV emergency buses are inoperable. Since each inoperable ( channel would have Required Action C.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated system of low pressure ECCS and DGs to be declared inoperable. However, since channels for both associated low pressure ECCS subsystems are inoperable (e.g., both CS subsystems), and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in the associated low pressure ECCS and DGs being concurrently declared inoperable. For Functions 1.d and 2.f, the affected portions are the associated low pressure ECCS pumps. In this situation (loss of redundant automatic initiation capability), the 24 hour allowance of Required Action C.2 is not appropriate and the feature (s) associated with the inoperable channels must be declared inoperable within I hour. As noted (Note 1 to Required Actions C.1), Required Action C.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the ECCS is not assumed and the probability of a LOCA occurring during (continued) t'~ Brunswick Unit 1 B 3.3-123 Revision No.
ECCS Instrumentatien B 3.3.5.1 V BASES ACTIONS C.1 and C.2 (continued) the period the channels are inoperable is low. Thus, a ' total loss of automatic initiation capability for 24 hours (as allowed by Required Action C.2) is allowed during l MODES 4 and 5. I Note 2 to Required Action C.1 states that it is only applicable for Functions 1.c,1.d, 2.c, 2.d, and 2.f. Required Action C.1 is not applicable to Function 3.c (which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic). This loss was considered during the development of Reference 7 and considered acceptable for the 24 hours allowed by Required Action C.2. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
" time ze.ro" for beginning the allowed outage time " clock."
For Required Action C.1, the Completion Time only begins e upon discovery that the same feature in both subsystems d (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels. Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acce) table (Ref. 7) to permit restoration of any inoperaale channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition G must be entered ! and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would either cause the initiation or it would not necessarily l result in a safe state for the channel in all events. l (continued) O Brunswick Unit 1 B 3.3-124 Revision No.
ECCS Instrumentaticn B 3.3.5.1 13 \g BASES ACTIONS D.I. D.2.1, and 0.2.2 (continued) Required Action 0.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic component initiation capability for the HPCI System. Automatic component initiation capability is lost if two Function 3.d channels or two Function 3.e channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions 0.2.1 and D.2.2 is not appropriate and the HPCI System must be declared inoperable within I hour after discovery of loss of HPCI initiation capability. As noted, Required Action 0.1 is only applicable if the HPCI pump suction is not aligned to the suppression pool, since, if aligned, the function is already performed. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock."
for Required Action D.1, the Completion Time only begins upon discovery that the HPCI System cannot be automatically aligned to the suppression pool due to two inoperable, (mV) untripped channels in the same function as described in the paragraph above. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 7) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1 or the suction source must be aligned to the suppression pool per Required Action D.2.2. Placing the inoperable channel in trip performs the intended function of the channel (shifting the suction source to the suppression pool). Performance of either of these two Required Actions will allow operation to continue. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the HPCI System piping remains filled with water. Alternately, if it is not (continued) Brunswick Unit 1 B 3.3-125 Revision No.
ECCS Instrumentatien B 3.3.5.1
'(X) BASES ACTIONS D.I. D.2.1. and D.2.2 (continued) l desired to perform Required Actions D.2.1 and 0.2.2 (e.g.,
as in the case where shifting the suction source could drain : down the HPCI suction piping), Condition G must be entered and its Required Action taken. ! E.1 and E.2 Required Action E.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within similar ADS trip system A and B Functions ; result in redundant automatic initiation capability being lost for the ADS. Redundant automatic initiation capability is lost if either (a) one Function 4.a channel and one Function 5.a channel are inoperable and untripped, or ! (b) one Function 4.c channel and one Function 5.c channel are inoperable and untripped. 1 In this situation (loss of automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required Action E.2 is not appropriate and all ADS valves must be (q ' declared inoperable within 1 hour after discovery of loss of ADS initiation capability. I l The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock."
For Required Action E.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable, untrippcd channels within similar ADS trip system Functions as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 7) to permit restoration of any inoperable channel to OPERABLE status if both HPCI and RCIC are OPERABLE. If either HPCI or RCIC is inoperable, the time is shortened to 96 hours. If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to (continued) Brunswick Unit 1 B 3.3-126 Revision No.
ECCS Instrumentation ( B 3.3.5.1-b lV BASES ACTIONS L1 and E.2 (continued) 96 hours, the 96 hours begins upon discovery of HPCI or RCIC ' inoperability. However, the total time for an inoperable, untripped channel cannot exceed 8 days. If the status of HPCI or RCIC changes such that the Completion Time changes from 96 hours to 8 dsys, the " time zero" for beginning the 8 day " clock" begins upon discovery of the inoperable, e untripped channel. If the inoperable channel cannot be I restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action E.2. Placing the inoperable channel in trip would conservatively compensate for the l inoperability, restore capability to accommodate a single ' failure, and allow operatien to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition G must be entered and its Required Action taken. F.1 and F.2 Required Action F.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within similar ADS trip system A and B Functions result in redundant automatic initiation capability being lost for the ADS. Redundant automatic initiation capability is lost if either (a) one Function 4.b channel and one Function 5.b channel are incperable, or (b) a combination of Function 4.d, 4.e, 5.d, and 5.e channels are inoperable such that channels associated with both CS pumps and one RHR pump in each LPCI loop are inoperable. In this situation (loss of automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required ! Action F.2 is not appropriate and all ADS valves must be declared inoperable within I hour after discovery of loss of ADS initiation capability. i The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This i Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock."
For Required Action F.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable channels within similar ADS trip t 3 (continued) Brunswick Unit 1 B 3.3-127 Revision No.
L ECCS Instrumentation l B 3.3.5.1: BASES ACTIONS F.1 and F.2' (continued) system Functions as described in the paragraph above. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels. ' Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 7) to permit restoration of any inoperable channel to OPERABLE status if both HPCI and RCIC are OPERABLE (Required Action F.2). If either HPCI or RCIC is inoperable,' the time shortens to 96 hours. If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to 96 hours, the 96 hours begins upon discovery of HPCI or RCIC inoperability. However, the total time for an inoperable channel.cannot exceed 8 days. If the status of HPCI or RCIC changes such that the Completion Time ' changes from 96 hours to 8 days, the " time zero" for beginning the 8 day " clock" begins upon discovery of the inoperable _ channel. If the inoperable channel cannot be. restored to OPERABLE status within the allowable out of O-service time, Condition G must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events. - G.1 With any Required Action and associated Completion Time not met, the associated feature (s) may be incapable of ! performing the intended function, and the supported j feature (s) associated with inoperable untripped channels ; must be declared inoperable immediately. SURVEILLANCE As noted (Note 1) in the beginning of the SRs, the SRs for ' i REQUIREMENTS each ECCS instrumentation function are found in the SRs column of Table 3.3.5.1-1. The Surveillances are modified by a Note (Note 2) to indicate that when a channel is placed in an inoperable status solely for performance of required Surve111ances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours as follows: (a) for Function 3.c; (continued) Brunswick Unit 1 B 3.3-128 Revision No.
1 l ECCS Instrumentatien l B 3.3.5.1 O BASES SURVEILLANCE and (b) for Functions other than 3.c provided the associated REQUIREMENTS Function or redundant Function maintains ECCS initiation (continued) capabiluy. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 7) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary. 1 SR 3.3.5.1.1 Performance of the CHANNEL CHECV, once every 24 hours ensures that a gross failure of instrumentation has not occurred. A ! CHANNEL CHECK is normally a comparison of the parameter ! indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations pd between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited to 24 hours; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CAllBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a cha.nnel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. I SR 3.3.5.1.2 and SR 3. 3. 5.1J l l A CHANNEL FUNCT!0NAL TEST is performed on each required channel to ensw e that the channel will perform the intended (continued) G \ Brunswick Unit 1 B 3.3-129 Revision No. i
l t ECCS Instrumentatien l B 3.3.5.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.5.1.2 and SR 3.3.5.1.6 (continued) b 1 function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.5.1.2 is based on the reliability analyses of Reference 7. The 24 month Frequency of SR 3.3.5.1.6 is based on engineering judgment and the reliability of the components. b SR 3.3.5.1.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than j the Allowable Value specified in Table 3.3.5.1-1. If the j trip setting is discovered to be less conservative than i accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety ; e analyses. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than the setting accounted for in the appropriate setpoint i methodology. The Frequency of 92 days is based on the reliability analysis of Reference 7. SR 3.3.5.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. (continued) 1 b Brunswick Unit 1 B 3.3-130 Revision No.
f_. ECCS Instrumentation B 3.3.5.1 i k BASES SURVEILLANCE SR 3.3.5.1.5 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic and simulated automatic operation for a specific channel. The system functional testing performed in LCO 3.5.1, LCO 3.5.2, LCO 3.8.1, and LCO 3.8.2 overlaps this Surveillance to complete testing of the assumed safety function. The 24 month frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency. REFERF,NCES 1. UFSAR, Section 5.2.
- 2. UFSAR, Section 6.3.
- 3. UFSAR, Chapter 15.
O)
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- 4. 10 CFR 50.36(c)(2)(ii).
- 5. NEDC-31624P, Brunswick Steam Electric Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis (Revision 2), July 1990.
- 6. GE-NE-187-26-1292, Power Uprate Transient Analysis for l Brunswick Steam Electric Plant Units 1 and 2, Revision 1, November 1995.
- 7. NEDC-30936-P-A, BWR Owners' Group Technical ;
Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation), Parts 1 and 2, December 1988. 4 O Brunswick Unit 1 B 3.3-131 Revision No.
i i RCIC System Instrumentaticn i B 3.3.5.2 l
,m
() B 3'3 INSTRUMENTATION B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation BASES i BACKGROUND The purpose of the RCIC System instrumentation is to l initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the i Reactor feedwater System is insufficient or unavailable, such that RCIC System initiation occurs and maintains sufficient reactor water level such that initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps does not occur. A more complete discussion of RCIC System operation is provided in the Bases of LCO 3.5.3, "RCIC System." The RCIC System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Level 2. The variable is monitored by four transmitters that are connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken Os twice logic arrangement. 1 The RCIC test line isolation valve is closed on a RCIC i initiation signal to allow full system flow. l The RCIC System also monitors the water levels in the condensate storage tank (CST) since this is the initial l source of water for RCIC operation. Reactor grade water in the CST is the normal source. Upon receipt of a RCIC initiation signal, the CST suction valve is automatically signaled to open. If the water level in the CST falls below a preselected level, first the RCIC suppression pool suction valves automatically open, and then the RCIC CST suction valve automatically closes. Two level switches are used to detect low water level in the CST. Either switch can cause the suppression pool suction valves to open and the CST suction valve to close (one-out-of-two logic). To prevent losing suction to the pump, the suction valves are interlocked so that one suction path must be open before the other automatically closes. (continued) O () Brunswick Unit 1 B 3.3-132 Revision No.
f i li RCIC System Instrumentaticn B 3.3.5.2 i BASES BACKGROUND The RCIC System provides makeup water to the reactor until (continued) the reactor vessel water level reaches the high water level trip (two-out-of-two logic), at which time the RCIC steam supply valve closes. -The RCIC System restarts if vessel ! level again drops to the low level initiation point l (Level 2). ; i 1 APPLICABLE .The function of the RCIC System to provide makeup ~ coolant to SAFETY ANALYSES, the reactor is used to respond to transient events. -The LCO, and RCIC System is not an Engineered Safety Feature System and APPLICA8ILITY no credit is taken in the safety analyses for RCIC System operation. Based on its contribution to the reduction of ! overall plant risk, however, the system, and therefore its tristrumentation, meets Criterion 4 of Reference 1. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the RCIC System instrumentation is , dependent upon the OPERABILITY of the individual 1 instrumentation channel Functions specified in Table 3.3.5.2-1. Each Function must have a required number i of OPERABLE channels with their setpoints within the specified Allowable Values, where appropriate. The actual ; setpoint is calibrated consistent with applicable setpoint ! methodology assumptions.
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Allowable Values are specified for each RCIC System instrumentation Function specified in Table 3.3.5.2-1. Trip setpoints are specified in the set)oint calculations. The , l setpoints are selected to ensure t1at the trip settings do l not exceed the Allowable Value between CHANNEL CALIBRATIONS. ) Operation with a trip setting less conservative than the . trip setpoint, but within its Allowable Value, is l acceptable. A channel is inoperable if its actual trip i setting is not within its required Allowable Value. Trip ! setpoints are those predetermined values of output at which l an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured out)ut value of the process parameter exceeds the setpoint, tie associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by (continued) Brunswick Unit 1 B 3.3-133 Revision No. ( ..
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l_ RCIC System Instrumentation-B 3.3.5.2 O BAS 150 psig since this is when RCIC is required to be OPERABLE. Refer to LCO 3.5.3 for Applicability Bases for the RCIC System.
The specific Applicable Safety Analyses, LCO, and i Applicability discussions are listed below on a Function by Function basis.
- 1. Reactor Vessel Water level-Low Level 2 Low reactor pressure vessel (RPV) water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the RCIC System is !
, initiated at Level 2 to assist in maintaining water level l above the top of the active fuel. Reactor Vessel Water Level-Low Level 2 signals are i initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. l The Reactor Vessel Water Level--Low Level 2 Allowable Value 1-is set high enough such that for complete loss of feedwater flow, the RCIC System flow with high pressure coolant injection assumed to fail will be sufficient to avoid ! initiation of low pressure ECCS at Level 3. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. (continued) O Brunswick Unit 1 B 3.3-134 Revision No.
i RCIC System Instrumentatien B 3.3.5.2 l v BASES . I l APPLICABLE 1. Reactor Vessel Water level-Low level 2 (continued) SAFETY ANALYSES, LCO, and Four channels of Reactor Vessel Water Level-Low Level 2 APPLICABILITY Function are available and are required to be OPERABLE when : RCIC is required to be OPERABLE to ensure that~no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.
- 2. Reactor Vessel Water level-Hiah High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the high water level signal is used to close the RCIC steam supply valve to prevent overflow into the main steam lines (MSLs).
Reactor Vessel Water Level--High signals for RCIC are initiated from two level transmitters from the narrow range water level measurement instrumentation, which sense the difference between.the pressure due to a constant column of water (reference leg) and the pressure due to the actual G water level (variable leg) in the vessel. (b The Reactor Vessel Water Level-High Allowable Value is high enough to preclude isolating the injection valve of the RCIC during normal operation, yet low enough to trip the RCIC System to prevent reactor vessel overfill. The Allowable i Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. Two channels of Reactor Vessel Want level-High Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases. l
- 3. Condensate Storace Tank level-Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source.
Normally, the suction valve between the RCIC pump and the CST is open and, upon receiving a RCIC initiation signal, water for RCIC injection would be taken from the CST. However, if the water level in the CST falls below a preselected level, first the suppression pool suction valves (continued) U l Brunswick Unit 1 8 3.3-135 Revision No.
l r l i p i l RCIC System Instrumentation
- l. B 3.3.5.2 BASES APPLICABLE 3. Condensate Storaae Tank Level-Low (centinued)
SAFETY ANALYSES, LCO, and automatically open, and then the CST suction valve APPLICABILITY automatically closes. This ensures that an adequate supply of makeup water is available to the RCIC pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes. Two level switches are used to detect low water level in the CST. The Condensate Storage Tank Level-Low Function Allowable Value is set high enough to ensure adequate pump suction head while water is being taken from the CST. Two channels of Condensate Storage Tank Level-Low Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to suppression pool source. , Refer to LC0 3.5.3 for RCIC Applicability Bases. l ACTIONS A Note has been provided to modify the ACTIONS related to l O RCIC System instrumentation channels. Section 1.3, O Completion Times, specifies that once a Condition has been : entered, subsequent divisions, subsystems, components, or - variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System instrumentation channels provide a)propriate compensatory measures for separate inoperable ciannels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel. A.1 i Required Action A.1 directs entry into the appropriate i Condition :eferenced in Table 3.3.5.2-1. The applicable i Condition referenced in the Table is Function dependent. ! Each time a channel is discovered to be inoperable, Condition A is entered for that channel and provides for ' transfer to the appropriate subsequent Condition. l (continued) . O I d i Brunswick Unit 1 B 3.3-136 Revision No. l l I
r RCIC System Instrumentatien B 3.3.5.2 O BASES ACTIONS. B.I and B.2 (continued) Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic initiation capability for the RCIC System. In this case, automatic initiation capability is lost if two Function I channels in the same trip system are inoperable and untripped. In this situation (loss of automatic initiation capability), the 24 hour allowance of Required Action B.2 is not appropriate, and the RCIC System must be declared inoperable within I hour after discovery of loss of RCIC initiation capability. Th~e Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This i Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock."
For Required Action B.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically initiated.due to two inoperable, untripped Reactor Vessel l Water Level-Low Level 2 channels in the hame trip system. The 1 hour Completion Time from discovery of loss of l O initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of 1 channels. Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours has been shown to be j acceptable (Ref. 2) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a : single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition E l must be entered and its Required Action taken. i (continued) .O
' Brunswick Unit 1 B 3.3-137 Revision No.
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RCIC System Instrumentation , B 3.3.5.2 l 'p i BASES ACTIONS .C_d (continued) A risk based analysis was performed and determined that an allowable out of service time of 24 hours (Ref. 2) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.1). A Required Action (similar to Required Action B.1) limiting the allowable out of service time, if a loss of automatic RCIC initiation capability exists, is not required. This Condition applies to the Reactor Vessel Water Level-High Function whose logic i , is arranged such that any inoperable channel will result in l a loss of automatic RCIC initiation capability (loss of high ' water level trip capability). As stated above, this loss of automatic RCIC initiation capability was analyzed and determined to be acceptable. One inoperable channel may . result in a loss of high water level trip capability but ' will not prevent RCIC System automatic start capability. However, the Required Action does not allow placing a channel in trip since this action would not necessarily result in a safe state for the channel in all events (a failure of the remaining channel could prevent a RCIC System start). . iQV 0.1. 0.2.1 and 0.2.2 Required Action 0.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in automatic component initiation capability being lost for the feature (s). For Required Action D.1, the RCIC System is the only associated feature. In this case, automatic initiation capability is lost if two Function 3 channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions D.2.1 and 0.2.2 is not appropriate, and the RCIC System must be declared inoperable within I hour from discovery of loss of RCIC initiation capability. As noted, Required Action D.1 is only applicable if the RCIC pump suction is not aligned to the suppression pool since, if aligned, the Function is already performed. The Completion Time is intended to allow the operator time I to evaluate and repair any discovered inoperabilities. This ' Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock." -
(continued) O Brunswick Unit 1 B 3.3-138 Revision No.
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RCIC System Instrumentaticn B 3.3.5.2 O BASES I ACTIONS Dil. D.2.1. and 0.2.2 (continued) For Required Action D.1, the Completion. Time only begins upon discovery that the RCIC System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The I hour ! Completion Time from discovery of loss of initiation 1 capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not ! assumed in any accident or transient analysis, an allowable j out of service time of 24 hours has been shown to be ! acceptable (Ref. 2) to permit restoration of any inoperable i channel to OPERABLE status. If the inoperable channel , cannot be restored to OPERABLE status within the allowable i out of service time, the channel must be placed in the tripped condition per Required Action 0.2.1, which performs the intended' function of the channel (shifting the suction source to the suppression pool). Alternatively, Required Action D.2.2 allows the manual alignment of the RCIC suction q to the suppression pool, which also performs the intended Q function. If Required Action D.2.1 or D.2.2 is' performed, measures should be taken to ensure that the RCIC System piping remains filled with water. If it is not desired to perform Required Actions 0.2.1 and D.2.2 (e.g., as in the 4 case where shifting the suction source could drain down the RCIC suction piping), Condition E must be entered and its Required Action taken. fal With any Required Action and associated Completion Tima not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately. SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RCIC REQUIREMENTS System instrumentation Function are found in the SRs column of Table 3.3.5.2-1. (continued) O Brunswick Unit 1 B 3.3-139 Revision No.
1 RCIC System Instrumentatien B 3.3.5.2 BASES-SURVEILLANCE The Surve111ances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for (continued) ' performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 2; and (b) for up to 6 hours for Functions 1 and 3, provided the associated-Function maintains RCIC initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.- This Note is based on the reliability analysis (Ref. 2) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the
' 6 hour. testing allowance does not significantly reduce the-probability that the RCIC will initiate when necessary.
SR 3.3.5.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a parameter on other similar O channels. It is based on the assumption that instrument' channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication'of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties,
-including indication and readability. If a channel is outside the criteria, it may be an indication that the ,
instrument has drifted outside its limit.
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The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. (continued) i O Brunswick Unit 1 B 3.3-140 Revision No.
l RCIC System Instrumentation l B 3.3.5.2 I) O BASES SURVEILLANCE SR 3.3.5.2.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 2. SR 3.3.5.2.3 The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.5.2-1. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint (O) must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 2. SR 3.3.5.2d > A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive I calibrations consistent with the plant specific setpoint methodology. 1 l The frequency is based upon the assumption of a 24 month i calibration interval in the determination of the magnitude l of equipment drift in the setpoint analysis. (continued) i j Brunswick Unit 1 B 3.3-141 Revision No. t
r RCIC System Instrumentatien B 3.3.5.2 l b] l BASES SURVEILLANCE SR 3.3.5.2.5 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel and includes simulated automatic actuation of the channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency. REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
- 2. GENE-770-06-2P-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, O December 1992.
l l A
. \g Brunswick Unit 1 B 3.3-142 Revision No.
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r . Pricary Containment Isolatien Instrumentation B 3.3.6.I B 3.3 INSTRUMENTATION B 3.3.6.1 Primary Containment Isolation Instrumentation. BASES l BACKGROUND The primary containment isolation instrumentation automatically initiates closure of appropriate primary containment isolation valves (PCIVs). The function of the PCIVs, in combination with other accident mitigation systems,,is to limit fission product release during and following postulated Design Basis Accidents (DBAs). Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in i the analyses for a DBA. j The isolation instrumentation includes the sensort, relays, ! and switches that are necessary to cause initiation of primary containment and reactor coolant pressure boundary l (RCPB) isolation. Most channels include electronic ! equipment (e.g., trip units) that compares measured input i q signals with pre-established setpoints. When the setpoint l Q is exceeded, the channel output relay actuates, which then outputs a primary containment isolation signal to the i isolation logic. Functional diversity is provided by , monitoring a wide range of independent parameters. The ! input parameters to the isolation logics are-(a) reactor vessel water level, (b) area ambient and differential i temperatures, (c) main steam line (MSL) flow measurement, (d) Standby liquid Control (SLC) System initiation, (e) condenser vacuum, (f) main steam line pressure, (g) high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) steam line flow, (h) drywell pressure, (1) HPCI and RCIC steam line pressure, (j) HPCI and RCIC turbine exhaust diaphragm pressure, (k) reactor water cleanup (RWCU) differential flow, (1) reactor steam dome pressure, (m) main stack radiation, and (n) reactor building exhaust radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation. The exceptions are SLC System initiation and main stack radiation. Primary containment isolation instrumentation has inputs to the trip logic of the isolation functions listed below. (continued) O Brunswick Unit I B 3.3-143 ' Revision No.
\; h l ! Pri::ary Containment Isolatten Instrumentaticn B 3.3.6.1 (q
,/ BASES J BACKGROUND 1. Main Steam Line Isolation (continued)
Most MSL Isolation Functions receive inputs from four channels. The outputs'from these channels are combined in a one-out-of-two taken twice logic to initiate isolation of all main steam isolation valves (MSIVs). The outputs from the same channels are arranged into two two-out-of-two logic trip systems to isolate all MSL drain valves. Each MSL drain line has two isolation valves with one two-out-of-two logic system associated with each valve. The exceptions to this arrangement are the Main Steam Line Flow-High Function and the Main Steam Isolation Valve Pit Temperature-High function. The Main Steam Line Flow-High function uses 16 flow channels, four for each steam line. One channel from each. steam line' inputs to one of the four trip strings. Two trip strings make up each trip system and both trip systems must trip to cause an MSL isolation. Each trip string has four inputs (one per MSL), any one of which will trip the trip string. The trip strings are arranged in a one-out-of-two taken twice logic. This is effectively a one-out-of-eight taken twice logic arrangement to initiate - isolation of the MSIVs. Similarly, the 16 flow channels are O connected into two two-out-of-two logic trip systems (effectively, two one-out-of-four twice logic), with each trip system isolating one of the two MSL drain valves on the associated steam line. The Main Steam Isolation Valve Pit Temperature-High Function consists of the four MSL tunnel temperature monitoring channels that sense temperature in the MSIV pit. 1 Each channel receives input from an individual temperature ; switch. The inputs are arranged in a one-out-of-two taken twice logic to isolate all MSIVs. Similarly, the inputs are arranged in two two-out-of-two logic trip systems, with each trip system required to isolate the two MSL drain valves per drain line. , MSL Isolation Functions isolate the Group 1 valves. (continued) 4 Brunswick Unit 1 B 3.3-144 Revision No.
1 Pri:ary Containment Isolatten Instrumentation B 3.3.6.1 g i I d BASES I BACKGROUND 2. Primary Containment Isolation Primary Containment Isolation Functions associated with Reactor Vessel Water Level-Low Level 1 and Drywell Pressure-High receive inputs from four channels. The outputs from these channels are arranged into one-out-of-two taken twice logics. One trip system initiates isolation of all inboard primary containment isolation valves, while the uther trip system initiates isolation of all outboard primary containment isolation valves. Each logic closes one of the two valves on each penetration, so that operation of either logic isolates the penetration. The Main Stack Radiation-High Function receives input from ; one channel. The output from this channel is provided to l each of two one-out-of-one logic trip systems. Each trip system isolates both valves in the associated penetration. The Reactor Building Radiation-High Function receives input from two channels. The outputs from these channels are i arranged into two one-out-of-one logic trip systems. Each trip system isolates one valve per associated penetration.
/ 3 V Primary Containment Isolation Drywell Pressure-High and Reactor Vessel Water Level-Low Level 1 Functions isolate the Group 2 and 6 valves. The Drywell Pressure-High Function in conjunction with reactor low pressure isolates 4 Group 10 valves. Primary Containment Isolation Main Stack Radiation-High Function isolates the containment purge and vent valves. Reactor Building Exhaust Radiation-High Function isolates the Group 6 valves.
- 3. 4. Hiah Pressure Coolant Iniection System Isolation and Reactor Core Isolation Coolina System Isolation Most functions that isolate HPCI and RCIC receive input from two channels, with each channel in one trip system using a one-out-of-one logic. Each of the two trip systems in each isolation group is connected to one of the two valves on each associated penetration.
The exceptions are the HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High, Steam Supply Line Pressure-Low, and Equipment Area Temperature-High Functions. These i Functions receive inputs from four turbine exhaust diaphragm pressure, four steam supply pressure, and four equipment g q (continued) C! Brunswick Unit 1 B 3.3-145 Revision No. I 1 l
F Primary Containment Isolatten Instrumentaticn 8 3.3.6.1 BASES BACKGROUND 3. 4. Hiah Pressure Coolant In.iection System Isolation and Reactor Core Isolation Coolina System Isolation (continued) area temperature channels for each system. The outputs from the turbine exhaust diaphragm pressure and steam supply b pressure channels are each connected to two two-out-of-two trip systems. The outputs from the equipment area temperature channels are connected to two one-out-of-two d trip systems. In addition, the output from one channel per trip system of the Steam Supply Line Pressure-Low Function coincident with a high drywell pressure signal will initiate isolation of the associated HPCI and RCIC turbine exhaust line vacuum breaker isolation valves. Each trip system isolates one valve per associated penetration. HPCI and RCIC functions isolate the Group 4, 5, 7, and 9 valves.
- 5. Reactor Water Cleanup System Isolation The Reactor Vessel Water Level-Low Level 2 Isolation A function receives input from four reactor vessel water level U channels. The outputs from the reactor vessel water level channels are connected into two two-out-of.-two trip systems.
The Differential Flow-High Function receives input from one channel. The output from this channel is provided to each of two one-out-of-one logic trip systems. The Piping Outside RWCU Rooms Area Temperature-High Function receives input from two channels with each channel in one trip system using a one-out-of-one logic. The Area Temperature-High function receives input from six temperature monitors, three to each trip system. The Area Ventilation Differential Temperature-High Function receives input from six differential temperature monitors, three in each trip system. These are configured so that any one input will trip the associated trip system. Each of the two trip systems is connected to one of the two valves on each RWCU penetration. The SLC System Initiation Function receives input from one channel. The output from this channel is provided to a one-out-of-one logic trip system. The trip , system isolates the RWCU suction outboard isolation valve. l l RWCU Functions isolate the Group 3 valves. l (continued) i Brunswick Unit 1 B 3.3-146 Revision No.
Prinary Containment Isolaticn Instrumentaticn B 3.3.6.1
- l. _
j BASES BACKGROUND 6. Shutdown Coolino System Isolation (continued) The Reactor Vessel Water Level-Low Level 1 Function receives input from four reactor vessel water level channels. The outputs from the reactor vessel water level channels are connected to two one-out-of-two taken twice logic trip systems. The Reactor Vessel Pressure-High Function receives input from two channels, with each channel in one trip system using a one-out-of-one logic. Each of the two trip systems is connected to one of the two valves on each shutdown cooling penetration. Shutdown Cooling System Isolation Functions isolate the Group 8 valves. APPLICABLE The isolation signals generated by the primary containment SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO, and safety analyses of References 1, 2, and 3 to initiate APPLICABILITY closure of valves to limit offsite doses. Refer to LCO 3.6.1.3, " Primary Containment Isolation Valves (PCIVs)," Applicable Safety Analyses Bases for more detail of the safety analyses. V Primary containment isolation instrumentation satisfies Criterion 3 of Reference 4. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the primary containment instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.6.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Each channel must also respond 4 within its assumed response time, where appropriate. Allowable Values are specified for each Primary Containment Isolation Function specified in Table 3.3;6.1-1. Trip i setpoints are specified in the setpoint calculations. The , setpoints are selected to ensure that the trip settings do j not exceed the Allowable Value between CHANNEL CALIBRATIONS. i Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip (continued) ,V 10 Brunswick Unit 1 B 3.3-147 Revision No.
L Pritary Centainment Isolation Instrumentation B 3.3.6.1
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,(d BASES APPLICABLE setting is not within its required Allowable Value. Trip l SAFETY ANALYSES, setpoints are those predetermined values of output at which LCO, and an action should take place. The setpoints are compared to APPLICABILITY the actual process parameter (e.g., reactor vessel water (continued) level), and when the measured out)ut value of the process parameter exceeds the setpoint, tie associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained > from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. Certain Emergency Core Cooling Systems (ECCS) and RCIC valves (e.g., LPCI injection) also serve the dual function of automatic PCIVs. The signals that isolate these valves are also associated with the automatic initiation of the ECCS and RCIC. The instrumentation requirements and ACTIONS associated with these signals are addressed in LCO 3.3.5.1, l
" Emergency Core Cooling Systems (ECCS) Instrumentation," and 1 LCO 3.3.5.2, " Reactor Core Isolation Cooling (RCIC) System Instrumentation," and are not included in this LCO.
In general, the individual Functions are required to be OPERABLE in MODES 1, 2, and 3 consistent with the Applicability for LC0 3.6.1.1, " Primary Containment." Functions that have different Applicabilities are discussed below in the individual Functions discussion. The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis. (continued) D (y/ , Brunswick Ui. 1 B 3.3-148 Revision No.
Primary Centainment Isolatien Instrumentatien B 3.3.6.1
,m U BASES APPLICABLE Main Steam Line Isolation SAFETY ANALYSES, LCO, and 1.a. Reactor Vessel Water Level-Low level 3 APPLICABILITY (continued) Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
Therefore, isolation of the MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits from being exceeded. The Reactor Vessel Water Level-Low Level 3 Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signals. The Reactor Vessel Water Level-Low Level 3 functicn associated with isolation is assumed in the analysis of the recirculation line break (Ref.1). The isolation of the MSLs on Level 3 supports actions to ensure that offsite dose limits are not exceeded for a DBA. Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water
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Level-Low Level 3 function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Reactor Vessel Water Level-Low Level 3 Allowable Value is chosen to be the same as the ECCS Level 3 Allowable Value (LCO 3.3.5.1) to ensure that the MSLs isolate on a potential l loss of coolant accident (LOCA) to prevent offsite doses from exceeding 10 CFR 100 limits. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. This Function isolates the Group 1 valves. 1.b. Main Steam Line Pressure-Low l Low MSL pressure indicates that there may be a problem with . the turbine pressure regulation, which could result in a low ] reactor vessel water level condition and the RPV cooling i down more than 100*F/hr if the pressure loss is allowed to 3 continue. The Main Steam Line Pressure-Low Function is I l directly assumed in the analysis of the pressure regulator (continued) 7x
.( )
Brunswick Unit 1 B 3.3-149 Revision No. l
p-L Prizary Containment Is31atten Instrumentaticn B 3.3.6.1 BASES APPLICABLE 1.b. Main Steam Line Pressure-Low (continued) SAFETY ANALYSES LCO, and ' failure (Ref. 2). For this event, the. closure of the MSIVs APPLICABILITY ensures that no significant thermal stresses are imposed on the RPV.~ In addition, this Function supports actions to ensure that Safety Limit 2.1.1.1 is not exceeded. (This Function closes the MSIVs prior to pressure decreasing below 785 psig, which results in a scram due to MSIV closure, thus reducing reactor power to < 25% RTP.) The MSL low pressure signals are initiated from four transmitters that are connected to the MSL header. The transmitters are arranged such that each transmitter is able to detect low MSL pressure. Four channels of Main Steam Line Pressure-Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value was selected to be far enough below normal turbine inlet pressures to avoid spurious isolations, yet high enough to provide timely detection of a pressure regulator malfunction. The Main Steam Line Pressure-Low Function is only required to be OPERABLE in MODE 1 since this is when the assumed transient can occur (Ref. 2). This Function isolates the Group 1 valves except for sample line isolation valves B32-F019 and B32-F020. l.c. Main Steam Line Flow-Hiah j 1 Main Steam Line Flow-High is provided to detect a break of the MSL and to initiate closure of the MSIVs. If the steam were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover. If the RPV water level decreases too far, fuel damage could occur. Therefore, the isolation is initiated on high flow i' to prevent or minimize core damage. The Main Steam Line Flow-High Function is directly assumed in the analysis of the main steam line break (MSLB) (Ref. 5). The isolation action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak cladding i temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 100 limits. ; (continued) O Brunswick Unit 1 B 3.3-150 Revision No.
I, L E Pricary Centainment Isolation Instrumentati:n B 3.3.6.1 BASES' 1 APPLICABLE 1.c. Main Steam Line Flow-Hiah (continued) SAFETY. ANALYSES, LCO, and The MSL flow signals are initiated from 16 transmitters that
'APPLICA8ILITY are connected to the four MSLs. The transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow. Four channels of Main Steam Line Flow-High function for each unisolated MSL (two channels per trip system) are available and are required to be.
OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL.
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The Allowable Value is chosen to be high enough to permit isolation of one main steam line for test at rated power without causing an automatic isolation of the rest of the steam lines, yet low enough to permit early detection of a gross steam line break.
.This function isolates the Group 1 valves except for sample line isolation valves B32-F019 and B32-F020.
1.d. Condenser Vacuum-tow The Condenser Vacuum-Low function is provided to prevent overpressurization of the main condenser in the event of a loss of the main condenser vacuum. Since the integrity of the condenser is an assumption in offsite dose calculations, the Condenser Vacuum-Low function is assumed to be OPERABLE and capable of initiating closure of the MSIVs. The closure of the MSIVs is initiated to prevent the addition of steam that would lead to additional condenser pressurization and possible rupture, thereby preventing a' potential radiation 4 leakage path following an accident. I Condenser vacuum pressure signals are derived from four pressure transmitters that sense the pressure in the condenser. Four channels of Condenser Vacuum-Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value is chosen to prevent damage to the condenser due to pressurization, thereby ensuring its integrity for offsite dose analysis. As noted (footnote (a) (continued) Brunswick Unit 1 B 3.3-151 Revision No. I
Prirary Containment Isolation Instrumentation l B 3.3.6.1 L lO lb BASES APPLICABLE 1.d. Condenser Vacuum-tow (continued)
- SAFETY ANALYSES, LCO, and to Table 3.3.6.1-1), the channels are not required to be APPLICABILITY OPERABLE its MODES 2 and 3 when all turbine stop valves (TSVs) are closed, since the potential for condenser
-overpressurization is minimized. Therefore, the channels may be bypassed when all TSVs are closed.
This Function isolates the Group 1 valves. 1.e. Main Steam Isolation Valve Pit Temperature-Hioh Main steam isolation valve pit temperature is provided to detect a leak in the RCPB and provides diversity to the high flow instrumentation. The isolation occurs when a very small leak has occurred in the main steam isolation valve pit. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. However, credit for these instruments is not taken in any transient or accident analysis in the UFSAR, since bounding analyses are performed for large breaks, such as MSLBs. Main steam isolation valve pit temperature signals are l(A initiated from temperature switches located in the area being monitored. Four channels of Main Steam Isolation
. Valve Pit Temperature-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The temperature switches are located or shielded so that they are sensitive to air temperature and not in the radiated heat from hot equipment.
The main steam isolation valve pit temperature monitoring Allowable Value is chosen to detect a leak equivalent to between 1% and 10% rated steam flow. This Function isolates the Group 1 valves except for sample line isolation valves B32-F019 and B32-F020. i Primary Containment Isolation 2.a. Reactor Vessel Water Level-low level 1 Low RPV water level indicates that the capability to cool the fuel may be threatened. The valves whose penetrations communicate with the primary containment are isolated to (continued) O.. Brunswick Unit 1 B 3.3-152 Revision No. L
l l Pri ary Containment Isolation Instrumentatien B 3.3.6.1 BASES APPLICABLE 2.a. Reactor Vessel Water Level-tow level 1 (continued) SAFETY ANALYSES, LCO, and limit the release of fission products. The isolation of the APPLICABILITY primary containment on Level 1 supports actions to ensure / that offsite dose limits of 10 CFR 100 are not exceeded. The Reactor Vessel Water Level-tow Level 1 Function associated with isolation is implicitly assumed in the UFSAR analysis as these leakage paths are assumed to be isolated post LOCA.
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Reactor Vessel Water Level-Low Level 1 signals are initiated from four level transmitters that sense the , difference between the pressure due to a constant column of ! water (reference leg) and the pressure due to the actual ! water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Level 1 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation 1 function, j The Reactor Vessel Water Level-Low Level 1 Allowable Value was chosen to be the same as the RPS Level 1 scram Allowable p Value (LCO 3.3.1.1), since isolation of these valves is not d critical to orderly plant shutdown. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. l This function isolates the Group 2 and 6 valves. l 2.b. Drywell Pressure-Hiah High drywell pressure can indicate a break in the RCPB inside the primary containment. The isolation of some of the primary containment isolation valves on high drywell pressure supports actions to ensure that offsite dose limits of 10 CFR 100 are not exceeded. The'Drywell Pressure-High Function, associated with isolation of the primary containment, is implicitly assumed in the UFSAR accident ; analysis as these leakage paths are assumed to be isolated ' post LOCA. High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. i (continued) j O Brunswick Unit 1 B 3.3-153 Revision No.
1 l Pri::ary Containment Isolation' Instrumentation 4 8 3.3.6.1 ( o U BASES APPLICABLE 2.b. Drywell Pressure-Hiah (continued) SAFETY ANALYSES, LCO, and The Allowable Value was selected to be the same as the ECCS APPLICABILITY Drywell Pressure-High Allowable Value (LCO 3.3.5.1), since this may be indicative of a LOCA inside primary containment. This Function isolates the Group 2 and 6 valves. This Function in conjunction with reactor low pressure also isolates Group 10 valves. 2.c. Main Stack Radiation-Hiqh High main stack radiation indicates increased airborne radioactivity levels in primary containment being released through the containment vent valves. Therefore, Main Stack Radiation-High Function initiates an isolation to assure timely closure of valves to protect against substantial releases of radioactive materials to the environment. However, this function is not assumed in any accident or transient analysis in the UFSAR because other leakage paths (e.g., MSIVs) are more limiting. n The main stack radiation signal is initiated from a y) radiation detector that is located in the main stack. The Allowable Value is establits ed in accordance with the methodology in the Offsite Dose Calculation Manual. This Function isolates the containment vent and purge ! valves. 2.d. Reactor Buildina Exhaust Radiation-Hiah High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the RCPB. When Reactor Building Exhaust Radiation-High is detected, valves whose penetrations commur.icate with the primary containment atmosphere are isolated to limit the release of fission products. The Reactor Building Exhaust Radiation-High signals are initiated from radiation detectors that are located on the ventilation exhaust piping coming from the reactor building. The signal from each detector is input to an individual (continued) Brunswick Unit 1 B 3.3-154 Revision No.
Pri;ary Centainment Isolaticn Instrumentati:n B 3.3.6.1 (rh) v BASES APPLICABLE 2.d. Reactor Buildina Exhaust Radiation-Hiah (continued) SAFETY ANALYSES, LCO, and monitor whose trip outputs are assigned to an isolation APPLICABILITY channel. Two channels of Reactor Building Exhaust-High Function are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are chosen to promptly detect gross failure of the fuel cladding. These Functions isolate the Group 6 valves. Hiah Pressure Coolant iniection and Reactor Core Isolation Coolina Systems isolation 3.a. 3.b., 4.a. 4.b. HPCI and RCIC Steam Line Flow-High and Time Delay Relays Steam Line Flow-High functions are provided to detect a break of the RCIC or HPCI steam lines and initiate closure of the steam line isolation valves of the appropriate system. If the steam is allowed to continue flowing out of (s. the break, the reactor will depressurize and the core can uncover. Therefore, the isolations are initiated on high flow to prevent or minimize core damage. The isolation action, along with the scram function of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Specific credit for these Functions is not assumed in any UFSAR accident analyses since the bounding analysis is performed for large breaks such as recirculation and MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding. The HPCI and RCIC Steam Line Flow-High signals are initiated after a short time delay from differential pressure switches (two for HPCI and two for RCIC) that are connected to the system steam lines. Two channels of both HPCI and RCIC Steam Line Flow-High Functions and the associated Time Delay Relays are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The time delay was (continued) Brunswick Unit 1 B 3.3-155 Revision No. I
Primary Containment Isolaticn Instrumentation i 8 3.3.6.1 BASES APPLICABLE 3.a. 3.b. 4.a. 4.b. HPCI and RCIC Steam Line Flow--Hiah-SAFETY ANALYSES, and Time Delay Relays (continued) LCO, and APPLICABILITY selected to prevent spurious isolation of HPCI and RCIC due to transient high steam flow during turbine starts and spurious operation during HPCI and RCIC operation. The Allowable Values are chosen to be low enough to ensure that the trip occurs to prevent fuel damage and maintains the MSLB event as the bounding event. l These Functions isolate the Group 4 and 5 valves, as appropriate. 3.c. 4.c. HPCI and RCIC Steam Supply Line Pressure-Low Low MSL pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue operation of the associated system's turbine and is indicative of a break of the HPCI or RCIC steam lines. These isolations provide a diverse signal to indicate a possible system break. The HPCI and RCIC Steam Supply Line Pressure-Low Functions are provided so that in the event a gross rupture of the HPCI or RCIC steam line occurred N upstream from the high flow sensing location, thus negating the Steam Line Flow-High functions, isolation would be effected on low pressure. The HPCI and RCIC Steam Supply Line Pressure-Low signals are initiated from pressure switches (four for HPCI'and four for RCIC) that are connected to the system steam line. Four channels of both HPCI and RCIC Steam Supply Line Pressure-Low Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are selected to be below the pressure at which the system's turbine can effectively operate. These Functions isolate the Group 4 and 5 valves, as l appropriate. ! (continued) o V Brunswick Unit 1 8 3.3-156 Revision No.
Primary containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 3.d. 4.d. HPCI and RCIC Turbine Exhaust Diaphraam SAFETY ANALYSCS, Pressure-High LCO, and APPLICABILITY High turbine exhaust diaphragm pressure could indicate that (continued) the turbine rotor is not turning, thus allowing reactor pressure to act on the turbine exhaust line. The HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High Functions initiate isolation to prevent overpressurization of the turbine exhaust line. These isolations are for equipment protection and are not assumed in any transient or accident analysis in the UFSAR. These instruments are included in the TS because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations. Therefore, they meet criterion 4 of Reference 4. The HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High signals are initiated from pressure switches (four for HPCI and four for RCIC) that are connected to the area between the rupture diaphragms on each system's turbine exhaust line. Four channels of both HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High Functions are available and are r required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are high enough to prevent isolation of HPCI or RCIC if the associated turbine is operating, yet low enough to effect isolation before the turbine exhaust line is unduly pressurized. These Functions isolate the Group 4 and 5 valves, as appropriate. 3.e. 4.e. Drywell Pressure-Hiah High drywell pressure can indicate a break in the RCPB. The HPCI and RCIC isolation of the turbine exhaust is arovided to prevent communication with the drywell when higi drywell pressure exists. A potential leakage path exists via the turbine exhaust. The isolation is delayed until the system becomes unavailable for injection (i.e., low steam line pressure). The isolation of the HPCI and RCIC turbine exhaust by Drywell Pressure-High is indirectly assumed in the UFSAR accident analysis because the turbine exhaust leakage path is not assumed to contribute to offsite doses. (continued) O V Brunswick Unit 1 B 3.3-157 Revision No.
1 Primary Centainment Isolaticn Instrumentatien B 3.3.6.1 I BASES l APPLICABLE 3.e. 4.e. Drywell Pressure-Hioh (continued) SAFETY ANALYSES, LCO, and High drywell pressure signals are initiated from pressure APPLICABILITY transmitters that sense the pressure in the drywell. Two l channels of both HPCI and RCIC Drywell Pressure-High l Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value was selected to be the same as the ECCS Drywell Pressure-High Allowable Value (LCO 3.3.5.1), since this is indicative of a LOCA inside primary containment. This Function isolates the Group 7 and 9 valves.
- 3. f. , 3.o . 3. h . 3.1. 4. f. , 4.o. 4. h . 4.1. 4. .i . 4. k .
Area and Differential Temperature-Hiah and Time Delav Area and differential temperatures are provided to detect a leak from the associated system steam piping. The isolation occurs to prevent excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier and is diverse to the hp high flow instrumentation. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. These Functions are not assumed in any UFSAR transient or accident analysis, since bounding analyses are i performed for large breaks such as recirculation or MSL l breaks. Area and Differential Temperature-High signals are initiated from thermocouples that are appropriately located to protect the system that is being monitored. Two instruments monitor each area. Two channels for each HPCI and RCIC Area and Differential Temperature-High Function, except for the HPCI and RCIC Equipment Area Temperature-High Function which are required to have four k channels each, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. In addition, a time delay is associated with the RCIC Steam Line Area Temperature-High, the RCIC
- Steam Line Tunnel Ambient Temperature-High, and the RCIC Steam Line Tunnel Differential Temperature-High Functions.
The time delay was selected to eliminate spurious isolations which might occur when switching from normal ventilation to standby ventilation. (continued)
.O (G Brunswick Unit 1 B 3.3-158 Revision No.
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Primary Containment Isolaticn Instrumentaticn B 3.3.6.1 I O V BASES APPLICABLE 3. f. 3.o. 3.h. 3.1. 4.f. , 4.o. 4.h. 4.1. 4.1. 4. k. , SAFETY ANALYSES, Area and Differential Temperature-Hiah and Time Delav LCO, and (continued) APPLICABILITY The Allowable Values are set high enough above anticipated i normal operating levels to avoid spurious isolation, yet low I enough to provide timely detection of a HPCI or RCIC steam line break. ' These Functions isolate the Group 4 and 5 valves, as appropriate. Reactor Water Cleanup System Isolation 5.a. 5.b. Differential Flow-Hiah and Time Delay The high differential flow signal is provided to detect a i' break in the RWCU System. Should the reactor coolant continue to flow out of the break, offsite dose limits may , be exceeded. Therefore, isolation of the RWCU System is initiated when high differential flow is sensed to prevent f] excessive loss of reactor coolant and release of significant Q amounts of radioactive material. A time delay is provided to prevent spurious trips during most RWCU operational transients. This function is not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large breaks such as MSLBs. The high differential flow signals are initiated from transmitters that are connected to the inlet (from the reactor vessel) and outlets (to condenser and feedwater) of the RWCU System. The outputs of the transmitters are compared (in a common summer) and the resulting output is sent to two high flow trip units. If the difference between
~the inlet and outlet flow is too large, each trip unit generates an isolation signal. Two channels of Differential Flow-High Function are available and are required to be OPEP.ABLE to ensure that no single instrument failure ,
downstream of the common summer can preclude the isolation ' function. The Differential Flow-High Allowable Value ensures that a break of the RWCU piping is detected. This function isolates the Group 3 valves. (continued) Brunswick Unit 1 B 3.3-159 Revision No. q l l
Primary Containment Isolatien Instrumentaticn B 3.3.6.1' L !
' BASES APPLICABLE 5.c. 5.d. 5.e. Area. Area Ventilation Differential. and SAFETY ANALYSES,- Pipino Outside RWCU Rooms Area Temperature-Hioh LCO, and
! APPLICABILITY RWCU area, area ventilation differential, and piping outside ! (continued) RWCU area temperatures are provided to detect a leak from the RWCU System. If the small leak continues without isolation, offsite dose limits may be reached. Credit for y these' instruments is not taken in any transient or accident L analysis in the UFSAR, since bounding analyses are performed l l for large breaks such as recirculation or MSL breaks. ! Area-and area ventilation differential temperature signals are initiated from temperature elements that are located in the room that is being monitored. Six thermocouples provide l input to the Area Temperature-High Function-(two per room). Six channels are required to be OPERABLE to ensure that no ; single instrument failure can preclude the isolation i function. l Twelve thermocouples provide input to the Area Ventilation Differential Temperature-High function. The output of , these thermocouples is used to determine the differential 1 l -temperature. Each channel consists of a differential I L temperature instrument that receives inputs from L thermocouples that are located in the inlet and outlet ducts which ventilate the RWCU System rooms for a total of six l available channels (two per room). However, only four channels are required to be OPERABLE. l Temperature signals are initiated from temperature elements , monitoring in the 20'/50' elevation RWCU System general ! piping areas located outside the RWCU System equipment rooms. Two thermocouples provide input to the Piping Outside RWCU Rooms Area Temperature-High Function. Two channels are required.to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Area and Area Ventilation Differential Temperature-High Allowable Values are set low enough-to provide timely
- detection of a break in the RWCU System within the associated room (s).
(continued) p e Brunswick Unit 1 B 3.3-160 Revision No.
Primary Containment Isolaticn Instrumentaticn l B 3.3.6.1 r~ ( BASES l l APPLICABLE 5.c., 5.d. 5.e. Area. Area Ventilation Differential, and SAFETY ANALYSES, Pipina Outside RWCU Rooms Area Temperature-High LCO, and (continued) l APPLICABILITY l The Piping Outside RWCU Rooms Area Temperature-High Function i Allowable Value is set low enough to isolate a design basis high energy line break at any point in the high temperature RWCU System piping located outside of the RWCU System equipment rooms. 1 These Functions isolate the Group 3 valves. 5.f. SLC System initiation The isolation of the RWCU System is required when the SLC System has been initiated to prevent dilution and removal of the boron solution by the RWCU System (Ref. 6). The SLC System initiation signal is initiated from the SLC pump start hand switch signal. l There is no Allowable Value associated with this Function since the channel is mechanically actuated based solely on
- the position of the SLC System initiation switch.
One channel of the SLC System Initiation Function is available and required to be OPERABLE only in MODES 1 and 2, l since these are the only MODES where the reactor can be i critical, and these MODES are consistent with the Applicability for the SLC System (LC0 3.1.7). l As noted (footnote (c) to Table 3.3.6.1-1), this Function is only required to close one of the RWCU isolation valves since the sign,als only provide input into one trip system. 5.a. Reactor Vessel Water level-Low level 2 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some interfaces with the reactor vessel occurs to isolate the potential sources of a break. The isolation of the RWCU System on Level 2 supports actions to ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Vessel Water Level-Low Level 2 Function associated with RWCU isolation is not directly (continued) Brunswick Unit 1 B 3.3-161 Revision No. i 1
Primary Containment Isolatien Instrumentatien B 3.3.6.1 l
=( h
(,) BASES APPLICABLE 5.c. Reactor Vessel Water Level-low Level 2 (continued) SAFETY ANALYSES, LCO, and assumed in the UFSAR safety analyses because the RWCU System APPLICABILITY line break is bounded by breaks of larger systems (recirculation and MSL breaks are more limiting). Reactor Vessel Water Level-Low Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Level 2 Function are-available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Reactor Vessel Water Level-Low Level 2 Allowable Value l was chosen to be the same as the ECCS Reactor Vessel Water Level-Low Level 2 Allowable Value (LC0 3.3.5.1), since the capability to cool the fuel may be threatened. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. O This Function isolates the Group 3 valves. MR Shutdown Coolina System Isolation 6.a. Reactor Steam Dome Pressure-Hiah The Reactor Steam Dome Pressure-High Function is provided to isolate the shutdown cooling portion of the Residual Heat Removal (RHR) System. This interlock is provided only for equipment protection to prevent an intersystem LOCA scenario, and credit for the interlock is not assumed in the accident or transient analysis in the UFSAR. The Reactor Steam Dome Pressure-High signals are initiated from two pressure switches that are connected to different taps on the RPV. Two channels of Reactor Steam Dome Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Function is only required to be OPERABLE in MODES 1, 2, and 3, since these (continued) i l b Brunswick Unit 1 B 3.3-162 Revision No. I
1 Primary Containment Isolation Instrumentation B 3.3.6.1 p ~ Q BASES APPLICABLE 6.a. Reactor Steam Dome Pressure-Hiah (continued) SAFETY ANALYSES, LCO, and are the.only MODES in which the reactor can be pressurized; APPLICABILITY thus, equipment protection is needed. The Allowable Value was chosen to be low enough to protect the system equipment from overpressurization. This Function isolates the Group 8 valves except for the LPCI injection valves Ell-F015A and Ell-F0158. 6.b. Reactor Vessel W1ter level-Low Level 1 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease - too far, fuel damage could result. Therefore, isalation of some reactor vessel interfaces occurs to begin isolating the potential sources of a break. The Reactor Vessel Water . Level--Low Level 1 Function associated with RHR Shutdown I Cooling System isolation is not directly assumed in safety I analyses because a break of the RHR Shutdown Cooling System is bounded by breaks of the recirculation and MSL. The RHR Shutdown Cooling System isolation on Level 1 supports , em actions to ensure that the RPV water level does not drop Q below the top of the active fuel during a vessel draindown event caused by a leak (e.g., pipe break or inadvertent l i valve opening) in the RHR Shutdown Cooling System. ' Reactor Vessel Water Level-Low Level 1 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of ; water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels (two channels per trip system) of the Reactor Vessel Water level-Low Level.1 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. As noted (footnote (d) to Table 3.3.6.1-1), only one channel per trip system (with an isolation signal available to one RHR shutdown cooling pump suction isolation valve) of the Reactor Vessel Water ; Level-Low level 1 Function is required to be OPERABLE in MODES 4 and 5, provided the RHR Shutdown Cooling System ds integrity is maintained. System integrity is maintained provided the piping is intact and no maintenance is being performed that has the potential for draining the reactor vessel through the system. (continued) .O v Brunswick Unit 1 B 3.3-163 Revision No.
t Primary Containment Isolation Instrumentatien B 3.3.6.1 l BASES APPLICABLE 6.b. Reactor Vessel Water level-Low level 1 (continued) SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level-Low Level 1 Allowable Value APPLICABILITY was chosen to be the same as the RPS Reactor Vessel Water Level-Low Level 1 Allowable Value (LC0 3.3.1.1), since the capability to cool the fuel may be threatened. The Allowable Values is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. The Reactor Vessel Water Level-Low Level 1 Function is only required to be OPERABLE in MODES 3, 4, and 5 to prevent this potential flow path from lowering the reactor vessel level to the top of the fuel. In MODES 1 and 2, another isolation (i.e., Reactor Steam Dome Pressure-High) and administrative controls ensure that this flow path remains isolated to prevent unexpected loss of inventory via this flow path. This Function isolates the Group 8 valves. ACTIONS A Note has been provided to modify the ACTIONS related to c primary containment isolation instrumentation channels. i Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, , subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also 9ecifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable primary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable primary 1 containment isolation instrumentation channel. ! IL l_ Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Functions 2.a, 2.b, and 6.b and 24 hours for functions other than functions 2.a 2.b, and 6.b has been shown to be g l (continued) O Brunswick Unit 1 B 3.3-164 Revision No.
l Primary Containment Isolation Instrumentation { B 3.3.6.1 l p ( . BASES ACTIONS A '.1 (continued) acceptable (Refs. 7 and 8) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). If the inoperable. channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.I. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an isolation), Condition C must be entered and its Required Action taken. B.1 Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped r O channels within the same function result in redundant automatic isolation capability being lost for the associated penetration flow path (s) The MSL Isolation Functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that both trip systems will generate a trip signal from the given function on a valid signal. The other isolation functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given Fut,alon on a valid signal. This ensures that one of the two PCIVs in the associated penetration flow path can receive an isolation signal from the given Function. For Functions 1.a. 1.b, l.d, and 1.e, this would require both trip systems to have a total of three channels OPERABLE or in trip. For Fenctions 2.a, 2.b, and 6.b, this would require both trip systems to have one channel OPERABLE or in trip. For Function 1.c, this would require both trip systems to have a total of three channels, associated with each MSL, OPERABLE or in trip. For functions 3.c, 3.d, 4.c, 4.d, and 5.g, this would require one trip system to have two channels, each OPERABLE or in trip. For Functions 2.c, 2.d,
- 3. a , 3.b, 3.e , 3. f, 3.g , 3. h, 3.1, 4. a, 4.b, 4.e, 4. f, 4.g ,
4.h, 4.1, 4.j, 4.k, 5.a, 5.b, 5.e, 5.f, and 6.a, this would (continued) Brunswick Unit 1 B 3.3-165 Revision No. i
y
)
r Primary Containment Isolation I'nstrumentatien B 3.3.6.1
- BASES ACTIONS D d (continued) require one trip system to have one channel OPERABLE or in trip. -For Functions 5.c and 5.d, each Function consists of
- channels that monitor severalL different locations.
Therefore, this would. require one channel per location to be i OPERABLE or in trip (the channels are not required to be in l the-same trip system). The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The I hour Completion Time is -acceptable because it minimizes risk while allowing time for restoration or tripping of channels. l C.1 Required Action C.1 directs entry into the appropriate l Condition referenced in Table 3.3.6.1-1. The applicable Condition specified in Table 3.3.6.1-1 is function and MODE or other specified condition dependent-and may change as the Required Action of a previous Condition is completed. Each
-time an inoperable channel has not met any Required Action of Condition A or B and the associated Completion -Time has expired, Condition C will be entered for that channel and provides for transfer to the appropriate subsequent Condition.
l I l D.l. D.2.1 and D.2.2 i If the channel is not restored to OPERABLE status or placed in. trip within the allowed Completion Time, the plant must i be placed in a MODE or other specified condition in which' the LCO does not apply. This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours (Required Actions D.2.1 and 0.2.2). Alternately, the associated MSLs may be isolated (Required Action D.1), ., and, if allowed (i.e., plant safety analysis allows' i operation with an MSL. isolated), operation with that MSL isolated may continue. Isolating the affected MSL accomplishes the safety function of the inoperable channel. ! The Completion Times are reasonable, based on operating - ' experience, to reach the required plant conditions from full power conditions in an orderly manner and without i challenging plant systems. (continued)
\
Brunswick Unit 1 B 3.3-166 Revision No. i.
i Primary Containment Isolation Instrumentation i B 3.3.6.1 !
,y lC BASES 1 ACTIONS L1 (continued)
If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LC0 does not apply. This is done by placing the plant in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems. F.1 , If the channel is not restored to OPERABLE statu; or placed in trip within the allowed Completion Time, plant operations ' may continue if the affected penetration flow path (s) is isolated. Isolating the affected penetration. flow path (s) accomplishes the safety function of the inoperable channels. rs For the RWCU Area and Area Ventilation Differential t Temperature-High functions, the affected penetration flow path (s) may be considered isolated by isolating only that portion of the system in the associated room monitored by the inoperable channel. That is, if the RWCU pump room A area channel is inoperable, the pump room A area can be isolated while allowing continued RWCU operation utilizing the B RWCU pump. g Alternately, if it is not desired to isolate the affected penetration flow path (s) (e.g., as in the case where { isolating the penetration flow path (s) could result in a l reactor scram), Condition G must be entered and its Required Actions taken, l The I hour Completion Time is acceptable because it > minimizes risk while allowing sufficient time for plant operations personnel to isolate the affected penetration flow path (s). (continued) L.) Brunswick Unit 1 B 3.3-167 Revision No.
i Primary Containment Isolation Instrumentation
'B 3.3.6.1 BASES ACTIONS G.1 and G.2 (continued)
If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, or the Required Action of Condition F is not met and the associated , Completion Time has expired, the plant must be placed in a i MODE or other specified condition in which the LCO does not apply. This is done by' placing the plant in at least MODE 3 . within 12 hours and in MODE 4 within 36 hours. The allowed l Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full , power conditions in an orderly manner and without ! challenging plant systems. i l H.1 and H.2 l If the channel is not restored to OPERABLE status or placed ! in trip within the allowed Completion Time, the associated SLC subsystem (s) is declared inoperable or the RWCU System is isolated. Since this Function is required to ensure that the SLC System performs its intended function, sufficient fg remedial measures are provided by declaring the associated Q SLC subsystems inoperable or isolating the RWCU System. I The I hour Completion Time is acceptable because it i minimizes risk while allowing sufficient time for personnel ito isolate the RWCU System. 1.1 and I.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated penetration flow psth should be closed. However, if the shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay heat removal capabilities so the penetration flow path can be isolated). Actions must continue until the channel is restored to OPERABLE status or the RHR Shutdown Cooling System is isolated. (continued) Brunswick Unit 1 B 3.3-168 Revision No.
Primary Containment Isolation Instrumentation B 3.3.6.1 l g (d BASES (continued). SURVEILLANCE As noted at the beginning of the SRs, the SRs for each' REQUIREMENTS Primary Containment Isrlation instrumentation Function are found in the SRs column of Table 3.3.6.1-1. The Surveillances are~ modified by a Note to indicate that u when a channel'is placed in an inoperable status. solely for ! performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Functions with a design that provides only one channel per trip system and (b) for up to 6 hours for all other Functions provided the associated Function maintains trip capability. Upon completion of the - Surveillance, or expiration of the 2 hour allowance for Functions with a design that provides only one channel ~per trip system or the 6 hour allowance for all other Functions, the channel must be returned to OPERABLE-status or the l applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 7 and 8) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the i 2 and 6 hour testing allowances do not significantly reduce the probability that the PCIVs will isolate the penetration flow path (s) when necessary. O SR 3.3.6.1.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channel s. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the i instrument has' drifted outside its limit. (continued) O V Brunswick Unit 1. B 3.3-169 Revision No. L'
i Primary Containment Isolation Instrumentation B 3.3.6.1 7'N h BASES SURVEILLANCE SR 3.3.6.1.1 (continued) REQUIREMENTS The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.6.1.2. SR 3.3.6.1.5 and SR 3.3.6.1.9 d A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint-methodology. The 92 day Frequency of SR 3.3.6.1.2 is based on the reliability analysis described in References 7 and 8. The 184 day Frequency of SR 3.3.6.1.5 and the 24 month Frequency /\ of SR 3.3.6.1.9 are based on engineering judgment and the L6 reliability of the components. O SR 3.3.6.1.3 ; Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety _ analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than that accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability antlysis of References 7 and 8. SR 3.3.6.1.4 and SR 3.3.6.1.6 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary j . (continued) O' LJ Brunswick Unit 1 B 3.3-170 Revision No.
L y L Primary Containment Isolaticn Instrumentation
- j. B 3.3.6.1
, . BASES l SURVEILLANCE- SR' 3.3.6.1.4 and SR 3.3.6.1.6 (continued)
' REQUIREMENTS . l l range and accuracy. CHANNEL CALIBRATION leaves the. channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint
methodology. The Frequency of SR 3.3.6.1.4 is based on the assumption of 1 a 92 day calibration interval in the determination of the magnitude of-equipment drift in the setpoint _ analysis. The Frequency of SR 3.3.6.1.6 is based on the assumption of a ~j 24 month calibration interval in the determination of the i magnitude of equipment drift in the setpoint analysis. ! SR 3.3.6.1.7 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel and includes simulated automatic operation of the channel. The system functional _ testing performed on PCIVs in LC0 3.6.1.3 overlaps this Surveillance to provide complete testing of the assumed safety function. The O 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an un)lanned transient if the Surveillance were performed with tie reactor at power.
- Operating experience has demonstrated these components will l usually pass the Surveillance when performed at the 24 month Frequency.
SR 3.3.6.1.8 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. Testing is performed only on channels where the. assumed response time does not correspond to the i diesel generator (DG) start time. For channels assumed to respond within the DG start time, sufficient margin exists in the 10 second start time when compared to the typical channel response time (milliseconds) so as to assure l adequate response without a specific measurement test i (Ref. 9). i (continued) y l
-Brunswick Unit 1 B 3.3-171 Revision No. l l.
i Primary Containment Isolation.Instrumentatien B 3.3.6.1 ,\ O BASES l SURVEILLANCE SR 3.3.6.1.8 (continued) REQUIREMENTS Note 1 to the Surveillance states that the radiation i
. detectors are excluded from ISOLATION INSTRUMENTATION RESPONSE TIME testing. This Note is necessary because of the difficulty of generating an appropriate detector input i signal and because the principles of detector operation !
virtually ensure an instantaneous response tine. Response , times for radiation detector channels shall be measured from 1 detector output or the input of the first electronic ! component in the channel. In addition, Note 2 to the l Surveillance states that the response time of the sensors ! for functions 1.a and 1.c may be assumed to be the design sensor response time and therefore, are excluded from the ISOLATION INSTRUMENTATION RESPONSE TIME testing. This is allowed since the sensor response time is a small part of the overall ISOLATION INSTRUMENTATION RESPONSE TIME (Ref. 9). j ISOLATION INSTRUMENTATION RESPONSE TIME tests are conducted ! on a 24 month STAGGERED TEST BASIS. The 24 month Frequency. l is consistent with the BNP refueling cycle and is based upon /~ ' plant operating experience that shows that random failures ( .of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. i REFERENCES 1. UFSAR, Section 6.3.
- 2. UFSAR, Chapter 15.
- 3. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
- 4. 10 CFR 50.36(c)(2)(ii).
- 5. UFSAR, Section 6.2.4.3. ;
- 6. UFSAR, Section 7.3.1.1.6.18.
- 7. NEDC-31677P-A, Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation, July 1990.
(continued) O Brunswick Unit 1 B 3.3-172 Revision No.
Primary Containment Isolatien Instrumentation B 3.3.6.1
- V- BASES REFERENCES 8. NEDC-30851P-A Supplement 2, Technical Specifications (continued) . Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
- 9. NED0-32291-A, System Analyses for Elimination of Selected Response Time Requirements, October 1995.
l l O l-I l l l , Brunswick Unit 1 B 3.3-173 Revision No. ; I
p 1 Secondary Containment Isolation Instrumentatien B 3.3.6.2 4 l U B 3.3 INSTRUMENTATION B 3.3.6.2 Secondary Containment Isolation Instrumentation l BASES BACKGROUND The secondary containment isolation instrumentation automatically initiates closure of appropriate secondary containment isolation dampers (SCIDs) and starts the Standby Gas Treatment (SGT) System. The function of these systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refs.1, 2, and 3). Secondary containment isolation and establishment of vacuum with the SGT System ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be OPERABLE are maintained within applicable limits. The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of secondary containment isolation. Most channels include (/',) electronic equipment (e.g., trip. units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a secondary containment isolation signal to the isolation logic. Functional diversity is provided by ] monitoring a wide range of independent parameters. The j input parameters .to the isolation logic are (1) reactor j vessel water level, (2) drywell pressure, and (3) reactor building exhaust radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation. The outputs of the channels associated with the Reactor Vessel Water Level-Low Level 2 Function and the Drywell l Pressure-High function are arranged in one-out-of-two taken j twice trip system logics. One trip system initiates isolation of one automatic secondary containment isolation damper in each penetration and starts both SGT subsystems while the other trip system initiates isolation of the other ' automatic secondary containment isolation damper in each penetration and starts both SGT subsystems. Each logic l closes one of the two dampers in each penetration and starts ; l both SGT subsystems, so that operation of either logic isolates the secondary containment and provides for the , necessary filtration of fission products. < p (continued) d Brunswick Unit 1 B 3.3-174 Revision No. .
Secondary Containment Isolation Instrumentatien B 3.3.6.2 0 \
-BASES BACKGROUND The outputs of the channels associated with the Reactor (continued) Building Exhaust Radiation-High function are arranged in two one-out-of-one trip system logics. Each trip system initiates isolation of one automatic secondary containment isolation damper in each penetration and starts both SGT subsystems while the other trip system initiates isolation of the other secondary containment isolation damper in each penetration and starts both SGT subsystems. Each logic closes one of the two dampers in each penetration and starts both SGT subsystems, so that operation of either logic isolates the secondary containment and provides for the necessary filtration of fission products.
APPLICABLE The isolation signals generated by the secondary containment SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO, and safety analyses of References 1, 2, and 3 to initiate APPLICABILITY closure of dampers and start the SGT System to limit offsite doses. Refer to LC0 3.6.4.2', " Secondary Containment Isolation Dampers (SCIDs)," and LC0 3.6.4.3, " Standby Gas Treatment n (SGT) System," Applicable Safety Analyses Bases for more detail of the safety analyses. The secondary containment isolation instrumentation satisfies Criterion 3 of Reference 4. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set within the specified Allowable Values, as shown in Table 3.3.6.2-1. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Allowable Values are specified for each function specified in the Table. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less (continued) CN Brunswick Unit 1 B 3.3-175 Revision No.
f l l Secondary Containment Isolation Instrumentation B 3.3.6.2 } O (V BASES l APPLICABLE conservative than the trip setpoint, but within its ! SAFETY ANALYSES, Allowable Value, is acceptable. A channel is inoperable if l LCO, and its actual trip setting is not within its required Allowable APPLICABILITY Value. (continued) Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation m uncertainties, process effects, calibration tolerances, I ' instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions when SCIDs and the SGT System are required. The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.
- 1. Reactor Vessel Water level-low Level 2 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
An isolation of the secondary containment and actuation of , the SGT System are initiated in order to minimize the potential of an offsite dose release. The Reactor Vessel (continued) v Brunswick Unit 1 B 3.3-176 Revision No.
Secondary Containment Isolation Instrumentation B 3.3.6.2 ( BASES APPLICABLE 1. Reactor Vessel Water Level-Low level 2 (continued) SAFETY ANALYSES, LCO, and Water f.evel-Low Level 2 Function is one of the Functions APPLICABILITY assumed to be OPERABLE and capable.of providing isolation and initiation signals. The isolation and initiation systems on Reactor Vessel Water Level-Low Level 2 support actions to ensure that any offsite releases are within the limits calculated in the safety analysis. Reactor Vessel Water Level-Low Level 2 signals are initiated from level transmitters that sense the difference i between the pressure due to a constant column of water ' (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Level 2 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. l l The Reactor Vessel Water Level-Low level 2 Allowable Value l was chosen to be the same at the High Pressure Coolant l Injection / Reactor Core Isolation Cooling (HPCI/RCIC) Reactor O Vessel Water Level-Low Level 2 Allowable Value (LC0 3.3.5.1 , V and LCO 3.3.5.2), since this could indicate that the capability to cool the fuel is being threatened. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. The Reactor Vessel Water Level-Low Level 2 Function is required to be OPERABLE in MODES 1, 2, and 3 where ccrs'derable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, this Function is not required.
- 2. Drywell Pressure-Hiah High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite (continued)
O yr Brunswick Unit 1 B 3.3-177 Revision No.
L I l I Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES-APPLICABLE 2. Drywell Pressure-Hiah (continued) SAFETY ANALYSES, LCO, and dose release. The isolation on high drywell pressure APPLICABILITY supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis. However, the Drywell Pressure-High Function associated with isolation is not assumed in any UFSAR accident or transient analyses. It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis. High drywell pressure signals are initiated fron, pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function. The Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Function Allowable Value (LCO 3.3.5.1) since this is indicative of a loss of coolant A accident (LCCA). The Drywell Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.
- 3. Reactor Buildina Exhaust Radiation-Hiah l
High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the RCPB or the refueling floor due to a fuel handling accident. When Reactor Building Exhaust Radiation-High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission products as assumed in the UFSAR safety analyses (Ref. 2). (continued) G U Brunswick Unit 1 B 3.3-178 Revision No.
Secondary Containment Isolation Instrumentatien B 3.3.6.2 (3 ' V BASES APPLICABLE 3. Reactor Buildina Exhaust Radiation-Hiah (continued) SAFETY ANALYSES, , LCO, and The Reactor Building Exhaust Radiation-High signals are l APPLICABILITY initiated from radiation detectors that are located in the I ventilation exhaust ductwork plenum coming from the reactor I building. The signal from each detector is input to an I individual monitor whose trip outputs are assigned to an isolation channel. Two channels of Reactor Building Exhaust Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are chosen to promptly detect gross failure of the fuel cladding. The Reactor Building Exhaust Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, the Function is not required. In addition, the Function is also sj required to be OPERABLE during C0P.E ALTERATIONS, OPDRVs, and movement of irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded. ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the l Condition continue to apply for each additional failure, I with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel. (continued) i ( ) Brunswick Unit 1 B 3.3-179 Revision No. l
Secondary Containment Isolation Instrumentation B 3.3.6.2 V BASES ACTIONS M (continued) Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for function 2, and 24 hours for functions other than Function 2, has been shown to be acceptable (Refs. 5 and 6) . to permit restoration of any inoperable channel to OPERABLE i status. This out of service time is only acceptable l provided the associated Function is still maintaining l isolation capability (refer to Required Action B.1 Bases). l If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.I. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore i capability to accommodate a single failure, and allow l operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where ; placing the inoperable channel in trip would result in an i isolation), Condition C must be entered and its Required l Actions taken. ' M Required Action B.1 is intended to ensure that appropriate I actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic isolation capability for the associated penetration flow path (s) or a complete loss of automatic initiation capability for the SGT System. A function is considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given Function on a valid signal. This ensures that one of the two SCIDs in the associated penetration flow path and both SGT subsystems can be initiated on an isolation signal from the given Function. For the functions with two one-out-of-two logic trip systems (Functions 1 and 2), this would require one trip system to have one channel OPERABLE or in trip. An inoperable channel need not be placed in the tripped condition where this would cause the trip Function to occur. In these cases, if the inoperable channel is not restored within the required Completion Time, Condition C shall be entered. (continued) Brunswick Unit 1 B 3.3-180 Revision No.
1 Secondary Containment Isolation Instrumentation l B 3.3.6.2 BASES I i ACTIONS L1 (continued)
]
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of , l channels. { C.I.l. C.1.2. C.2.1. and C.2.2 If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate -the j secondary containment and start the SGT System cannot be ; ensured. Therefore, further actions.must be performed to - ensure the ability to maintain the secondary containment function.- Isolating the associated dampers and starting the associated SGT subsystam (Required Actions C.1.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue. Alternately, declaring the associated SCIDs or SGT subsystem (s) inoperable (Required Actions C.1.2 and C.2.2) O is also acceptable since the Required Actions of the respective LCOs (LC0 3.6.4.2 and LCO 3.6.4.3) provide appropriate actions for the inoperable components. One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems. SURVEILLANCE As noted at the beginning of'the SRs, the SRs for each
' REQUIREMENTS Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1.
The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Function 3 and (b) for up to 6 hours for functions 1 and 2 provided the associated Function maintains isolation capability. Upon completion of ; the Surveillance, or expiration of the 2 hour allowance for ' (continued) O Brunswick Unit 1 B 3.3-181 Revision No. 1 o 1
Secondary Containment Isolation Instrumentatien B 3.3.6.2 V BASES SURVEILLANCE Function 3 or the 6 hour allowance for Functions 1 and 2, REQUIREMENTS the channel must be returned to OPERABLE status or the (continued) applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 5 and 6) assumption of the average time required to perform channel surveillance. That analysis demonstrated the 2 hour testing allowance for Function 3 and the 6 hour testing allowance for functions 1 and 2 do not significantly reduce the probability that the SCIDs will isolate the associated penetration flow paths and that the SGT System will initiate when necessary. SR 3.3.6.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read A approximately the same value. Significant deviations between the instrument channels could be an indication of d excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. 3 Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO. (continued) O V Brunswick Unit 1 B 3.3-182 Revision No.
i Secondary Containment Isolation Instrumentation I B 3.3.6.2 O O BASES l SURVEILLANCE SR 3.3.6.2.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The frequency of 92 days is based on the reliability analysis of References 5 and 6. SR 3.3.6.2.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.2-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, performance is still within the requirements of the plant safety analysis. Under n these conditions, the setpoint must be readjusted to be Q equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability ) analysis of References 5 and 6. I SR 3.3.6.2.4 4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive ; calibrations consistent with the plant specific setpoint methodology. The frequency is based on the assumption of a 24 month , calibration interval in the determination of the magnitude ' of equipment drift in the setpoint analysis. (continued) O Brunswick Unit I B 3.3-183 Revision No.
l l Secondary Containment Isolation Instrumentation l B 3.3.6.2 l f i V BASES SURVEILLANCE- SR 3.3.6.2.5 i REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed on SCIDs and the SGT System in LC0 3.6.4.2 and LC0 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function. The 24 month Frequency is based on the need to perform this ) Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency. REFERENCES 1. UFSAR, Section 15.6.4.
- 2. UFSAR, Section 15.7.1.
I 3. NEDC-32466P, Power Uprate Safety Analysis Report for l Brunswick Steam Electric Plant Units 1 and 2, September 1995. I
- 4. 10 CFR 50.36(c)(2)(ii).
- 5. NEDC-31677P-A, Technical Specification Improvement !
Analysis for BWR Isolation Actuation Instrumentation, : July 1990. .
- 6. NEDC-30851P-A Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
I Brunswick Unit 1 B 3.3-184 Revision No.
l l l CREV System Instrumentatien B 3.3.7.1 /~T d 'B 3.3 INSTRUMENTATION B 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation BASES BACKGROUND The CREV System is designed to provide a radiologically ; controlled environment to ensure the habitability of the i control room for the safety of control room operators under all plant conditions. Two independent CREV subsystems are each capable of fulfilling the stated safety function. The instrumentation and controls for the CREV System l automatically initiate a*ction to pressurize the main control ' room (MCR) to minimize the consequences of radioactive material in the control room environment. In the event of a Control Building Air Intake Radiation-High signal, the CREV System is automatically started in the radiation / smoke protection mode. Air is then recirculated through the charcoal filter, and sufficient outside air is drawn in through the normal intake to maintain the MCR slightly pressurized with respect to outside atmosphere. The CREV System instrumentation has two tria systems, either j of which can initiate the CREV System. Eac1 trip system ' receives input from the two Control Building Air Intake Radiation-High Function channels. The Control Building Air g Intake Radiation-High Function is arranged in a ' one-out-of-two logic for each trip system. The channels % include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel ; output relay actuates, which then outputs a CREV System initiation signal to the initiation logic. ; l APPLICABLE The ability of the CREV System to maintain the habitability SAFETY ANALYSES, of the MCR is explicitly assumed for the design basis LCO, and accident as discussed in the UFSAR safety analyses (Ref.1). APPLICABILITY CREV System operation ensures that the radiation exposure of control room personnel, through the duration of any one of the postulated accidents, does not exceed the limits set by GDC 19 of 10 CFR 50, Appendix A. i CREV System instrumentation satisfies Criterion 3 of Reference 2. (continued) O Brunswick Unit 1 B 3.3-185 Revision No.
1 CREV Systea Instrumentatien j B 3.3.7.1 i () BASES < APPLICABLE The OPERABILITY of the CREV System instrumentation is SAFETY ANALYSES, dependent upon the OPERABILITY of the Control Building Air l LCO, and Intake Radiation-High instrumentation channel Function. APPLICABILITY The Function must have a required number of OPERABLE (continued) channels, with their setpoints within the specified Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Allowable Values are specified for Control Building Air Intake Radiation-High function. Trip setpoints are specified in the set)oint calculations. The setpoints are selected to ensure t1at the trip settings do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., control building air intake radiation), and when the measured output value of the process pa :: meter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process (OV) parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. 1 The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument urift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe i environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. The control building air intake radiation monitors measure radiation levels in the control building air intake plenum. , A high radiation level may pose a threat to MCR personnel; thus, automatically initiating the CREV System. (continued) O Brunswick Unit 1 B 3.3-186 Revision No. l
CREV System Instrumentaticn B 3.3.7.1 L BASES APPLICABLE The Control Building Air Intake Radiation-High Function SAFETY ANALYSES, consists of two independent monitors. Two channels per trip A LCO, and system of Control Building Air Intake Radiation-High are ab l -- APPLICABILITY available and are required to be OPERABLE to ensure that no } (continued) single instrument failure can preclude CREV System l initiation. The Allowable Value was selected to ensure j .. protection of the control room personnel. The Control Building Air Intake Radiation-High Function is i required to be OPERABLE in MODES 1, 2, and 3 and during CORE r ALTERATIONS, OPDRVs, and movement of irradiated fuel assemblies in the secondary containment, to ensure that l- control room personnel are protected during a LOCA, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress L' (e.g., CORE ALTERATIONS), the probability of a LOCA, main steam line break accident, control rod drop accident, or fuel damage is low; thus, the Function is not required. l I ACTIONS A Note has been provided to modify the ACTIONS related to l CREV System. instrumentation channels. Section 1.3, j Completion Times, specifies that once a Condition has been ,(' entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that l Recuired Actions of the Condition continue to apply for each l adcitional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoper41e CREV System instrumentation channels provide appropmte compensatory measures for separate inoperable l channels. As such, a Note has been provided that. allows separate Condition entry for each inoperable CREV System i instrumentation channel. L Ad b E Because of the redundancy of sensors available to provide initiation signals and the redundancy of the CREV System , design,'an allowable out of service time of 7 days is ! provided to permit restoration of any inoperable channel to l' OPERABLE status. This out of service time is only acceptable provided the Control Building Air Intake b Radiation-High Function is still maintaining CREV System initiation capability (refer to Required Action B.1 Bases). b (continued) Brunswick Unit I B 3.3-187 Revision No.
L CREV System Instrumentaticn , B 3.3.7.1 l ly U BASES ACTIONS M (continued) g If the Function is not maintaining CREV System initiation capability, Condition B must be entered. g If the inoperable channel cannot be restored to OPERABLE status within the 7 day allowable out of service time, one CREV subsystem must be placed in the radiation / smoke protection mode of operation per Required Action A.1. The method used to place the CREV subsystem in operation must provide for automatically re-initiating the subsystem upon restoration of power following a loss of power to the CREV subsystem. Placing one CREV subsystem in the radiation / smoke protection mode of operation provides a suitable compensatory action to ensure that the automatic radiation protection function of the CREV System is not lost. U b Required Action B.1 is intended to ensure that appropriate
, action is taken if multiple, inoperable, untripped channels
( result in the Control Building Air Intake Radiation-High Function not maintaining CREV System initiation capability. The Function is considered to be maintaining CREV System initiation capability when sufficient channels are OPERABLE or in trip such that one trip system will generate an initiation signal for one CREV subsystem from the Function on a valid signal. For the Control Building Air Intake Radiation-High Function, this would require one trip system j to have one channel OPERABLE or in trip. With CREV System i initiation capability not maintained, one CREV subsystem l must be placed in the radiation / smoke protection mode of i operation per Required Action B.1 to ensure that control room personnel will be protected in the event of a Design Basis Accident. The method used to place the CREV subsystem in operation must provide for automatically re-initiating
!b the subsystem upon restoration of power following a loss of power to the CREV subsystem. $l The 1 hour Completion Time is intended to allow the operator time to place the CREV subsystem in operation. The 1 hour $,
Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels, or for placing one CREV subsystem in operation. g (continued) l(Q_/ Brunswick Unit 1 B 3.3-188 Revision No.
CREV System Instrumentatien B 3.3.7.1 o
.V BASES (continued)
SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS . when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains CREV System initiation capability. Upon completion of the l Surveillance, or expiration of the 6 hour allowance, the
' channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
l This Note is based on the reliability analysis (Ref. 3) l assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the CREV System will initiate when necessary. l i SR 3.3.7.1.1 l Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other ,I channels. It is based on the assumption that instrument channels monitoring the same parameter should read , approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the
- instrument has drifted outside its limit.
The Frequency is based upon operating experience that i demonstrates channel failure is rare. The CHANNEL CHECK l supplements less formal, but more frequent, checks of l channels during normal operational use of the displays associated with channels required by the LCO. l (continued)
)
Brunswick Unit 1 B 3.3-189 Revision No. ! i
I l CREV System Instrumentatien 1 j B 3.3.7.1 ('N O BASES l 1 SURVEILLANCE SR 3.3.7.1.2 ' REQUIREMENTS (continued)- A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint , methodology. I The Frequency of 92 days is based on the reliability analyses'of Reference 3. { l SR 3.3.7.1.3 A CHANNEL CALIBRATION is a complete check of the instrument i loop and the sensor. This test verifies the channel l responds to the measured parameter within the necessary ' range and accuracy. CilANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. 'O The Frequency is based upon the assumption of a 24 month Q calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.7.1.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.7.3, " Control Room Emergency Ventilation (CREV) System," overlaps this Surveillance to provide complete testing of the assumed safety function. While this surveillance can be performed with the reactor at power, operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability standpoint. (continued) 4 [ q Brunswick Unit 1 B 3.3-190 Revision No. l l 4
L j CREV Systcc Instrumentatien i 8 3.3.7.1 bO BASES (continued) L REFERENCES 1. UFSAR, Section 15.6.4.5.5. l 2. 10 CFR 50.36(c)(2)(ii). l
- 3. GENE-770-06-1-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, December 1992.
1 O V l l
,,o o Brunswick Unit'l B 3.3-191 Revision No.
1
Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 'n b 8 3.3 INSTRUMENTATION B 3.3.7.2 Condenser Vacuum Pump Isolation Instrumentation BASES BACKGROUND The condenser vacuum pump isolation instrumentation I initiates a trip of the respective condenser vacuum pump and isolation of the common isolation valve following events in which main steam radiation monitors exceed a predetermined value. Tripping and isolating the condenser vacuum pumps limits control room dose in the event of a control rod drop l accident (CRDA). The condenser vacuum pump isolation instrumentation includes i sensors, relays and switches that are necessary to cause initiation of condenser vacuum pump isolation. The channels include electronic equipment that compares measured input ' signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then i outputs an isolation signal to the condenser vacuum pump I isolation logic. l f') (,/ The isolation logic consists of two independent trip systems, with two channels of the Main Steam Line b Radiation-High Function in each trip system. Each trip system is a one-out-of-two logic for this Function. Thus, either channel of the Main Steam Line Radiation-High function in each trip system are needed to trip a trip system. The outputs of the channels in a trip system are arranged in a logic so that both trip systems must trip to result in an isolation signal. There are two condenser vacuum pumps and one isolation valve associated with this function. APPLICABLE The condenser vacuum pump isolation is assumed in the safety SAFETY ANALYSES analysis for the CRDA. The condenser vacuum pump isolation instrumentation initiates an isolation of the condenser vacuum pumps to limit control room doses resulting from fuel cladding failure in a CRDA. The condenser vacuum pump isolation instrumentation satisfies Criterion 3 of Reference 1. l (continued) lO Brunswick Unit 1 B 3.3-192 Revision No.
l Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 IO lV BASES (continued) 1 I LCO The OPERABILITY of the condenser vacuum pump isolation ) instrumentation is dependent on the OPERABILITY of the ! individual Main Steam Line Radiation-High Function ; instrumentation channels, which nust have a required number : of OPERABLE channels in each trip system, with their " setpoints within the specified Allowable Value of ! SR 3.3.7.2.3. The actual setpoint is calibrated consistent i with applicable setpoint methodology assumptions. Channel l OPERABILITY also includes the condenser vacuum pump trip breakers and isolation valve. Allowable Values are specified for the condenser vacuum pump isolation Function specified in'the LCO. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the actual trip settings do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value,-is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are g compared to the actual process parameter (i.e., main steam line radiation), and when the measured output value of the i !- process parameter exceeds the setpoint, the associated i device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process. parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip . setpoint values, by accounting for the calibration based , errors. These calibration based errors are limited to l instrument drift, errors associated with measurement and test equipment, and calibration tolerances of loop _ components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and-severe environment errors (for channels that must function in harsh l environments as defined by 10 CFR 50.49) are accounted for . and appropriately applied for the instrumentation. I (continued) O Brunswick Unit 1 B 3.3-193 Revision No.
Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 O b BASES (continued) APPLICABILITY The condenser vacuum pump isolation is required to be OPERABLE in MODES I and 2 when the condenser vacuum pump is in service to mitigate the consequences of a postulated CRDA. In this condition, fission products released during a CRDA could be discharged directly to the environment. Therefore, the condenser vacuum pump isolation is necessary to assure conformance with the radiological evaluation of the CRDA. In MODE 3, 4 or 5 the consequences of a control rod drop are insignificant, and are not expected to result in any fuel damage or fission product releases. When the condenser vacuum pump is not in operation in MODE 1 or 2, fission product releases via this pathway would not occur. ACTIONS A Note has been provided to modify the ACTIONS related to condenser vacuum pump isolation instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the p Condition continue to apply for each additional failure, /A t.
' with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable condenser vacuum pump isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable condenser vacuum pump isolation instrumentation channel.
A.1 and A.2 With one or more channels inoperable, but with condenser vacuum pump isolation capability maintained (refer to Required Actions B.1, B.2, and B.3 Bases), the condenser vacuum pump isolation instrumentation is capable of performing the intended function. However, the reliability and redundancy of the condenser vacuum pump isolation instrumentation is reduced, such that a single failure in one of the remaining channels could result in the inability of the condenser vacuum pump isolation instrumentation to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to (continued) i p ! d Brunswick Unit 1 B 3.3-194 Revision No.
i: 1 l Condenser Vacuum Pump Isolation Instrumentaticn B 3.3.7.2 f3 l C) BASES ACTIONS A.1 and A.2 (continued) OPERABLE status. Because of the low probability of extensive number of inoperabilities affecting multiple channels, and the low probability of an event requiring the initiation of condenser vacuum pump isolation,12 hours has been shown to be acceptable (Ref. 2) to permit restoration of any inoperable channel to OPERABLE status (Required Action A.1). Alternately, the inoperable channel, or associated trip system, may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable condenser vacuum pump trip breaker or isolation valve, since this may not adequately compensate for the ) inoperable condenser vacuum pump trip breaker or isolation valve (e.g., the trip breaker may be inoperable such that it will not trip). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in loss of condenser vacuum), or if the inoperable channel is the result of an inoperable condenser (ov) vacuum pump trip breaker or isolation valve, Condition 8 must be entered and its Required Actions taken. d l B.I. B.2, and B.3 With any Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE or l other specified condition in which the LC0 does not apply. To achieve this status, the plant must be brought to at I least H00E 3 with 12 hours (Required Action B.3). Alternately, the associated condenser vacuum pumps may be removed from service since this performs the intended function of the instrumentation (Required Action B.1). An additional option is provided to isolate the main steam lines (Required Action B.2), which may allow operation to continue. Isolating the main steam lines effectively provides an equivalent level of protection by precluding fission product transport to the condenser. This isolation is accomplished by isolation of all main steam lines and main steam line drains which bypass the main steam isolation valves. (continuedl l b(% Brunswick Unit 1 8 3.3-195 Revision No. l 1 I
I l l l Condenser Vacuum Pump Isolation Instrumentation I B 3.3.7.2 '
/'~'s
() BASES ACTIONS B.I. B.2. and B.3 (continued) Condition B is also intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in the function not maintaining condenser vacuum pump isolation capability. The Function is ( considered to be maintaining condenser vacuum pump isolation l capability when sufficient channels are OPERABLE or in trip such that the condenser vacuum pump isolation instruments lj will generate a trip signal from a valid Main Steam Line ! Radiation-High signal, and the condenser vacuum pumps will i trip. This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the condenser vacuum pump trip breakers to be OPERABLE. SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains condenser vacuum pump isolation trip capability. Upon (q
'"j completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status d
or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 2) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the condenser vacuum pumps will isolate when necessary. SR 3.3.7.2.1 Performance of th; (llANNEL CHECK once every 24 hours ensures I that a gross failure of instrumentation has not occurred. A ! CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other l channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of (continued) (O L) Brunswick Unit 1 B 3.3-196 Revision No.
l t Condenser Vacuum Pump Isolatien Instrumentation B 3.3.7.2 I f3 lV BASES SURVEILLANCE SR 3.3.7.2.1 (continued) REQUIREMENTS excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect i gross channel failure; thus, it is key to verifying the J instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on combination of the channel instrument uncertainties, including indication and readability. If a channel is L outside the criteria, it may be an indication that the instrument has' drifted outside its limit. i The Frequency is based on the CHANNEL CHECK Frequency requirement of other instrumentation. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays 1 L associated with the required channels of this LCO.
- SR 3.3.7.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required -
! channel to ensure that the channel will perform the intended i l function. Any setpoint adjustment shall be consistent with -- l the assumptions of the current plant ' specific setpoint L methodology. l-l The Frequency of 92 days is based on the reliability l analysis of Reference 2. SR 3.3.7.2.3 A CHANNEL CALIBRATION is a complete check of the instrument l loop and the sensor. This test verifies the channel l responds to the measured parameter within the necessary
- range and accuracy. CHANNEL CALIBRATION leaves the channel l adjusted to account for instrument drifts between successive l calibrations consistent with the plant specific setpoint methodology.
The Frequency is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. (continued) Brunswick Unit 1 B 3.3-197 Revision No.
f j Condenser Vacuum Pump Isolation Instrumentation l , 8 3.3.7.2 I n b BASES i SURVEILLANCE SR 3.3.7.2.3 (continued) l REQUIREMENTS for the purposes of this SR, background is the dose level l experienced at 100% RATED THERMAL POWER with hydrogen water i chemistry at the maximum injection rate. Under these ' conditions, an Allowable Value of s 6 x background will ensure that General Design Criterion 19 limits of 10 CFR 50, Appendix A, will not be exceeded in the control room in the event of a CRDA. SR 3.3.7.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers and actuation of the associated isolation valve are included as part of this Surveillance and overlaps the LOGIC SYSTEM { FUNCTIONAL TEST to provide complete testing of the assumed j safety function. Therefore, if a breaker is incapable of operating or the isolation valve is incapable of actuating, the instrument channel would be inoperable. r' b i)
' The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
l REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
- 2. NEDC-30851P-A, Supplement 2. Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
I O v i Brunswick Unit 1 B 3.3-198 Revision No.
LOP Instrumentation B 3.3.8.1 W l U -B 3.3 INSTRUMENTATION l B 3.3.8.1 Loss of Power (LOP) Instrumentation l BASES BACKGROUND- Successful operation of the required safety functions of the Emergency Core Cooling Systems (ECCS) is dependent upon the availability of adequate power sources for energizing the various components such as pump motors, motor operated valves, and the associated control components. The LOP instrumentation monitors the 4.16 kV emergency buses. Offsite power is the preferred source of power for the 4.16 kV emergency buses. If the monitors determine that insufficient power is available, the buses are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources. Each 4.16 kV emergency bus has its own independent LOP instrumentation and associated trip logic. The voltage for each bus is monitored at two levels, which can be considered as two different undervoltage Functions: Loss of Voltage and 4.16 kV Emergency Bus Undervoltage Degraded Voltage. O Each Function causes various bus transfers and disconnects. The Loss of Voltage Function is monitored by.one inverse time delay undervoltage relay (27/59E) and the Degraded Voltage function is monitored by three definite time undervoltage relays (270VA, 270VB, and 27DVC) for each i emergency bus. The Loss of Voltage Function is a one-out-of-one logic configuration and the Degraded Voltage Function output is arranged as a two-out-of-three logic configuration. The channels include electronic equipment l (e.g., internal relay contacts, coils, etc.) that compares measured input signals with pre-established setpoints. When ; the setpoint is exceeded, the channel'out)ut relay actuates, ' which then outputs a LOP trip signal to t1e trip logic. < APPLICABLE The LOP instrumentation is required for Engineered Safety SAFETY ANALYSES, Features to function in any accident with a loss of offsite LCO, and power. The required channels of LOP instrumentation ensure APPLICABILITY that the ECCS and other assumed systems powered from the DGs, provide plant protection in the event of any of the Reference 1 and 2 analyzed accidents in which a loss of offsite power is assumed. The initiation of the DGs on loss l of offsite power, and subsequent initiation of the ECCS, ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. i (continued) Brunswick Unit 1 B 3.3-199 Revision No.
LOP Instrumentation B 3.3.8.1 l'V : BASES APPLICABLE Accident analyses credit the loading of the DG based on the SAFETY ANALYSES, loss of offsite power during a loss of coolant accident. LCO, and The diesel starting and loading times have been included in APPLICABILITY the delay time associated with each safety system component (continued) requiring DG supplied power following a loss of offsite power. The LOP instrumentation satisfies Criterion 3 of Reference 3. The OPERABILITY of the LOP instrumentation is dependerit upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.8.1-1. Each Function must have a required number of OPERABLE channels per 4.16 kV emergency bus, with their setpoints within the specified Allowable Values. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. l The Allowable Values are specified for each Function in the Table. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less O) (
conservative than the trip setpoint, but within the Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., degraded voltage), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g.,
trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for (continued) O Brunswick Unit 1 B 3.3-200 Revision No. 1
LOP Instrumentatien B 3.3.8.1 Ch V BASES APPLICABLE channels that must function in harsh environments as defined SAFETY ANALYSES, by 10 CFR 50.49) are accounted for and appropriately applied LCO, and for the instrumentation. APPLICABILITY (continued) The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by
. Function basis.
- 1. 4.16 kV EmerQency Bus UndervoltaQe (Loss of VoltaQe)
Loss of voltage on a 4.16 kV emergency bus indicates that offsite power may be completely lost to the respective emergency bus and is unable to supply sufficient power for proper operation of the applicable equipment. Therefore, l the power supply to the bus is transferred from offsite i power to DG power when the voltage on the bus drops below the Loss of Voltage Function Allowable Values (loss of voltage with a short time delay). This ensures that adequate power will be available to the required equipment. ; 1 The Bus Undervoltage Allowable Values are low enough to , q prevent inadvertent power supply transfer, but high enough i Q to ensure that power is available to the required equipment. The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that power is available ; to the required equipment. ' One channel of 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) Function per associated emergency bus is only required to be OPERABLE when the associated DG is required to be OPERABLE to ensure that no single instrument failure can preclude the start of three of the four DGs. (One channel inputs to each of the four DGs.) Refer to LC0 3.8.1, "AC Sources-Operating," and 3.8.2, "AC Sources-Shutdown," for Applicability Bases for the DGs.
- 2. 4.16 kV EmerQency Bus UndervoltaQe (DeQraded VoltaQe)
A reduced voltage condition on a 4.16 kV emergency bus indicates that, while offsite power may not be completely lost to the respective emergency bus, available power may be (continued) Brunswick Unit 1 B 3.3-201 Revision No.
r-
]
LOP Instrumentatien 8 3.3.8.1 (3
/ BASES APPLICABLE 2. 4.16 kV Emeraency Bus Undervoltaae (Dearaded Voltaae)
SAFETY ANALYSES, (continued) LCO, and APPLICABILITY insufficient for starting large ECCS motors without risking i damage to the motors that could disable the ECCS function. ' Therefore, the power supply to the bus is transferred from offsite power to onsite DG power when the voltage on the bus drops below the Degraded Voltage function Allowable Values (degraded voltage with a time delay). This ensures that adequate power will be available to the required equipment. The Bus Undervoltage Allowable Values are low enough to prevent inadvertent power supply transfer, but high enough to ensure that sufficient power is available to the required equipment. The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that sufficient , power is available to the required equipment. i Three channels of 4.16 kV Emergency Bus Undervoltage l (Degraded Voltage) Function per associated bus are only required to be OPERABLE when the associated DG is required to be OPERABLE to ensure that no single instrument failure (q can preclude the DG function. (Three channels input to each 1 of the four emergency buses and DGs.) Refer to LCO 3.8.1 l and LCO 3.8.2 for Applicability Bases for the OGs. ACTIONS A Note has been provided to modify the ACTIONS related to ! LOP instrumentation channels. Section 1.3, Completion I Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable LOP instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable LOP instrumentation channel. (continued) , O Brunswick Unit 1 B 3.3-202 Revision No.
LOP. Instrumentation B 3.3.8.1 m BASES ACTIONS Ad (continued) With one or more channels of a function iroperable, the function is not capable of performing the intended function. Therefore, only I hour is allowed to restare the inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.I. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure (within the LOP instrumentation), and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in a DG initiation), Condition B must be entered and its Required Action taken. The Completion Time is intendad to allow the operator time to evaluate and repair any discovered inoperabilities. The I hour Completion Time is acceptable because it minimizes l risk while allowing time for restoration or tripping of channels. O i B.1 If any Required Action and associated Completion Time are not met, the associated Function is not capable of performing the intended function. Therefore, the associated DG(s) is declared inoperable immediately. This requires entry into applicable Conditions and Required Actions of l LCO 3.8.1 and LCO 3.8.2, which provide appropriate actions { for the inoperable DG(s). i 1 SURVEILLANCE As noted at the beginning of the SRs, the SRs for each LOP REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.8.1-1. ! The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated i Conditions and Required Actions may be delayed for up to (continued) A I Brunswick Unit 1 B 3.3-203 Revision No. i
F 1 j LOP Instrumentation l- B 3.3.8.1
/~N l V BASES j l
l SURVEILLANCE 2 hours provided: (a) for Function 1, the associated REQUIREMENTS Function maintains initiation capability for three DGs; and (continued) (b) for Function 2, the associate Function maintains DG initiation capability. For Function 1, the loss of function for one DG for this short period is appropriate since only three of four DGs are required to start within the required times and because there is no appreciable impact on risk. Also, upon completion of the Surveillance, or expiration of the 2 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. SR 3.3.8.1.1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 31 days is based on operating experience with regard to channel OPERABILITY and drift, which (n-) demonstrates that failure of more than one channel of a given function is a rare event. j j SR 3.3.8.1.2 and SR 3.3.8.1.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive l calibrations consistent with the plant specific setpoint methodology. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequencies of SR 3.3.8.1.2 and SR 3.3.8.1.3 are based upon the assumptions of 18 and 24 month calibration intervals, respectively, in the determination of the magnitude of equipment drift in the setpoint analyses. (continued)
.m -
Brunswick Unit 1 B 3.3-204 Revision No.
LOP Instrumentaticn B 3.3.8.1 ! /~ BASES i SURVEILLANCE .SR 3.3.8.1.4 ) REQUIREMENTS l (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed in LCO 3.8.1 and LC0 3.8.2 overlaps this Surveillance to provide complete testing of the assumed safety functions.
)
The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated these components will pass the Surveillance when performed at the 24 month Frequency. REFERENCES 1. UFSAR, Section 6.3.
- 2. UFSAR, Chapter 15.
- 3. 10 CFR 50.36(c)(2)(ii).
%/ i l ?\ V Brunswick Unit 1 B 3.3-205 Revision No. j
-C
[ RPS Electric Power Monitoring B 3.3.8.2 B 3.3 INSTRUMENTATION. B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring BASES i l BACKGROUND RPS Electric Power Monitoring System is provided to isolate l the RPS bus from the motor generator (MG) set or an alternate power supply in the event of overvoltage, undervoltage, or underfrequency. This system protects the loads connected to the RPS bus against unacceptable voltage and frequency conditions (Ref.1) and forms an important part of the primary success path of the essential safety circuits. Some of the essential equipment powered from the RPS buses includes the RPS logic and scram solenoids. I RPS electric power monitoring assembly will detect any abnormal high or low voltage or low frequency condition in the outputs of the two MG sets or the alternate power supply and will de-energize its respective RPS bus, thereby causing all safety functions normally powered by this bus to de-energize. t In the event of failure of an RPS Electric Power Monitoring System (e.g., both in series electric power monitoring assemblies), the RPS loads may experience significant effects from the unregulated power supply. Deviation.from the nominal conditions can potentially cause damage to the scram solenoids and other Class IE devices. In the event of a low voltage condition for an extended period of time, the scram solenoids can chatter and potentially lose their pneumatic control capability, l resulting in a loss of primary scram action. In the event of an overvoltage condition, the RPS logic relays and scram solenoids may experience a voltage higher than their design voltage. If the overvoltage condition ; persists for an extended time period, it may cause equipment i degradation and the loss of plant safety function. ! Two redundant Class IE circuit breakers are connected in ! series between each RPS bus and its MG set, and between each ! RPS bus and the alternate power supply. Each of these i circuit breakers has an associated independent set of l Class IE overvoltage, undervoltage, and underfrequency
- . sensing logic. .Together, a circuit breaker and its sensing logic constitute an electric power monitoring assembly. If i (continued) !
!O Brunswick Unit-1 B 3.3-206 Revision No.
RPS Electric Power Monitoring B 3.3.8.2 10 Q BASES BACKGROUND the output of the MG set or the alternate power supply (continued) exceeds predetermined limits of overvoltage, undervoltage, or underfrequency, a trip coil driven by this logic circuitry opens the circuit breaker, which removes the associated power supply from service. APPLICABLE The RPS electric power monitoring is necessary to meet the SAFETY ANALYSES assumptions of the safety analyses by ensuring that the RPS equipment powered from the RPS buses can perform its intended function. RPS electric power monitoring provides protection to the RPS components, by acting to disconnect the RPS from the power supply under specified conditions that could damage the RPS equipment. RPS electric power monitoring satisfies Criterion 3 of Reference 2. LCO The OPERABILITY of each RPS electric power monitoring assembly is dependent on the OPERABILITY of the overvoltage, undervoltage, and underfrequency logic, as well as the OPERABILITY of the associated circuit breaker. Two electric hyv power monitoring assemblies are required to be OPERABLE for each inservice power supply. This provides redundant protection against any abnormal voltage or frequency conditions to ensure that no single RPS electric power monitoring assembly failure can preclude the function of RPS components. Each of the inservice electric power monitoring assembly trip logic setpoints is required to be within the . specified Allowable Value. The actual setpoint is ! calibrated consistent with applicable setpoint methodology l assumptions. l Allowable Values are specified for each RPS electric power monitoring assembly trip logic (refer to SR 3.3.8.2.2 l and SR 3.3.8.2.3). Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CAllBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., overvoltage), and when the measured output value of i (continued) O O Brunswick Unit 1 B 3.3-207 Revision No.
I ! RPS Electric Power Monit: ring l B 3.3.8.2 1
-q h BASES LCO the process parameter exceeds the setpoint, the associated
.(continued) device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors.
l The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to I instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. l The Allowable Values for the instrument settings of the normal power supply (RPS MG set) electric power monitoring assembly are based on the RPS MG sets providing a 57 Hz and ) 117 V i 10%. The Allowable Values for the instrument i h) v settings of the alternate power supply electric power ' monitoring assembly are based on the alternate power supply providing a 57 Hz and 120 V i 10%. The most limiting l voltage requirement and associated line losses determine the ! settings of the electric power monitoring instrument ! channels. The settings are calculated based on the line ; resistance losses at the downstream locations of the solenoids and relays. APPLICABILITY The operation of the RPS electric power monitoring assemblies is essential to disconnect the RPS components from the MG set or alternate power supply during abnormal voltage or frequency conditions. Since the degradation of a nonclass IE source supplying power to the RPS bus can occur as a result of any random single failure, the OPERABILITY of the RPS electric power monitoring assemblies is required when the RPS components are required to be OPERABLE. This results in the RPS Electric Power Monitoring System OPERABILITY being required in MODES I and 2; and in MODES 3, 4, and 5 with any control rod withdrawn from a core cell g containing one or more fuel assemolies. (continued) O b Brunswick Unit 1 B 3.3-208 Revision No.
RPS Electric Power Monitoring B 3.3.8.2 .r\ V BASES (continued) ACTIONS M If one RPS electric power monitoring assembly for an inservice power supply (MG set or alternate) is inoperable, or one RPS electric power monitoring assembly on each inservice power supply is inoperable, the OPERABLE assembly will still provide protection to the RPS components under degraded voltage or frequency conditions. However, the reliability and redundancy of the RPS Electric Power Monitoring System is reduced, and only a limited time (72 hours) is allowed to restore the inoperable assembly to OPERABLE status. If the inoperable assembly cannot be restored to OPERABLE status, the associated power supply (s) must be removed from service (Required Action A.1). This places the RPS bus in a safe condition. An alternate power supply with OPERABLE powering monitoring assemblies may then be used to power the RPS bus. The 72 hour Completion Time takes into account the remaining OPERABLE electric power monitoring assembly and the low probability of an event requiring RPS electric power monitoring protection nccurring during this period. It n allows time for plant operations personnel to take (') corrective actions or to place the plant in the required condition in an orderly manner and without challenging plant I systems. Alternately, if it is not desired to remove the power supply from service (e.g., as in the case where removing the power supply (s) from service would result in a scram or isolation), Condition C or D, as applicable, must be entered and its Required Actions taken. M If both power monitoring assemblies for an inservice power supply (MG set or alternate) are inoperable or both power monitoring assemblies in each inservice power supply are inoperable, the system protective function is lost. In this condition, I hour is allowed to restore one assembly to OPERABLE status for each inservice power supply. If one inoperable assembly for each inservice power supply cannot be restored to OPERABLE status, the associated power (continued) Brunswick Unit 1 B 3.3-209 Revision No.
RPS Electric Power Monitoring B 3.3.8.2 Q BASES
. ACTIONS M (continued) supply (s) must be removed from service within I hour (Required Action B.1). An alternate power supply with ,
OPERABLE' assemblies may then be used to power one RPS bus. I The 1 hour Completion Time is sufficient for the plant ! operations personnel to take corrective. actions and is- ' acceptable because it minimizes risk while allowing time for restoration or removal from service of the electric power monitoring assemblies. Alternately, if it is not desired to remove the power supply (s) from service (e.g., as in the case where removing i the power supply (s) from service would result in a scram or isolation), Condition C or D, as applicable,'must be entered and its Required Actions taken. l C.1 and C.2 If any Required Action and associated Completion Time of Condition A or B are not met in MODE 1 or 2, a plant shutdown must be performed. This places the plant in a O condition where minimal equipment, powered through the inoperable RPS electric power monitoring assembly (s), is required and ensures that the safety function of the RPS (e.g., scram of control rods) is not required. -The plant shutdown is accomplished by placing the plant in MODE 3 within 12 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. I M If any Required Action and associated Completion Time of Condition A or B are not met in MODE 3, 4, or 5 with any control rod withdrawn from a core cell containing one or b l more fuel assemblies, the operator must immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Required Action D.1 results in the least reactive condition for the reactor core and ensures'that the safety function of the RPS : (e.g., scram of control rods) is not required. (continued) l l Brunswick Unit 1 B 3.3-210 Revision No.
p i RPS Electric Power M:nitoring B 3.3.8.2 L BASES (continued)' I SURVEILLANCE SR 3.3.8.2.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed on each overvoltage, undervoltage, and underfrequency channel to ensure that the .. channel will perform the intended function. Any setpoint l , adjustment shall be consistent with the assumptions of the l current plant specific setpoint methodology. . 1 As noted in the Surveillance, the CHANNEL FUNCTIONAL TEST is only required to be performed while the plant is.in a q condition in which the loss of the RPS bus will not jeopardize steady state power operation (the design of the system is such that the power source must be removed from service to conduct the Surveillance). The 24 hours is intended to indicate an outage of sufficient duration to allow for scheduling and proper performance of the J Surveillance. I l The 184 day Frequency and the Note in the Surveillance are j based on guidance provided in Generic letter 91-09 (Ref. 3). SR 3.3.8.2.2 and SR 3.3.8.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel' responds to the measured parameter within the necessary l range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations ' consistent with the plant specific setpoint methodology. The Frequencies are based on the assumption of a 24 month
- l. calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
SR 3.3.8.2.4 l Performance of a system functional test demonstrates that, with a required system actuation (simulated or actual) signal, the logic of the system.will automatically trip open l the associated power monitoring assembly. Only one signal (continued) ! LO L Brunswick Unit 1 B 3.3-211 Revision No.
t RPS Electric Power Monitoring B 3.3.8.2 l -BASES l . SURVEILLANCE SR 3.3.8.2.4 (continued) REQUIREMENTS per power monitoring assembly is required to be tested. This Surveillance overlaps with the CHANNEL CALIBRATION to provide. complete testing of the safety function. The system functional test of the class IE circuit breakers is included as part of_ this test to provide complete testing of the safety function. If the breakers are incapable of operating, the associated electric power monitoring assembly would be inoperable. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency. REFERENCES 1. UFSAR, Section 7.2.1.1.1.3.
- 2. _ 10 CFR 50.36(c)(2)(ii).
O 3. NRC Generic Letter 91-09, Modification of Surveillance Interval for the Electrical Protective Assemblies in Power Supplies for the Reactor Protection System. l l 1 j O Brunswick Unit 1 B 3.3-212 Revision No.
SRM Instrumentaticn 3.3.1.2 y/ . SURVEILLANCE REQUIREMENTS
-------------------------------------NOTE-------------------------------------
Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified condition. SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK. 12 hours SR 3.3.1.2.2 -------------------NOTES----------------- n
- 1. Only required to be met during CORE QC ALTERATIONS.
- 2. One SRM may be used to satisfy more A than one of the following. /O-Verify an OPERABLE SRM detector is 12 hours located in:
- a. The fueled region;
- b. The core quadrant where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region; and g
- c. A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region.
SR 3.3.1.2.3 Perform CHANNEL CHECK. 24 hours (continued) O Brunswick Unit 2 3.3-13 Amendment No. i
SRM Instrumentation 3.3.1.2 O Q SURVEILLANCE REQUIREMENTS (continued) i SURVEILLANCE FREQUENCY SR 3.3.1.2.4 ------------------NOTES------------------
- 1. Not required to be met with less than or equal to four fuel assemblies -
adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
- 2. Not required to be met during a core spiral offload.
Verify count rate is 2: 3.0 cps. 12 hours during CORE ALTERATIONS A_N_0 24 hours g SR 3.3.1.2.5 Perfctm CHANNEL FUNCTIONAL TEST. 7 days i SR 3.3.1.2.6 -------------------NOTE------------------ Not required to be performed until l 12 hours after IRMs on Range 2 or below. p Perform CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.1.2.7 ------------------NOTES------------------ l
- 1. Neutron detectors are excluded. i
- 2. Not required to be performed until 12 hours after IRMs on Range 2 or below.
Perform CHANNEL CALIBRATION. 24 months Brunswick Unit 2 3.3-14 Amendment No.
1 SRM Instrumentation 3.3.1.2 O Q Tabte 3.3.1.2-1 (page 1 of 1) source Range Monitor Instrumentation APPLICABLE MODES OR OTHER REQUIRED SURVEILLANCE FUNCil0N- SPECIFIED CONDIT!DNS CHANNELS REQUIREMENTS
- 1. Source Range Monitor 2(a) 3 sa 3.3.1.2.1 SR 3.3.1.2.4 SR 3.3.1.2.6 st 3.3.1.2.7 3,4 2 SR 3.3.1.2.3 SR. 3.3.1.2.4 st 3.3.1.2.6 st 3.3.1.2.7 5 2(b) sR 3.3.1.2.1 SR 3.3.1.2.2 SR 3.3.1.2.4
- SR 3.3.1.2.5 SR 3.3.1.2.7 (a) With IRMs on Range 2 or below.
(b) special movable detectors may be used in place of SRMs if connected to normal SRM circuits. k l O Brunswick Unit 2. 3.3-15 Amendment No. 1.
PBDS 3.3.1.3 3.3 INSTRUMENTATION 3.3.1.3 Period Based Detection System (PBDS) LC0 3.3.1.3 One channel of PBDS instrumentation shall be OPERABLE. MD Each OPERABLE channel of PBDS instrumentation shall not indicate High-High Alarm. APPLICABILITY: THERMAL POWER and core flow in the Restricted Region specified in the COLR, THERMAL POWER and core flow in the Monitored Region specified in the COLR. l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A A. Any OPERABLE PBDS A.I Manually scram the Immediately . Q channel indicating reactor. High-High Alarm. B. Required PBDS channel B.1 --------NOTE--- ---- inoperable while in Only applicable .f the Restricted Region. RPS Function 2.b, APRM flow Biased Simulated Thermal Power-High, Allowable Value is
" Setup" l Initiate action to Immediately l
exit the Restricted l Region. l l B i l (continued) I i I Brunswick Unit 2 3.3-16 Amendment No.
l PBDS
]
3.3.1.3 J ( (N) ACTIONS
)
CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Manually scram the Immediately reactor. C. Required PBDS channel C.1 Initiate action to 15 minutes inoperable while in exit the Monitored the Monitored Region. Region. I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.3.1 Verify each OPERABLE channel of PBDS 12 hours p instrumentation not in High-High Alarm V SR 3.3.1.3.2 Perform CHANNEL CHECK. 12 hours ! 1 SR 3.3.1.3.3 Perform CHANNEL FUNCTIONAL TEST. 24 months ! j -C) V Brunswick Unit 2 3.3-17 Amendment No.
i l Control Rod Bleck Instrumentaticn ! l 3.3.2.1 l O (/ 3.3 INSTRUMENTATION 3.3.2.1 Control Rod Block Instrumentation LCO 3.3.2.1 The control rod block instrumentation for each Function in Table 3.3.2.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.2.1-1. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One rod block monitor A.1 Restore RBM channel 24 hours (Rr>M) channel to OPERABLE status. inoperable. B. Pequired Action and 8.1 Place one RBM channel I hour l A associated Completion in trip. V Time of Condition A not met. E Two RBM channels inoperable. C. Rod worth minimizer C.1 Suspend control rod Immediately . (RWM) inoperable movement except by during reactor scram. ! startup. E (continued) 3.3-18 Brunswick U91t 2 Amendment No. I
Control Rod Block Instrumentation 3.3.2.1 '.(V'S ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.1.1 Verify 2: 12 rods Immediately : withdrawn. E C.2.1.2 Verify by - Immediately administrative methods that startup with RWM inoperable, for reasons other than bypassed control rod (s), has not been performed in the last calendar year. AND C.2.2 Verify movement of During control bypassed control rod movement rod (s) is in compliance with banked position V(n) withdrawal sequence (BPWS) by a second licensed operator or other qualified member of the technical staff. D. RWM inoperable during D.1 Verify movement of During control reactor shutdown. bypassed control rod movement rod (s) is in i accordance with BPWS I by a second licensed operator or other qualified member of , the technical staff. i (continued) l Brunswick Unit 2 3.3-19 Amendment No. l 1 I t
1 l Control Rod Bleck Instrumentaticn 3.3.2.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor E.1 Suspend control rod Immediately Mode Switch-Shutdown withdrawal. Position channels inoperable. AND E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more feel assemblies. SURVEILLANCE REQUIREMENTS __...._____...._____.........N0TES------------------------------------
- 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.
- 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.
SURVEILLANCE FREQUENCY SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. 92 days (continued) I j Brunswick Unit 2 3.3-20 Amendment No. ,
1 control Rod Block Instrumentaticn 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.2.1.2 ------------------NOTE------------------- Not required to be performed until I hour after any control rod is withdrawn at s 10% RTP in MODE 2. Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.3 ------------------NOTE------------------- Not required to be performed until I hour after THERMAL POWER is s 10% RTP in MODE 1. Perform CHANNEL FUNCTIONAL TEST. 92 days (' SR 3.3.2.1.4 ------------------NOTE------------------- Neutron detectors are excluded. Verify the RBM: 24 months
- a. Low Power Range-Upscale Function is not bypassed when THERMAL POWER is 2: 29% RTP and s Intermediate Power Range Setpoint specified in the COLR. 4
- b. Intermediate Power Range-Upscale Function is not bypassed when THERMAL POWER is > Intermediate Pcwer Range Setpoint specified in the COLR and s High Power Range Setpoint specified in the COLR.
- c. High Power Range-Upscale function is not bypassed when THERMAL POWER is
> High Power Range Setpoint specified in the COLR.
(continued) O Brunswick Unit 2 3.3-21 Amendment No.
i f' Control Rod Block Instrumentatien 3.3.2.1 l .O V SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.2.1.5 Verify the RWM is not bypassed when 24 months THERMAL POWER is s 10% RTP. SR 3.3.2.1.6 ------------------NOTE------------------- Not required to be performed until I hour after reactor mode switch is in the shutdown position. Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.2.1.7 ------------------NOTE------------------- Neutron detectors are excluded. Perform CHANNEL CALIBRATION. 24 months A SR 3.3.2.1.8 Verify control rod sequences input to the Prior to RWM are in conformance with BPWS. declaring RWM OPERABLE following loading of sequence into RWM l l Brunswick Unit 2 3.3-22 Amendment No. j i
Control Rod Block Instrumentation 3.3.2.1 x Table 3.3.2.1 1 (page 1 of 1) control Rod Block Instrunentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLAECE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE
- 1. Rod eIock Monitor
- a. Low Power Range --4Jpscale (a) 2 SR 3.3.2.1.1 (h)
SR 3.3.2.1.4 SR 3.3.2.1.7
- b. Intermediate Power (b) 2 SR 3.3.2.1.1 (h)
Range --Upscale SR 3.3.2.1.4 SR 3.3.2.1.7
- c. High Power Range @cate (c),(d) 2 SR 3.3.2.1.1 (h)
SR 3.3.2.1.4 SR 3.3.2.1.7
- d. Inop (d),(e) 2 SR 3.3.2.1.1 NA
- e. Downecat e (d),(e) 2 SR 3.3.2.1.1 NA SR 3.3.2.1.7 h
- 2. Rod Worth Minimizer 1(I) 2(Il
, 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.8
- 3. Reactor Mode Switch -shutdown (g) 2 SR 3.3.2.1.6 NA Position (a) THERMAL POWER t 29% RTP and 5 Intermediate Power Range Setpoint specified in the COLR and MCPR < 1.70.
(b) THERMAL POWER > Intermediate Power Range Setpoint specified in the COLR and 5 High Power Range Setpoint specified in the COLR and MCPR < 1.70. (c) THERMAL POWER > High Power Range Setpoint specified in the COLR and < 90% RTP and MCPR < 1.70. (d) THERMAL POWER t 90% RTP and MCPR < 1.40.
-(e) THERMAL POWER t 29% and < 90% RTP and MCPR < 1.70.
(f) With YHERMAL POWER 5 10% RTP. (g) Reactor mode switch in the shutdown position. (h) Allowable Value specified in the COLR. O v Brunswick Unit 2 3.3-23 Amendment No.
.Ferdwater and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 3.3 INSTRUMENTATION 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation LCO 3.3.2.2 Three channels of feedwater and main turbine high water level trip instrumentation shall be OPERABLE.
APPLICABILITY: THERMAL POWER 2: 25% RTP. ACTIONS
.............___..................---NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One feedwater and main A.1 Place channel in 7 days turbine high water trip. level trip channel O inoperable. B. Two or more feedwater B.1 Restore feedwater and 4 hours and main turbine high main turbine high water level trip water level trip channels inoperable. capability.
-C. Required Action and C.1 Reduce THERMAL POWER 4 hours associated Completion to < 25% RTP.
Time not met. O Brunswick Unit 2 3.3-24 Amendment No. l
Feedwater and Hain Turbine High Water Level Trip Instrumentation 3.3.2.2 l v SURVEILLANCE REQUIREMENTS
-------------------------------------NOTE-------------------------------------
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided feedwater and main turbine high water level trip capability is maintained. SURVEILLANCE FREQUENCY SR 3.3.2.2.1 Perform CHANNEL CHECK. 24 hours SR 3.3.2.2.2 Perform CHANNEL CALIBRATION. The 24 months Allowable Value shall be :s; 207 inches. SR 3.3.2.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST, 24 months including valve actuation. O V n V Brunswick Unit 2 3.3-25 Amendment No.
L PAM Instrumentation 3.3.3.1
,~
V 3.3 INSTRUMENTATION 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation l > LCO 3.3.3.1 The PAM instrumentation for each function in Table 3.3.3.1-1 - shall be OPERABLE. APPLICABILITY. MODES 1 and 2. ACTIONS
-------------------------------------NOTES------------------------------------
- 1. LC0 3.0.4 is not applicable.
- 2. Separate Condition entry is allowed for each Function. j CONDITION REQUIRED ACTION COMPLETION TIME e A. One or more Functions A.1 Restore required 30 days b with one required channel to OPERABLE ;
channel inoperable. status. l l 1 B. Required Action and B.1 Initiate action in Immediately associated Completion accordance with Time of Condition A Specification 5.6.6. not met. C. One or more functions with two required C.1 Restore one required channel to OPERABLE 7 days b channels inoperable. status. (continued) I
,~.
I { Brunswick Unit 2 3.3-26 Amendment No.
l PAM Instrumentatien ' O 3.3.3.1 ! ACTIONS (continued)
' CONDITION REQUIRED ACTION COMPLETION TIME l
D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C Table 3.3.3.1-1 for - not met. the ch.inel . E. As required by E.1 Be in MODE 3. 12 hours Required Action D.1 and referenced in Table 3.3.3.1-1. F. As required by F.1 Initiate action in Immediately Required Action D.1 accordance with and referenced in Specification 5.6.6. Table 3.3.3.1-1. O SURVEILLANCE REQUIREMENTS
-------------------------------------NOTE-------------------------------------
These SRs apply to each Function in Table 3.3.3.1-1. SURVEILLANCE FREQUENCY I l SR 3.3.3.1.1 Perform CHANNEL CHECK. 31 days g SR 3.3.3.1.2 Perform CHANNEL CALIBRATION of the 92 days Drywell and Suppression Chamber H z and 02 Analyzers. (continued) O . Brunswick Unit 2 3.3-27 Amendment No. I
l PAM Instrumentation 3.3.3.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.3.1.3 Perform CHANNEL CALIBRATION for each 24 months required PAM Instrumentation channel except for Drywell and Suppression Chamber 2H and 02 Analyzers. O O Brunswick Unit 2 3.3-28 Amendment No.
PAM Instrumentation 3.3.3.1 l N Table 3.3.3.1 1 (page 1 of 1) Post Accident Nonitoring Instrumentation l CONDITIONS REFERENCED REQUIRED FRON REQUIRED
' FUNCil0N CHANNELS ACTION D.1 l
l
- 1. Reactor vesset Pressure ' 2 E
- 2. Reactor Vessel Water Level
- a. 150 inches to +150 inches 2 E
. b. O inches to +210 inches 2 E
- c. +150 inches to +550 inches 2 E
- 3. Suppression Chanber Water Level 2 E
- 4. Suppression Chaneer Water Tenperature 2 E
- 5. Suppression Chamber Pressure 2 C
- 6. Drywell Pressure 2 E
- 7. Drywell Tenperature 2 E
- 8. PCIV Position 2perpenetgatg flow path E
- 9. Drywell and Suppression Chanber H, & 0, Anatyrer 2 E k 10. Drywell Area Radiation 2 F (a) Not required for isolation valves Whose associated penetration flow path is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow <
through the valve secured.
)
(b) only one position indication channel is required for penetration flow paths with only one instatted control room indication channet. j l 1
~
( O Brunswick Unit 2 ~3.3-29 Amendment No.
Remote Shutdown Monitoring Instrumentation 3.3.3.2 O-i 3.3 INSTRUMENTATION 3.3.3.2 Pemote Shutdown Monitoring Instrumentation LCO 3.3.3.2 The Remote Shutdown Monitoring Instrumentation Functions shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS
.......__.__...___..____....____.....N0TES-------------------- ---____-__....-
-1. LCO 3.0.4 is not applicable.
- 2. Separate Condition entry is allowed for each function.
1 l CONDITION REQUIRED ACTION COMPLETION TIME fs A. One or more required A.1 Restore required 30 days d@s ' () Functions inoperable. Function to OPERABLE status. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized. (continued) O l Brunswick Unit 2 3.3-30 Amendment No.
i i Remote Shutdown Monitoring Instrumentation 3.3.3.2 i SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.3.2.2 Perform CHANNEL CALIBRATION for each 24 months required instrumentation channel. O i O Brunswick Unit 2 3.3-31 Amendment No.
r ATWS-RPT Instrumentaticn 1 3.3.4.1 3.3 INSTRUMENTATION I 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip i (ATWS-RPT) Instrumentation LCO 3.3.4.1 Two channels per trip system for each ATWS-RPT instrumentation Function listed below shall be OPERABLE-
}
- a. Reactor Vessel Water Level-Low Level 2; and
- b. Reactor Vessel Pressure-High.
APPLICABILITY: MODE 1. ACTIONS
.__........................----------NOTE--------------------------------.----
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME 0 V i l A. One or more channels A.1 Restore channel to 14 days I inoperable. OPERABLE status. 03 A.2 --------NOTE--------- Not applicable if inoperable channel is , the result of an ' inoperable breaker. Place channel in 14 days trip. (continued) O Brunswick Unit 2 3.3-32 Amendment ilo
ATWS-RPT Instrumentation 3.3.4.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. One Function with B.1 Restore ATWS-RPT trip 72 hours , ATWS-RPT trip capability. ' capability not maintained. C. Both Functions with C.1 Restore ATWS-RPT trip I hour ATWS-RPT trip capability for one capability not Function. maintained. D. Required Action and D.1 Remove the associated 6 hours ! associated Completion recirculation pump (s) Time not met. from service. l 0H D.2 Be in MODE 2. 6 hours SURVEILLANCE REQUIREMENTS
.......__.........___.........__.----NOTE-------------------------------------
When a channel is placed in an inoperable status solely for performance of required Surve111ances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability. SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL CHECK. 24 hours (ccntinued)
/^\
V - Brunswick Unit 2 3.3-33 Amendment No.
l ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.4.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days . I SR 3.3.4.1.3 Calibrate the trip units. 92 days SR 3.3.4.1.4 Perform CHANNEL CALIBRATION. The 24 months Allowable Values shall be:
- a. Reactor Vessel Water Level-Low Level 2: 2: 101 inches; and
- b. Reactor Vessel Pressure-High:
s 1147 psig. 3.3.4.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST 24 months
, ,) SR including breaker actuation. -(J i
l i l Brunswick Unit 2 3.3-34 Amendment No.
ECCS Instrumentation 3.3.5.1 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.5.1-1. ACTIONS
..............................-------NOTE-------------------------------------
Separate Condition entry is allowed for each channe). CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately
- inoperable. referenced in Table 3.3.5.1-1 for O the channel.
B. As required by B.1 --------NOTES-------- Required Action A.1 1. Only applicable and referenced in in MODES 1, 2, Table 3.3.5.1-1. and 3.
- 2. Only applicable for Functions 1.a, 1.b, 2.a, and 2.b.
Declare supported I hour from feature (s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature (s) in both divisions A10 (continued) O Brunswick Unit 2 3.3-35 Amendment No. I
ECCS Instrumentatien 3.3.5.1 OV ACTIONS CONDITION- REQUIRED ACTION COMPLETION TIME B. (continued) 8.2 ----
---NOTE---------
Only applicable for Functions 3.a and 3.b. Declare High Pressure I hour from Coolant Injection discovery.of (HPCI) System loss of HPCI inoperable. initiation capability AND B.3 Place channel in 24 hours trip. C. As required by C.1 --------NOTES-------- Required Action A.1 1. Only applicable f" and referenced in in MODES 1, 2, Q] Table 3.3.5.1-1. and 3.
- 2. Only applicable for Functions 1.c, l.d. 2.c, 2.d, and 2.f.
Declare supported I hour from feature (s) inoperable discovery of when its redundant loss of-feature ECCS initiation initiation capability capability for is inoperable, feature (s) in both divisions AND C.2 Restore channel to 24 hours OPERABLE status. (continued) O). \,/ ' Brunswick Unit 2 3.3-36 Amendment No. W
ECCS Instrumentatien 3.3.5.1
. ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME l l O. As required by D.1 --------NOTE--------- ! Required Action A.1 Only applicable if and referenced in HPCI pump suction is-Table 3.3.5.1-1. not aligned to the suppression pool. l Declare HPCI System I hour from l inoperable. discovery of I loss of HPCI initiation capability l MD 1 0.2.1 Place channel in 24 hours trip. OE D.2.2 Align the HPCI pump 24 hours suction to the Q,o suppression pool. (continued) 1 l l 1 l l O Brunswick Unit 2 3.3-37 Amendment No.
i4 t ECCS Instrumentation 3.3.5.1 ACTIONS (continued) CONDITION- REQUIRED ACTION COMPLETION TIME E. .As required by E.1 Declare Automatic 1 hour from Required Action A.l' Depressurization discovery of and referenced in System (ADS) valves loss of ADS
. Table 3.3.5.1-1. inoperable . initiation capability in both trip systems AND E.2 Place channel in 96 hours from
~ trip. discovery of inoperable channel concurrent with HPCI or reactor core isolation cooling (RCIC) inoperable AND 8 days (continued)
S 8
ECCS Instrumentatien 3.3.5.1 O. V ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME F. As required by F.1 Declare ADS valves 1 hour from Required Action A.1 inoperable. discovery of and referenced in loss of ADS Table 3.3.5.1-1. initiation capability in j both trip systems F.2 Restore channel to 96 hours from OPERABLE status. discovery of inoperable channel concurrent with HPCI or RCIC inoperable AND 8 days G. Required Action and G.1 Declare associated Immediately associated Completion supported feature (s) Time of Condition B, inoperable. C, D, E, or F not met. i l i
%)
Brunswick Unit 2 3.3-39 Amendment No. ~
f: I i ECCS Instrumentaticn 1: 3.3.5.1 ! -q j SURVEILLANCE REQUIREMENTS ~
-------------------------------------NOTES------------------------------------ !
- 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS !
Function. I
- 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as fo'ilows: (a) for up to 6 hours for Function 3.c; I and (b) for up to 6 hours for Functions other than 3.c provided the l associated Function or the. redundant function maintains ECCS initiation l capability. J SURVEILLANCE FREQUENCY j SR- 3.3.5.1.1 Perform CHANNEL CHECK. 24 hours SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days !
O
-C/ SR 3.3.5.1.3 Calibrate the trip unit. 92 days i
i 1 SR 3.3.5.1.4 Perform CHANNEL CALIBRATION. 24 months j i SR 3.3.5.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.5j.6 Perform CHANNEL FUNCTIONAL TEST. 24 months b-l
*D Brunswick Unit 2 3.3-40 Amendment No.
i l l l ECCS Instrumentatien 3.3.5.1 1 ' .( p)j Table 3.3.5.1 1 (page 1 of 4) V Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS DKlDES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
- 1. Core Spray System
- a. Reactor vessel Water 1,2,3, 4 B SR 3.3.5.1.1 e 13 inches Level --tow Level 3 SR 3.3.5.1.2 I
4 *), 5(*I SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
- b. Drywell 1,2,3 4 B SR 3.3.5.1.1 s 1.8 psig Pressure - 4tish SR 3.3.5.1.2 SR 3.3.5.1.3
{ SR 3.3.5.1.4 i SR 3.3.5.1.5 l l
- c. Reactor Steam Dome 1,2,3 4 C SR 3.3.5.1.1 t 402 psig Pressure -Low SR 3.3.5.1.2 and SR 3.3.5.1.3 5 425 psig SR 3.3.5.1.4 SR 3.3.5.1.5 4(a)* $(a) 4 8 SR 3.3.5.1.1 t 402 psig SR 3.3.5.1.2 and SR 3.3.5.1.3 s 425 psig SR 3.3.5.1.4 SR 3.3.5.1.5
- d. Core Spray Punp 1,2,3, 2 C SR 3.3.5.1.4 t 14 seconds Start -Time Delay 1 per punp SR 3.3.5.1.5 and Relay 4(a), $(a) SR 3.3.5.1.6 s 16 seconds
- 2. Low Pressure Coolant injection (LPCI) Systeca
- a. Reactor Vessel Water Level --Low Level 3 1,2,3, 4 B SR 3.3.5.1.1 .t 13 inches SR 3.3.5.1.2 4(a),5(a) sg 3,3,5,3,3 SR 3.3.5.1.4 SR 3.3.5.1.5
- b. Drywell 1,2,3 4 B SR 3.3.5.1.1 5 1.8 psis Pressure --High SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5 (continued)
(a) When associated othsystem(s) are required to be OPERABLE. l
\
l l Brunswick Unit 2 3.3-41 Amendment No.
'l j
i ECCS Instrumentation j 3.3.5.1 l -0
\j Table 3.3.5.1 1 (page 2 of 4)
Emergency Core Cootlng System Instrunentation I APPLICA8tE CONDITIONS j
. MODES REQUIRED REFERENCED l
- OR OTHER CHANNELS FROM i SPECIFIED PER REQUIRED SURVEILLANCE ALLOWA8LE I FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
- 2. LPCI System (continued)
- c. Reactor Steam Dame Pressure.4 ow 1,2;3 4 C SR 3.3.5.1.1 a 402 pelo 1 SR 3.3.5.1.2 and 1 SR 3.3.5.1.3 5 425 psig i SR 3.3.5.1.4 !
SR 3.3.5.1.5 4(*), 5("I- 4 B SR 3.3.5.1.1 t 402 psis SR 3.3.5.1.2 and SR 3.3.5.1.3 5 425 psig SR 3.3.5.1.4 SR 3.3.5.1.5 I
- d. Reactor Steam Dome 1M ,2(b) , 4 .C SR 3.3.5.1.1 t 302 psig i' Pressure --Low SR 3.3.5.1.2 (Recirculatlon Puy 3(b) gg 3,3,3,g,3 j Discharge Valve SR 3.3.5.1.4 Permissive) SR 3.3.5.1.5
- e. Reactor Vessel Shroud 1,2,3 2 B SR 3.3.5.1.1 t -50 inches ,
Level SR 3.3.5.1.2 l SR 3.3.5.1.3 i SR 3.3.5.1.4 l 4 SR 3.3.5.1.5 l
\ f. RnR Ptmp Start -Time 1,2,3, 4 C SR 3.3.5.1.4 t 9 seconds Delay Relay 1 per punp SR 3.3.5.1.5 and
, 4(a), 3(a) SR 3.3.5.1.6 5 11 seconds
- 3. High Pressure Coolant j Injectior. (HPCI) System
- a. Reactor Vesset Water 1, 4 8 SR 3.3.5.1.1 t 101 inches Level -tow Level 2 SR 3.3.5.1.2 2(*), 3(*) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.5.5.1.5
- b. Drywett 1, 4 B SR 3.3.5.1.1 s 1.8 psig l Preseure --High SR 3.3.5.1.2 2(*),3(*) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5 (continued) ta) When associated subsystem (s) are required to be OPERABLE.
(b) With associated recirculation pump discharge valve or recirculation punp discharge bypass valve open. (c) With reactor steam dome pressure > 150 psto. v\ Brunswick Unit 2 3.3-42 Amendment No.
ECCS Instrumentatien 3.3.5.1 l < Tabte 3.3.5.1 1 (page 3 of 4) ' Emergency Core Coollne system Instrumentation APPLICABLE , CONo!TIONS MtBES OR REQUIRED REFERENCED OTNER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWRELE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
- 3. NPCI system (continued)
- c. Reactor Vessel Water Level --Nigh 1, 2 C st 3.3.5.1.1 s 207 inches sR 3.3.5.1.2 2(*),3(c) g, 3,3,3,3,3 SR 3.3.5.1.4 i sa 3.3.5.1.5 .!
-d. Condensate storage- 1, 2 0 SR 3.3.5.1.2 2 23 feet Tank Level -Low - st 3.3.5.1.4 4 inches 2f*),3(*) sR 3.3.5.1.5
- e. suppression Chamber 1, 2 D r.R 3.3.5.1.2 s 2 feet-Water Level -Nigh SR 3.3.5.1.4 2(*), 3(*) SR 3.3.5.1.5 a
4 Automet{c Depressur{tation -j system (ADS) Trip system A
)
- a. Reactor vessel Water 1, 2 E SR 3.3.5.1.1 L 13 inches Levet -Low Level 3 sR 3.3.5.1.2 -
2(*), 3(*) st 3.3.5.1.3
- SR 3.3.5.1.4 sR 3.3.5.1.5 a
- b. ADS Timer 1, 1 F st 3.3.5.1.4 s 108 seconds 1
~ '
SR 3.3.5.1.5
- c. Reactor *. < seet Water
- 2(*), 3(c) 1, 1 E sR 3.3.5.1.6 sR 3.3.5.1.1 t 153 Inches
[I j tevet -L< - Level 1 sR 3.3.5.1.2 2(*), 3(*) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
- d. Core spray P g 1 2 F SR 3.3.5.1.2 t 102 psig j Discharge SR 3.3.5.1.4 and 1 Pressure -Nigh 2(*), 3(C) $R 3.3.5.1.5 s 130 pois
- e. RNR (LPCI Mode) Pump 1, 4 F SR 3.3.5.1.2 t 102 psig .l Discharge 2 per pump SR 3.3.5.1.4 and )
Pressure -4tish 2(c), 3(c) st 3.3.5.1.5 s 130 psis (continued) j (c) With reactor steam dame pressure > 150 psig. I l I i Brunswick Unit.2 3.3-43 Amendment No.
-)
l ECCS Instrumentatien , 3.3.5.1 ! Table 3.3.5.1 1 (page 4 of 4) l Emergency Core Cooling system Instrtmentation APPLICABLE -CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCil0N ACil0N A.1 REQUIREMENTS VALUE
- 5. Aos Trip system B
- a. Reactor vessel Water 1,. 2 E SR 3.3.5.1.1 t 13 inches Levet -Low Level 3 sR 3.3.5.1.2 2(*), 3(*) sR 3.3.5.1.3 SR 3.3.5.1.4 sR 3.3.5.1.5
- b. ADS Timer 1, 1 F SR 3.3.5.1.4 s 108 seconds sR 3.3.5.1.5 A.
2(CI, 3(c) sR 3.3.5.1.6 O
- c. Reactor vessel Water 1, 1 E sR 3.3.5.1.1 1 153 inches Level --Low Level 1 SR 3.3.5.1.2 2(*), 3(*) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5 ;
- d. Core spray Ptsap 1, 2 F SR 3.3.5.1.2 t 102 psis Discharge SR 3.3.5.1.4 and i Pressure -High 2(*), 3(*) sR 3.3.5.1.5 s 130 psis
- e. RHR (LPCI Mode) Punp 1, 4 F SR 3.3.5.1.2 t 102 pois Discharge 2 per pump sR 3.3.5.1.4 and Pressure --High 2(C), 3(C) SR 3.3.5.1.5 5 130 pals D\
V '- (c) With reactor steam dome pressure > 150 psig, l I l l Brunswick Unit 2 3.3-44 Amendment No.
RCIC Systes Instrumentaticn 3.3.5.2 q i 1 b 3.3 INSTRUMENTATION 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3.3.5.2 The RCIC System instrumentation for each Function in ! Table 3.3.5.2-1 shall be OPERABLE. APPLICABILITY: MODE 1, { MODES 2 and 3 with reactor steam dome pressure > 150 psig. ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME ] l l A. One or more channels A.1 Enter the Condition Immediately A inoperable. referenced in bl Table 3.3.5.2-1 for the channel. B. As required by B.1 Declare RCIC System I hour from Required Action A.1 inoperable. discovery of and referenced in loss of RCIC Table 3.3.5.2 1. initiation capability AND B.2 Place channel in 24 hours trip. C. As required by C.1 Restore channel to 24 hours Required Action A.1 OPERABLE status. and referenced in Table 3.3.5.2-1. (continued) Brunswick Unit 2 3.3-45 Amendment No.
RCIC System Instrumentation 3.3.5.2
/3 kj ACTIONS.(continued)
CONDITION REQUIRED ACTION COMPLETION TIME I D. As required by D.1 ----.-.. NOTE--------- l Required Action A.1 Only applicable if (- and referenced in RCIC pump suction is Table 3.3.5.2-1, not aligned to the suppression pool. I Declare RCIC System I hour from inoperable, discovery of loss of RCIC initiation capability-AND D.2.1 Place channel in 24 hours trip. 0_B D.2.2 Align RCIC pump 24 hours (qj suction to the suppression pool. E. Required Action and E.1 Declare RCIC System Immediately associated Completion inoperable. Time of Condition B, C, or D not met. Brunswick Unit 2 3.3-46 Amendment No.
RCIC System Instrumentaticn 3.3.5.2 I" (,) SURVEILLANCE REQUIREMENTS
-------------------------------------MOTES------------------------------------
.l. Refer to Table 3.3.5.2-1 to determine which SRs apply for each RCIC Function.
- 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 2; and (b) for up to 6 hours for Functions 1 and 3 provided the associated function maintains RCIC initiation capability.
SURVEILLANCE FREQUENCY SR- 3.3.5.2.1 Penform CHANNEL CHECK. 24 hours SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days /~'\ SR 3.3.5.2.3 Calibrate the trip units. 92 days V SR 3.3.5.2.4 Perform CHANNEL CALIBRATION. 24 months SR 3.3.5.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months n Brunswick Unit 2 3.3-47 Amendment No. l l 1
.i . -
RCIC System Instrumentatien 3.3.5.2 0 t Table 3.3.5.2 1 (page 1 of 1)
% Reactor Core Isolation Cooling System Instrumentation q C0KClfl0NS REQUIRED REFERENCED CHANNELS FROM REQUIRED. SURVElLLANCE ALLOWABLE FUNCTION PER FUNCTION ACTION A.1 REQUIREMENTS VAltE
- 1. Reactor Vesset Water 4 8 SR 3.3.5.2.1 a 101 inches Level -Low Level 2 SR 3.3.5.2.2 SR 3.3.5.2.3 -
SR 3.3.5.2.4 SR' 3.3.5.2.5
- 2. atector Vessel Water 2 C SR 3.3.5.2.1 5 207 inches Level -High SR 3.3.5.2.2 SR 3.3.5.2.3 SR 3.3.5.2.4
. SR 3.3.5.2.5
- 3. Condensate Storage 7ank 2 D SR 3.3.5.2.2 1 23 feet
' Level -Low SR -3.3.5.2.4 SR 3.3.5.2.5
/
\
\
Brunswick Unit 2 3.3-48 Amendment No.
Primary Containment Isolation Instrumentation 3.3.6.1 O Q 3.3 INSTRUMENTATION i 3.3.6.1 Primary Containment Isolation Instrumentation j LCO 3.3.6.1 The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABl.E. APPLICABILITY: According to Table 3.3.6.1-1. , 1 ACTIONS
..__.___.....____....__...__....-----NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in 12 hours for channels inoperable. trip. Functions 2.a, a p 2.b, and 6.b CO d g 24 hours for Functions other than Functions 2.a. 2.b, and a 6.b @ B. One or more Functions B.1 Restore isolation 1 hour with isolation capability. capability not maintained. C. Required Action and C.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A Table 3.3.6.1-1 for ; or B not met. the channel. (continued) A 1 Q 3runswick Unit 2 3.3-49 Amendment No. 1 l
I l Primary Containment Isolation Instrumentation 3.3.6.1 1 l ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME l D. As required by D.1 Isolate associated 12 hours , Required Action C.1 main. steam line ! and referenced in (MSL). Table 3.3.6.1-1. Og 0.2.1 Be in MODE 3. 12 hours l AND I D.2.2 .De in MODE 4. 36 hours i 1 E. As required by E1 Be in MODE 2. G hours ! Required Action C.1 1 and referenced in Table 3.3.6.1-1. l 'L F. As required by
' Required Action C.1 F.1 Isolate the affected penetration flow I hour and referenced in path (s).
Table 3.3.6.1-1. l < G. Required Action and G.1 Be in MODE 3. 12 hours ! associated Completion Time for Condition F AND not met. l G.2 Be in MODE 4. 36 hours l E
'As required by Required Action C.1 and referenced in Table 3.3.6.1-1.
(continued) O Brunswick Unit 2 3.3-50 Amendment No.
1 Primary Containment Isolation Instrumentation 3.3.6.1 I v ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME H. As required by H.1 Declare associated I hour Required Action C.1 standby liquid and referenced in control subsystem Table 3.3.6.1-1. (SLC) inoperable. OB H.2 Isolate the Reactor I hour
-Water Cleanup (RWCU)
System. I. As required by I.1 Initiate action to Immediately Required Action C.1 restore channel to and referenced in OPERABLE status. ! Table 3.3.6.1-1. ] OR 1 1.2 Initiate action to Immediately isolate the Residual ( Heat Removal (RHR) Shutdown Cooling (SDC) System. O Brunswick Unit 2 3.3-51 Amendment No.
Primary Containment Isolation Instrumentation 3.3.6.1 ' f'% U SURVEILLANCE REQUIREMENTS __..................__.__....._------NOTES------------------------------------
- 1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.
- 2. When a channel is placed in an inoperable status. solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Functions 2.c, 2.d. 3.a. 3.b, 3.e, 3.f, 3.g, 3.h, 4.a 4.b, 4.e, 4.f. 4.g, A 4.h, 4.1, 4.k, 5.a, 5.b, 5.e, 5.f. and 6.a; and (b) for up to 6 hours for E all other Functions provided the associated Function maintains isolation capability.
SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK. 24 hours SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.6.1.3 Calibrate the trip unit. 92 days SR 3.3.6.1.4 Perform CHANNEL CAllBRAT10N. 92 days SR 3.3.6.1.5 Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.6.1.6 Perform CHANNEL CALIBRATION. 24 months SR 3.3.6.1.7 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months (continued) .I V Brunswick Unit 2 3.3-52 Amendment No. 1 l
Primary Containment Isolatten Instrumentaticn 3.3.6.1 f3 , Q SURVEILLANCE REQUIREMENTS (continued) i SURVEILLANCE FREQUENCY SR 3.3.6.1.8 ------------------NOTES-----------------
- 1. Radiation detectors are excluded. g
- 2. The sensor response time for Functions 1.a, 1.c, and 1 f may be assumed to be the design sensor response time.
Verify the ISOLATION INSTRUMENTATION 24 months on a
' RESPONSE TIME is within limits. STAGGERED TEST BASIS SR 3.3.6.1.9 Perform CHANNEL FUNCTIONAL TEST. 24 months A i
)
I l (3 Brunswick Unit 2 3.3-53 Amendment No.
Primary Containment Isolation Instrumentatien 3.3.6.1 O ( table 3.3.6.1 1 (page 1 of 5) Primary Containment Isolation Instrunentation i 1 APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVE!LLANCE ALLOWABLE FUNCTION COND!il0NS SYSTEM ACil0N C.1 REQUIREMENTS VALUE
- 1. Main Steam Line Isolation
- a. Reactor vessel Water 1,2,3 2 0 SR 3.3.6.1.1 t 13 inches Level -Low Level 3 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 SR 3.3.6.1.8
- b. Main Steam Line 1 2 E SR 3.3.6.1.1 t 825 psig Pressure -Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- c. Main Steam Line 1,2,3 2 per D SR 3.3.6.1.1 5 138% rated Flow -High MSL SR 3.3.6.1.2 steam flow SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 i SR 3.3.6.1.8 l
- d. Condenser Vacuun -Low 1, 2 D SR 3.3.6.1.1 t 7.5 inches SR~ 3.3.6.1.2 Hg vacuun n' 2(a), 3(a) SR 3.3.6.1.3
'/ SR 3.3.6.1.6 SR 3.3.6.1.7 ,
1
- e. Main Steam Isolation 1,2,3 2 D SR 3.3.6.1.2 5 197'F I Valve Pit SR 3.3.6.1.6 Tenperature -High SR 3.3.6.1.7
- f. Main Steam Line 2.3 2 0 SR 3.3.6.1.1 5 33% rated l Flow --High (Not in SR 3.3.6.1.2 steam flow i Run) SR 3.3.6.1.3 SR 3.3.6.1.6 '
SR 3.3.6.1.7 SR 3.3.6.1.8
- 2. Primary Containment Isolation
- a. Reactor vessel Water 1,2,3 2 G SR 3.3.6.1.1 2 153 inches Level -Low Level 1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- b. Drywell Pressure -High 1, .!,3 2 C SR 3.3.6.1.1 5 1.8 paig SR 3.3.6.1.2 SR 3.5.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 (continued)
(a) With any turbine stop valve not closed. A Brunswick Unit 2 3.3-54 Amendment No.
F Primary Containment Isolation Instrumentation 3.3.6.1
'Oj fable 3.7.6.1 1 (page 2 of 5)
Primary Containment Isotation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CON 0lTIONS STSTEM ACTION C.1 REQUIREMENTS VALUE
- 2. Primary containment isolation (continued)
- c. Main Stack 1,2,3 1 F SR 3.3.6.1.2 (b)
Radiation --High SR 3.3.6.1.6 SR 3.3.6.1.7 SR 3.3.6.1.8
- d. Reactor Building 1,2,3 1 C SR 3.3.6.1.1 s 16 W/hr Exhaust SR 3.3.6.1.2 Radiation -High SR 3.3.6.1.6 SR 3.3.6.1.7
- 3. High Pressure Coolant l Injection (HPCI) Syste.n I Isolation I
- a. HPCI Steam L6ne 1,2,3 1 F SR 3.3.6.1.1 5 275% rated Flow --High SR 3.3.6.1.2 steam flow SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- b. HPCI Steam Line 1,2,3 1 F SR 3.3.6.1.6 t 4 seconds and Flow -High Time Delay SR 3.3.6.1.7 s 12 seconds Relay SR 3.3.6.1.9 D c. HPCI Steam Supply L!ne 1,2,3 2 F SR 3.3.6.1.2 a 104 psig Pressure -low SR 3.3.6.1.4 SR 3.3.6.1.7
- d. HPCI Turbine Exhaust 1,2,3 2 F SR 3.3.6.1.2 s 9 psig Olaphragm SR 3.3.6.1.6 Pressure -High SR 3.3.6.1.7
- e. Drywett Pressure -Hlph 1,2,3 1 F SR 3.3.6.1.1 5 1.8 psis SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- f. HPCI Steam Line Area 1,2,3 1 F SR 3.3.6.1.5 5 200*F Tenperature --High SR 3.3.6.1.6 SR 3.3.6.1.7
- g. HPCI Steam Line Tunnel 1,2,3 1 F SR 3.3.6.1.5 s 200*F Ambient SR 3.3.6.1.6 Temperature -High SR 3.3.6.1.7
- h. HPCI Steam Line Tunnel 1,2,3 1 F SR 3.3.6.1.5 5 50*F Differential SR 3.3.6.1.6 Temperature --High SR 3.3.6.1.7 (continued)
(b) Allowable Value established in accordance with the methodology in the Offsite Dose Calculation Manual. 9 l r O C/ l Brunswick Unit 2 3.3-55 Amendment No. I
Primary Containment Isolation Instrumentation 3.3.6.1' [
\ Tabte 3.3.6.1 1 (page 3 of 5)
Primary Contairveent Isolation Instrumentation APPLICABLE- CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWASLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE
- 3. HPCl System isolation (continued)
I. HPCI Equipment Area 1,2,3 2 F SR 3.3.6.1.5 5 175'F Temperature -High SR 3.3.6.1.6 SR 3.3.6.1.7
- 4. Reactor Core Isolation CootIng (RCIC) Systaa Isolation
- a. RCIC Steam Line Flow -High 1,2,3 1 F SR 3.3.6.1.1 5 275% rated SR 3.3.6.1.2 steam flow SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- b. RCIC Steam Line 1,2,3 1 F SR 3.3.6.1.6 2 4 seconds Flow -High Time Delay SR 3.3.6.1.7 and Retsy SR 3.3.6.1.9 s 12 seconds
- c. RCIC Steam Stpply Line 1,2,3 2 F SR 3.3.6.1.2 5 53 psig Pressure -Low SR 3.3.6.1.4 SR 3.3.6.1.7 (v) d. RCIC Turbine Exhaust Diaphragm 1,2,3 2 F SR 3.3.6.1.2 SR 3.3.6.1.6 5 6 psig Pressure --High SR 3.3.6.1.7
- e. Drywell Pressure -High 1,2,3 1 F SR 3.3.6.1.1 s 1.8 psig SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7
- f. RCIC Steam Line Area 1,2,3 1 SR 3.3.6.1.5 s 175'F Temperature -High F
SR 3.3.6.1.6 -d SR 3.3.6.1.7
- s. RCic Steam Line Tunnet 1,2,3 1 F SR 3.3.6.1.5 5 200'F Ambient SR 3.3.6.1.6 f eeperature -High SR 3.3.6.1.7
- h. RCIC Steam Line Tunnet 1,2,3 1 F SR 3.3.6.1.5 s 30 minutes and Area SR 3.3.6.1.6 Temperature -High Time SR 3.3.6.1.7 Delay
- 1. RCIC Steam Line Tunnel 1,2,3 1 F SR 3.3.6.1.5 5 50'F Differential SR 3.3.6.1.6 Temperature -High SR 3.3.6.1.7 (continued; k
Brunswick Unit 2 3.3-56 Amendment No. l
i l I t Primary Containment Isolaticn Instrumentation { 3.3.6.1 ! , m I Table 3.3.6.1 1 (pnge 4 of 5)
~rimary containment Isolation Instrumentation APPLICA8LE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE
- 4. RCIC System Isolation (continued)
J. RCIC Equipment Area 1,2,3 2 F SR 3.3.6.1.5 5 175'F Teeperature .-High SR 3.3.6.1.6 - SR 3.3.6.1.7
- k. RCIC Equipment Area 1,2,3 1 F SR 3.3.6.1.5 s 50*F Differential SR 3.3.6.1.6 Tenperature --Nigh SR 3.3.6.1.7
- 5. Reactor Water cleanup (RWCU) System Isolation
- a. Differential 1,2,3 1 F SR 3.3.6.1.5 5 73 gpm Flow -High SR 3.3.6.1.6 SR 3.3.6.1.7 1
- b. Differential 1,2,3 1 F SR 3.3.6.1.5 s 30 minutes i Flow -High Time Delay SR 3.3.6.1.6 '
SR 3.3.6.1.7
- c. Area 1,2,3 3 F SR 3.3.6.1.5 s 150*F i fg Tenperature --High 1 per SR 3.3.6.1.6 j V, room SR 3.3.6.1.7 !
- d. Area Ventilation 1,2,3 2 F SR 3.3.6.1.5 5 50'F i Differential SR 3.3.6.1.6 Temperature --High SR 3.3.6.1.7
- e. Piping outside RWCU 1,2,3 1 F SR 3.3.6.1.5 s 120*F I Rooms Area SR 3.3.6.1.6 l Tenperature -High SR 3.3.6.1.7 j
- 1. SLC System Initiation 1,2 1(C) H SR 3.3.6.1.7 NA
- g. Reactor vessel Water 1,2,3 2 F SR 3.3.6.1.1 t 101 inches Level -Low Level 2 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 (continued)
(c) SLC System Initiation only inputs into one trip system. p. iv Brunswick Unit 2 3.3-57 Amendment No.
Primary Containment Isolation Instrumentation i 3.3.6.1 .q kj Table 3.3.6.1 1 (page 5 of 5) Primary Contalrunent isolation Instrumentation APPLICABLE CONDIT10NS MODES OR REQUIRED REFERENCED OTHER CHANNELS TROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWASLE FUNCTION' CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE
- 6. RNR Shutdown Cooling System Isolation
- a. Reactor Steam Dome - 1,2,3 I 1 F SR 3.3.6.1.2 5 137 psig i Pressure -High SR 3.3.6.1.4 SR 3.3.6.1.7
- b. Reactor vessel Water 3,4,5 2(d) i SR 3.3.6.1.1 t 153 Inches Level -Low Level 1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 (d) In MODES 4 and 5, provided RHR Shutdown Cooling System integrity maintalned, only one channel per trip system with an isolation signal available to one RHR shutdown cooling pump suction isolation valve Is required.
l l l N I C Brunswick Unit 2 3.3-58 Amendment No.
Secondary Containment Isolation Instrumentaticn 3.3.6.2
/7 V. 3.3 INSTRUMENTATION 3.3.6.2 Secondary Containment Isolation Instrumentation 4
LC0 3.3.6.2 The secondary containment isolation instrumentation for each Function in Table 3.3.6.2-1 shall be OPERABLE. i 1 APPLICABILITY: According to Table 3.3.6.2-1. 1 ACTIONS I
.......____............--------------NOTE-------------------------.-------..--
Separate Condition entry is allowed for each channel. l 1 CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in ' 12 hours for inoperable. trip. Function 2 O i i AND
%/
24 hours for Functions other than Function 2 B. One or more Functions B.1 Restore isolation 1 hour with isolation capability. capability not maintained. C. Required Action and C.1.1 Isolate the 1 hour associated Completion associated Time not met. penetration flow paths. 08 i (continued) q ', Brunswick Unit 2 3.3-59 Amendment No.
1 Secondary Containment' Isolation Instrumentation 3.3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. -(continued) C.I.2 Declare associated I hour secondary containment isolation dampers inoperable. ! btLD C.2.1 Place the associated I hour standby gas treatment (SGT) subsystem (s) in operation. O_B C.2.2 Declare associated I hour-SGT subsystem (s) inoperable. LO SURVEILLANCE REQUIREMENTS
.___________....____.------N0TES------------------------------------
- 1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.
- 2. When a channel is placed in an inoperable status solely for performance of required _Surveillances, entry into associated Conditions and Requireo Actions may be delayed as follows: (a) for up to 2 hours for Function 3 and (b) for up to 6 hours for Functions 1 and 2 provided the associated Function maintains isolation capability.
SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. 24 hours (continued) f I p Brunswick Unit 2 3.3-60 Amendment No. L i
Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.6.2.3 Calibrate the trip unit. 92 days SR 3.3.6.2.4 Perform CHANNEL CALIBRATION. 24 months SR 3.3.6.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months O 1 I O Brunswick Unit 2 3.3-61 Amendment No.
Secondary Containment Isolation Instrumentation 3.3.6.2
.:( Table 3.3.6.2 1 (page 1 of 1)
(' Secondary Contaltunent isolation Instrtmentation APPLICA8LE ~ HODES OR AEQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE - ALLOWASLE FUNCTION CONDITIONS TRIP SYSTEN REQUIREMENTS VALUE
- 1. Reactor vessel Water 1,2,3, 2 SR 3.3.6.2.1 2 101 inches Level -Low Level 2 SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4 SR 3.3.6.2.5
- 2. Drywell Pressure -4tigh 1,2,3 2 SR 3.3.6.2.1 5 1.8 psig SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4 SR 3.3.6.2.5
- 3. Reactor Building Exhaust 1,2,3, 1 SR 3.3.6.2.1 5 16 mR/hr Radiatlon --High sa),(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.5 (a)~ During operations with a potential for draining the reactor vessel.
(b) During CORE ALTERATIONS and during aiovement of irradiated fuel assemblies in secondary containment. I l l 1 i i l IA. Brunswick Unit 2 3.3-62 Amendment No. l b o
L CREV System Instrumentaticn 3.3,7.1 { 3.3 INSTRUMENTATION 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation LCO 3.3.7.1 Two channels per trip system of the Control Building Air /d Intake Radiation-High function shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ACTIONS
...................................--NOTE-------------------------------------
Separate Condition entry is Wwed for each channel. , CONDITION REQUIRED ACTION COMPLETION TIME V A. One or more channels A.1 Place one CREV 7 days . inoperable. subsystem in the ! radiation / smoke protection mode of operation, b'
- B. CREV System initiation B.1 Place one CREV 1 hour capability not subsystem in the maintained. radiation / smoke protection mode of operation.
l l' o (" ) Brunswick Unit 2 3.3-63 Amendment No. t L
i CREV System Instrumentation 3.3.7.1 ( ,) _ ' SURVEILLANCE REQUIREMENTS
~
..................--NOTE----------------------------------- -
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains CREV initiation capability. SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK. 24 hours SR 3.3.7.1.2 Perform CHANNEL FUhCTIONAL TEST. 92 days SR 3.3.7.1.3 Perform CHANNEL CAllBRATION. The 24 months Allowable Value shall be s 27 mR/hr.
~O \' / SR 3.3.7.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months 1
i l l 1 i l l' l V Brunswick Unit 2 3.3-64 Amendment No. f
Condenser Vacuum Pump Isolation Instrumentation 3.3.7.2 i
~( 3.3 INSTRUMENTATION 3.3.7.2 Condenser Vacuum Pump Isolation Instrumentation LCO 3.3.7.2 Four channels of the Main Steam Line Radiation-High Function for condenser vacuum pump isolation shall be OPERABLE.
APPLICABILITY: MODES 1 and 2 with a condenser vacuum pump in service. b ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels inoperable. A.1 Restore channel to OPERABLE status. 12 hours $ 03 A.2 --------NOTE--------- Not applicable if inoperable channel is the result of an inoperable condenser vacuum pump trip breaker or isolation valve. Place channel or 12 hours associated trip g system in trip. -- (continued) l lO Brunswick Unit 2- 3.3-65 Amendment No.
Condenser Vacuum Pump Isolation Instrumentatien 3.3.7.2
/3
- Q ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Isolate condenser 12 hours associated Completion vacuum pumps. Time of Condition A not met. QR OJ B.2 Isolate main steam 12 hours lines. Condenser vacuum pump isolation capability QR not maintained. B.3 Be in MODE 3. 12 hours SURVEILLANCE REQUIREMENTS
.....................................N0TE---------------------------------- --
When a channel is placed in an inoperable status solely for performance of d O required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains condenser vacuum pump isolation capability. SURVEILLANCE FREQUENCY SR 3.3.7.2.1 Perform CHANNEL CHECK. 24 hours SR 3.3.7.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.7.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Value shall be s 6 x background. (continued) Brunswick Unit 2 3.3-66 Amendment No.
4 i Condenser Vacuum Pump Isolation Instrumentatien 3.3.7.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l
- _ ~ .
SR 3.3.7.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 24 months including condenser vacuum pump trip ; breaker and isolation valve actuation. ! I O a l l I I
)
O Brunswick Unit 2 3.3-67 Amendment No.
U. t i LOP Instrumentation j 3.3.8.1 ,Q-- Q 3.3 INSTRUMENTATION'
'3.3.8.1 Loss of Power (LOP) Instrumentation
. LCO .3.3.8.1 The LOP instrumentation for each function in Table 3.3.8.I-1 shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3, When the associated diesel generator is required to be OPERABLE by LCO 3.8.2, "AC Sources-Shutdown." ACTIONS
......................__.............N0TE----------------------------------..-
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME
\
A. One or more channels A.1 Place channel in I hour-inoperable, trip. l B. Required Action and B.1 Declare associated Immediately , associated Completion diesel generator (DG) ! d Time not met. inoperable.
)
Brunswick Unit 2 3.3-68 Amendment No.
LOP Instrumentation 3.3.8.1 (x ( ,) SURVEILLANCE REQUIREMENTS
-------------------------------------NOTES------------------------------------
- 1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.
- 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours provided: (a) for Function 1, the associated Functions maintains initiation capability for three DGs; and (b) for Function 2, the associated Function maintains DG initiation capability.
SURVEILLANCE FREQUENCY SR 3.3.8.1.1 Perform CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.8.1.2 Perform CHANNEL CALIBRATION. 18 months n ! ) 's_ / SR 3.3.8.1.3 Perform CHANNEL CALIBRATION. 24 months SR 3.3.8.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months C'\
\,)
Brunswick Unit 2 3.3-69 Amendment No.
LOP Instrumentation !~ 3.3.8.1 I: Table 3.3.8.1-1 (page 1 of 1) . Loss of Power Instrumentation l REQUIRED CHANNELS .SURVE!LLANCE ALLOWA8LE FUNCTION PER BUS- REQUIREMENTS VALUE
- 1. 4.16 kV Emergency Bus Undervoltage
- ., .(Loss of Voltage)
- a. Bus Undervoltage 1 SR 3.3.8.1.2 t 3115 y and 5 3400 V SR 3.3.8.1.4
- b. Time Delay 1 SR 3.3.8.1.2 t 0.5 seconds and SR 3.3.8.1.4 s 2.0 seconds
- 2. 4.16 kV Emergency Bus Undervoltage (Degraded Voltage)
- a. Bus Undervoltage 3 SR 3.3.8.1.1 2 3706 V and 5 3748 Y SR 3.3.8.1.3 SR 3.3.8.1.4
- b. Ilme Delay 3 SR 3.3.8.1.1 2 9.0 sce n e and SR 3.3.8.1.3 s 11.0 seconds SR 3.3.8.1.4 I
l ! i i l o v Brunswick Unit 2 3.3-70 Amendment No.
i RPS Electric Power Monitcring ' 3.3.8.2 O (/ 3.3 INSTRUMENTATION 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring LCO 3.3.8.2 Two RPS electric power monitoring assemblies shall be OPERABLE for each inservice RPS motor generator set or alternate power supply. APPLICABILITY: MODES I and 2, MODES 3, 4, and 5 with any control rod withdrawn from a core d; cell containing one or more fuel assemblies. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or both inservice A.1 Remove associated 72 hours power supplies with inservice power one electric power supply (s) from monitoring assembly service. . inoperable. B. One or both inservice B.1 Remove associated I hour power supplies with inservice power both electric power supply (s) from monitoring assemblies service. i inoperable. l C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion ; Time of Condition A ' or B not met in MODE 1 or 2. (continued) Brunswick Unit 2 3.3-71 Amendment No.
, RPS Electric Pcwer M:nitoring-
- l. 3.3.8.2
) ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME . t l D. Required Action and D.1 Initiate action to Immediately l associated Completion fully insert all - Time of Condition A insertable control-or B not met in rods in core cells MODE 3, 4, or 5 with any control rod containing one or more fuel assemblies. g . withdrawn from a core I cell containing one or J more fuel assemblies. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l l SR 3.3.8.2.1 ------------------NOTE------------------- Only required to be performed prior to t , v entering MODE 2 from MODE 3 or 4, when in 1 MODE 4 for a 24 hours. Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.8.2.2 Perform CHANNEL CALIBRATION for each RPS 24 months motor generator set electric power monitoring assembly. The Allowable Values shall be: l
- a. Overvoltage s 129 V.
- b. Undervoltage a 105 V.
- c. Underfrequency a 57.2 Hz.
(continued) i L (~h r Brunswick Unit 2 3.3-72 Amendment No. I
RPS Electric Power Monttcring 3.3.8.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.8.2.3 Perform CHANNEL CALIBRATION for each RPS 24 months alternate power supply electric power monitoring assembly. The Allowable Values shall be: l a. Overvoltage s 132 V.
- b. Undervoltage a 108 V.
- c. Underfrequency 2 57.2 Hz.
SR 3.3.8.2.4 Perform a system functional test. 24 months O I l I i l 1 I O Brunswick Unit 2 3.3-73 Araendment No. i l' !
SRkInstrumentatfor. B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.1 and SR 3.3.1.2.3 (continued) REQUIREMENTS The Frequency of once every 12 hours for SR 3.3.1.2.1 is
' based on operating experience that demonstrates channel failure is rare. While in MODES 3 and 4, reactivity changes are not expected; therefore, the 12 hour Frequency is relaxed to 24 hours for SR 3.3.1.2.3. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by tLe LCO.
SR 3.3.1.2.2 To provide adequate coverage of potential reactivity changes in the core, one SRM is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed, and the other OPERABLE SRM must be in an adjacent quadrant containing fuel. Note 1 states that the SR is required to be met only during CORE ALTERATIONS. It is not required to 84 be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a i review of plant logs to ensure that SRMs required to be OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE. Note 2 clarifies that more than one of the three A requirements can be met by the same OPERABLE SRM. The /D - 12 hour Frequency is based upon operating experience and supplements operational controls over refueling activities that include steps to ensure that the SRMs required by the LC0 are in the proper quadrant. SR 3.3.1.2.4 This Surveillance consists of a verification of.The'SRM instrument readout to ensure that the SRM reading is greater
- than a specified minimum count rate with the detector inserted to the normal operating level,- which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient " source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate.
(conting_edl e O Brunswick Unit 2 B 3.3-42 Revision No.
SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.4 (continued) REQUIREMENTS To accomplish ~this, the SR is modified by Note 1 that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical. In addition, Note 2 states that this requirement does not have to be met during a core spiral offload. A core spiral offload encompasses offloading a cell on the edge of a continuous fueled region (the core cell can be offloaded in any sequence). If the core is being unloaded in this manner, the various core configurations encountered will not be critical. The Frequency is based upon channel redundancy and other information available in the control room, and ensures that , the required channels are frequently monitored while core l reactivity changes are occurring. When no reactivity i changes are in progress, the Frequency. is relaxed from dl 12 hours to 24 hours. l SR 3.3.1.2.5 and SR 3.3.1.2.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the . associated channel will function properly. SR 3.3.1.2.5 is ' required in MODE 5, and the 7 day Frequency ensures that the ' channels are OPERABLE while core reactivity changes could be in progress. This Frequency is reasonable, based on operating experience and on other Surveillances (such as a l CHANNEL CHECK), that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. SR 3.3.1.2.6 is required to be met in MODE 2 with IRMs on Range 2 or below, and in MODES 3 and 4. Since core reactivity changes do not normally take place in MODES 3 and 4 and core reactivity changes are due only to control rod movement in MODE 2, the Frequency is extended from 7 days to 31 days. The 31 day Frequency is based on operating experience and on other Surveillances (such as b CHANNEL CHECK) that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. b (continued) O Brunswick Unit 2 B 3.3-43 Revision No.
SRM Instrumentatisn 4 B 3.3.1.2 G/ BASES SURVEILLANCE SR 3.3.1.2.5 and SR 3.3.1.2.6 (continued) REQUIREMENTS The Note to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability (THERMAL POWER decreased to IRM Range 2 or below). The SR must be perfomed within 12 hours after IRMs are on Range 2 or below. The allowance to enter the Applicability with the 31 day Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to 8 perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the pla'nt would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances. i SR 3.3.1.2.7 s Performance of a CHANNEL CALIBRATION at a Frequency of 4
) 24 months verifies the performance of the SRM detectors and associated circuitry. The Frequency considers the plant '
conditions required to perform the test, the ease of ; performing the test, and the likelihood of a change in the j system or component status. The neutron detectors are ) excluded from the CHANNEL CALIBRATION (Note 1) because they ) cannot readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant sensitivity over the range and with an accuracy specified for a fixed useful life. Note 2 to the Surveillance allows the Surveillance to be delayed untti entry into the specified condition of the Applicability. The SR must be performed in MODE 2 within 12 hours of entering MODE 2 with IRMs on Range 2 or below. l The allowance to enter the Applicability with the 24 month I Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, (continued) O O Brunswick Unit 2 B 3.3-44 Revision No.
SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.7 (continued) REQUIREMENTS the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances. REFERENCES None. O l i l O Brunswick Unit 2 B 3.3-45 Revision No.
PBDS B 3.3.1.3
'8 3.3 INSTRUMENTATION B 3.3.1.3 Period Based Detection System (PBDS)
BASES BACKGROUND ' General Design Crl'teria 12 requires protection'of fuel thermal safety limits from conditions caused by neutronic/ thermal hydraulic instability. Neutronic/ thermal hydraulic instabilities can result in power oscillations which could result in exceeding the MCPR Safety Limit (SL). The MCPR SL ensures that at least 99.97, of the fuel rods avoid boiling transition during normal operation and during an anticipated operational occurrence (A00) (refer to the Bases for SL 2.1.1.2). The PBDS provides the operator with an indication that conditions consistent with a significant degradation in the stability performance of the reactor core has occurred and the potential for imminent onset of neutronic/ thermal hydraulic _ instability may exist. Indication of such degradation is cause for.the operator to initiate an immediate reactor scram if the reactor'is being operated in either the Restricted Region or Monitored Region.. The O Restricted Region and Monitored Region are defined in the COLR. - The PBDS instrumentation of the Neutron Monitoring System (NMS) consists of two channels. PBDS channel A includes input from 13 local power range monitors (LPRMs) within the , reactor core and PBDS channel B includes input from 11 LPRMs i within the reactor core. All LPRMs are utilized T om each i of the axial levels except for the D level detectm. These ') inputs are continually monitored by the PBDS for variations ' in the neutron flux consistent with the onset of neutronic/ thermal hydraulic instability. Each channel includes separate local indication and separate control room ; High-High Alarms. While, this LC0 specifies OPERABILITY requirements only for one monitoring and indication chtnnel of the PBDS, if both are OPERABLE, a High-High Alarm from either channel results in the need for the operator to take actions. The primar) PBDS component is a card in the NMS with analog inputs and digital processing. The PBDS card has an automatic self-test feature to periodically test the hardware circuit. The self-test functions are executed , during their allocated portion of the executive loop 1 (continuedl Brunswick Unit 2 B 3.3-46 Revision No. i
PBDS B 3.3.1.3 (3 U BASES BACKGROUND sequence. Any self-test failure indicating loss of critical (continued) function results in a common control room " Inoperative" alarm. The inoperable condition is also displayed by an indicating light on the card front panel. A manually initiated internal test sequence can be actuated via a recessed push button. This internal test consists of simulating alarm and inoperable conditions to verify card OPERABILITY. Further descriptions of the PBDS are provided in References 1 and 2.
' Actuation of the PBDS High-High Alarm is not postulated to occur due to neutronic/ thermal hydraulic instability during operation outside the Restricted Region and the Monitored Region. Periodic perturbations can be introduced into the thermal hydraulic behavior of the reactor core from external sources such as recirculation system components and the pressure and feedwater control systems. These perturbations can potentially drive the neutron flux to oscillate within a frequency range expected for neutronic/ thermal hydraulic instability. The presence of such oscillations may be recognized by the period based algorithm of the PBDS and could result in a High-High Alarm. Actuation of the PBDS High-High Alarm outside the Restricted Region and the Monitored Region indicate the presence of a source external to the reactor core and are not indications of neutronic/ thermal hydraulic instability.
APPLICABLE Analysis, as described in Section 4 of Reference 1, SAFETY ANALYSES confirms that A00s initiated from outside the Restricted Region without stability control and from within the Restricted Region with stability control are not expected to result in neutronic/ thermal hydraulic instability. The stability control applied in the Restricted Region (refer to LC0 3.2.3, " Fraction of Core Boiling Boundary (FCBB)") is established to prevent neutronic/ thermal hydraulic instability during operation in the Restricted Region. Operation in the Monitored Region is only susceptible to instability under operating conditions beyond those analyzed in Reference 1. The types of transients specifically evaluated are loss of flow and coolant temperature decrease which are limiting for the onset of instability. The initial conditions assumed in the analysis are reasonably conservative and the immediate post-event reactor conditions are significantly stable. However, these assumed initial conditions do not bound each individual parameter p (continued) G Brunswick Unit 2 8 3.3-47 Revision No.
e PBDS l B 3.3.1.3-
)
BASES , APPLICABLE which impacts stability performance (Ref. 1). The PBOS ! SAFETY ANALYSES instrumentation provides the o>erator with an indication (continued) that conditions consistent.witi a significant degradation in the stability performance of the reactor core has occurred l and the potential for.iminent onset of neutronic/ thermal l ' hydraulic instability may exist. Such conditions are only postulated to result from events initiated from initial conditions beyond the conditions assumed in the safety analysis- (refer to Section 4, Ref.1).
~
l The PBDS has no safety function and is not assumed to i function during any UFSAR design basis accident or transient analysis. However, the PBDS provides the only indication of I the iminent onset of neutronic/ thermal hydraulic l instability during operation in regions of the operating i domain potentially susceptible to instability. Therefore, the PBDS is included in the Technical Specifications.
\
LC0 One PBDS channel is required to be OPERABLE with a minimum-of eight LPRM inputs to monitor reactor neutron flux for. 4 indications of iminent onset of neutronic/ thermal hydraulic l instability. A PBDS channel may be considered OPERABLE with A six LPRM inputs when the distribution of OPERABLE LPRMs , provides: a) at least one OPERABLE LPRM in each core l quadrant or b) at least two OPERABLE LPRMs in the core ! quadrant opposite any core quadrant with no OPERABLE LPRMs. The required distribution of the LPRMs when a PBDS channel is considered OPERABLE with as few as six OPERABLE LPRMs ensures a minimum of two OPERABLE LPRMs in opposite core quadrants. This distribution ensures that,-for all ; postulated orientations and modes of oscillation, there are i at least two OPERABLE LPRMs in the core quadrants in which the local neutron flux will oscillate with a frequency ; within the range monitored by the PBDS. OPERABILITY ; requires the ability for the operator to be imediately alerted to a High-High Alarm. This is accomplished by the ; instrument channel control room alarm. The LCO also I requires reactor operation be such that the High-High Alarm j is not actuated by any OPERABLE PBDS instrumentation channel. i
-APPLICABILITY At least one of two PBDS instrumentation channels is required to be OPERABLE during operation in either the Restricted Region or the Monitored Region specified in the 4 i
(continued) l O Brunswick Unit 2 B 3.3-48 Revision No. I
1 PBDS B 3.3.1.3 q b BASES-i APPLICABILITY COLR. Similarly, operation with the PBDS High-High Alarm j (continued) of any OPERABLE PBDS instrumentation channel is not allowed ' in the Restricted Region or the Monitored Region. Operation in these regions is susceptible to . instability (refer to the Bases for LCO 3.2.3 and Section 4 of Ref.1). OPERABILITY of at least one PBDS instrumentation channel and operation with no indication of a PBDS High-High Alarm from any 1 OPERABLE PBDS instrumentation channel is therefore required during operation in these regions. The boundary of the Restricted Region in the Applicability of this LCO is analytically established in terms of thermal power and core flow. The Restricted Region is defined by the APRM Flow Biased Simulated Thermal Power-High Control Rod Block setpoints, which are a function of reactor recirculation drive flow. The Restricted Region Entry Alarm (RREA) signal is generated by the Flow Control Trip Reference (FCTR) card using the APRM Flow Biased Simulated Thermal Power-High Control Rod Block setpoints. As a result, the RREA is coincident with the Restricted Region boundary under all anticipated operating conditions when the setpoints are not " Setup," and provides the indication regarding entry into the Restricted Region. However, APRM O Flow Biased Simulated Thermal Power-High Control Rod Block signals provided by the FCTR card, that are not coincident with the Restricted Region boundary, do not generate a valid RREA. The Restricted Region boundary for this LCO
-Applicability is specified in the COLR.
When the APRM Flow Biased Simulated Thermal Power-High
, Control Rod Block setpoints are " Setup" the applicable setpoints used to generate the RREA are moved to the interior boundary of the Restricted Region to allow ,
controlled operation within the Restricted Region. While i the setpoints are " Setup" the Restricted Region boundary remains defined by the normal APRM Flow Biased Simulated Thermal Power-High Control Rod Block setpoints. Parameters such as reactor power and core flow available at the reactor controls, may be used to provide immediate confirmation that entry into the Restricted Region could reasonably have occurred. The Monitored Region in the Applicability of this LCO is analytically established in terms of thermal power
# and core flow. However, unlike the Restricted Region boundary the Monitored Region is not specifically monitored by plant instrumentation to provide automatic indication of entry into the region. Therefore, the Monitored Region (continued)
O Brunswick Unit 2 B 3.3-49 Revision No.
PBDS B 3.3.1.3 BASES L
. APPLICABILITY buundary is defined solely in terms of thermal- power and l
(continued) core flow. The Monitored Region boundary for this LCO l Applicability is specified in the COLR. Operation outside the Restricted Region and the Monitored Region is not susceptible to neutronic/ thermal hydraulic ! instability even under extreme postulated conditions.
' ACTIONS M If at any time while in the Restricted Region or Monitored Region, an OPERABLE PBDS instrumentation channel indicates a High-High Alarm, the operator is required to initiate an immediate reactor scram. Verification that the High-High-Alarm is valid may be performed without delay against another output from a PBDS card observable from the reactor controls in the control room prior to the manual reactor scram. This provides assurance that core conditions leading to neutronic/ thermal hydraulic instability will be-mitigated. This Required Action and associated Completion Time does not allow for evaluation of circumstances leading to the High-High Alarm prior to manual initiation of i- O reactor scram.
I B.1 and B.2 Operation with the APRM Flow Biased Simulated Thermal Power-High Function (refer to LC0 3.3.1.1, Table 3.3.1.1-1, Function 2.b) " Setup" requires the stability control applied in the Restricted Region (refer to LCO 3.2.3) to be met. Requirements for operation with the stability control met are established to prevent reactor thermal hydraulic instability during operation in the Restricted Region. When the APRM Flow Biased Simulated Thermal Power-High Control Rod Block setpoints are not " Setup" uncontrolled entry into the Restricted Region is identified by receipt of a valid RREA. Immediate confirmation that the RREA is valid and indicates an actual entry into the Restricted Region may be performed without delay. Immediate confirmation constitutes observation that plant parameters immediately available at the reactor controls (e.g., core power and core flow) are reasonably consistent with entry into the Restricted Region. This immediate confirmation may also constitute recognition that plant parameters are rapidly changing during a (continued) O Brunswick Unit 2 B 3.3-50 Revision No.
I PBDS B 3.3.1.3 g) - BASES ACTIONS B.1 and B.2 (continued) transient (e.g., a recirculation ) ump trip) which could reasonably result in entry into tie Restricted Region. While the APRM Flow Biased Simulated Thermal Power-High Control Rod Block setpoints are " Setup," operation in the Restricted Region may be confirmed by use of plant parameters such as reactor power and core flow available at the reactor controls. With the required PBDS channel inoperable while in the Restricted Region, the ability to monitor conditions indicating the potential for imminent onset of neutronic/ thermal hydraulic instability as a result of unexpected transients is lost. Therefore, action must be immediately initiated to exit the Restricted Region. Exit of the Restricted Region can be accomplished by control rod insertion and/or recirculation flow increases. Actions to restart an idle recirculation loop, withdraw control rods or reduce recirculation flow may result in unstable reactor 5 conditions and are not allowed to be used to comply with this Required Action. p) (" The time required to exit the Restricted Region will depend on existing plant conditions. Provided efforts are begun without delay and continued until the Restricted Region is exited, operation is acceptable based on the low probability of a transient which degrades stability performance occurring simultaneously with the required PBDS channel inoperable. Required Action B.1 is modified by a Note that specifins that initiation of action to exit the Restricted Region only applies if the APRM Flow Biased Simulated Thermal Power-High Function is " Setup". Operation in the Restricted Region without the APRM Flow Biased Simulated Thermal Power-High Function " Setup" indicates uncontrolled entry into the Restricted Region. Uncontrolled entry is consistent with the occurrence of unexpected transients, which, in combination with the absence of stability controls being met may result in significant degradation of stability performance. Under these conditions with the required PBDS instrumentation channel inoperable, the ability to monitor conditions indicating the potential for imminent onset of neutronic/ thermal hydraulic instability is lost and continued operation is not justified. Therefore, Required Action B.2 requires immediate reactor scram. (continued) (O U Brunswick Unit 2 B 3.3-51 Revision No. { l
- _ _ _ _ - _ . \
PBDS-B 3.3.1.3 BASES ACTIONS _ [d
;(continued)
In the Monitored Region the PBDS High-High Alarm provides indication of. degraded stability performance. Although not anticipated, operation in the Monitored Region is susceptible to neutronic/ thermal hydraulic instability under postulated conditions exceeding those previously assumed in the safety analysis. With the required PBDS channel inoperable while in the Monitored Region, the ability to.
~ monitor conditions: indicating the potential for imminent' onset of neutronic/ thermal hydraulic instability is lost.
Therefore, action must be initiated to exit the Monitored
-Region.
Actions to restart an idle recirculation loop, withdraw control rods or reduce recirculation flow may result in approaching unstable reactor conditions and are not allowed to be used to comply with this Required Action. - Exit of the Monitored Region is accomplished by control rod insertion and/or recirculation flow increases. However, actions which reduce recirculation flow are _ allowed provided the FCBB is recently (within 15 minutes) verified to be s 1.0. Recent verification of FCBB_ being met, provides assurance that with iO the PBDS ino)erable, planned decreases in recirculation drive flow siould not result in significant degradation of core stability. performance. The Completion Time of 15 minutes ensures timely operator action to exit the region consistent with the low probability that reactor conditions exceed the initial conditions assumed in the safety analysis. The time required to exit the Monitored Region will depend on j existing plant conditions. Provided efforts are begun within 15 minutes and continued until the Monitored Region is exited, operation is acceptable based on the low probability of a transient which degrades stability performance occurring simultaneously with the required PBDS channel inoperable. SURVEILLANCE SR 3.3.1.3.1 i 2 REQUIREMENTS' During operation in the Restricted Region or the Monitored l Region the PBDS High-High Alarm is relied upon to indicate ! conditions consistent with the onset of neutronic/ thermal hydraulic instability. Verification that each OPERABLE (continued) O < Brunswick Unit 2 B 3.3-52 Revision Nc. l l 1 1
.PBDS ,
8 3.3.1.3 BASES' SURVEILLANCE SR 3.3.1.3.1 (continued) REQUIREMENTS . . channel of PBDS instrumentation is not in High-High Alarm every 12 hours provides assurance of the proper indication. of the. alarm during operation in the Restricted Region or the Monitored Region. The 12 hour Frequency supplements less formal, but more frequent, checks ' tiara status during operation. SR 3.3.1.3.2 Performance of the CHANNEL CHECK every 12 hours ensures that a gross failure of instrumentation has not occurred. This CHANNEL CHECK is normally a comparison of the PBDS indication to the state of the annunciator, as well as comparison to the same parameter on the other channel if it is available. It is based on the assumption that the instrument channel indication agrees with the immediate indication available to the operator, and that instrument channels monitoring the same parameter should read similarly. Deviations between the instrument channels could be an indication of instrument component failure. A CHANNEL O CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication - and readability. The 12 hour Frequency is based on the CHANNEL CHECK Frequency requirement of similar Neutron Monitoring System components. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by. ! the LCO. 1 i SR 3.3.1.3.3 A CHANNEL FUNCTIONAL TEST is performed for each required PBDS channel to ensure that the system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the PBDS includes manual initiation of an internal test sequence and verification of appropriate alarms and inop conditions being reported. (continued)
~ I O Brunswick Unit 2 B 3.3-53 Revision No. jl
PBDS B 3.3.1.3 BASES SURVEILLANCE SR 3.3.1.3.3 (continued) REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST at a Frequency of 24'. months verifier, the performance of the PBDS and associated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the' likelihood of a change in the system or component status. The alarm circuit is designed to operate for over 24 months with sufficient accuracy.on signal amplitude and signal timing considering environment, initial calibration, and accuracy drift (Ref. 2). REFERENCES 1. NEDO 32339-A, Reactor Stability Long Term Solution: Enhanced Dption I-A, July 1994.
- 2. NEDC-32339, Supplement 2, Reactor Stability Long Ters Solution: Enhanced Option I-A Solution Design.
April 1995. O v i I I I I I O Brunswick Unit 2 8 3.3-54 Revision No.
C:ntrol Rod Block Irstrumentation B 3.3.2.1 8 3.3 INSTRUMENTATION' 8 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Control rods provide the primary means for control of ! reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch-Shutdown Position Function ensure that all control-rods remain inserted to prevent inadvertent criticalities. The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint O during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above
- the low power range setpoint specified in the COLR. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One R8N channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A signal from one average power range monitor (APRM) channel assigned to each Reactor Protection System (RPS) tri > system supplies a reference signal for the RBM channel in tie same trip system. This reference signal is used to determine which RBM range set)oint (low, intermediate, or high) is enabled. If tie APRM is indicating less than the low power range setpoint, the RBM is automatically bypassed. The R8N is also automatically bypassed if a peripheral control rod is selected (Ref.1).
A rod block signal is also generated if an RBM downscale trip or an inoperable trip occurs, since this could indicate a problem with the RBN channel. The downscale trip will (continued) Brunswick Unit 2 8 3.3-55 Revision No.
4 Control Rod Block Instrumentation-B 3.3.2.1 0 ' () - BASES BACKGROUND occur if the R8N channel signal decreases below the (continued) downscale trip setpoint after the RBM channel has been normalized. The inoperable trip will occur during the nulling (normalization) sequence, if the R8N channel fails to null, too few LPRM inputs are available, if a module is not plugged in, or the function switch is moved to any position other than " Operate." The purpose of the RWM is to control rod patterns during startup and shutdown, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. The sequences effectively ~ 11mit the potential amount and - rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses steam flow signals to determine when the reactor power is
. above the preset power level at which the RWM is automatically bypassed. The RWM is a single channel system that provides input into the RMCS rod withdraw permissive O circuit.
With the reactor mode switch in the shutdown position,.a , control rod withdrawal block is applied to all control rods { to ensure that the shutdown condition is maintained. This 1 Function prevents inadvertent criticality as the result of a- l control rod withdrawal during MODE 3 or 4, or during MODE 5 ! when'the reactor mode switch is required to be in the shutdown position. The reactor mode switch has two
-channels, each inputting into a separate RMCS rod block i circuit. A rod block in either RMCS circuit will provide a !
control rod block to all control rods. ; APPLICABLE 1. Rod Block Monitor
' SAFETY ANALYSES, LCO, and The RBM is designed to prevent violation of the MCPR l APPLICABILITY SL and the cladding 1% plastic strain fuel design limit that i may result from a single control rod withdrawal error (RWE) l event. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 2. A statistical analysis of RWE events was performed to '
(continued) O Brunswick Unit 2 B 3.3-56 Revision No.
Control Rod Block Instrumentation C ').3.2.1 (N V BASES APPLICABLE 1. Rod Block Monitor (continued) SAFETY ANALYSES, LCO, and determine the RBM response for both channels for each event. APPLICABILITY From these responses, the fuel thermal performance as a function of RBM Allowable Value was determined. The Allowable Values are chosen as a function of power level. Based on the specified Allowable Values, operating limits are established. The RBM Function satisfies Critsrion 3 of Reference 3. Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Value for the associated power range, to ensure that no single instrument failure can preclude a rod block from this Function. The actual setpoints are calibrated consistent with applicable setpoint methodology. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setting less (] conservative than the trip setpoint, but within its v Allowable Value, is teceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint value, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. (continued) U Brunswick Unit 2 B 3.3-57 Revision No.
Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE '1. Rod Block Monitor (continued) SAFETY ANALYSES, LCO, and The R8M is assumed to mitigate the consequences _of an RWE APPLICABILITY event when operating = 29% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE , (Ref. 2). When operating < 90% RTP, analyses (Ref. 2) have l shown that with an initial MCPR = 1.70, no RWE event will result in exceeding the MCPR SL. Also, the analyses demonstrate that when operating at m'90% RTP with i MCPR = 1.40, no RWE event will result in exceeding the MCPR SL (Ref. 2). Therefore, under these conditions, the RBM is also not required to be OPERABLE.
- 2. -Rod Worth Minimizer The RWM enforces the banked position withdrawal sequence l (CPWS) to ensure that the initial conditions of the CRDA analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in References 4, 5, and 6.. The BPWS requires that control rods p
y be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, " Rod Pattern Control." The RWM Function satisfies Criterion 3 of Reference 3. I The RWM is a microprocessor-based system with the principle I task to reinforce procedural control to limit the reactivity _ i wcrth of control rods under lower power conditions. Only one channel of the RWM is available and required to be OPERABLE. Special circumstances provided for in the ; Required Action of LC0 3.1.3,'" Control Rod OPERABILITY," and LC0 3.1.6 may necessitate bypassing the RWM to allow l continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. As required by these conditions, one or more control rods may be bypassed in the RWM or the RWM may be bypassed. However, the RWM must be considered inoperable and the Required Actions of this LCO followed since the RWM can no longer enforce compliance with the BPWS. (continued) ; O Brunswick Unit 2 B 3.3-58 Revision No.
Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 2. Rod Worth Minimizer (continued) SAFETY ANALYSES, LCO, and Compliance with the BPWS, and therefore OPERABILITY of the APPLICABILITY RWM, is required in MODES I and 2 when THERMAL POWER is
- s; 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a CRDA (Refs. 5 and 6). In MODES 3 and 4, all control rods are required to be inserted into the core; therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDN ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.
- 3. Reactor Mode' Switch-Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch-Shutdown Position control rod withdrawal block i iO ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.
The Reactor Mode Switch-Shutdown Position Function satisfies Criterion 3 of Reference 3. Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required. There is no Allowable Value for this function since the channels are mechanically actuated based solely on reactor mode switch position. During shutdown conditions (MODE 3, 4, or 5), no positive i reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod withdrawal block is required to be OPERABLE. During MODE 5 with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LCO 3.9.2, " Refuel Position One-Rod-Out Interlock") provides the required control rod withdrawal blocks. (continued) O Brunswick Unit 2 B 3.3-59 Revision No.
Control Rod tilock Instrumentation B 3.3.2.1 .(' .\ BASES (continued) ACTIONS M With one RBM channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function; however, overall reliability is reduced because a single failure in the remaining OPERABLE channel can result in no control rod block capability for the RBM. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Completion Time of 24 hours is based on the low probability of an event ~ occurring coincident with a failure in the remaining OPERABLE channel. M If Required Action A.1 is not met and the associated Cempletion Time has expired, an RBM channel must be placed in the tripped condition within I hour. If both RBM channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip within I hour. This initiates a control rod withdrawal block, thereby ensuring that the RBM y function is met. The I hour Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inopes*able channels. C.I. C.2.1.1. C.2.1.2. and C.2.2 With the RWM Function inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence. However, the overall reliability is reduced because a single operator error can result in violating the control rod sequence. Therefore, control rod movement must be immediately suspended except by scram. Alternatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with the RWM inoperable, for reasons other than one or more control rods bypassed in the RWM, was not performed in the last 12 months. Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant (continued) O Brunswick Unit 2 B 3.3-60 Revision No.
i L Control Rod Block Instrumentation B 3.3.2.1 ,O l V BASES ACTIONS C.1. C.2.1.1. C.2.1.2 and C.2.2 (continued) logs and control room indications. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, 1 control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2. Required Action C.2.2 allows for the RWM function to be performed manually and requires double verification of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. ] One or more control rods may be bypassed in the RWM or the i RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LC0 3.1.3 and LC0 3.1.6 may require bypassing one or more control rods in the RWM or bypassing the RWM, during which l time the RWM must be considered inoperable with Condition C entered and its Required Actions taken. In the event one or l more control rods are bypassed in the RWM (up to 8 control i rods'may be bypassed in accordance with the RWM design), Required Action C.2.1.2 does not restrict startup. b u With the RWM Function inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence. Required Action D.1 allows for the RWM Function to be performed manually and requires double verification of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical L staff. One or more control rods may be bypassed in the RWM or the RWM may be bypassed under these conditions to allow , the reactor shutdown to continue. ! l l E.1 and E.2 With one Reactor Mode Switch-Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE l channel is adequate to perform the control rod withdrawal ! block function. However, since the Required Actions are (continued) O Brunswick Unit 2 B 3.3-61 Revision No. l.
i Control Rod Block Instrumentation B 3.3.2.1-BASES ACTIONS E.1 and E.2 (continued)_ consistent with the normal action of an OPERABLE Reactor Mode Switch-Shutdown Position Function (i.e., maintaining all control rods inserted), tjere is no distinction between having one or two channels inoperable. In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing l one or more fuel assemblies.will ensure that the core is j subcritical with adequate'SDM ensured by LCO 3.1.1. Control ; rods in core cells containing no fuel assemblies do not i effect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Control Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1. The Surve111ances are modified by a Note to indicate that when an RBM channel is placed in an inoperable status solely _ for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be retilrned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 7) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour. testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary. SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel I to ensure that the channel will perform the intended i function. It includes the Reactor Manual Control System (continued) f O Brunswick Unit 2 B 3.3-62 Revision No. l i
Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.1 (continued) REQUIREMENTS input. Any setpoint adjustment shall be consistent with the j
- assumptions of the current plant specific setpoint methodology.
The Frequency of 92 days is based on reliability analyses (Ref. 8). SR 3.3.2.1.2 and SR 3.3.2.1.3 4 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the RWM is performed by selecting a control rod not in compliance with the prescribed sequence and verifying proper annunciation of the
-selection error, and by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until I hour after any control rod is withdrawn in MODE 2. As noted, SR 3.3.2.1.3 is not required to be performed until I hour O after THERMAL POWER is s 10% RTP in MODE 1. This allows entry into MODE 2 for SR 3.3.2.1.2, and entry into MODE 1 when THERMAL POWER is s 10% RTP for SR 3.3.2.1.3, to perform the required Surveillance-if the 92 day frequency is not met i
per SR 3.0.2. The 1 hour allowance is based on. operating experience and in consideration of providing a reasonable time in which to complete the SRs. Operating experience has
~ demonstrated these components will usually pass the Surveillances when performed at the 92 day Frequency.
Therefore, the Frequency is acceptable from a reliability l standpoint. l SR 3.3.2.1.4 i The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The L power at which the control rod block Allowable Values l automatically change are based on the' APRM signal's input to (. each RBM channel. Below the minimum power range setpoint, l the RBM is automatically bypassed. These power range setpoints (low power range setpoint, intermediate power range setpoint, and high power range setpoint) must be (continued) l Brunswick Unit 2 B 3.3-63 Revision No.
i Control Rod Block Instrumentation B 3.3.2.1
~
BASES SURVEILLANCE SR 3.3.2.1.4 (continued) REQUIREMENTS l verified periodically to be less than or equal to the j specified Allowable Values in the COLR. If any power range l setpoint is nonconservative, then the affected RBM channel ! is considered inoperable. Alternatively, the RBM power i range channel can be placed in the conservative condition l (i.e., enabling the proper RBM setpoint). If placed in this condition, the SR is met and the RBM channel is not considered inoperable. As noted, neutron detectors are excluded from the Surveillance because they are passive ! devices, with minimal drift, and because of the difficulty l of simulating a meaningful signal. Neutron detectors.are l adequately tested in SR 3.3.1.1.3 and SR 3.3.1.1.8. The 24 month Frequency is based on the actual trip setpoint methodology utilized for these channels. 1 l l SR 3.3.2.1.5 I The RWM is automatically bypassed when power is above a specified value. The power level is determined from steam flow signals. The automatic bypass setpoint must be O verified periodically to be > 10% RTP. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can I be placed in the conservative condition (nonbypass). If ! placed in the nonbypassed condition, the SR is. met and the ! RWM is not considered inoperable. The Frequency is based on j the trip setpoint methodology utilized for the low power l setpoint channel. l SR 3.3.2.1.6 A CHANNEL FUNCTIONAL TEST is perfor::.ed for the Reactor Mode Switch-Shutdown Position Function to ensure that the channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch-Shutdown Position function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown position and verifying a control, rod block occurs. As noted in the SR, the Surveillance is not required to be performed until I hour after the reactor mode switch is in the shutdown . position, since testing of this interlock with the reactor mode switch in any other position cannot be i (continued) O Brunswick Unit 2 B 3.3-64 Revision No. l l t
Control Rod Block Instrumentation B 3.3.2.1 BASES i SURVEILLANCE SR 3.3.2.1.6 (continued) REQUIREMENTS performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the 24 month Frequency is not met per SR 3.0.2. The I hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The 24 month frequency is based on the need to perform this ) Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. SR 3.3.2.1.7 A CHANNEL CALIBRATION is a complete check of the instrument i loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary i range and accuracy. CHANNEL CALIBRATION leaves the channel l adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint (s. methodology. The CHANNEL CALIBRATION may be performed Q electronically. As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.3 and SR 3.3.1.1.8. The Frequency is based upon the assumption of a 24 month , calibration interval in the determination of the magnitude 1 of equipment drift in the setpoint analysis. ' SR 3.3.2.1.8 1 The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible. (continued) p-u Brunswick Unit 2 B 3.3-65 Revision No.
Control Rod Block Instrumentation , B 3.3.2.1 l
- O.
V ' BASES (continued) REFERENCES 1. UFSAR, Section 7.6.1.1.5. L 2.- NEDC-31654P, Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant, February 1989.
- 3. 10CFR50.36(c)(2)(ii).
- 4. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Unit I and 2,
' September 1995.
- 5. UFSAR'Section 15.4. l
- 6. NRC SER, Acceptance for Referencing of Licensing Topical Report NEDE-24011-P-A; General Electric Standard Application for Reactor Fuel, Revision 8 Amendment 17, December 27, 1987.
- 7. GENE-770-06-1-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for !
Selected Instrumentation Technical Specifications, j December 1992. '
- 8. NEDC-30851P-A, Supplement 1 Technical Specification ,
Improvement Analysis for BWR Control Rod Block i Instrumentation, October 1988. Brunswick Unit 2 B 3.3-66 Revision No.
Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 i B 3.3 INSTRUMENTATION B.3.3.2.2 feedwater and Main Turbine High Water Level Trip Instrumentation BASES BACKGROUND The feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow. With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level setting causing the trip of the two feedwater pump turbines and the main turbine. High water levels signals are provided by three narrow range sensors of the Digital Feedwater Control System. These three level sensors sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vessel (variable leg). The three level signals are input into a digital control computer. The digital control p computer provides three output signals to the high water D)' t level trip channels. Each high water level trip channel consists of a relay whose contacts form the trip logic. The high water level trip logic is arranged as a two-out-of-three logic, that trips the two feedwater pump turbines and the main turbine. The digital control computer processes the reactor vater level input signals and compares them to , pre-establis..., setpoints. When the setpoint is exceeded, l the associated channel output relay actuates, which then outputs to the main turbine and feedwater pump trip initiation logic. A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop valves protects the turbine from damage due to water entering the turbine. APPLICABLE The feedwater and main turbine high water level trip SAFETY ANALYSES instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref. 1). (continued) T Brunswick Unit 2 8 3.3-67 Revision No.
Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 f BASES APPLICABLE The high water level trip indirectly initiates a reactor SAFETY ANALYSES scram from the main turbine trip (above 30% RTP) and trips (continued) the feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction in MCPR. Feedwater and main turbine high water level trip instrumentation satisfies Criterion 3 of Reference ~2. LC0 The LCO requires three channels of the reactor vessel high water level instrumentation to be OPERABLE to ensure that { the feedwater pump turbines and main turbine trip on a valid high water level signal. Two of the three channels are needed to' provide trip signals in order for the feedwater and main turbine trips to occur. Each channel must have its setpoint set within the specified Allowable Value of SR 3.3.2.2.2. The Allowable Value is set to ensure that the thermal limits are not exceeded during the event. The actual setpoint is calibrated to be consistent with the applicable setpoint methodology assumptions. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that_the trip settings do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. O Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those ) redetermined values of output at which an action should tace place. The setpoints are compared to the actual process )arameter (e.g., reactor vessel water level), and when tie measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process - l parameters obtained from the safety analysis. The trip ; setpoints are determined from the analytic limits corrected ' for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip .i setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection (continued) O Brunswick Unit 2 B 3.3-68 Revision No.
feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES LCO because instrumentation uncertainties, process effects, (continued) calibration tolerances, instrument drift, and severe l environment errors-(for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. ! APPLICABILITY The feedwater and main turbine high water level trip instrumentation is required to be OPERABLE at 2: 25% RTP to ensure that the fuel cladding integrity Safety Limit and the cladding 1% plastic strain limit are not violated during the feedwater controller failure, maximum demand event. . As discussed in the Bases for LCO 3.2.1, " Average. Planar Linear Heat Generation Rate (APLHGR)," and LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," sufficient margin to these limits exists below 25% RTP; therefore, these requirements-are only necessary when operating at or above this power level. ACTIONS A Note has been provided to modify the ACTIONS related to feedwater and main turbine high water level trip instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent O- divisions, subsystems, compor,ents, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional-failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable feedwater and main turbine high water level trip instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable feedwater and main turbine high water level trip instrumentation channel. 8.d With one channel inoperable, the remaining two OPERABLE channels can provide the required trip signal. However, overall instrumentation reliability is reduced because a single failure in one of the remaining channels concurrent (continued) O Brunswick Unit 2 B 3.3-69 Revision No.
Feedwater and Main Turbine High Water Level Trip Instrumentation 8 3.3.2.2 j BASES-i ACTIONS M (continued) l with feedwater controller failure, maximum demand event, may I result in the instrumentation not being able to perform its ! intended function. . Therefore, continued operation is only I allowed for a limited time with one channel inoperable. If- j the inoperable channel cannot be restored to OPERABLE status ' within the Completion Time, the channel must be placed in the tripped condition per Required Action A.I. Placing the inoperable channel in trip would conservatively compensate i for the inoperability, restore capability to accommodate a i single failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel in trip (e.g., as.in the case where ' placing the inoperable channel in trip would result in a feedwater or main turbine trip), Condition C must be entered and its Required Action taken. The Completion Time of 7 days is based on the low probability of the event occurring coincident with a single i failure in a remaining _0PERABLE channel. O u . With two or mo.re channels inoperable, the feedwater and main i' turbine high water level trip instrumentation cannot perform its design function (feedwater and main turbine high water level trip capability is not maintained). Therefore, continued operation is_only permitted for a 4 hour period, during which feedwater and main turbine high water level trip capability must be restored. The trip capability is considered maintained when sufficient channels are OPERABLE or in trip such that the feedwater and main turbine high water level trip logic will generate a trip signal on a valid signal. This requires two channels to each be OPERABLE or in trip.- If the required channels cannot be restored to OPERABLE status or placed in trip, Condition C must be entered and its Required Action taken. The 4 hour Completion Time is sufficient for the operator to take corrective action, and takes into account the likelihood of an event requiring actuation of feedwater_ and i main turbine high water level trip instrumentation occurring (continued) O Brunswick Unit 2 B 3.3-70 Revision No.
l l Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES 1
-ACTIONS L 1 '(continued) during this period. It is also consistent with the 4 hour Completion Time provided in LCO 3.2.2 for Required 3
<~ Action A.1, since this instrumentation's purpose is to preclude a MCPR violation. i ful With the required channels not restored to OPERABLE status or placed in trip, THERMAL POWER must be reduced to l
< 25% RTP within 4 hours. As discussed in the Applicability i section of the Bases, operation below 25% RTP results in sufficient margin to the required limits, and the feedwater and main turbine high water level trip instrumentation is not required to protect fuel integrity during the feedwater controller failure, maximum demand event. The allowed Completion Time of 4 hours is based on operating experience to reduce THERMAL POWER to < 25% RTP from full power conditions in an orderly manner and without challenging plant systems.
O-I SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains feedwater and main turbine high water level trip capability. Upon i completion of the Surveillance, or expiration of the 6 hour ' allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on~the reliability analysis (Ref. 3) assumption that 6 hours is the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the feedwater pump turbines and main turbine will trip when necessary. SR 3.3.2.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A l CHANNEL CHECK is normally a comparison of the parameter (continued 1 O Brunswick Unit 2 B 3.3-71 Revision No.
Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE SR 3.3.2.2.1 (continued) REQUIREMENTS' indicated on one channel to a siellar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels, or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, .it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limits. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of r channel status during normal operational use of the displays i associated with the channels required by the LCO. SR 3.3.2.2.2 CHANNEL CALIBRATION is a complete check of the instrument loopland the sensor. This test verifies the channel responds to the measured parameter within the necessary , range and accuracy. CHANNEL CALIBRATION leaves the channel l adjusted to account for instrument drifts between successive l E calibrations consistent with the plant specific setpoint methodology. The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. ; i SR 3.3.2.2.3 l The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the feedwater and (continued) O Brunswick Unit 2 B 3.3-72 Revision No.
Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES l SURVEILLANCE SR 3.3.2.2.3 (continued) REQUIREMENTS main turbine valves is included as part of this Surveillance l 2nd overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide j complete testing of the assumed safety function. Therefore, ! if a valve is incapable of operating, the associated instrumentation would also be inoperable. The 24 month < Frequency is based on the need to perform this Surveillance ! under the conditions that apply during a plant outage and the potential for an unplanned. transient if the Surveillance , were performed with the reactor at power. ) i l REFERENCES 1. .UFSAR, Section 15.1.2.
- 2. 10 CFR 50.36(c)(2)(ii).
- 3. GENE-770-06-1-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-0f-Service Times for Selected Instrumentation Technical Specifications, December 1992.
O I 1 l i i O O Brunswick Unit 2 B 3.3-73 Revision No. l l j
PAM Instrumentation B 3.3.3.1 (
\ 8 3.3 INSTRUMENTATION B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation BASES-BACKGROUND The )rimary purpose of the PAM instrumentation is to display in tie control room plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Events. The instruments that monitor these variables are designated as Type A, Category I, and non-Type A, Category I, in accordance with Regulatory Guide 1.97 (Ref.1).
The OPEPABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected plant parameters to monitor and assess plant status and behavior following an accident. This capability is consistent with the recommendations of Reference 1. O APPLICABLE The PAM instrumentation LC0 ensures the OPERABILITY of SAFETY ANALYSES Regulatory Guide 1.97, Type A variables so that the control room operating staff can: Perform the diagnosis specified in the Emergency Operating Procedures (E0Ps). These variables are restricted to preplanned actions for the primary success path of Design Basis Accidents (DBAs), (e.g., loss of coolant accident (LOCA)), and
- Take the specified, preplanned, manually controlled actions for which no automatic control is provided, which are required for safety systems to accomplish their safety function.
The PAM instrumentation LC0 also ensures OPERABILITY of Category I, non-Type A, variables so that the control room operating staff can:
- Determine whether systems important to safety are performing their intended functions; (continued)
~O k/ . Brunswick Unit 2 B 3.3-74 Revision No.
i PAM Instrumentation B 3.3.3.1 rG (,/ BASES-APPLICABLE
- Determine the potential for causing a gross breach of SAFETY ANALYSES the barriers to radioactivity release; (continued)
- Determine whether a gross breach of a barrier has occurred; and
- Initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.
The plant specific Regulatory Guide 1.97 Analysis (Ref. 2) documents the process that identified Type A and Category I, non-Type A, variables. Accident monitoring instrumentation that satisfies the , definition of Type A in Regulatory Guide 1.97 meets ! Criterion 3 of Reference 3. Category I, non-Type A, instrumentation is retained in Technical Specifications (TS) because they are intended to assist operators in minimizing the consequences of accidents. Therefore, these Category I l variables are important for reducing public risk. LCO LCO 3.3.3.1 requires two OPERABLE channels for all but one Function to ensure that no single failure prevents the operators from being presented with the information necessary to determine the status of the plant and to bring the plant to, and maintain it in, a safe condition following that accident. Furthermore, providing two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information. The exception to the two channel requirement is primary containment isolation valve (PCIV) position. In this case, the important information is the status of the primary containment penetrations. The LC0 requires one position indicator for each active (e.g., automatic) PCIV. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of passive valve or via system boundary status. If a normally active PCIV is known to be closed and deactivated, position indication is not needed to determine status. Therefore, the position indication for closed and deactivated valves is not required to be OPERABLE. (continued) ( Brunswick Unit 2 B 3.3-75 Revision No.
PAM Instrumentation B 3.3.3.1 ; I DASES LCC The following list is a. discussion of the specified (continued) instrument Functions listed in Table 3.3.3.1-1 in the accompanying LCO.
- 1. Reactor Vessel Pressure Reactor vessel pressure.is a Type A and Category I variable provided to support monitoring of Reactor Coolant System (RCS) integrity and to verify operation of the Emergency Core Cooling Systems (ECCS).. Two independent pressure transmitters with a range of 0 )sig to 1500 psig monitor ,
pressure and are indicated in tie control room. . Wide range !
-instruments are the primary indication used by the operator l during an accident. Therefore, the PAM Specification deals i specifically with this portion of the instrument channel. I 2.a. 2.b. 2.c. Reactor Vessel Water level'
' Reactor vessel water level is a Type A and Category I I variable provided to support monitoring of core cooling and n to verify operation of the ECCS. Channels from three l g different ranges of water level provide the PAM Reactor Vessel Water Level Function. The water level channels ,
. measure from -150 inches to +550 inches. Water level is measured by independent differential pressure transmitters for each required channel. The output from these channels is recorded on independent recorders or read on indicators, I which are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
- 3. SuDDression Chamber Water level Suppression chamber water level is a Type A and Category I variable provided to detect a breach in the reactor coolant pressure' boundary (RCPB). This variable is also used to verify and provide long term surveillance of ECCS function.
The wide range suppression pool water level measurement provides the operator with sufficient information to assess the status of both the RCPB and the water supply to the ECCS. The wide range water level indicators are capable of monitoring the suppression pool water level from the bottom (continuedl b-v Brunswick Unit 2 B 3.3-76 Revision No.
e PAM Ins'trumentation B 3.3.3.1 D(,/ BASES LCO -3. Suppression Chamber Water Level (continued) of the ECCS suction lines to 5 feet above the normal pool water level. Two wide range suppression pool water level signals are transmitted from separate differential pressure transmitters for each channel. The out >ut of one of these channels is recorded on a recorder in tie control room. The output of the other channel is read on an indicator in the control room. These instruments are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
- 4. Suppression Chamber Water Temperature Suppression chamber water temperature is a Type A and Category I variable provided to detect a condition that i
could potentially lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach. The suppression chamber water temperature instrumentation, which measures from 40*F to j 240*F, allows operators to detect trends in suppression pool O water temperature in sufficient time to take action to prevent steam quenching vibrations in the suppression pool. I Suppression pool temperature is monitored by 12 pairs of ; temperature sensors spaced around the suppression pool. A j pair of sensors is located near each of the quenchers on the ' discharge lines of the 11 safety / relief valves. Each pair of sensors is located so as to sense the representative ! temperature of that sector of the suppression pool even with the associated safety / relief valve open. The outputs for the sensors are indicated on two microprocessors in the ' control room. The signals from the sensors are conditioned by the two microprocessors to provide an average water temperature. Average water temperature is recorded on two independent recorders in the control room. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channels.
- 5. Suppression Chamber Pressure Suppression chamber pressure is a Type A and Category I variable provided to detect a condition that could potentially lead to containment breach and to verify the (continued)
Brunswick Unit 2 B 3.3-77 Revision No.
PAM Instrumentation B 3.3.3.1 A h BASES LCO 5. Suppression Chamber Pressure (continued)- effectiveness of ECCS actions.taken to prevent containment breach. Suppression chamber pressure is indicated in the control room from two separate pressure transmitters. The range of indication is O psig to 75 asig. These instruments are the primary indication used by tie operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
- 6. Drvwell Pressure Drywell pressure. is a Type A and Category I variable provided to detect breach of the RCPB and to verify ECCS functions that operate to maintain RCS integrity. Two wide range drywell pressure signals are transmitted from separate pressure transmitters for each channel. The output of one of these channels is recorded on a recorder in a control room. The output of the other channel is read on an indicator in the control room. The pressure channels measure from -5 psig to 245 psig. These instruments are the primary indication used by the operator during an accident.
O Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
- 7. Drywell Temperature Drywell temperature is a Type A and Category I variable .
provided to detect a breach of the RCPB and to verify the effectiveness of ECCS functions that operate to maintain RCS l integrity. Sixteen temperature sensors are located in the drywell to monitor drywell temperature. The sensors are divided into two divisions for redundancy. The signals from these sensors are conditioned by two divisionalized microprocessors. Drywell temperature is recorded by two pairs of divisionalized recorders in the control room. The range of the recorders is from 40*F to 440*F. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel. (continued) O Brunswick Unit 2 8 3.3-78 Revision No.
PAM Instrumentation B 3.3.3.1 BASES LCO - 8. Primary Containment Isolation Valve (PCIV) Position (continued) FCIV position, a Category I variable, is provided for verification of containment integrity. In the case of PCIV position, the important information is the isolation status of the containment penetration. The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active PCIV in a containment penetration flow path, i.e.,-two total channels of PCIV position indication for..a penetration flow path with two active valves. For containment penetrations with'only one active PCIV having control room indication, Note (b) requires a single channel of valve position indication to be OPERABLE. This is sufficient to redundantly verify the isolation status of each isolable penetration via indicated status of the active valve, as applicable, and prior knowledge of passive valve or system boundary status. If a penetration flow path is isolated, position indication for the PCIV(s) in the associated penetration flow path is not needed to determine status. Therefore, the position indication for valves in an isolated penetration flow path is not required to be OPERABLE. The PCIV position PAM instrumentation consists of position switches, associated wiring and control room indication for active PCIVs (check valves and manual valves are not required to have position indication). Therefore, the PAM Specification deals specifically with these instrument channel s.
- 9. Drywell'and Suppression Chamber Hydrocen and 0xvaen Analyzers i
Drywell and suppression chamber hydrogen and oxygen ! analyzers are Type A and Category I instruments provided to I detect high hydrogen or oxygen concentration conditions that i represent a potential for containment breach. This variable 1 is also important in verifying the adequacy of mitigating j actions. The drywell and suppression chamber hydrogen and : oxygen analyzers PAM instrumentation consists of two I independent gas analyzer systems. Each gas analyzer system l con.ists of a hydrogen analyzer and an oxygen analyzer. The analyzers are capable of determining hydrogen concentration , in the range of 0% to 30% and oxygen concentration in the range of 0% to 25%. Each gas analyzer system must be (continued) l O U Brunswick' Unit 2 B 3.3-79 Revision No. 1
1 PAM Instrumentation B 3.3.3.1 BASES LCO 9. Drywell and Suporession Chamber Hydroaen and Oxvaen , Analyzers (continued) l capable of sampling the drywell and the suppression chamber. There are two independent recorders in the control room to display the results. Therefore, the PAN Specification deals specifically with these portions of the analyzer channels.
- 10. Drywell Area Radiation Drywell area radiation is a Category I variable provided to monitor the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans. Post accidentdrywellarearadiationlevelskremonjtoredbyfour instruments, each with a range of 1 R/hr to 10 R/hr. The outputs of these channels are indicated and recorded in the control room. Therefore, the PAM S)ecification deals specifically with this portion of tle instrument channel. 1 APPLICABILITY The.PAM instrumentation LC0 is a)plicable in MODES 1 and 2.
O These variables are related to tie diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1 and 2. In MODES 3, 4, and 5, i i plant conditions are such that the likelihood of an event i that would require PAM instrumentation is extremely low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES.
)
ACTIONS Note I has been added to the ACTIONS to exclude the MODE change' restriction of LC0 3.0.4. This exception allows entry into the applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the passive function of the instruments, the operator's ability to ' diagnose an accident using alternative instruments and methods, and the low probability of an event requiring these instruments. Note 2 has been provided to modify the ACTIONS related to PAM instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables (continuedi O Brunswick Unit 2 B 3.3-80 Revision No.
PM Instrumentaticn B 3.3.3.1 BASES ACTIONS expressed in the Condition discovered to be inoperable or (continued) not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable PM instrumentation channels provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable PM Function. Ad When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account d the remaining OPERABLE channels, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval. O u If a channel has not been restored to OPERABLE status in 30 days, this Required Action specifies initiation of action in accordance with Specification 5.6.6, which requires a b writ (en report to be submitted to the NRC. This report discusses the results of the root cause evaluation of the inoperability and identifies prcposed restorative actions. This Required Action is appropriate in lieu of a shutdown requirement, since another OPERABLE channel is monitoring the Function, and given the likelihood of plant conditions that would require information, provided by this instrumentation. F
.Gd Wnen one or more Functions have two required channels that are inoperable (i.e., two channels inoperable in the same b
Function), one channel in the Function should be restored to OPERABLE status within 7 days. The Completion Time of b 7 days is based on the relatively low probability of an event requiring PM instrument operation and the (continued) Brunswick Unit 2 B 3.3-81 Revision No. I
PAN Instrumentatica l 8 3.3.3.1 i BASES ACTIONS M (continued)- availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAN instrumentation. Therefore, requirir.g restoration of one inoperable channel of the Function limits the risk that the PAN Function will be in a degraded condition should an accident occur. M This Required Action directs entry into the appropriate Condition referenced in Table 3.3.3.1-1. The applicable Condition referenced in the Table is Function dependent. Each time an inoperable channel has not met the Required Action of Condition C and the associated Completion Time has expired, Condition D is entered for that channel and provides for transfer to the appropriate subsequent Condition. u For the majority of Functions in Table 3.3.3.1-1, if any Required Action and associated Completion Time of Condition C is not met, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. f_d Since alternate means of monitoring primary containment area radiation are available, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.6. These alternate means may be temporarily installed if the normal PAN channel cannot be restored to OPERABLE status within the allotted time. The (continued) O Brunswick Unit 2 B 3.3-82 Revision No.
PAM Instrumentation B 3.3.3.1 V BASES ACTIONS f_d (continued) report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels. SURVEILLANCE As noted at the beginning of the SRs, the following SRs REQUIREMENTS apply to each PAM instrumentation function in Table 3.3.3.1-1. SR 3.3.3.1.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should O read approximately the same value. Significant deviations 's' between instrument channels could be an indication of excessive instrument driftlin one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar plant instruments located throughout the plant. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. The frequency of 31 days is based upon plant operating experience, with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of those displays associated with the channels required by the LCO. (continued) O Brunswick Unit 2 B 3.3-83 Revision No.
PAM Instrumentation B 3.3.3.1 r"N U BASES SURVEILLANCE SR 3.3.3.1.2 and SR 3.3.3.1.3 REQUIREMENTS (continued) These SRs require a CHANNEL CALIBRATION to be performed. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies the channel responds to measured parameter with the necessary range and accuracy. For Function 9, the CHANNEL CALIBRATION shall be performed using standard gas samples containing a nominal:
- a. Zero volume percent hydrogen, balance nitrogen;
- b. Seven to ten volume percent hydrogen, balance nitregen;
- c. Twenty-five to thirty volume percent hydrogen, balance nitrogen;
- d. Zero volume percent oxygen, balance nitrogen;
- e. Seven to ten volume percent oxycen, balance nitrogen; and q
(j
- f. . Twenty to twenty-five volume percent oxygen,' balance nitrogen.
For Function 10, the CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one point calibration check of the detector below 10 R/hr with an installed or portable gamma source. The 92 day Frequency for CHANNEL CALIBRATION of the drywell and suppression chamber hydrogen and oxygen analyzers is based on operating experience. The 24 month Frequency for CHANNEL CALIBRATION of all other PAM Instrumentation of Table 3.3.3.1-1 is based on operating experience and consistency with the BNP refueling cycles. REFERENCES 1. Regulatory Guide 1.97, Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Revision 2, December 1980. (continued) O Brunswick Unit 2 B 3.3-84 Revision No.
)
PAM Instrumentation i B 3.3.3.1 1 BASES REFERENCES 2. NRC Safety Evaluation Report, Conformance to l (continued) , Regulatory Guide 1.97, Rev. 2. Brunswick Steam ' Electric Plant, Units 1 and 2, May 14, 1985.
- 3. 10CFR50.36(c)(2)(ii). I l
I l I O i 4 i O Brunswick Unit 2 B 3.3-85 Revision No.
)
Remote Shutdown Monitoring Instrumentation B 3.3.3.2 B 3.3 INSTRUMENTATION B 3.3.3.2' Remote Shutdown Monitoring Instrumentation , i BASES BACKGROUND The remote shutdown monitoring instrumentation provides the control room operator with sufficient instrumentation to support placing and maintaining the plant in a safe shutdown , condition from a location other than the control room. This capability is necessary to protect against the possibility of the control _ room becoming inaccessible. A safe shutdown condition is defined as MODE 3. With the plant in MODE 3, the Reactor Core Isolation Cooling (RCIC) System, the safety / relief valves, and the Residual Heat Removal (RHR) System can be used to remove core decay heat and meet all safety requirements. The long term supply of water for the RCIC System and the ability to operate shutdown cooling from outside the control room allow extended operation in MODE 3. In the event that the control room becomes inaccessible, the operators can monitor the status of the reactor and primary containment and the operation of the RCIC and RHR Systems at i A the remote shutdown panel and place and maintain the plant i b in MODE 3. Controls and selector switches will have to be operated locally at the switchgear, motor control panels, or other local stations. The plant is in MODE 3 following a plant shutdown and can be maintained safely in MODE 3 for an ; extended period of time. ' The OPERABILITY of the remote shutdown monitoring instrumentation Functions ensures that there is sufficient information available on selected )lant parameters to place and maintain the plant in MODE 3 siould the control room become inaccessible. APPLICABLE The remote shutdown monitoring instrumentation is required SAFETY ANALYSES to provide equipment at appropriate locations outside the control room with a design capability to monitor the prompt shutdown of the reactor to MODE 3, including the' necessary instrumentation to support maintaining the plant in a safe condition in. MODE 3. The criteria governing the design and the specific system requirements of the remote shutdown monitoring instrumentation are located in the UFSAR (Ref.1). (continued) O Brunswick Unit 2 B 3.3-86 Revision No.
Remote Shutdown Monitoring Instrumentation B 3.3.3.2 i: BASES l APPLICABLE The Remote Shutdown Monitoring Instrumentation is considered SAFETY ANALYSES an important contributor to reducing the risk of accidents; ! (continued) as such, it meets Criterion 4 of Reference 2. l LCO The Remote Shutdown Monitoring Instrumentation LC0 provides the requirements for the OPERABILITY of the monitoring instrumentation necessary to support placing and maintaining the plant in MODE 3 from a location other'than the control room. The monitoring instrumentation required are listed in Table B 3.3.3.2-1. The monitoring instrumentation are those required for:
- Reactor pressure vessel (RPV) pressure control; l
- Decay heat removal; and
- RPV inventory control.
The remote shutdown monitoring instrumentation is OPERABLE if all instrument channels needed to support the remote I shutdown monitoring function are OPERABLE with readouts g displayed external to the control room. The remote shutdown monitoring instruments covered by this LCO do not need to be energized to be considered OPERABLE. This LCO is intended to ensure that the instruments will be OPERABLE if plant conditions require that the remote shutdown monitoring instrumentation be placed in operation. l APPLICABILITY The Remote Shutdown Monitoring Instrumentation LC0 is l applicable in MODES 1 and 2. This is required so that the l plant can be placed and maintained in MODE 3 for an extended period of time from a location other than the control room. This LCO is not applicable in MODES 3, 4, and 5. In these MODES, the plant is already subc itical and in a condition of reduced Reactor Coolant System energy. Under these conditions, considerable time is available to restore (continued) O Brunswick Unit 2 B 3.3-87 Revision No. l
Remote Shutdown Monitoring Instrumentation B 3.3.3.2 BASES APPLICABILITY 'necessary instrument Functions if control room instruments (continued) or control becomes unavailable. Consequently, the LCO does not require OPERABILITY in MODES 3, 4, and 5. ACTIONS A Note is included that excludes the MODE change restriction of LCO 3.0.4. This exception allows entry into an applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require a plant shutdown. This exception is acce) table due to the low probability of an ! event requiring tits system. ' Note 2 has been provided to modify the ACTIONS related to Remote Shutdown Monitoring Instrumentation Functions. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, r l subsystems, coinponents, or variables cxpressed in the Condition, discovered to be inoperable or not within limits, I will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. - However, the Required Actions for inoperable ,O Remote Shutdown Monitoring Instrumentation Functions provide appropriate compensatory measures for separate Functions. , As such, a Note has been provided that allows separate l Condition entry for each inoperable Remote Shutdown i Monitoring Instrumentation Function. 1 M l Condition A addresses the situation where one or more required Functions of the remote shutdown monitoring instrumentation is inoperable. This includes any Function l listed in Table B 3.3.3.2-1. The Required Action is to restore the Function (all required channels) to OPERABLE status within 30 days. The Completion Time is based on operating experience and the low probability of an event b that would require evacuation of the control room. (continued) r O Brunswick Unit 2 B 3.3-88 Revision No.
Remote Shutdown Monitcring Instrumentation B 3.3.3.2 BASES ACTIONS JL1 (continued) If the Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time is reasonable, based I on operating experience, to reach the required MODE from
! full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.3.3.2.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 31 days ensures l that a gross failure of instrumentation has not occurred. A I CHANNEL CHECK is normally a comparison of the parameter l indicated on one channel to a siellar parameter on other l channels. It is based on the assumption that instrument
- channels monitoring the same parameter should read l approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect lO gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, l including indication and readability. If a channel is i outside the criteria, it may be an indication that the j sensor or the signal processing equipment has drifted i outside its limit. As specified in the Surveillance, a i CHANNEL CHECK is only required for those channels that are l normally energized. For Function 2 of Table B 3.3.3.2-1, . , the CHANNEL CHECK requirement does not apply to the N017 ! ! instrument loo) since this instrument loop has no displayed j indication. Tie CHANNEL CHECK requirement does apply to the ! remaining instruments of Function 2. The Frequency is based upon plant operating experience that demonstrates channel failure is rare. (continued) O Brunswick Unit 2 B 3.3-89 Revision No. e
I l Remote Shutd:wn Monitoring Instrumentation B 3.3.3.2 l l Ch V BASES l SUP.VEILLANCE SR 3.3.3.2.2 REQUIREMENTS (continued) CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifles the channel responds to measured parameter values with the necessary range and l. l accuracy. l l The 24 month Frequency is based upon operating experience l and consistency with the BNP refueling cycle. l REFERENCES 1. UFSAR, Section 7.4.4.
- 2. 10 CFR 50.36(c)(2)(ii).
l l .O l t i i l l O Brunswick Unit 2 B 3.3-90 Revision No.
Remote Shutdown Monitoring Instrumentation i B 3.3.3.2 j Table B 3.3.3.2 1 (page 1 of 1) l .. Remote Shutdown Monitorirg Instruusntatlan l
."IQUIRED i l
READGJT NUMBER OF FUNCTION LOCATION CHANNELS 1
- 1. Reactor Vesset Pressure (a) 1
- 2. Reactor Vessel Water Levet (a) 1
- 3. Swession Chamber Water Levet (a) 1
- 4. St@pression Chamber Water Temperature (a) 1
- 5. Drywell Pressure (a) 1 l 6. Drywett Temperature (a) 1
- 7. Residual Neat Removat System flow (a) 1 (a) Remote Shutdown Panel, Reactor Building 20 ft. Elevation.
4 l i i 1 l Brunswick Unit 2 B 3.3-91 Revision No. 1 l I 4 l
L l l ATWS-RPT Instrumentation l B 3.3.4.1 l B 3.3 INSTRUMENTATION B 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation
- l BASES l
BACKGROUND The ATWS-RPT System initiates an RPT, adding negative reactivity, following events in which a scram does not, but should occur, to lessen the effects of an ATWS event. Tripping the recirculation pumps adds negative reactivity from.the increase in steam voiding in the core area as core i flow decreases. When Reactor Vessel Water Level-Low Level 2 or Reactor Vessel Pressure-High setpoint is reached, the recirculation pump drive motor breakers trip. The ATWS-RPT System (Ref.1) includes sensors, relays, and l circuit breakers that are nocessary to cause initiation of an RPT. The channels include electronic equipment (e.g., trip units) that compare measured input signals with l pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic. i The ATWS-RPT two channels consists of Reactor of two Vessel independent trih and twosystems, w Pressure-Hig ' channels of Reactor Vessel Water Level-Low Level 2 in each l trip system. Each ATWS-RPT trip system is a two-out-of-two ! logic for each Function. Thus, either two Reactor Water i Level-Low Level 2 or two Reactor Vessel Pressure-High signals are needed to trip a trip system. The outputs of the channels in a trip system are combined in a logic so that either trip system will trip both recirculation pumps (by tripping the respective drive motor breakers). There is one drive motor breaker provided for each of the two recirculation pumps for a total of two breakers. The output of each trip system is provided to these recirculation pump breakers. l APPLICABLE The ATWS-RPT is not assumed to mitigate any &ccident or SAFETY ANALYSES, transient in the safety analysis. The ATWS-RPT initiates an g LCO, and RPT to aid in preserving the integrity of the fuel cladding APPLICABILITY following events in which a scram does not, but should, occur. Based on its contribution to the reduction of overall plant risk, however, the instrumentation meets Criterion 4 of Reference 2. (continued) v t Brunswick Unit 2 8 3.3-92 Revision No.
i e
' ATWS-RPT Instrumentation B 3.3.4.1 BASES-APPLICABLE The OPERABILITY of the ATWS-RPT is dependent on the SAFETY ANALYSES, OPERABILITY of the individual instrumentation channel LCO, and ' Functions. Each Function must have a required number of APPLICABILITY OPERABLE channels in each trip system, with their setpoints (continued) within the specified Allowable Value of SR 3.3.4.1.4. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the associated recirculation pump drive motor breakers.
Allowable Values are specified for each ATWS-RPT Function specified in the LCO. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated O device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the design ~ analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe er;vironment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. The individual Functions are required to be OPERABLE in MODE 1 to protect against common mode failures of the ! Reactor Protection System by providing a diverse trip to l mitigate the consequences of a postulated ATWS event. The Reactor Vessel Pressure-High and Reactor Vessel Water (continued) ( Brunswick Unit 2 8 3.3-93 Revision No.
ATWS-RPT Instrumentation B 3.3.4 I BASES APPLICABLE Level-Low Level 2 Functions are required to be OPERABLE in SAFETY ANALYSES, MODE 1, since the reactor is producing significant power and LCO, and ' the recirculation system could be at high flow. During this APPLICABILITY MODE, the potential exists for pressure increases or low (continued) water level, assuming an ATWS event. In MODE 2, the reactor is at low power and the recirculation system is at low flow; thus, the potential is low for a pressure increase or low water level, assuming an ATWS event. Therefore, the ATWS-RPT is not necessary. In MODES 3 and 4, the reactor is shut down with all control rods inserted; thus, an ATWS event is not significant and the possibility of a significant pressure increase or low water level is negligible. In MODE 5, the one rod out interlock ensures that the reactor remains subcritical; thus, an ATWS event is not significant. In addition, the reactor pressure vessel (RPV) head is not fully tensioned and no pressure transient threat to the reactor coolant pressure boundary (RCPB) exists. The specific Applicable Safety Analyses and LCO discussions l are listed below on a Function by Function basis. '
- a. Reactor Vessel Water level-tow level 2 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
Therefore, the ATWS-RPT System is initiated at Level 2 to aid in maintaining level above the top of the active fuel. The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the i rate of coolant boiloff. 4 Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a conttant column of water ' (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-tow Level 2, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Fun ~ction on a valid signal. The Reactor Vessel Water Level-Low Level 2 Allowable Value is (continued) Brunswick Unit 2 B 3.3-94 Revision No.
f ATWS-RPT Instrumentation B 3.3.4.1 ! /~'} b BASES APPLICABLE a. Reactor Vessel Water level-Low level 2 (continued) l SAFETY ANALYSES, LCO, and chosen so that the system will not be initiated after APPLICABILITY a Level 1 scram with feedwater still available, and l for convenience with the reactor core isolation l cooling initiation. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. I
- b. Reactor Vessel Pressure-Hiah 1 Excessively high RPV pressure may rupture the RCPB.
An increase in the RPV pressure during reactor operation compresses the steam voids and results in a : l positive reactivity insertion. This increases neutron ' l flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Vessel Pressure-High Function initiates an RPT for transients that result in a pressure increase, ! counteracting the pressure increase by rapidly ! reducing core power generation. For the 'n overpressurization event, the RPT aids in the 'Q termination of the ATWS event and, along with the safety / relief valves, limits the peak RPV pressure to less than the ASME Section III Code Service Level C limits (1500 psig). The Reactor Vessel Pressure-High signals are initiated i from four pressure transmitters that monitor reactor vessel pressure. Four channels of Reactor Vessel Pressure-High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an AM-RPT from this function on a valid signal. The Reactor Vessel Pressure-High Allowable Value is chosen to provide ar adequate margin to the ASME Section III Code Service Level C allowable Reactor Coolant System , pressure. l l ACTIONS A Note has been provided to modify the ACTIONS related te ATWS-RPT instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into (continued) k Brunswick Unit 2 B 3.3-95 Revision No. I l l
l l ATWS-RPT Instrumentation B 3.3.4.1 i BASES ACTIONS the Condition. Section 1.3 also specifies that Required (continued) Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable ATWS-RPT instrumentation channels provide appropriate compensatory nieasures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable ATWS-RPT instrumentation channel. A.1 and A.2 With one or more channels inoperable, but with ATWS-RPT capability for each Function maintained (refer to Required Actions B.1 and C.1 Bases), the ATWS-RPT System is ccpable of performing the intended function. However, the reliability and redundancy of the ATWS-RPT instrumentation is reduced, such that a single failure in the remaining trip system could result in the inability of the ATWS-RPT System to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to OPERABLE status. Because of the diversity of sensors ( available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of ATWS-RPT, 14 days is provided to restore the inoperable channel (Required Action A.1). Alternately, the inoperable channel may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable breaker, since this may not adequately compensate for the inoperable breaker (e.g., the breaker may be inoperable such that it will not open). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in an RPT), or if the inoperable channel is the result of an inoperable breaker, Condition D must be entered and its Required Actions taken. ! (continued) i l l l O Brunswick Unit 2 B 3.3-96 Revision No. 1
ATWS-RPT Instrumentation B 3.3.4.1 BASES ACTIONS M (continued) Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in the Function not maintaining ATWS-RPT trip capability. A Function is considered to be maintaining ATWS-RPT trip capability when sufficient channels are OPERABLE or in trip such that the ATWS-RPT System will generate a triisignal from the given Function on a valid signal, and boti recirculation pumps can be tripped. This requires two channels of the Function in the same trip system to each be OPERABLE or in trip, and the recirculation pump drive motor breakers to be OPERABLE or in trip. The 72 hour Completion Time is sufficient for the operator to take corrective action (e.g., restoration or tripping of ' channels) and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period and that one Function is still maintaining ATWS-RPT trip capabliity. M Required Action C.1 is intended to ensure that appropriate Actions are taken if multiple, inoperable, untripped channels within both Functions result in both Functions not maintaining ATWS-RPT trip capability. The description of a function maintaining ATWS-RPT trip capability is discussed in the Bases for Required Action B.1 above. The I hour Completion Time is sufficient for the operator to take corrective action and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period. D.1 and 0.2 With any Required Action and associated Completion Time not met, the plant must be brought to a MODE or other specified ; condition in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours (Required Action D.2). Alternately, the associated recirculation pump (s) may be removed from service since this (continued) l O . Brunswick Unit 2 B 3.3-97 Revision No. l j l i l
l ATWS-RPT Instrumentatt'on i B 3.3.4.1 BASES-L ACTIONS D..I and 0.2 (continued) performs the intended function of the instrumentation (Required Action D.1). The allowed Completion' Time of 6 hours is reasonable, based on operating experience, both to reach MODE 2 from full power. conditions and to remove a recirculation pump from service in an orderly manner and without challenging plant systems. SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for l performance of required Surve111ances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the a)plicable Condition ontered and Required Actions taken. T11s Note is based on the reliability analysis (Ref. 3) i i assumption of the average time required to perform channel ' Surveillance. .That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the 'Os probability that the recirculation pumps will trip when necessary. l l SR 3.3.4.1.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the pr.rameter indicated on one channel to a siellar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument-drift in one of the channeis or something even mor.e serious. A CHANNEL CHECK will detect l gross channel failure; thus, it is key to verifying the ! instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is (continued) O Brunswick Unit 2 B 3.3-98 Revision No. l
ATWS-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.1 (continued) REQUIREMENTS outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO. SR 3.3.4.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended < function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 3. SR 3.3.4.1.3 I Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in SR 3.3.4.1.4. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the
. plant design analysis. Under these conditions the setpoint must be readjusted to be equal to or more conse,rvative than accounted for in the appropriate setpoint methodology.
The Frequency of 92 days is based on the reliability analysis of Reference 3. SR' 3.3.4.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary (continuedi Brunswick Unit 2 B 3.3-99 Revision No.
I l ' ATWS-RPT Instrumentation 4 B 3.3.4.1 , BASES SURVEILLANCE SR 3.3.4.1.4 (continued) REQUIREMENTS , range and accuracy. CHANNEL-CALIBRATION leaves the channel ! adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint j methodology. ! The Frequency is based upon the assumption of a 24 month l calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. 1 SR 3.3.4.1.5 ; The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the ; design function. Therefore, if a breaker is incapable of l operating, the associated instrument channel (s) would be inoperable. The 24 month Frequency is based on the need to perform this i Surveillance under the conditions that apply during a plant l outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month ; Frequency. REFERENCES 1. UFSAR Sections 5.4.1.2.4 and 7.6.1.3.1.
- 2. 10 CFR 50.36(c)(2)(ii).
- 3. GENE-770-06-1-A, Bases for Changes To Surveillance Test Intervals and Allowed Out-of-Service Times For ;
Selected Instrumentation Technical Specifications, December 1992. iO l Brunswick Unit 2 B 3.3-100 Revision No. I
i ECCS Instrumentation B 3.3.5.1 B 3.3 INSTRUMENTATION B 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation BASES { i i BACKGROUND The purpose of the ECCS instrumentation is to initiate I appropriate responses from the systems to ensure that the fuel is adequately cooled in the event of a design basis accident or transient. i l For most anticipated operational occurrences and Design Basis Accidents (DBAs), a wide range of dependent and independent parameters are monitored. The ECCS instrumentation actuates core spray (CS), the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System, high pressure coolant injection (HPCI), Automatic Depressurization System (ADS), and the diesel generators (DGs). The equipment involved with each of these systems is described in the Bases for LC0 3.5.1, "ECCS-Operating" or LCO 3.8.1, "AC Sources-Operating." Core Soray System The CS System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Level 3 or Drywell Pressure-High coincident with Reactor Steam Dome Pressure-Low. Each of these diverse variables is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic (i.e., two trip systems) l for each Function. I The CS System initiation signal is a sealed in signal and must be manually reset. The CS System can be reset if reactor water level and high drywell pressure have been , restored. Upon receipt of an initiation signal, the CS l pumps are started approximately 15 seconds after power is j available to limit the loading of the AC power sources. ; (continued) i I O Brunswick Unit 2 8 3.3-101 Revision No. i
ECCS Instrumentation B 3.3 5 1 O .ASpression pool suction valves automatically open, and then tie CST suction ,
valve automatically closes. Two level switches are used to '
detect low water level in the CST. Either switch can cause the suppression pool suction valves to open and the CST l suction valve to close. Two level switches are also used to detect high water level in the suppression pool. Either switch can cause an automatic swap of the HPCI pump suction valves. The suppression pool suction valves also automatically open and the CST suction valve closes if high water level is detected in the suppression pool. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction path must be open before the CST suction path is automatically l isolated.
The HPCI System provides makeup water to the reactor until O the reactor vessel water level reaches the Reactor Vessel Water Level-High trip, at which time the HPCI turbine trips, which causes the turbine's stop valve and the injection valve to close. This variable is monitored by two transmitters, which are, in turn, connected to two trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a two-out-of-two logic to provide high reliability of the HPCI System. The HPCI System automatically restarts if a Reactor Vessel Water Level-Low Level 2 signal is subsequently received.
Automatic Depressurization System The ADS may be initiated by either automatic or manual means. Automatic initiation occurs when signals indicating Reactor Vessel Water Level-Low Level 3; and confirmed Reactor Vessel Water Level-Low Level 1; and CS or RHR (LPCI Mode) Pump Discharge Pressure-High are all present and the ADS Timer has timed out. There are two transmitters for Reactor Vessel Water Level-Low Level 3 and one transmitter for confirmed Reactor Vessel Water Level-Low Level 1 in (continued)
Brunswick Unit 2 B 3.3-104 Revision No.
ECCS Instrumentation B 3.3.5.1 BASES BACKGROUND Automatic Depressurization System (continued) each of the two ADS trip systems. Each of these transmitters connects to a trip unit, which then drives a relay whose contacts form the initiation logic. .
l Each ADS tri) system includes a time delay between i satisfying tie initiation logic and the actuation of the ADS valves. The ADS Timer time delay setpoint chosen is long enough that the HPCI System has sufficient operating time to recover to a level above Reactor Vessel Water Level-Low Level 3, yet not so long that the LPCI and CS Systems are unable to adequately cool the fuel if the HPCI System fails to maintain that level. An alarm in the control room is annunciated when either of the timers is timing. Resetting the ADS initiation signals resets the ADS Timers.
The ADS also monitors the discharge pressures of the four LPCI pumps and the two CS pumps. Each ADS trip system includes two discharge pressure permissive switches from one ;
CS pump and from each LPCI pump in a Division (i.e., j Division II LPCI subsystems B and D input to ADS trip l g system A, and Division I LPCI subsystems A and C input to '
ADS trip system B). The signals are used as a permissive i for ADS actuation, indicating that there is a scurce of core j coolant available once the ADS has depressurized the vessel.
One CS pump or two RHR pumps in a LPCI loop are-sufficient to permit automatic depressurization.
l The ADS logic in each trip system is arranged in two i strings. .Each string has a contact from Reactor Vessel ,
Water Level-tow Level 3. One of the two strings in each !
trip system also has a confirmed Reactor Vessel Water Level-Low Level'1 contact and an ADS Timer. All contacts in both logic strings must close, the ADS timer must time '
out, and a CS or LPCI pump discharge pressure signal must be present to initiate an ADS trip system. Either the A or B trip system will cause all the ADS relief valves to open.
Once the ADS Timer has timed out and the ADS initiation signal is present, the trip system is sealed in until manually reset. '
Manual inhibit switches are provided in the control room for the ADS; however, their function is not required for ADS OPERABILITY (provided ADS is not inhibited when required to be OPERABLE). l (continued)-
O Brunswick Unit 2 B 3.3-105 Revision No.
ECCS Instrumentation B 3.3.5.1
. BASES BACKGROUND Diesel Generators (continued)
The DGs may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Level 3 or Drywell Pressure-High coincident with Reactor Steam Dome Pressure-Low. The DGs are also initiated upon loss of volta'ge signals. (Refer to the Bases for LCO 3.3.8.1, " Loss Instrumentation," for a discussion of these of PowerEac signals.) (LOP)h of these diverse variables is monitored b four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the four trip units are connected to relays whose contacts are connected to a one-out-of-two taken twice logic to initiate all DGs. The DGs receive their initiation signals from the CS System initiation logic. The DGs can also be started manually from the control room and locally from the associated DG room. Upon receipt of a loss of coolant accident (LOCA) initiktion signal, each DG is automatically started, is ready to load within 10 seconds, and will run in standby conditions (rated voltage and frequency, with the DG output breaker open). The DGs will only energize their respective 4.16 kV emergency buses if a loss of offsite power occurs. (Refer O to Bases for LCO 3.3.8.1.) APPLICABLE The actions of the ECCS are explicitly assumed in the safety SAFETY ANALYSES, analyses of References 1, 2, and 3. The ECCS is initiated LCO, and to preserve the integrity of the fuel cladding by limiting APPLICABILITY the post LOCA peak cladding temperature to less than the 10 CFR 50.46 limits. ECCS instrumentation satisfies Criterion 3 of Reference 4. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the ECCS instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.1-1. Each function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, I where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. (continued) l O Brunswick Unit 2 B 3.3-106 Revision No. i
l 1 ECCS Instrus.entation
'B 3.3.5.1 O
t/ BASES i APPLICABLE Allowable Values are specified for each ECCS Function SAFETY ANALYSES, specified in the table. Trip setpoints are specified in the LCO, and setpoint calculations. The setpoints are selected to ensure APPLICABILITY that the trip settings do not exceed the Allowable Value
.(continued) between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if 1 its actual trip setting is not within its required Allowable Value. Tri > setpoints are those predetermined values of output at witch an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors.
The Allowable Values are then determined, based on the trip l setpoint values, by accounting for calibration based errors. I These calibration based errors are limited to instrument I drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. O The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions that may require ECCS (or DG) initiation to mitigate the consequences of a design basis transient or accident. Table 3.3.5.1-1 footnotes (a), (b), and (c) specifically indicate other conditions when certain ECCS Instrumentation Functions are required to be OPERABLE. To ensure reliable ECCS and DG function, a combination of Functions is required to provide primary and secondary initiation signals. The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a function by Function basis. (continued) O Brunswick Unit 2 B 3.3-107 Revisien No. i l
L ECCS Instrumentation B 3.3.5.I~ BASES , APPLICABLE Core Sorav and Low Pressure Coolant Iniection Systems SAFETY-ANALYSES . 1.a. 2.a. Reactor Vessel Water Level-tow Level 3 LCO, and APPLICABILITY (continued)~ Low reactor pressure vessel-(RPV) water level indicates that the capability to cool the fuel may be thre:tened. Should l ' RPV water level decrease too far, fuel damage could result. The low pressure ECCS and associated DGs are initiated at Reactor Vessel Water Level-Low Level 3 to ensure that core < spray and flooding functions are available to prevent or ! minimize fuel damage. - The Reactor Vessel Water Level-Low Level 3 is one of the Functions assumed to be OPERABLE and capable of initiating the ECCS and associated DGs during the transients analyzed in References I and 3. In addition, the Reactor Vessel Water. Level-Low Level 3 Function is directly assumed in the analysis of the recirculation line break (Ref. 5). The core cooling function of the ECCS, along with ! the scram action of the Reactor Protection System (RPS), 1- ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Reactor Vessel Water Level-Low Level 3 signals are initiated from four level transmitters that sense the O difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) .in the vessel. The Reactor Vessel Water Level-Low Level 3 Allowable Value is chosen to allow time for the low pressure core flooding systems to activate and provide adequate cooling. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. Four' channels of Reactor Vessel Water Level--Low Level 3 Function are only required to be OPERABLE when the ECCS or DG(s) are required to be OPERABLE to ensure that no single instrument failure can preclude ECCS and DG initiation. Refer to LCO 3.5.1 and LCO 3.5.2, "ECCS-Shutdown," for Applicability Bases for the low pressure ECCS subsystems; and LC0 3.8.1 and LCO 3.8.2, "AC Sources-Shutdown," for Applicability Bases for the DGs. 1.b. 2.b. Drywell Pressure-Hich' High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). The low pressure
- (continued)
O Brunswick Unit 2 B 3.3-108 Revision No. 1
l I l l ECCS Instrumentation ) B 3.3.5.1 l r ( BASES 1 APPLICABLE 1.b. 2.b. Drywell Pressure-Hiah (continued) SAFETY ANALYSES, LCO, and ECCS and associated DGs are initiated upon receipt of the APPLICABILITY Drywell Pressure-High Function coincident with Reactor Steam Dome Pressure-Low function in order to minimize the possibility of fuel damage. The Drywell Pressure-High function is directly assumed in the analysis of the recirculation line break (Ref. 5). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible to be indicative of a LOCA inside primary containment. The Drywell Pressure-High Function is required to be OPERABLE when the ECCS or DG is required to be OPERABLE in conjunction with times when the primary containment is rcquired to be OPERABLE. Thus, four channels of the CS and LPCI Drywell Pressure-High Functions are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single (m~) instrument failure can preclude ECCS and DG initiation. In H0 DES 4 and 5, the Drywell Pressure-High function is not required, since there is insufficient energy in the reactor to pressurize the primary containment to Drywell Pressure-High setpoint. Refer to LCO 3.5.1 for Applicability Bases for the low pressure ECCS subsystems and to LC0 3.8.1 for Applicability Bases for the DGs. 1.c. 2.c. Reactor Steam Dome Pressure-tow i Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS subsystems. This ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure. The low reactor steam dome 3ressure signals are also used in the Drywell Pressure-Higi logic circuits to distinguish high drywell pressure caused by a LOCA from that caused by loss of drywell cooling. The Reactor Steam Dome Pressure-Low is one of the functions assumed to be OPERABLE and capable of permitting initiation of the ECCS and i associated DGs during the transients analyzed in References 2 and 3. In addition, the Reactor Steam Dome (continued) i Brunswick Unit 2 B 3.3-109 Revision No. l
ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 1.c. 2.c. Reactor Steam'na-. Pressure-tow (continued) SAFETY ANALYSES, LCO, and Pressure-Low Function is directly assumed in the analysis APPLICABILITY of the recirculation line break (Ref. 5). The core cooling function of the ECCS, along with the scram action of the RPS, ensures-that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Steam Dome Pressure-Low signals are initiated-from four pressure transmitters that sense the reactor done pressure. The Allowable Value is low enough to prevent overpressuring the equipment in the low pressure ECCS, but high enough to . ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of 10 CFR 50.46. Four channels of Reactor Steam Dome Pressure-Low function are only required to be OPERABLE when the ECCS or DG(s) are required to be OPERABLE to ensure that no single instrument failure can preclude ECCS and DG initiation. Refer to LC0 3.5.1 and LC0 3.5.2 for Applicability Bases for the low O pressure ECCS subsystems; and LC0 3.8.1 and LC0 3.8.2 for Applicability Bases for the DGs. 1.d. 2. f. Core Sorav and RHR Pumo Start-Time Delav Relays The purpose of these time delays is to stagger the start of the CS and RHR pumps that are in each of Divisions I and II, thus limiting the starting transients on the 4.16 kV emergency buses. These Functions are necessary when power is being supplied from either the normal power sources (offsite power) or the standby power sources (DGs). The Core. Spray Pump Start-Time Delay Relays and the RHR Pum) Start-Time Delay Relays are assumed to be OPERABLE in t te accident and transient analyses requiring ECCS initiation. That is, the analyses assume that the pumps will initiate when required and excess loading will not cause failure of the power sources. There are eight RHR Pump Start-Time Delay Relays, two channels in each of the RHR pump start logic circuits. There are six CS pump start timers arranged such that there are four separate channels of the Core Spray Pump Start
' Time-Delay Relay Function, two channels in each of the CS (continued) 1 Brunswick Unit 2 B 3.3-110 Revision No.
ECCS Instrumentatton B 3.3.5.1 0 #SeS APPLICABLE 1.d.'t.f. Core Sorav and RHR Pumo Start-Time Delav Relays SAFETY ANALYSES (continued) LCO, and APPLICABILITY pump start logic circuits. Each channel consists of an individual lo second timer and a 5 second timer. The 5 second timer is common to both channels associated with a CS pump start logic circuit. Each 10 second timer associated with a CS pump start logic channel is shared with an RHR pump start logic channel. While two time delay relay channels are dedicated to a single CS pump start logic, a single failure of a 5 second CS pump timer could result in the failure of the two low pressure ECCS pum>s, powered from the same 4.16 kV emergency bus, to perform tieir intended function within the assumed ECCS RESPONSE TIME (e.g., as in the case where both ECCS pumps on one 4.16 kV emergency bus start simultaneously due to an inoperable time delay relay). This still leaves four of the six low pressure ECCS pumps OPERABLE. Additionally, a failure of both shared time delay relay channels in an RHR and CS pump start logic circuit would also leave four of the six low pressure ECCS pumps OPERABLE as described above. As a result, to satisfy the single failure criterion (i.e., O loss of one instrument does not preclude ECCS initiation), only one channel per pump of the Core Spray and RHR Pump Start-Time Delay Relay Functions are required to be OPERABLE when the associated ECCS subsystem is required to be OPERABLE. Refer to LCO 3.5.1 and LCO 3.5.2 for Applicability Bases for the ECCS subsystems. The Allowable Values for the Core Spray and RHR Pump Start-Time Delay Relays are chocen to be long enough so that most of the starting transient of the previously started pump is complete before starting a subsequent pump on the same 4.16 kV emergency bus and short enough so that ECCS operation is not degraded. 1.d. Reactor Steam Dome Pressure-Low (Recirculation Pumo Discharoe Valve Permissive) Low reactor steam dome pressure signals are used as permissives for recirculation pum) discharge valve closure and recirculation' pump discharge aypass valve closure. This ensures that the LPCI subsystems inject into the proper RPV location assumed in the safety analysis. The Reactor Steam (continued) Brunswick Unit 2 8 3.3-111 Revision No. 1
ECCS Instrumentation B 3.3.5.1 O V BASES APPLICABLE 2.d. Reactor Steam Dome Pressure-Low (Recirculation Pumo SAFETY ANALYSES Discharae Valve Permissive) (continued) LCO, and APPLICABILITY Dome Pressure-Low is one of the Functions assumed to be OPERABLE and capable of closing the valve (s) during the transients analyzed in References 2 and 3. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Steam Dome Pressure-Low function is directly assumed in the analysis of the recirculation line break (Ref. 5). The Reactor Steam Dome Pressure-Low signals are initiated from four pressure transmitters that sense the reactor dome pressure. The Allowable Value is chosen to ensure that the valves I close prior to commencement of LPCI injection flow into the core, as assumed in the safety analysis. Four channels of the Reactor Steam Dott.e Pressure-Low function are only required to be OPERABLE in MODES 1, 2, and 3 with the associated recirculation pump discharge valve O open or the associated recirculation pum) discharge bypass valve open. With the valve (s) closed, t.ie function of instrumentation has been performed; thus, the Function is not required. In MODES 4 and 5, the loop injection location is not critical since LPCI injection through the j recirculation loop in either direction will still ensure that LPCI flow reaches the core (i.e., there is no significant reactor steam dome back pressure). 2.e. Reactor Vessel Shroud Level The Reactor Vessel Shroud Level Function is provided as a permissive to allow the RHR System to be manually aligned from the LPCI mode to the suppression pool cooling / spray or drywell spray modes. The permissive ensures that water in the vessel is at least two thirds core height before the manual transfer is allowed. This ensures that LPCI is available to prevent or minimize feel damage. This function may be overridden during accident conditions as allowed by plant procedures. The Reactor Vessel Shroud Level function is implicitly assumed in the analysis of the recirculation (continued) O Brunswick Unit 2 8 3.3-112 Revision No.
ECCS Instrumentation B 3.3.5.1 BASES-
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I APPLICABLE 2.e. Reactor Vessel Shroud Level (continued) SAFETY ANALYSES, _ LCO, and line break (Ref. 5) since the analysis assumes that no LPCI APPLICABILITY J flow diversion occurs when reactor water level is below the Reactor Vessel Shroud Level. Reactor Vessel Shroud Level' signals are initiated from two level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Reactor Vessel Shroud Level Allowable Value is chosen to allow the low pressure core flooding systems to activate and provide adequate cooling before allowing a manual transfer. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. Two channels of the Reactor Vessel Shroud Level Function are 1 only required to be OPERABLE in MODES 1, 2, and 3. In ! MODES 4 and 5, the specified initiation time of the LPCI ' subsystems is not assumed, and other administrative controls are adequate to control the valves that this Function
) isolates (since the systems that the valves are opened for are not required to be OPERABLE in MODES 4 and 5 and are normally not used).
HPCI System 3.a. Reactor Vessel Water level-Low Level 2 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the HPCI System is initiated at Level 2 to maintain level above the top of the active fuel. The Reactor Vessel Water Level-Low Level 2 is one of the Functions assumed ~ to be OPERABLE and
. capable of initiating HPCI during the transients analyzed in References 2, 3, and 6.
Reactor Vessel Water Level-Low Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. (continued) O Brunswick Unit 2 B 3.3-113 Revision No.
1 ECCS Instrumentation ) B 3.3.5.1 BASES APPLICABLE 3.a. Reactor Vessel Water Level-Low Level 2 (continued) SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level-Low. Level 2 Allowable Value APPLICABILITY is low enough to avoid a HPCI System start from normal reactor level transients (e.g., a reactor scram without the ] loss of feedwater flow) and high enough to avoid. initiation of low pressure ECCS at Reactor Vessel Water Level-Low Level 3 during a transient resulting from a complete loss of feedwater flow. The Allowable Value is referenced from l reference level zero. Reference level zero is 367 inches l above the vessel zero point. . i Four channels of Reactor Vessel Water Level-Low Level 2 Function are required to be OPERABLE only when HPCI is i required to be OPERABLE to ensure that no single instrument failure can preclude HPCI initiation. Refer to LC0 3.5.1 for HPCI Applicability Bases. , 3.b. Drywell Pressure-Hiah High pressure in the drywell could indicate a break in the RCPB. The HPCI System is initiated upon receipt of the Drywell Pressure-High Function in order to minimize the O possibility of fuel damage. The Drywell Pressure-High - Function-is not assumed in accident or transient analyses. It is retained since it is a potentially significant contributor to risk. High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible to be indicative of a LOCA inside primary containment. Four channels of the Drywell Pressure-High Function are required to be OPERABLE when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude , HPCI initiation. Refer to LCO 3.5.1 for the Applicability Bases for the HPCI System, i 3.c. Reactor Vessel Water Level-Hiah ; High RPY water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the Reactor Vessel Water Level-High signal is used to trip the HPCI turbine to i prevent overflow into the main steam lines (MSLs) which precludes an unanalyzed event. (continued) O Brunswick Unit 2 B 3.3-114 Revision No. j
1 ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE 3.c. Reactor Vessel Water Level-Hiah (continued) l SAFETY ANALYSES, L LCO, and Reactor Vessel Water Level-High signals for HPCI are i APPLICABILITY initiated from two level transmitters from the narrow range water level measurement instrumentation. Both Reactor Vessel Water Level-High signals are required in order to close the HPCI turbine stop valve. This ensures that no-single instrument failure can preclude HPCI initiation. The Reactor Vessel Water Level-High Allowable Value is high enough to avoid interfering with HPCI System operation during reactor water level recovery resulting from low reactor water level events and low enough to prevent flow from'the HPCI System from overflowing into the MSLs. -The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. ! Two channels of Reactor Vessel Water Level-High Function are required to be OPERABLE only d en HPCl is required to be OPERABLE. Refer to LC0 3.5.1 for HPCI Applicability Bases. !o L l 3.d Condensate Storaae Tank level-Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source. Normally the suction valves between HPCI and the CST are open and, upon receiving a HPCI initiation signal, water for HPCI injection would be taken from the CST. However, if the l water level in the CST falls below a preselected level,. first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes. This ensures that an adequate supply of makeup water is available to the HPCI pump. To prevent losing suction to the pump, the suction valves are interlocked so that.the suppression pool suction valves must be open before the CST l suction valve automatically closes. The function is ! implicitly assumed in the accident and transient analyses (which take credit for HPCI) since the analyses assume that the HPCI suction source is the suppression pool. i The Condensate Storage Tank Level-Low signal is initiated from two level _ switches. The logic is arranged such that either level switch can cause the suppression pool suction valves to open and the CST suction valve to close. The (continued) Brunswick Unit 2 B 3.3-115 Revision No. i
l ECCS Instrumentation B 3.3.5.1 I n . BASES APPLICABLE 3.d Condensate Storace Tank level-tow (continued) SAFETY ANALYSES LCO, and Condensate Storage Tank Level-Low function Allowable Value APPLICABILITY is high enough to ensure adequate pump suction head while - water is being taken from the CST. Two channels of the Condensate Storage Tank Level-Low Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument l failure can preclude HPCI swap to suppression pool source. Refer to LCO 3.5.1 for HPCI Applicability Bases. 3.e. Suporession Chamber Water level-Hiah Excessively high suppression pool water could impact operation of the HPCI and Reactor Core Isolation Cooling (RCIC) exhaust vacuum breakers resulting in an inoperable. HPCI or RCIC System. Therefore, signals indicating high suppression pool water level are used to transfer the suction source of HPCI from the CST to the suppression pool to eliminate the possibility of HPCI continuing to provide additional water from a source outside containment. To
,o prevent losing suction to the pump, the suction valves are
(') interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes. l l This Function is implicitly assumed in the accident and transient analyses (which take credit for HPCI) since the analyses assume that the HPCI suction source is the suppression pool. The Suppression Chamber Water Level-High signal is initiated from two level switches. The logic is arranged such that either switch can cause the suppression pool 4 suction valves to open and the CST suction valve to close. The Allowable Value for. the Suppression Chamber Water Level-High Function is chosen to ensure that HPCI will be aligned ~f 6i suctiori from the suppression pool before the water level reaches the point at which the HPCI and RCIC exhaust vacuum breakers become inoperable. The Allowable , Value is referenced from the suppression chamber water level ' zero. Suppression chamber water level zero is one inch below the torus centerline. (continued) l l lO l Brunswick Unit 2. B 3.3-116 Revision No. u
I ECCS Instrumentation B 3.3.5.1 l BASES l APPLICABLE 3.e. Suporession Chamber Water Level-Hiah (continued) SAFETY ANALYSES' LCO, and. Two channels of Suppression Chamber Water Level-High APPLICABILITY Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI swap to suppression pool source. Refer to LCO 3.5.1 for HPCI Applicability Bases. Automatic Depressurization System (ADS) l 4.a. 5.a. Reactor Vessel Water Level-Low Level 3 Low RPV water level indicates that the capability to cool i the fuel may be threatened. Should RPV water level decrease i too far, fuel damage could result... Therefore, ADS receives oie of the signals necessary for initiation from this Function. The Reactor Vessel Water Level-Low Level 3 is one of the functions assumed to be OPERABLE and capable of initiating the ADS during the accident analyzed in References 2 and 5. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the l fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Reactor Vessel Water Level-Low Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level-(variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Level 3 Function are required to be OPERABLE only when ADS is required to be L OPERABLE to ensure that no single instrument failure can L preclude ADS initiation. Two channels input to ADS trip system A, while the other two channels input to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases. J The Reactor Vessel Water Level-Low Level 3 Allowable Value I is chosen to alinw time for the low pressure core flooding
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systems to initiate and provide adequ' ate cbolin6 lhe ~~~ Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero l point. !' (continued) j l G < Brunswick Unit 2 B 3.3-117 Revision No. o
F ECCS Instrumentation i B 3.3.5.1 BASES APPLICABLE 4.b. 5.b. ADS Timer SAFETY ANALYSES l LCO, and- - The purpose of the ADS Timer is to delay depressurization of APPLICABILITY the reactor vessel to allow the HPCI System time to maintain
-(continued) reactor vessel water level. Since the rapid depressurization caused by ADS operation is one of the most severe transients on the reactor vessel, its occurrence should be limited. By delaying initiation of the ADS Function, the operator is given the chance to monitor the success or failure of the HPCI System to maintain water level, and then to decide whether or not to allow ADS to initiate, to delay initiation further by recycling the timer, or to inhibit initiation permanently. The ADS Timer Function is assumed to be OPERABLE for the accident analyses of References 2 and 5 that require ECCS initiation and assume failure of the HPCI System.
There are two ADS Timer relays, one in each of the two ADS trip systems. The Allowable Value for the ADS Timer is chosen to be long enough to allow HPCI to start and avoid an inadvertent blowdown yet short enough so that there is still time after depressurization for the low pressure ECCS subsystems to provide adequate core cooling. O Two channels of the ADS Timer Function are only required to l be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. One channel inputs to ADS trip system A, while the other channeT inputs to ADS trip system B. Refer to LC0 3.5.1 for ADS App 1tcability Bases. 4.c. 5.c. Reactor Vessel Water Level-Low level 1 The Reactor Vessel Water Level-Low Level 1 Function is used by the ADS only as a confirmatory low water level signal. ADS receives one of the signals necessary for initiation l from Reactor Vessel Water Level-Low Level 3 signals. In order to prevent spurious initiation of the ADS due to spurious Level 3 signals, a Level 1 signal must also be I received before ADS initiation commences. I Reactor Vessel Water Level-Low Level 1 signals are initiated from two level transmitters that sense the l difference between the pressure due to a constant column of I water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Allowable (continued) O Brunswick Unit 2 B 3.3-118 Revision No.
ECCS Instrumentation BASES-APPLICABLE 4.c. 5.c. Reactor Vessel Water level-Low Level 1 SAFETY ANALYSES, (continued) LCO, and APPLICABILITY Value for Reactor Vessel Water Level--Low Level 1 is selected at the RPS Level 1 scram Allowable Value for convenience. Refer to LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation," for the Bases discussion of this Function. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. Two channels of Reactor Vessel Water Level-Low Level 1 Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. One channel inputs to ADS trip system A, while the other channel inputs to ADS trip system B. Refer to LC0 3.5.1 for ADS Applicability 1 Bases. 4.d. 4.e. 5.d. 5.e. Core Soray and RHR (LPCI Model Pump Discharae Pressure-Hiah The Pump Discharge Pressure-High signals from the CS and O RHR pumps are used as permissives for ADS initiation, indicating that there is a source of low pressure cooling l water available once the ADS has depressurized the vessel. Pump Discharge Pressure-High is one of the Functions assumed to be OPERABLE and capable of permitting ADS initiation during the events analyzed in References 2 and 5 with an assumed HPCI failure. For these events the ADS depressurizes the reactor vessel so that the low pressure ECCS can perform the n re cooling functions. This core , cooling function of t'i ECCS, along with the' scram action of ' the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Pump discharge pressure signals are initiated from twelve pressure switches, two on the discharge side of each of the six low pressure ECCS pumps. In order to generate an ADS permissive in one trip system, it is necessary that only one CS pump (both channels for the pump) indicate the high discharge pressure conditten or two RHR pumps in one LPCI loop (one channel for each pump) indicate a high discharge pressure condition. The Pump Discharge Pressure-High Allowable Value is less than the pump discharge pressure when the pump is operating at all t~1ow ranges and high (continued) O Brunswick Unit 2 . B 3.3-119 Revision No.
3 ECCS Instrumentatien B 3.3.5.1 , IN lV BASES l APPLICABLE 4.d. 4.e. 5.d. 5.e. Core Sorav and RHR (LPCI Mode) Pumo l SAFETY ANALYSES, Discharae Pressure-Hiah (continued) LCO, and APPLICABILITY enough to avoid any condition that results in a discharge pressure permissive when the CS and LPCI pumps are aligned for injection and the pumps are not running. The actual ; operating point of this function is not assumed in any transient or accident analysis. Twelve channels of Core Spray and RHR (LPCI Mode) Pump Discharge Pressure-High Functions are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two CS channels associated with CS pump B and four LPCI channels associated with RHR pumps B and D are required for trip system A. Two CS channels associated with CS pump A and four LPCI channels associated with RHR pumps A and C are required for trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases. ACTIONS A Note has been provided to modify the ACTIONS related to ECCS instrumentation channels. Section 1.3. Completion O Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables i expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into 4 the condition. However, the Required Actions for inoperable l ECCS instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable ECCS instrumentation channel. 8.d . Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.1-1. The applicable Condition referenced in the Table is function dependent. Each time a channel is discovered inoperable. Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition. (continued) l O Brunswick Unit 2 B 3.3-120 Revision No.
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS B.I. B.2. and B.3 (continued) Required Actions B.1 and B.2 are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in redundant automatic initiation capability being lost for the feature (s). Required Action B.1 features would be those that are initiated by Functions 1.a. 1.b. 2.a and 2.b (e.g.,lowpressureECCS). The Required Action B.2 system would be HPCI. For Required Action B.1, redundant automatic initiation capability is lost if (a) two Function 1.a channels are inoperable and untripped in the same trip system, (b) two Function 2.a channels are inoperable and untripped in the same trip system,-(c) two Function 1.b channels are inoperable and untripped in the same system, or (d) two Function 2.b channels are inoperable and untripped in the same trip system. For low pressure ECCS, since each inoperable channel would have Required Action B.1 applied separately (refer to ACTIONS Note), each ino)erable channel would only require the affected portion of tie associated system of low pressure ECCS and DGs to be decla'red inoperable. However, since channels in both associated low pressure ECCS subsystems (e.g., both CS subsystems) are O inoperable and untripped, and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in the associated low pressure ECCS and DG being concurrently declared inoperable. For Required Action B.2, redundant automatic initiation capabliity is lost if two Function 3.a or two Function 3.b channels are inoperable and untripped in the same trip system. In this situation (loss of redundant automatic initiation capability), the 24 hour allowance of Required Action B.3 is ' not appropriate and the feature (s) associated with the ino)erable, untripped channels must be declared inoperable witiin I hour. As noted (Note I to Required Action B.1), Required Action B.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the low pressure ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of initiation capability for 24 hours (as allowed by Required Action B.3) is allowed during MODES 4 and 5. There is no similar Note provided for Required Action B.2 since HPCI instrumentation is not required in MODES 4 and 5; thus, a Note is not necessary. ' (continued) O Brunswick Unit 2 B 3.3-121 Revision No.
ECCS Instrumentation B 3.3.5.1 BASES . ACTIONS B.I. B.2. and B.3 (continued) Notes are also provided (Note 2 to Required Action B.1 and the Note to Required Action B.2) to delineate which Required Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable. This ensures that the proper loss of initiation capability check is performed. Required Action B.1 (the Required Action for certain inoperable channels in the low pressure ECCS subsystems) is not applicable to Function 2.e, since
- i. this Function provides backup to administrative controls l ensuring that operators do not divert LPCI flow from injecting into the core when needed. Thus, a total loss of Function 2.e capability for 24 hours is allowed, since the LPCI subsystems remain capable of performing their intended i function.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock."
For Required Action B.1, the Completion Time only begins upon discovery that a redundant feature in the same system O (e.g., both CS subsystems) cannot be automatically initiated due to inoperable, untripped channels within the same Function as described in the paragraph above. For Required Action B.2, the Completion Time only begins upon discovery that the HPCI System cannot be automatically initiated due to two inoperable, untripped channels for the associated Function in the same trip system. The I hour Completion I Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. l Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acce) table (Ref. 7) to permit restoration of any inopera)1e channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.3. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. (continued) l O Brunswick Unit 2 B 3.3-122 Revision No. l l
I i l ECCS Instrumentation l 8 3.3.5.1 BASES ACTIONS B.I. B.2. and B.3 (continued) Alternately, if it is not desired to place the channel in l trip (e.g., as in the case where pir.cing the inoperable ' channel in trip would result in an initiation), Condition G ! I must be entered and its Required Action taken. ' 1 C.1 and C.2 Required Action C.1 is intended to ensure that appropriate l actions are taken if multiple, inoperable channels within l the same Function result in redundant automatic initiation capability being lost for the feature (s). Required Action C.1 features would be those that are initiated by Functions 1.c, l.d, 2.c, 2.d, and 2.f (i.e., low pressure ECCS). Redundant automatic initiation capability is lost if either (a) two Function 1.c channels are inoperable in the same trip system, (b) two Function 2.c channels are inoperable in the same trip system, (c) two Function 2.d channels are inoperable in the same trip system, or (d) two or more required Function 1.d and 2.f channels associated with low pressure ECCS pumps powered from separate 4.16 kV ' O emergency buses are inoperable. Since each inoperable channel would have Required Action C.1 applied separately (refer to ACTIONS Note), each ino)erable channel would only require the affected portion of tie associated system of low pressure ECCS and DGs to be declared inoperable. However, since channels for both associated low pressure ECCS subsystems are inoperable (e.g., both CS subsystems), and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in the associated low pressure ECCS and DGs being concurrently l declared inoperable. For Functions 1.d and 2.f the l affected portions are the associated low pressure ECCS , l pumps. ! In this situation (loss of redundant automatic initiation l capability), the 24 hour allowance of Required Action C.2 is not appropriate and the feature (s) associated with the ; inoperable channels must be declared inoperable within ! I hour. As noted (Note 1 to Required Actions C.1), hquired Action C.1 is only applicable in MODES 1, 2, and 3. In i MODES 4 and 5, the specific initiation time of the ECCS is ! not assumed and the probability of a LOCA occurring during l j (continued) l Brunswick Unit 2 B 3.3-123 Revision No.
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS C.1 and C.2 (continued) the period the channels are inoperable is low. Thus, a total loss of automatic initiation capability for 24 hours (as allowed by Required Action C.2) is allowed during MODES 4 and 5. Note 2 to Required Action C.1 states that it is only applicable for Functions 1.c, 1.d 2.c, 2.d, and 2.f. Required Action C.1 is not applicable to function 3.c'(which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic). This loss was considered during the development of Reference 7 and considered acceptable for the 24 hours allowed by Required Action C.2. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock."
For Required Action C.1, the Completion Time only begins upon discovery that the same feature in both subsystems O (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels. Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acce) table (Ref. 7) to permit restoration of any inopera ale channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition G must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would either cause the initiation or it would not necessarily result in a safe state for the channel in all events. (continued) I O Brunswick Unit 2 B 3.3-124 Revision No. l l J
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS D.I. D.2.1. and 0.2.2 l (continued) Required Action D.1 is intended to ensure that appropriate l actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic component initiation capability for the HPCI i System. Automatic component initiation capability is lost if two Function 3.d channels or two Function 3.e channels ( are inoperable and untri In this situation loss of automatic suction swap),pped.the 24 hour allowance of(Requ Actions D.2.1 and D.2.2 is not appropriate and the HPCI System must be declared inoperable within 1 hour after ! discovery of loss of HPCI initiation capability. As noted, Required Action D.1 is only applicable if the HPCI pump suction is not aligned to the suppression pool, since, if aligned, the Function is already performed. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock."
For Required Action D.1, the Comple_ tion Time only begins upon discovery that the HPCI System cannot be automatically i O aligned to the suppression pool due to two inoperable, untripped channels in the same Function as described in the l paragraph above. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while. allowing time for resto ation or tripping of channels. Because of the diversity of sensors availab1' e to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has-been shown to be acceptable (Ref. 7) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel mJst be placed in the tripped condition per Required Action D.2.1 or the suction source must be aligned to the suppression pool per Required Action D.2.2. Placing the inoperable channel in trip performs the intended function of the channel (shifting the suction source to the suppression pool). Performance of either of these two Required Actions will allow operation to continue. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the HPCI System piping remains filled with water. Alternately, if it is not (continued) l O Brunswick Unit 2 B 3.3-125 Revision No. l L
i ECCS Instrumentation B 3.3.5.1 BASES ACTIONS D.1.'D.2.1. and D.2.2 (continued) desired to perform Required Actions D.2.1 and D.2.2 (e.g., as in the case where shifting the suction source could drain down the HPCI suction piping), Condition G sust be entered and its Required Action taken. E.1 and E.2 Required Action E.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within similar ADS trip system A and B Functions result in redundant automatic initiation capability being lost for the ADS. Redundant automatic initiation capability is lost if either (a) one Function 4.a channel and one Function 5.a channel are inoperable and untripped, or (b) one Function 4.c channel and one Function 5.c channel are inoperable and untripped. In this situation (loss of automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required Action E.2 is not appropriate and all ADS valves must be O declared inoperable within 1 hour after discovery of loss of ADS initiation capability. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
" time zero" for begira.kg the allowed outage time " clock."
For Requbs! Action i.1 the Completion Time only begins upon disc)ven that ti 2 ADS cannot be automatically initiated iue a im rable, untripped channels within similar ADS trip system Functions as described in the paragraph abo m The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk whits allowing time for restoration or tripping of channelt. Because of the diversity of sensors available to provide ! initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 7) to permit restoration of any inoperable i channel to OPERABLE status if both HPCI and RCIC are l OPERABLE. If either HPCI or RCIC is inoperable, the time is 1 shortened to 96 hours. If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to (continued) i O Brunswick Unit 2 8 3.3-126 Revision No.
ECCS Instrumentation B 3.3.5.1 l BASES
~
ACTIONS E.1 and E.2 (continued) 96 hours, the 96 hours begins upon discovery of HPCI or RCIC inoperability. However, the total time for an inoperable, untripped channel cannot exceed 8 days. If the status of HPCI or RCIC changes such that the Completion Time changes L from.96 hours to 8 days, the " time zero" for beginning the 3 day " clock" begins upon discovery of the inoperable, l untripped channel. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action E.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is net desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in' trip would result in an initiation), Condition G aust be entered and i its Required Action taken. F.1 and F.2 Required Action F.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within similar ADS trip system A and B Functions result in redundant automatic initiation capability being lost for the ADS. Redundant automatic initiation capability is lost if either (a) one function 4.b channel and one Function 5.b chantiel are inoperable, or-(b) a combination of Function 4.d, 4.e, 5.d, and 5.e channels are inoperable such that channels associated with both CS pumps and one RHR pump in each LPCI loop are inoperable. In this situation (loss of ' automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required Action F.2 is not appropriate and all ADS valves must be declared inoperable within I hour after diccovery of loss of ADS initiation capability. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock." '
For Required Action F.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable channels within similar ADS trip i (continued) O Brunswick Unit 2 B 3.3-127 Revision No.
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS F.1 and F.2 (continued) system Functions as described in the paragraph above. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels. Because of the diversity of sensors available to provide initiation ~ signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 7) to permit restoration of any inoperable channel to OPERABLE status if both HPCI and RCIC are OPERABLE (Required Action F.2). If either HPCI or RCIC is inoperable, the time shortens to 96 hours. If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to 96 hours, the 96 hours begins upon discovery of HPCI or RCIC inoperability. . However, the total time for an inoperable channel cannot exceed 8 days. If the status of HPCI or RCIC changes such that the Completion Time changes from 96 hours to 8 days, the " time zero" for beginning the 8 day " clock" begins upon discovery of the inoperable channel. If the ino)erable channel cannot be restored to OPERABLE status witiin the allowable out of O service time, Condition G must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events, ful With any Required Action and associated Completion Time not met, the associated feature (s) may be incapable of ; performing the intended function, and the supported ' feature (s) associated with inoperable untripped channels must be declared inoperable immediately. 1 SURVEILLANCE As noted (Note 1) in the beginning of the SRs, the SRs for REQUIREMENTS each ECCS instrumentation Function are found in the SRs column of Table 3.3.5.1-1 .+. ,, The Surve111ances are modified by a Note (Note 2) to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours as follows: (a) for Function 3.c; (continued) O Brunswick Unit 2 B 3.3-128 Revision No.
ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE and (b) for Functions other than 3.c provided the associated REQUIREMENTS Function or redundant Function maintains ECCS initiation (continued) capability. Upon completion of the Surveillance, or expiration ~ of the 6 hour allowance, the channel must be returned to OPERABLE status or the a entered and Required Actions taken. pplicable This NoteCondition is based on the reliability analysis (Ref. 7) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary. SR 3.3.5.1.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of O excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited to 24 hours; thus, it is key to verifying the instrumentation-continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The~ Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.5.1.2 and SR 3.3.5.1.6 b A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended (continued) O Brunswick Unit 2 8 3.3-129 Revision No.
ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE SR 3.3.5.1.2 and SR 3.3.5.1.6 (continued) b REQUIREMENTS function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.5.1.2 1s based on the reliability analyses of Reference 7. The 24 moath Frequency a of SR 3.3.5.1.6 is based on engineering judgment and the Cts reliability of the components. SR 3.3.5.1.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.5.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analyses. Under these conditions, the setpoint must be O readjusted to be equal to or more conservative than the setting accounted for in the appropriate setpoint methodology. The frequency of 92 days is based on the reliability analysis of Reference 7. SR 3.3.5.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. (continued) O Brunswick Unit 2 B 3.3-130 Revision No.
ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE SR 3.3.5.1.5 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic and simulated automatic operation for a specific channel. The system functional testing performed in LC0 3.5.1, LC0 3.5.2, LC0 3.8.1, and LC0 3.8.2 overlaps this Surveillance to complete testing of the assumed safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency. REFERENCES 1. UFSAR, Section 5.2.
- 2. UFSAR, Section 6.3.
- 3. UFSAR, Chapter 15.
- 4. 10 CFR 50.36(c)(2)(ii).
- 5. NEDC-31624P, Brunswick Steam Electric Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis (Revision 2), July 1990. l
- 6. GE-NE-187-26-1292, Power Uprate Transient Analysis for 1 Brunswick Steam Electric Plant Units 1 and 2, l Revision 1, November 1995.
- 7. 'NEDC-30936-P-A, BWR Owners' Group Technical Specification Improvement Methodology (With ,
Demonstration for BWR ECCS Actuation Instrumentation), ' Parts 1 and 2, December 1988. 1 O Brunswick Unit 2 B 3.3-131 Revision No.
RCIC System Instrumentaticn B 3.3.5.2 8 3.3 INSTRUMENTATION B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation BASES BACKGROUND The purpose of the RCIC System instrumentation is to j initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary haat sink (the main condenser) and normal coolant makeup flow from the Reactor Feedwater System is insufficient or unavailable, such that RCIC System initiation occurs and maintains sufficient reactor water level such that initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps does not occur. A more complete discussion of RCIC System operation is provided in the Bases of LCO 3.5.3, "RCIC System." The RCIC System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-tow Level 2.~ The variable is monitored by four transmitters that are connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken O twice logic arrangement. The RCIC test line isolation valve is closed on a RCIC initiation signal to allow full system flow. The RCIC System also monitors the water levels in the condensate storage tank (CST) since this is the initial source of water for RCIC operation. Reactor grade water in the CST is the normal source. Upon receipt of a RCIC initiation signal, the CST suction valve is automatically signaled to open. If the water level in the CST falls belou a preselected level, first the RCIC suppression pool suctio valves automatically open, and then the RCIC CST suction I valve automatically closes. Two level switches are used to detect low water level in the CST. Either switch can cause the suppression pool suction valves to open and the CST suction valve to close (one-out-of-two logic). To prevent losing suction to the pump,-the suction valves are interlocked so that one suction path must be open before the other automatically closes. (continued) O Brunswick Unit 2 B 3.3-132 Revision No.
RCIC System Instrumentation B 3.3.5.2 h b BASES BACKGROUND The RCIC System provides makeup water to the reactor until (continued) the reactor vessel water level reaches the high water level trip (two-out-of-two logic), at which time the RCIC steam supply valve closes. The RCIC System restarts if vessel level again drops to the low level initiation point (Level 2). APPLICABLE The function of the RCIC System to provide makeup coolant to SAFETY ANALYSES, the reactor is used to respond to transient events. The LCO, and RCIC System is not an Engineered Safety Feature System and APPLICABILITY no credit is taken in the safety analyses for RCIC. System operation. Based on its contribution to the reduction of overall plant risk, however, the system, and therefore its instrumentation, meets Criterion 4 of Reference 1. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the RCIC System instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.2-1. Each Function must have a required number of OPERABLE channels with their setpoints within the t specified Allowable Values, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Allowable Values are specified for each RCIC System instrumentation Function specified in Table 3.3.5.2-1. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by (continued) Brunswick Unit 2 B 3.3-133 Revision No.
RCIC Systc2 Instrumentation L B 3.3.5.2 l BASES l l APPLICABLE accounting for calibration based errors. These calibration
- SAFETY ANALYSES, based errors are limited to instrument drift, LCO, and errors associated with measurement and test equipment, and APPLICABILITY calibration tolerance of loop components. The trip (continued) setpoints and Allowable Values determined in this manner provide adequate prote: tion because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.
The individual Functions are required to be OPERABLE in MODE 1, and in MODES 2 and 3 with reactor steam dome pressure > 150 psig since this is when RCIC is. required to be OPERABLE. Refer to LCO 3.5.3 for Applicability Bases for the RCIC System. The specific Applicabic Safety Analyses, LCO, and Applicability discussions are listed below on a Function by function basis.
- 1. Reactor Vessel Water Level-Low Level 2 Low reactor pressure vessel (RPV) water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the RCIC System is initiated at Level 2 to assist in maintaining water level above the top of the active fuel, Reactor Vessel Water Level-Low Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. ;
The Reactor Vessel Water Level-Low Level 2 Allow 6le Value is set high enough such that for complete loss of feedwater flow, the RCIC System flow with high pressure coolant j injection assumed to fail will be sufficient to avoid ; initiation of low pressure ECCS at Level 3. The Allowable j Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. (continued) lO i Brunswick Unit 2 B 3.3-134 Revision No. I I
RCIC System Instrumentation B 3.3.5.2 BASES APPLICABLE 1. Reactor Vessel Water Level-tow level 2 (continued) SAFETY ANALYSES, LCO, and Four channels of Reactor Vessel Water Level-Low Level 2 APPLICABILITY Function are available and are required to be OPERABLE when RCIC is required to be OPERAELE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LC0'3.5.3 for RCIC Applicability Bases.
- 2. Reactor Vessel Water Level-Hioh High RPV water level indicates that sufficient cooling water inventory 9.ists in the reactor vessel such that there is no danger to the fuel. Therefore, the high water level signal is used to close the RCIC steam supply valve to prevent overflow into the main steam lines (MSLs).
Reactor Vessel Water Level-High signals for RCIC are initiated from two level transmitters from the narrow range water level measurement instrumentation, which sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Reactor Vessel Water Level-High Allowable Value is high enough to preclude isolating the injection valve of the RCIC during normal operation, yet low enough to trip the RCIC System to prevent reactor vessel overfill. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. Two channels of Reactor Vessel Water Level-High Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.
- 3. Condensate Storaae Tank Level-Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source.
Normally, the suction valve between the RCIC pump and the CST is open and, upon receiving a RCIC initiation signal, water for RCIC injection would be taken from the CST. However, if the water level in the CST falls below a preselected level, first the suppression pool suction valves (continued) Brunswick Unit 2 8 3.3-135 Revision No.
RCIC System Instrumentatica B 3.3.5.2 BASES APPLICABLE 3. Condensate Storace Tank Level-Low (continued) SAFETY ANALYSES, LCO, and automatically open, and then the CST suction valve APPLICABILITY automatically closes. This ensures that an adequate supply of makeup water is available to the RCIC pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes. Two level switches are used to detect low water level in the CST. The Condensate Storage Tank Level-Low Function Allowable Value is set high enough to ensure adequate pump suction head while water is being taken from the CST. Two channels of Condensate Storage Tank Level-Low Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to suppression pool source. Refer to LC0 3.5.3 for RCIC Applicability Bases. ACTIONS A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels. Section 1.3, g Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System instrumentation channels provide a)propriate compensatory measures for separate inoperable ciannels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel. Ad Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.2-1. The applicable Condition referenced in the Table is Function dependent. Each time a channel is discovered to be inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition. (continued) O !v Brunswick Unit 2 B 3.3-136 Revision No.
RCIC System Instrumentation B 3.3.5.2 BASES ACTIONS B.1 and B.2 )- -(continued) Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete _ loss of automatic initiation capability for the RCIC System. In this case, automatic initiation capability is lost if two Function I channels in the same trip system are inoperable and untripped. In this situation (loss of automatic initiation capability), the 24 hour allowance of Required ; Action B.2 is not appro>riate, and the RCIC System must be- l declared inoperable witiin I hour after discovery of loss of RCIC initiation capability. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock."
For Required Action B.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically initiated due to two inoperable, untripped Reactor Vessel j Water Level-Low Level 2 channels in the same trip system.
/ The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 2) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition E must be entered and its Required Action taken. 1 (continued) O Brunswick Unit 2 B 3.3-137 Revision No.
RCIC System Instrumentation B 3.3.5.2 , BASES ACTIONS Ed (continued) A risk based analysis was performed and determined that an allowable out of service time of 24 hours (Ref. 2) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.1). A Required Action (siallar to Required Action B.1) limiting the allowable out of service time, if a loss of automatic RCIC initiation capability exists, is not required. This Condition applies to the Reactor Vessel Water I.evel-High Function whose logic is arranged such that any inoperable channel will . result in a loss of automatic RCIC initiation capability (loss of high water level trip capability). As stated above, this loss of automatic RCIC initiation capability wa:, analyzed and determined to be acceptable. One inoperable channel may result in a loss of high water level trip capability but will not prevent RCIC System automatic start capability. However, the Required Action does not allow placing a channel in trip since this action would not necessarily result in a safe state for the channel in all events (a failure of the remaining channel could prevent a RCIC System start). O D.I. D.2.1. and D.2.2 Required Action D.1 is intended to ensure that appropriate actions are taken if nultiple, inoperable, untripped channels within the stme Function result in automatic component initiation capability being lost for the feature (s). For Required Action 0.1, the RCIC System is the only associated feature. In this case, automatic initiation capability is lost if two Function 3 channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions D.2.1 and 0.2.2 is not appropriate, and the RCIC System must be declared inoperable within I hour from discovery of loss of RCIC initiation capability. As noted, Required Action D.1 is only applicable if the RCIC pump suction is not aligned to the suppression pool since, if aligned, the Function is already performed. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
" time zero" for beginning the allowed outage time " clock."
(continued) O Brunswick Unit 2 B 3.3-138 Revision No. 1 i
RCIC System Instrumentation B 3.3.5.2 O b BASES ACTIONS D.I.'D.2.1. and D.2.2 (continued) For Required Action D.1, the Completion Time only begins l upon discovery that the RCIC System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The I hour Completion Time from discovery of loss of initiation capability is acceptable'because it minimizes risk while allowing time for restoration or tripping of channels. Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 2) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1, which performs the intended function of the channel (shifting the suction source to the suppression pool). Alternatively, Required i Action D.2.2 allows the manual alignment of the RCIC suction C to the suppression pool, which also performs the intended function. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the RCIC System piping remains filled with water. If it is not desired to perform Required Actions D.2.1 and D.2.2 (e.g., as in the case wherc shifting the suction source could drain down the RCIC suction piping), Condition E must be entered and its Required Action taken. I f.d With any Required Action and associated Completion Time not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared , inoperable immediately. ! SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RCIC l REQUIREMENTS System instrumentation Function are found in the SRs column of Table 3.3.5.2-1. (continued) ; i O Brunswick Unit 2 B 3.3-139 Revision No.
I l l RCIC System Instrumentation B 3.3.5.2 p b BASES SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for (continued) performance of required Surve111ances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 2; and (b) for up to 6 hours for Functions 1 and 3, provided the associated Function maintains RCIC initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the ap)11 cable Condition entered and Required Actions taken. Tiis Note is based on the reliability analysis (Ref. 2) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary. SR 3.3.5.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter ( indicated on one channel to a parameter on other similar channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a, combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. (continued) O Brunswick Unit 2 B 3.3-140 Revision No.
L- i 1 L 1 RCIC System Instrumentation B 3.3.5.2 l BASES SURVEILLANCE SR 3.3.5.2.2 l REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST.is performed on each required , channel to ensure that the channel will perform the intended !
- function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint l '
methodology. The Frequency of 92 days is based on the reliability analysis of Reference 2. SR 3.3.5.2.3 1 The calibration of trip units provides a check of the actual ! trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than i the Allowable Value specified in Table 3.3.5.2-1. If the ! trip setting is discovered to be less conservative than the l setting accounted'for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the e plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal.to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 2. SR 3.3.5.2.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. .This test verifies the channel' responds to the measured parameter within the necessary range and accuracy. CP WNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive- , calibrations consistent with the plant specific setpoint ' methodology. l The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift. in the setpoint analysis. (continued) 'O Brunswick Unit 2 B 3.3-141 Revision No.
RCIC System Instrumentatica B 3.3.5.2 BASES SURVEILLANCE SR 3.3.5.2.5 REQUIREMENTS l (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel and includes simulated automatic actuation of the ' channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient-if the Surveillance were performed with the reactor at power.
- l. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the
, 24 month Frequency. ! REFERENCES 1. 10 CFR 50.36(c)(2)(ii).
- 2. GENE-770-06-2P-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for l Selected Instrumentation Technical Specifications, ;
December 1992. '
=
!O ! Brunswick Unit 2 B 3.3-142 Revision No.
L Primary Centainment Isolation Instrumentatien B 3.3.6.I l B 3.3. INSTRUMENTATION B 3.3.6.1 Primary Containment Isolation Instrumentation BASES BACKGROUND The primary containment isolation instrumentation automatically initiates closure of appropriate primary containment isolation valves (PCIVs). The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs). Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DBA. The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of primary containment and reactor coolant pressure boundary (RCPB)isolatica Most channels include electronic equipment (e.g.. t rip units) that compares measured input signals with 4 e apablished setpoints. When the setpoint , is exceeded, th .L.mel output relay actuates, which then outputs a prisio ;;ontainment isolation signal- to the isolation logic. Functional diversity is provided by . , monitoring a wide. range of independent parameters. The input parameters to the isolation logics are (a) reactor
- vessel water level, (b) area ambient and differential temperatures, (c) main steam line (MSL) flow measurement, (d) Standby Liquid Control (SLC) System initiation, (e) condenser vacuum, (f) main steam line pressure, (g) high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) steam line flow, (h) drywell pressure, (1) HPCI and RCIC steam line pressure, (j) HPCI and RCIC turbine exhaust diaphragm pressure, (k) reactor water cleanup (RWCU) differential flow, (1). reactor steam dome pressure, (m) main stack radiation, and (n) reactor building exhaust radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation. The
- exceptions are SLC Syst a initiation and main stack
! radiation. Primary containment isolation instrumentation has inputs to the trip logic of the isolation functions listed below. (continued) O Brunswick Unit 2 B 3.3-143 Revision No.
Prizary Containment Isolation Instrumentati n B 3.3.6.I 10 V BASES BACKGROUND 1. Main Steam Line Isolation (continued) Most MSL Isolation Functions receive inputs from four channels. The outputs from these channels are combined in a one-out-of-two taken twice logic to initiate isolation of all main steam isolation valves (MSIVs). The outputs from the same channels are arranged into two two-out-of-two logic trip systems to isolate all MSL drain valves. Each MSL drain line has two isolation valves with one two-out-of-two logic system associated with each valve. The exceptions to this arrangement are the Main Steam Line Flow-High Function and the Main Steam Isolation Valve Pit Temperature-High Function. The Main Steam Line Flow-High function uses 16 flow channels, four for each steam line. One channel from each steam line inputs to one of the four trip strings. Two trip strings make up each trip system and both trip systems must trip to cause an MSL isolation. Each trip string has four inputs (one per MSL), any one of which will trip the trip string. The trip strings are arranged in a one-out-of-two taken twice logic. This is effectively a one-out-of-eight taken twice logic arrangement to initiate isolation of the MSIVs. Similarly, the 16 flow channels are connected into two two-out-of-two logic trip systems (effectively, two one-out-of-four twice logic), with each trip system required to isolate one of the two MSL drain valves on the associated steam line. The Main Steam Line Flow-High (Not in Run) function uses four flow channels, one for each steam line. The out)uts from these channels are arranged in one-out-of-two ta cen twice logic trip systems to isolate all MSIVs and MSL drain valves. The Main Steam Isolation Valve Pit Temperature-High function consists of the four MSL tunnel temperature monitoring channels that sense temperature in the MSIV pit. Each channel receives input from an individual temperature switch. The inputs are arranged in a one-out-of-two taken twice logic to isolate all MSIVs. Similarly, the inputs are arranged in two two-out-of-two logic trip systems, with each trip system required to isolate the two MSL drain valves per drain line. MSL Isolation Functions isolate the Group 1 valves. (continued) O Brunswick Unit 2 B 3.3-144 Revision No.
Prinary Containment Isolaticn Instrumentation B 3.3.6.1 BASES BACKGROUND 2. Primary Containment Isolation (continued) Primary Containment Isolation Functions associated with Reactor Vessel Water Level-Low Level 1 and Drywell Pressure-High receive inputs from four channels. The outputs from these channels are arranged into one-out-of-two taken twice logics. One trip system initiates isolation of all inboard primary containment isolation valves, while the other trip system initiates isolation of all outboard l primary containment isolation valves. Each logic closes one of the two valves on each penetration, so that operation of either logic isolates the penetration. The Main Stack Radiation-High Function receives input from one channel. The output from this channel is provided to each of two one-out-of-one logic trip systems. Each trip system isolates both valves in the associated penetration. The Reactor Building Radiation-High Function receives input i from two channels. The outputs from these channels are l arranged into two one-out-of-one logic trip systems. Each l trip system isolates one valve per associated penetration. 'O Q Primary Containment Isolation Drywell Pressure-High and Reactor Vessel Water Level-Low Level 1 Functions isolate the Group 2 and 6 valves. The Drywell Pressure-High Function in conjunction with reactor low pressure isolates Group 10 valves. Primary Containment Isolation Main Stack Radiation-High Function isolates the containment purge and vent valves. Reactor Building Exhaust Radiation-High Function isolates the Group 6 valves.
- 3. 4. Hiah Pressure Coolant Iniection System Isolation and L Reactor Core Isolation Coolina System Isolation Most functions that isolate HPCI and RCIC receive input from two channels, with each channel in one trip system using a one-out-of-one logic. Each of the two trip systems in each !
isolation group is connected to one of the two valves on ) y each associated penetration. i l L The exceptions are the HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High, Steam Supply Line Pressure-Low, and Equipment Area Temperature-High Functions. These A Functions receive inputs from four turbine exhaust diaphragm pressure, four steam supply pressure, and four equipment g! (continued) Brunswick Unit 2 B 3.3-145 Revision No. ,
Prl:ary Centainment Isolatten Instrumentation B 3.3.6.1 BASES BACKGROUND 3. 4. Hiah Pressure Coolant In.iection System Isolation and Reactor Core Isolation Coolina System Isolation (continued) area . temperature channels for each system. The outputs from the turbine exhaust diaphragm pressure and steam supply & pressure channels are each connected to two two-out-of-two trip systems. The outputs from the equipment area temperature channels are connected to two one-out-of-two g trip systems. In addition, the output from one channel per trip system of the Steam Supply Line Pressure-Low Function coincident with a high drywell pressure signal will initiate isolation of the associated HPCI and RCIC turbine exhaust line vacuum breaker isolation valves. Each trip system isolates one valve per associated penetration. HPCI and RCIC Functions isolate the Group 4, 5, 7, and 9 valves.
- 5. Reactor Water Cleanup System Isolation The Reactor Vessel Water Level-Low Level 2 Isolation Function receives input from four reactor vessel water level
'O channels. The outputs from the reactor vessel water level channels are connected into two two-out-of-two trip systems. The Differential Flow-High Function receives input from one channel. The output from this channel is provided to each of two one-out-of-one logic trip systems. The Piping Outside RWCU Rooms Area Temperature-High Function receives input from two channels with each channel in one trip system using a one-out-of-one logic. The Area Temperature-High Function receives input from six temperature monitors, three to each trip system. The Area Ventilation Differential Temperature-High Function receives input from six differential temperature monitors, three in each trip system. These are configured so that any one input will trip the associated trip system. Each of the two trip systems is connected to one of the two valves on each RWCU penetration. The SLC System Initiation Function receives input from one channel. The output from this channel is provided to a one-out-of-one logic trip system. The trip system isolates the RWCU suction outboard isolation valve. RWCU Functions isolate the Group 3 valves. (continued) l l ! l O Brunswick Unit 2 B 3.3-146 Revision No. l
L 1 Prl::ary Containment Isolation Instrumentation 8 3.3.6.1 BASES BACKGROUND 6. Shutdown Coolina System Isolation l (continued) The Reactor Vessel Water Level-Low Level 1 Function receives input from four reactor vessel water level l channels. The outputs from the reactor vessel water level ! channels are connected to two one-out-of-two taken twice logic trip systems. The Reactor Vessel Pressure-High Function receives input from two channels, with each channel in one trip system using a one-out-of-one logic. Each of the two trip systems is connected to one of the two valves on each shutdown cooling penetration. 1 Shutdown Cooling System Isolation Functions isolate the Group 8 valves. APPLICABLE The isolation :;4;nals generated by the primary containment SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO, and safety analyses of References 1, 2, and 3 to initiate APPLICABILITY closure of valves to limit offsite doses. Refer to LC0 3.6.1.3, " Primary Containment Isolation Valves (PCIVs)," Applicable Safety Analyses Bases for more detail of the safety analyses. Primary containment isolation instrumentation satisfies Criterion 3 of Reference 4. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the primary containment instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.6.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. The actual set)oint is calibrated corsistent with applicable setpoint metiodology assumptions. Each channel must also respond within its assumed response time, where appropriate. , Allowable Values are specified for each Primary Containment Isolation Function specified in Table 3.3.6.1-1. Trip setpoints are specified in the setpoint calculattens. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. i Operation with a trip setting less conservative than the i trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip (continued) i Brunswick Unit 2 8 3.3-147 Revision No. 1
Pricary Containment Isolation Instrumentation i B 3.3.6.1 ! BASES i APPLICABLE setting is not within its required Allowable Value. Trip SAFETY ANALYSES, setpoints are those predetermined values of output at which LCO and an action should take place. The setpoints are compared to APPLICABILITY the actual process parameter (e.g., reactor vessel water (continued) level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., 4 trip unit) changes state. The analytic limits are derived ' from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration i based instrument errors are limited to instrument drift, I errors associated with measurement snd test equipment, and ' calibration tolerance of loop components. The trip setpoints and Allowable Values datermined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. Certain Emergency Core Cooling Systems (ECCS) and RCIC valves (e.g., LPCI injection) also serve the dual function of automatic PCIVs. The signals that isolate these valves are also associated with the automatic initiation of the ECCS and RCIC. The instrumentation requirements and ACTIONS associated with these signals are addressed in LCO 3.3.5.1,
" Emergency Core Cooling Systems (ECCS) Instrumentation," and u LC0 3.3.5.2, " Reactor Core Isolation Cooling (RCIC) System !
Instrumentation," and are not included in this LCO. I { In general, the individual Functions are required to be j OPERABLE in MODES 1, 2, and 3 consistent with the l Applicability for LCO 3.6.1.1, " Primary Containment." { Functions that have different Applicabilities are discussed j below in the individual Functions discussion, j The specific Applicable Safety Analyses, LCO, and l App 1tcability discussions are listed below on a Function by ! Function basis, i (continued) l l O Brunswick Unit 2 B 3.3-148 Revision No.
Primary Containment Isolaticn Instrumentation B 3.3.6.1 BASES APPLICABLE Main Steam Line Isolation SAFETY ANALYSES, LCO, and 1.a. Reactor Vessel Water Level-Low Level 3 APPLICABILITY (continued) Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of the MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits - from being exceeded. The Reactor Vessel Water Level-Low Level 3 Function is one of the many Functions ~ assumed to be OPERABLE and capable of providing isolation signals. . The Reactor Vessel Water Level-tow Level 3 Function associated with isolation is assumed in the analysis of.the recirculation line break (Ref.1). The isolation of the MSLs on Level 3 supports actions to ensure that offsite dose limits are not exceeded for a DBA. Reactor vessel water level signals are initiated.from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water O Level-Low Level 3 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function, The Reactor Vessel Water Level-Low Level 3 Allowable Value is chosen to be the same as the ECCS Level 3 Allowable Value (LCO 3.3.5.1) to ensure that the MSLs isolate on a potential loss of coolant accident ~ (LOCA) to prevent offsite doses from exceeding 10 CFR 100 limits. The Allowable Value is : referenced from reference level zero. Reference level zero is-367 inches above the vessel zero point. ' This Function isolates the Group 1 valves. 1.b. Main Steam Line Pressure-tow Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV ~ cooling down more than 100*F/hr if the pressure loss is allowed to continue. The Main Steam Line Pressure-Low function is directly assumed in the analysis of the pressure regulator (continued) O Brunswick Unit 2 B 3.3-149 Revision No.
Primary Containment Isolatten Instrumentation B 3.3.6.1 BASES APPLICABLE- 1.b. Main Steam Line Pressure-Low (continued) SAFETY ANALYSES LCO, and failure (Ref.2). For this event, the closure of the MS!Vs
. APPLICABILITY ensures that no significant thermal stresses are imposed on the RPV. In addition, this' Function supports actions to ensure that Safety Limit 2.1.1.1 is not exceeded.- (This Function closes the MSIVs prior to pressure decreasing below 785 psig, which results in a scram due to MSIV closure, thus reducing reactor power to < 25% RTP.)
The MSL low pressure signals are initiated from four transmitters that are connected-to the MSL header. The transmitters are arranged such that each-transmitter is able to detect low MSL pressure. Four channels of Main Steam. Line Pressure-Low function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value was selected to be far enough below I normal turbine inlet pressures to avoid spurious isolations, yet high enough to provide timely detection of a pressure regulator malfunction. The Main Steam Line Pressure-Low Function is only required to be OPERABLE in MODE 1 since this is when the assumed transientcanoccur(Ref.2). This function isolates the Group 1 valves except for sample line isolation valves B32-F019 and B32-F020. 1.c. Main Steam Line Flow-Hioh Main Steam Line Flow-High is provided to detect a break of the MSL and to initiate closure of the MSIVs. If the steam were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover. If the RPV water level decreases too far, fuel damage could occur. Therefore, the isolation is initiated on high flow to prevent or minimize core damage. The Main Steam Line Flow-High Function is directly assumed in the analysis of the main steam line break (MSLB) (Ref. 5). The isolation i action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 100 limits. (continued) O Brunswick Unit 2 B 3.3-150 Revision No, a
Prizary Centainment Isolation Instrumentation 8 3.3.6.1 BASES. APPLICABLE 1.c. Main Steam Line Flow-Hiah (continued) , SAFETY ANALYSES,. LCO, and The MSL flow signals are initiated from 16 transmitters that APPLICABILITY are connected to the four MSLs. The transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to I detect the high flow. Four channels of Main Steam Line Flow-High Function for each unisolated MSL (two channels per trip system) are available and are required to be OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL. The Allowable Value is chosen to be high enough to permit L isolation of one main steam line for test at rated power without causing an automatic isolation of the rest of the I steam lines, yet low enough to permit early detection of a gross steam line break. This Function isolates the Group 1 valves except for sample line isolation valves B32-F019 and 832-F020. 1.d. Condenser Vacuum-Low The Condenser Vacuum-Low Function is provided to prevent overpressurization of the main condenser in the event of a loss of the main condenser vacuum. Since th'e integrity of the condenser is an assumption in offsite dose calculations, the Condenser Vacuum-Low Function is assumed to be OPERABLE and capable of initiating closure of the MSIVs. The closure of the MSIVs is initiated to prevent the addition of steam that would lead to additional condenser pressurization and possible rupture, thereby preventing a potential radiation leakage path following an accident.
]
Condenser vacuum pressure signals are derived from four pressure transmitters that sense the pressure in the I' condenser. Four channels of Condenser Vacuum-Low function are available and are required to be OPERABLE to ensure that { l no single instrument failure can preclude the isolation i function. l The Allowable Value is chosen to prevent damage to the condenser due to pressurization, thereby ensuring its integrity for offsite dose analysis. As noted (footnote (a) (continued) O Brunswick Unit 2 B 3.3-151 Revision No.
e Prizary Containment Isolation Instrumentatica B 3.3.6.1
. BASES APPLICABLE 1.d. Condenser Vacuum-Low (continued)
SAFETY ANALYSES, LCO, and to Table 3.3.6.1-1), the channels are not required to be APPLICABILITY OPERABLE in MODES 2 and 3 when all turbine stop valves (TSVs) are closed, since the potential for condenser overpressurization is minimized. Therefore, the channels may be bypassed when all TSVs are closed. This function isolates the Group 1 valves. d 1.e. Main Steam Isolation Valve Pit Temperature-Hioh Main steam isolation valve pit temperature is provided to detect a leak in the RCPB and provides diversity to the high flow instrumentation. The isolation occurs when a very small leak has occurred in the main steam isolation valve pit. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. However, credit for these instruments is not taken in any transient or accident analysis in the UFSAR, since bounding analyses are performed for large breaks, such as MSLBs. Main. steam isolation valve pit temperature signals are O initiated from temperature switches located in the area being monitored. Four channels of Main Steam Isolation Valve Pit Temperature-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The temperature switches are located or shielded so that they are sensitive to air temperature and not in the radiated ' heat from hot equipment. The main steam isolation valve pit temperature monitoring Allowable Value is chosen to detect a leak equivalent to between 1% and 10% rated steam flow. This function isolates the Group 1 valves except for sample line isolation valves B32-F019 and B32-F020. 1.f. Main Steam Line Flow-Hiah (Not in Run) Main Steam Line Flow-High (Not in Run) is provided to detect a break of the MSL and to initiate closure of the MSIVs. . Malfunction of the pressure regulator for the main turbine bypass valves during MODE 2 or 3 can cause depressurization of the reactor vessel. Therefore, (continued) O Brunswick Unit 2 B 3.3-152 Revision No.
Prigary Containment Isolation Instrumentation B 3.3.6.1 ( BASES I APPLICABLE Main Steam Line Flow-Hiah (Not in Run) 1.f. SAFETY ANALYSES, (continued) LCO, and isolation of the main steam lines is initiated on high steam l APPLICABILITY line flow when the mode switch is not in run to ensure significant thermal stresses are not imposed on the RPV. This Function is bypassed when the mode switch is in the run < mode. With the mode switch in the run mode, l depressurization protection is provided by the Main Steam i Line Pressure-tow function. The MSL flow signals are initiated from four transmitters, each connected to one of the MSLs. Four channels of the Main Steam Line Flow-High (Not in Run) Function (two channels per trip system) are available and required to be OPERABLE so that no single instrument failure will preclude the isolation function. The Allowable Value was selected to be low enough to provide timely detection of a pressure regulator malfunction. This function isolates the Group 1 valves except for sample line isolation valves B32-F019 and B32-F020. Primary Containment Isolation 2.a. Reactor Vessel Water Level-tow level 1 Low RPV water level indicates that the ca) ability to cool the fuel may be threatened. The valves wiose penetrations communicate with the primary containment are isolated to limit the release of fission products. The isolation of the primary containment on Level I supports actions to ensure i that offsite dose limits of 10 CFR 100 are not exceeded. The Reactor Vessel Water Level-Low Level 1 Function associated with isolation is implicitly assumed in the UFSAR analysis as these leakage paths are assumed to be isolated post LOCA. Reactor Vessel Water Level-Low Level I signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of (continued) ! Brunswick Unit 2 B 3.3-153 Revision No.
Pri ary Containment Isolation Instrumentatica j B 3.3.6.1 1 BASES APPLICABLE 2.a. Reactor Vessel Water Level-Low Level 1 (continued) 1 SAFETY ANALYSES, LCO, and Reactor Vessel Water Level-Low Level 1 Function are l APPLICABILITY available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation- j function. The Reactor Vessel Water Level-Low Level 1 Allowable Value was chosen to be the same as the RPS Level 1 scram Allowable Value (LCO 3.3.1.1), since isolation of these valves is not critical to orderly plant shutdown. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. This Function isolates the Group 2 and 6 valves. 2.b. Drywell Pressure-Hiah High drywell pressure can indicate a break in the RCPB inside the primary containment. The isolation of some of the primary containment isolation valves on high drywell pressure supports actions to ensure that offsite dose limits of 10 CFR 100 are not exceeded.. The Drywell Pressure-High Q Function, associated with isolation of the primary containment, is implicitly assumed in the UFSAR accident analysis as these leakage paths are assumed to be isolated post LOCA. High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. 1 The Allowable Value was selected to be the same as the ECCS , Drywell Pressure-High Allowable Value (LCO 3.3.5.1), since i this may be indicative of a LOCA inside primary containment. ' This function isolates the Group 2 and 6 valves. This ! Function in conjunction with reactor low pressure also isolates Group 10 valves. (continued) , O Brunswick Unit 2 B 3.3-154 Revision No.
Pri;:ary Containment Isolaticn Instrumentation B 3.3.6.1 ,O V BASES l l APPLICABLE 2.c. Main Stack Radiation-Hiah I SAFETY ANALYSES, LCO, and High main stack radiation indicates increased airborne APPLICABILITY radioactivity levels in primary containment being released (continued) through the containment vent valves. Therefore, Main Stack Radiation-High function initiates an isolation to assure timely closure of valves to protect against substantial releases of radioactive materials to the environment. l However, this function is not assumed in any accident or transient analysis in the UFSAR because other leakage paths (e.g., MSIVs) are more limiting. The main stack radiation signal is initiated from a radiation detector that is located in the main stack. The Allowable Value is established in accordance with the l methodology in the Offsite Dose Calculation Manual. ! This function isolates the containment vent and purge valves. l 2.d. Reactor Buildina Exhaust Radiation-Hiah l High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding. The release may have originated from the primary containment due to a break in the RCPB. When Reactor Building Exhaust ' Radiation-High is detected, valves whose penetrations communicate with the primary containment atmosphere are isolated to limit the release of fission products. The Reactor Building Exhaust Radiation-High signals are
. initiated from radiation detectors that are located on the ventilation exhaust piping coming from the reactor building.
The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Two channels of Reactor Butiding Exhaust-High Function are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are chosen to promptly detect gross failure of the fuel cladding. These Functions isolate the Group 6 valves. (continued) O Brunswick Unit 2 B 3.3-155 Revision No.
Prinary Containment Isolaticn Instrurentation B 3.3.6.1 BASES l APPLICABLE Hiah Pressure Coolant In.iection and Reactor Core Isolation SAFETY ANALYSES, Coolina Systems Isolation LCO, and APPLICABILITY 3.a. 3.b..'4.a. 4.b. HPCI and RCIC Steam Line Flow-Hiah (continued) and Time Delav Relays Steam Line Flow-High Functions are provided to detect n' break of the RCIC or HPCI steam lines and' initiate closure of the steam line isolation valves of the appropriate system. If the steam is allowed to continue flowing out of the break, the reactor will depressurize and the core can uncover. Therefore, the isolations are initiated on high flow to prevent or minimize core damage. The isolation action, along with the scram function of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Specific credit for these Functions is not assumed in any.UFSAR accident analyses since the bounding analysis is performed for large breaks such as recirculation and MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming . bounding. I o The HPCI and RCIC Steam Line flow-High signals are ed initiated after a short time delay from differential pressure switches (two for HPCI and two for RCIC) that are connected to the system steam lines. Two channels of both
' HPCI and RCIC Steam Line Flow-High Functions and the associated Time Delay Relays are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The time delay was selected to prevent spurious isolation of HPCI and RCIC due to transient high steam flow during turbine starts and spurious operation during HPCI and RCIC operation. ;
The Allowable Values are chosen to be low enough to ensure , that the trip occurs to prevent fuel damage and maintains the MSLB event as the bounding event. These Functions isolate the Group 4 and 5 valves, as appropriate. (continued) O Brunswick Unit 2 B 3.3-156 Revision No.
Pri=ary Containment Isolatien Instrumentation B 3.3.6.1 BASES I APPLICABLE 3.c. 4.c. HPCI and RCIC Steam Supply Line Pressure-Low SAFETY ANALYSES, LCO, and Low MSL pressure indicates that the pressure of the steam in I APPLICABILITY the HPCI or RCIC turbine may be too low to continue (continued) operation of the associated system's turbine and is i indicative of a break of the HPCI or RCIC steam lines. ; l These isolations provide a diverse signal to indicate a l possible system break. The HPCI and RCIC Steam Supply Line l Pressure-Low functions are provided so that in the event a l l gross rupture of the HPCI or RCIC steam line occurred l upstream from the high flow sensing location, thus negating l the Steam Line Flow-High Functions, isolation would be
- effected on low pressure.
The HPCI and RCIC Steam Supply Line Pressure-Low signals are initiated from pressure switches (four for HPCI and four for RCIC) that are connected to the system steam line. Four
- _ channels of both HPCI and RCIC Steam Supply Line l
Pressure-Low Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are selected to be below the pressure O at which the system's turbine can effectively operate. These Functions isolate the Group 4 and 5 valves, as appropriate. 3.d. 4.d. HPCI and RCIC Turbine Exhaust Diaohraam Pressure-Hiah High turbine exhaust diaphragm pressure could indicate that the turbine rotor is not turning, thus allowing reactor pressure to act on the turbine exhaust line. The HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High Functions initiate isolation to prevent overpressurization of the turbine exh'aust line. These isolations are for equipment protection and are not assumed in any transient or accident l analysis in the UFSAR. These instruments are included in i the TS because of the potential for risk due to possible i failure of the instruments preventing HPCI and RCIC L initiations. Therefore, they meet Criterion 4 of Reference 4. (continued) O Brunswick Unit 2 B 3.3-157 Revision No.
l l Prinary Centainment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 3.d., 4.d. HPCI and RCIC Turbine Exhaust DiaDhraam SAFETY ANALYSES, Pressure-Hiah (continued) LCO, and APPLICABILITY The HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High signals are initiated from pressure switches (four for HPCI and four for RCIC) that are connected to the area between the rupture diaphragms on each system's turbine exhaust line. Four channels of both HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are high enough to prevent isolation of HPCI or RCIC if the associated turbine is operatir.g, yet low enough to effect isolation before the turbine exhaust line is unduly pressurized. These Functions isolate the Group 4 and 5 valves, as appropriate. 3.e. 4.e. Drywell Pressure-Hiah p High drywell pressure can indicate a break in the RCPB. The ( HPCI and RCIC isolation of the turbine exhaust is provided l to prevent communication with the drywell when high drywell ! pressure exists. A potential leakage path exists via the ! turbine exhaust. The isolation is delayed until the system becomes unavailable for injection (i.e., low steam line pressure). The isolation of the HPCI and RCIC turbine ; exhaust by Drywell Pressure-High is indirectly assumed in ; the UFSAR accident analysis because the turbine exhaust leakage path is not assumed to contribute to offsite doses. High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Two channels of both HPCI and RCIC Drywell Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value was selected to be the same as the ECCS Drywell Pressure-High Allowable Value (LC0 3.3.5.1), since this is indicative of a LOCA inside primary containment. This function isolates the Group 7 and 9 valves. (continued) O Brunswick Unit 2 B 3.3-158 Revision No.
Prinary Containment Isolatten Instrumentation 8 3.3.6.1 BASES APPLICABLE 3.f. 3.a. 3.h. 3.1. 4.f. 4.a. 4. h. 4.1. 4.1. 4.k. SAFETY ANALYSES, Area and Differential Temperature-Hia l and Time Delav LCO, and APPLICABILITY . Area and differential temperatures are provided to detect a (continued) leak from the associated system steam piping. The isolation occurs to prevent excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier and is diverse to the high ' flow instrumentation. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. These Functions are not assumed in any UFSAR transient or accident analysis, since bounding analyses are
)erformed for large breaks such as recirculation or MSL areaks.
Area and Differential Temperature-High signals are initiated from thermocouples that are appropriately located to protect the system that is being monitored. Two instruments monitor each area. Two channels for each HPCI and RCIC Area and Differential Temperature-High Function, except for the HPCI and RCIC Equipment Area
- Temperature-High Function which are required to have four d channels each, are available and are required to be OPERABLE O.
to ensure that no single instrument failure can preclude the isolation function. In addition, a time delay is associated with the RCIC Steam Line Area Temperature-High, the RCIC Steam Line Tunnel Ambient Temperature-High, and the RCIC Steam Line Tunnel Differential Temperature-High Functions. The time delay was selected to eliminate spurious isolations which might occur when switching from normal ventilation to ! standby ventilation. l The Allowable Values are set high enough above anticipated normal operating levels to avoid spurious isolation, yet low enough to provide timely detection of a HPCI or RCIC steam line break. These functions isolate the Group 4 and 5 valves, as appropriate. (continued) G Brunswick Unit 2 B 3.3-159 Revision No.
L; I L Priaary Centainment Isolation Instrumentaticn B 3.3.6.1 i BASES 1 APPLICABLE- Reactor Water Cleanuo System Isolation ! SAFETY ANALYSES, LCO, and 5.a. 5.b. Differential Flow-Hiah and Time Delay l APPLICABILITY (continued) The high differential flow signal is provided to detect a break in the RWCU System. Should the reactor coolant continue to flow out of the break, offsite dose limits may be exceeded. Therefore, isolation of the RWCU System is initiated when high differential flow is sensed to prevent excessive loss of reactor coolant and release of significant amounts of radioactive material. A time delay is provided to prevent spurious trips during most RWCU operational transients. This function-is not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large breaks such as MSLBs. The high differential flow signals are initiated from transmitters that are connacted to the inlet (from the reactor vessel) and outlets (to condenser and feedwater) of I the RWCU System. The outputs of the transmitters are compared (in a common summer) and the resulting output is sent to two high flow trip units. If the difference between the inlet and outlet flow is too large, each trip unit O generates an isolation signal. Two channels of Differential Flow-High Function are available and are required to be OPERABLE to ensure that no single instrument failure downstream of the common summer can preclude the isolation function. The Differential Flow-High Allowable Value ensures that a break of the RWCU piping'is detected. This Function isolates the Group 3 valves.
},c u_5.d. 5.e. Area. Area Ventilation Differential and Pipina Outside RWCU Rooms Area Temperature-Hiah RWCU ~ area, area ventilation differential, and piping outside RWCU area temperatures are provided to detect a leak from the RWCU System. If the small leak continues without isolation, offsite dose limits may be reached. Credit for these instruments is not taken in any transient or accident analysis in the UFSAR, since bounding analyses are performed for large breaks such as recirculation or MSL breaks.
(continued) O Brunswick Unit 2 B 3.3-160 Revision No. l
i Prigary Containment Isolatien. Instrumentation B 3.3.6.1 j i BASES i APPLICABLE 5.c. 5.d. 5.e. Area. Area Ventilation Differential. and i SAFETY ANALYSES, >ioino Outside RWCU Rooms Area Temoerature-Hioh i LCO, and (continued) APPLICABILITY 1 Area and area ventilation differential temperature signals are initiated from temper &ture elements that are located in the room that is being monitored. Six thermocouples provide input to the Area Temperature-High Function (two per room). Six channels are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation . function. Twelve thermocouples provide input to the Area Ventilation Differential Temperature-High Function. The output of these thermocouples is used to determine the differential temperature. Each channel consists of a differential temperature instrument that receives inputs from thermocouples that are located in the inlet and outlet ducts , L which ventilate the RWCU System rooms for a total of six ' available channels (two per room). However, only four channels are required to be OPERABLE. Temperature signals are initiated from temperature elements O monitoring in the 20'/50' elevation RWCU System general piping areas located outside the RWCU System equipment rooms. Two thermocouples provide input to the Piping Outside RWCU Rooms Area Temperature-High function. Two i channels are required to be OPERABLE to ensure that no l single instrument failure can preclude the isolation ! function. The Area and Area Ventilation Differential Temperature-High Allowable Values are set low enough to provide timely detection of a break in the RWCU System within the associated room (s). l The P1)ing Outside RWCU Rooms Area Temperature-High Function l Allowaale Value is set low enough to isolate a design basis high energy line break at any point in the high temperature RWCU System piping located outside of the RWCU System i equipment rooms. These Functions isolate the Group 3 valves. (continued) i i O Brunswick Unit 2 8 3.3-161 Revision No. 1
l l Primary Containment Isolatica Instrumentation B 3.3.6.1 ,O . V BASES APPLICABLE 5.f. SLC System Initiation. SAFETY ANALYSES, LCO, and The isolation of the RWCU System is required when the SLC APPLICABILITY System has been initiated to prevent dilution and removal of (continued) the boron solution by the RWCU System (Ref. 6). The SLC System initiation signal is initiated from the SLC pump start hand switch signal. There is no Allowable' Value associated with this function since the channel is mechanically actuated based solely on the position of the SLC System initiation switch. One channel of the SLC System Initiation Function is available and required to be OPERABLE only in MODES' I and 2, since these are the only MODES where the reactor can be critical, and these MODES are consistent with the Applicability for the SLC System (LCO 3.1.7). As noted (footnote (c) to Table 3.3.6.1-1), this Function is only required to close one of the RWCU isolation valves since the signals only provide input into one trip system. 5.a. Reactor Vessel Water Level-Low Level 2 \ Low RPV water level indicates that the capability to cool the fuel.may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some interfaces with the reactor vessel occurs to isolate the potential sources of a break. The isolation of the RWCU System on Level 2 supports actions to ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Vessel Water Level-Low Level 2 Function associated with RWCU isolation is not directly assumed in the UFSAR safety analyses because the RWCU System line break is bounded by breaks of larger systems (recirculation and MSL breaks are more limiting). Reactor Vessel Water Level-Low Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Level 2 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. (continued) O Brunswick Unit 2 B 3.3-162 Revision No. i
Primary Centainment Isolation Instrumentation l-B 3.3.6.1 j BASES l l APPLICABLE 5.a. Reactor Vessel Water Level-Low Level 1 (continued) l SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level-Low Level 2 Allowable Value APPLICABILITY was chosen to be the same as the ECCS Reactor Vessel Water Level-Low Level 2 Allowable Value (LCO 3.3.5.1), since the i l capability to cool the fuel may be threatened. The ' Allowable Value is referenced from reference level zero.
. Reference level zero is 367 inches above the vessel zero point.
This Function isolates the Group 3 valves. RHR Shutdown Coolina System Isolation 6.a. Reactor Steam Dome Pressure-Hiah The Reactor Steam Dome Pressure-High function is provided l to isolate.the shutdown cooling portion of the Residual Heat Removal (RHR) System. This interlock is provided only for equipment protection to prevent an intersystem LOCA scenario, and credit for the interlock is not assumed in the accident or transient analysis in the UFSAR. The Reactor Steam Dome Pressure-High signals are initiated from two )ressure switches that'are connected to different taps on tie RPV. Two channels of Reactor Steam Dome Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Function is only required to be OPERABLE in MODES 1, 2, and 3, since these are the only MODES in which the reactor can be pressurized; thus, equipment protection is needed. The Allowable Value was chosen to be low enough to protect the system equipment from overpressurization. This function isolates the Group 8 valves except for the LPCI injection valves E11-F015A and E11-F015B. 6.b. Reactor Vessel Water Level-Low level 1 l Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease i too far, fuel damage could result. Therefore, isolation of some reactor vessel interfaces occurs to begin isolating the potential sources of a break. The Reactor Vessel Water (continued) l O Brunswick Unit 2 B 3.3-163 Revision No. l j
Pri:ary Centainment Isolatten Instrumentation 8 3.3.6.1
. BASES l APPLICABLE 6.b. Reactor Vessel Water Level -Low Level 1 SAFETY ANALYSES, (continued)
LCO, and Level-Low Level.1 Function associated with RHR Shutdown
-APPLICABILITY Cooling' System isolation is not directly assumed in safety analyses because a break of the RHR Shutdown Cooling System is bounded by breaks of the recirculation and MSL. The RHR Shutdown Cooling System isolation on Level 1 supports actions to ensure that the RPV water level does not drop below the top of the active fuel during a vessel draindown l event caused by a leak (e.g., pipe break or inadvertent
'. valve opening) in the RHR Shutdown Cooling System. Reactor Vessel-Water Level-Low Level I signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels (two channels per trip system) of the Reactor Vessel Water Level-Low Level 1 function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. As noted (footnote (d) to Table 3.3.6.1-1), only one channel per trip system (with an isolation signal available to one RHR shutdown cooling O pump suction isolation valve) of the Reactor Vessel Water Level-Low Level 1 Function is required to be OPERABLE in MODES 4 and 5, provided the RHR Shutdown Cooling System integrity is maintained. System integrity is maintained d provided the piping is intact and no maintenance is being performed that has the potential for draining the reactor vessel through the system. The Reactor Vessel Water Level-Low Level 1 Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level-Low Level 1 Allowable Value (LCO 3.3.1.1), since the capability to cool the fuel may be' threatened. The Allowable Values is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. The Reactor Vessel Water Level-Low Level 1 Function is only required to be OPERABLE in MODES 3, 4, and 5 to prevent this potential flow )ath from lowering the reactor vessel level to the top of tie fuel. In MODES I and 2, another isolation I (continued) O Brunswick Unit 2 8 3.3-164 Revision No. 1
Prinary Containment isolatitn Instrumentaticn B 3.3.6.1 G b BASES APPLICABLE 6.b. Reactor Vessel Water Level-Low Level 1 (continued) SAFETY ANALYSES, LCO, and (i.e., Reactor Steam Dome Pressure-High) and administrative APPLICABILITY controls ensure that this flow path remains isolated to prevent unexpected loss of inventory via this flow path. This function isolates the Group 8 valves. ACTIONS A Note has been provided to modify the ACTIONS related to primary containment isolation instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will ne result in separate entry into the Condition. Sectir .3 also specifies that Required Actions of the Cond' ... continue to apply for each additional failure, wit. uompletion Times based on initial entry into the Condition. However, the Required Actions for inoperable primary containment isolation instrumentation channels provide appropriate compensatory measures for separate f inoperable channels. As such, a Note has been provided that ( allows separate Condition entry for each ino erable primary containment isolation instrumentation channe . Ad Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Functions 2.a. 2.b and 6.b and 24 hours for Functions other than Functions 2.a. 2.b and 6.b has been shown to be g acceptable (Refs. 7 and 8) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.I. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single (continued) Brunswick Unit 2 8 3.3-165 Revision No.
l Prl::ary Containment Isolatten Instrumentation ! l B 3.3.6.1 l BASES I ACTIONS. M (continued) failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel in trip (e.g.,.as in the case where placing the inoperable channel in trip would result in an isolation), Condition C must be entered and its Required Action taken. u Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same function result in redundant automatic isolation capability being lost for the associated penetration flow path (s). The MSL Isolation Functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that both trip systems will generate a trip signal from the given Function on a valid signal. The other isolation functions-are considered to be maintaining isolation capability when , sufficient channels are OPERABLE or in trip, such that one L trip system will generate a trip signal from the given Function on a valid signal. This ensures that one of the
.two PCIVs in the associated penetration flow path can receive an isolation signal from the given Function. For Functions 1.a. 1.b, 1.d, and 1.e, this would require both trip systems to have a total of three channels OPERABLE or in trip. For Functions 2.a 2.b, and 6.b, this would ,
require both trip systems to have one channel OPERABLE or in ' trip. For Function 1.c, this would require both trip systems to have a total of three channels, associated with each MSL, OPERABLE or in trip. For Function 1.f, this would require both trip systems to have one channel, associated with each MSL, OPERABLE or in trip. For Functions 3.c, 3.d. 4.c, 4.d, and 5.g this would require one trip system to have two channels, each OPERABLE or in trip. For Functions 2.c, 2.d 3.a 3.b, 3.e, 3.f, 3.g, 3.h, 3.1, 4.a,
. 4.b, 4.e, 4.f, 4.g 4.h, 4.1, 4.j , 4.k. 5.a. 5.b, 5.e, 5.f, and 6.a. this would require one trip system to have one channel OPERABLE or in trip. For Functions 5.c and 5.d, each Function consists of channels that monitor several different locations. Therefore, this would require one
- channel per location to be OPERABLE or in trip (the channels are not required to be in the same trip system).
i (continued) O Brunswick Unit 2 B 3.3-166 Revision No. I
Pri::ary Containment Isolaticn bstrumentation B 3.3.6.1 BASES-ACTIONS' L 1 (continued) l-The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The I hour completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. ful-Required Action C.1 directs entry into the appropriate Condition referenced in Table 3.3.6.1-1. The applicable l ' Condition specified in Table 3.3.6.1-1 is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed.- Each l time an inoperable channel has not met any Required Action of Condition A or B and the associated Com>1etion Time has expired, Condition C will be entered for tiat channel and provWes for transfer to the appropriate subsequent Condition. D.1. D.2.1. and D.2.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does nct apply. This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours (Required Actions D.2.1 and 0.2.2). Alternately, the associated MSLs may be isolated (Required Action D.1), and, if allowed (i.e., plant safety analysis allows operation with an MSL isolated), operation with that MSL isolated may continue. Isolating the affected MSL accomplishes the safety function of the inoperable channel. The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. (continued) l l O Brunswick Unit 2 B 3.3-167 Revision No. 1 l ( .- L
Prirary Containment Isolation Instrumentatien B 3.3.6.1 BASES ACTIONS [d (continued) If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based en operating experience, to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems. Ed If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, plant operations may continue if the affected penetration flow path (s) is isolated. Isolating the affected penetration flow path (s) accomplishes the safety function of the inoperable channels. For the RWCU Area and Area Ventilation Differential f]cs Temaerature-High functions, tne affected penetration flow pat 1(s) may be considered isolated by isolating only that
, portion of the system in the associated room monitored by the inoperable channel. That is, if the RWCU pump room A area channel is inoperable, the pump room A area can be isolated while allowing continued RWCU operation utilizing the B RWCU pump.
g Alternately, if it is not desired to isolate the affected penetration flow path (s) (e.g., as in the case where isolating the reactor scram) penetration flow path (s) could result in a, Conditio Actions taken. The I hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for plant operations personnel to isolate the affected penetration flowpath(s). (continued) l Brunswick Unit 2 B 3.3-168 Revision No. l I l
l l. Prt:ary Containment Isolaticn Instrumentation B 3.3.6.1 O-i U BASES l ACTIONS G.1 and G.2
.(continued)
If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, or the Required Action of Condition F is not met and the associated Completion Time has expired, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 3 i within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience,-to reach the required plant conditions from full l power conditions in an orderly manner and without l challenging plant systems. l H.1 and H.1 i If the channel is not restored to OPERABLE status or placed in tri) within the allowed Completion Time, the associated SLC su) system (s) is declared inoperable or the RWCU System is isolated. Since this Function is required to ensure that the SLC System performs its intended function, sufficient remedial measures are provided by declaring the associated O SLC subsystems inoperable or isolating the RWCU System.
- l. The I hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for personnel i to isolate the RWCU System.
l 1.1 and I.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated i i penetration flow path should be closed. However, if the ' shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay ; heat removal capabilities so the penetration flow path can ' I be isolated). Actions must continue until the channel is l l restored to OPERABLE status or the RHR Shutdown Cooling System is isolated. (continued) O Brunswick Unit 2 B 3.3-169 Revision No. 1
Pricary Cont.ainment Isolation Instrumentation B 3.3.6.1
)
v BASES (continued) SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Primary Containment Isolation instrumentation Function are found in the SRs column of Table 3.3.6.1-1. , The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 2 hours for Functions with a design that ! provides only one channel per trip system and (b) for up to 6 hours for all other Functions provided the associated function maintains trip capability. Upon completion of the Surveillance, or expiration of the 2 hour allowance for Functions with a design that provides only one channel per trip system or the 6 hour allowance for all other Functions, the channel must be returned to OPERADLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 7 and 8) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 2 and 6 hour testing allowances do not significantly reduce the probability that the PCIVs will isolate the penetration n flow path (s) when necessary. U SR 3.3.6.1.1 Performance of the CHANNEL CllECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are detetmined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. (continued) Brunswick Unit 2 B 3.3-170 Revision No.
Pri ary Containment Isolation Instrumentaticn B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.1 (continued) REQUIREMENTS The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent,- checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.6.1 2. SR 3.3.6.1.5. and SR 3.3.6.1.9 b A CHANNEL FUhCTIONAL TEST is performed on each required channel to ersure that the channel will perform the intended function. Aoy setpoint adjustment shall be consistent with the assumptions of the current plant' specific setpoint methodology. The 92 day frequency of SR 3.3.6.1.2 is based on the reliability analysis described in References 7 and 8. The l 184 day Frequency of SR 3.3.6.1.5 and the 24 month Frequency of SR 3.3.6.1.9 are based on engineering judgment and the reliability of the components. b O 1R 3.3.6.1.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than 1 the Allowable Value specified in Table 3.3.6.1-1. If the ' trip setting is discovered to be less conservative than l accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance ; is still within the requirements of the plant safety ' analysis. Under these conditions, the setpoint must be ! readjusted to be equal to or more conservative than that accounted for in the appropriate setpoint methodclogy. l l The Frequency of 92 days is based on the reliability ! analysis of References 7 and 8. SR 3.3.6.1.4 and SR 3.3.6.1.6 A CHANNEL CALIBRATION is a complete check of the instrument . loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary (continued) Brunswick Unit 2 B 3.3-171 Revision No. 4
c: L ! Primary Containment Isolation Instrumentaticn I B 3.3.6.1
-BASES SURVEILLANCE- SR 3.3.6.1.4 and SR 3.3.6.1.6 (continued)
REQUIREMENTS range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency of SR 3.3.6.1.4 is based on the assumption of a 92 day. calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. The-Frequency of SR 3.3.6.1.6 is based on the assumption of a 24 month calibration interval in-the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.6.1.7 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed on PCIVs in LC0 3.6.1.3 overlaps this Surveillance to provide complete testing of the assumed safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply ditring a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated ihese components will usually pass the Surveillance when per/ormed at the 24 month Frequency. SR 3.3.6.1.8 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. Testing is performed only on channels where the assumed response time does not correspond to the diesel generator (DG) start time. For channels assumed to respond within the DG start time, sufficient margin exists in the 10 second start time when compared to the typical channel response time (milliseconds) so as to assure adequate response without a specific measurement test (Ref. 9). (continued) O -Brunswick Unit 2 B 3.3-172 Revision No.
Primary Containment Isolatien Instrumentation ' , B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.8 (continued) REQUIREMENTS Note 1.to the Surveillance states that the radiation detectors are excluded from ISOLATION INSTRUMENTATION RESPONSE TIME testing. This Note is necessary because of the difficulty of generating an appropriate detector input /A signal and because the principles of detector operation E virtually ensure an instantaneous response time. Response times for radiation detector channels shall be measured from detector output or the input of the first electronic component in the channel. In addition, Note 2 to the Surveillance states that the response time of the sensors for Functions 1.a, 1.c, and 1.f may be assumed to be the design sensor response time and therefore, are excluded from the ISOLATION INSTRUMENTATION RESPONSE TIME testing. This is allowed since the sensor response time is a small part of the overall ISOLATION INSTRUMENTATION RESPONSE TIME (Ref. 9). ISOLATION INSTRUMENTATION RESPONSE TIME tests are conducted on a 24 month STAGGERED TEST BASIS. The 24 month Frequency is consistent with the BNP refueling cycle and is based upon plant operating experience that shows that random failures O- of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. REFERENCES 1. UFSAR, Section 6.3.
- 2. UFSAR, Chapter 15.
- 3. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.
- 4. 10 CFR 50.36(c)(2)(ii).
- 5. UFSAR, Section 6.2.4.3. -
- 6. UFSAR, Section 7.3.1.1.6.18.
- 7. NEDC-31677P-A, Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation, July 1990.
(continued) O Brunswick Unit 2 B 3.3-173 Revision No.
Prl:ary Containment Isolation Instrumentation I B 3.3.6.1 'l BASES 1
\
REFERENCES 8. NEDC-30851P-A Supplement 2. Technical Specifications (continued) Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
- 9. NEDO-32291-A, System Analyses for Elimination of Selected Response Time Requirements October 1995.
O i l O Brunswick Unit 2 B 3.3-174 Revision No.
Secondary Containment Isolation Instrumentatica B 3.3.6.2 B 3.3 INSTRUMENTATION B 3.3.6.2 Secondary Containment Isolation Instrumentation BASES BACKGROUND The secondary containment isolation instrumentation automatically initiates closure of appropriate secondary containment isolation dampers (SCIDs) and starts the Standby Gas Treatment (SGT) System. The function of these systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refs.1, 2, and 3). Secondary containment isolation and establishment of vacuum with the SGT System ensures that fission products that' leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be OPERABLE are maintained within applicable limits. The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of secondary containment isolation. Most channels include O electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the set)oint is exceeded, the channel output relay actuates, which taen outputs a secondary containment isolation signal to the isolation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logic are (1) reactor vessel water level, (2) drywell pressure, and (3) reactor building exhaust radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation. The outputs of the channels associated with the Reactor Vessel Water Level-Low Level 2 Function and the Drywell Pressure-High Function are arranged in one-out-of-two taken twice trip system logics. One trip system initiates isolation of one automatic secondary containment isolation damper in each penetration and starts both SGT subsystems while the other trip system initiates isolation of the other automatic secondary containment isolation damper in each penetration and starts both SGT subsystems. Each logic closes one of _the two dampers in each penetration and starts both SGT subsystems, so that operation of either logic isolates the secondary containment and provides for the necessary filtration of fission products. (continued) Brunswick Unit 2 B 3.3-175 Revision No. I
1 Secondary Containment Isolation Instrumentation ~l B 3.3.6.2 BASES BACKGROUND' The outputs of the channels associated with the Reactor (continued) Building Exhaust Radiation-High Function are arranged in two one-out-of-one trip system logics. Each trip system initiates isolation of one automatic secondary containment l isolation damper in each penetration and starts both SGT subsystems while the other trip system initiates isolation of the other secondary containment isolation damper in each penetration and starts both SGT subsystems. Each logic closes one of the two dampers in each penetration and starts i both SGT subsystems, so that operation of either logic isolates the secondary containment and provides for the necessary filtration of fission products. APPLICABLE The isolation signals generated by the secondary containment j SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO, and safety analyses of References 1, 2, and 3 to initiate j APPLICABILITY closure of dampers and start the SGT System to limit offsite ! daes. Refer to LC0 3.6.4.2, " Secondary Containment Isolation Dampers (SCIDs)," and LCO 3.6.4.3, " Standby Gas Treatment (SGT) System," Applicable Safety Analyses Bases for more O detail of the safety analyses. The secondary containment isolation instrumentation satisfies Criterion 3 of Reference 4. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set within the specified Allowable Values, ; as shown in Table 3.3.6.2-1. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Allowable Values are specified for each Function specified in the Table. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less (continued) Brunswick Unit 2 B 3.3-176 Revision No.
Secondary Containment Isolation Instrumentatien B 3.3.6.2 BASES @ APPLICABLE conservative than the trip setpoint, but within its t SAFETY ANALYSES, Allowable Value, is acceptable. A channel is inoperable if LCO,'and its actual trip setting is not within its required Allowable APPLICABILITY Value. (continued) Trip.setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor
.'. vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit). changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors.
The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration' based errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, O instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) kre accounted for and appropriately applied for the instrumentation. In general, the individual functions are required to be OPERABLE in the MODES or other specified conditions when SCIDs and the SGT System are required. The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a function by Function basis.
- 1. Reactor Vessel Water Level-Low Level 2 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The Reactor Vessel (continued) O Brunswick Unit 2 B 3.3-177 Revision No.
Secondary Centainment Isolation Instrumentaticn 8 3.3.6.2 BASES APPLICABLE 1. Reactor Vessel Water Level-Low Level 2 (continued) SAFETY ANALYSES. LCO, and Water Level-Low Level 2 Function is one of the Functions APPLICABILITY assumed to be OPERABLE and capable of providing isolation and initiation signals. The isolation and initiation systems on Reactor. Vessel Water Level-Low Level 2 support actions to ensure that any offsite releases are within the limits calculated in the safety analysis. Reactor Vessel Water Level-Low Level 2 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Level 2 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Reactor Vessel Water Level-Low Level 2 Allowable Value was chosen to be the same at the High Pressure Coolant Injection / Reactor Core Isolation Cooling (HPCI/RCIC) Reactor Vessel Water Level-Low Level 2 Allowable Value (LCO 3.3.5.1 O and LC0 3.3.5.2), since this could indicate that the capability to cool the fuel is being threatened. The Allowable Value is referenced from reference level zero. Reference level zero is 367 inches above the vessel zero point. The Reactor Vessel Water Level-Low Level 2 Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these
, events are low due to the RCS pressure and temperature limitations of these MODES; thus, this function is not required.
- 2. Drywell Pressure-Hiah High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite (continued)
O Brunswick Unit 2 B 3.3-178 Revision No.
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 2. Drywell Pressure-Hioh (continued)
' SAFETY ANALYSES, LCO, and dose release. The isolation on high drywell pressure APPLICABILITY supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis.
However, the Drywell Pressure-High Function associated with isolation is not assumed in any UFSAR accident or transient analyses. It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis. High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function. The Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Function Allowable Value (LC0 3.3.5.1) since this is indicative of a loss of coolant accident (LOCA). The Drywell Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.
- 3. Reactor Buildina Exhaust Radiation-Hioh High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.
The release may have originated from the primary containment due to a break in the RCPB or the refueling floor due to a fuel handling accident. When Reactor Building Exhaust Radiation-High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission products as assumed in the UFSAR safety analyses (Ref. 2). (continued) O Brunswick Unit 2 B 3.3-179 Revision No.
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 3. Reactor Buildina Exhaust Radiation-Hiah (continued) SAFETY ANALYSES. . LCO, and The Reactor Building Exhaust Radiation-High signals are APPLICABILITY initiated from radiation detectors that are located in the ventilation exhaust ductwork plenum coming from the reactor building. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Two channels of Reactor Building Exhaust Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are chosen to promptly detect gross failure of the fuel cladding. 1he Reactor Building Exhaust Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, the Function is not required. In addition, the Function is also O required to be OPERABLE during CORE ALTERATIONS, OPDRVs, and movement of irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation , releases due to fuel failures (due to fuel uncovery or ! dropped fuel assemblies) must be provided to ensure that ; offsite dose limits are not exceeded. ' ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels. > Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel. (continued) O Brunswick Unit 2 B 3.3-180 Revision No.
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASFS ACTIONS M , I (continued) l Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Function 2, and 24 hours for Functions other than l Function 2, has been shown to be acceptable (Refs. 5 and 6) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). I If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in tri) (e.g., as in the case where placing the inoperable ciannel in trip would result in an isolation), Condition C must be entered and its Required Actions taken. O M Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic isolation capability for the associated penetration flow path (s) or a complete loss of automatic l initiation capability for the SGT System. A Function is I considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given Function on a valid signal. This ensures that one of the two SCIDs in the associated penetration flow path and both SGT subsystems can be initiated on an isolation signal from the given Function. For the Functions with two one-out-of-two logic trip systems (Functions 1 and 2), this would require one trip system to have one channel OPERABLE or in trip. An ino)erable channel need not be placed in the ! tripped condition w1ere this would cause the trip Function l to occur. In these cases, if the inoperable channel is not , ! restored within the required Completion Time, Condition C ! shall be entered. (continued) Brunswick Unit 2 B 3.3-181 Revision No.
Secondary Containment Isolation Instrumentaticn B 3.3.6.2 BASES l l l ACTIONS R d (continued) The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. C.1.1. C.I.2. C.2.1. and C.2.2 i I If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate the secondary containment and start the SGT System cannot be ensured. Therefore, further actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the associated dampers and starting the associated SGT subsystem (Required Actions C.1.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue. Alternately, declaring the associated SCIDs or SGT subsystem (s) inoperable (Required Actions C.I.2 and C.2.2) O is also acceptable since the Required Actions of the respective LCOs (LC0 3.6.4.2 and LCO 3.6.4.3) provide appropriate actions for the inoperable components. One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems. SURVEILLANCE As noted at the beginning of the SRs, the SRs for each ; REQUIREMENTS Secondary Containment Isolation instrumentation Function are i located in the SRs column of Table 3.3.6.2-1. ! The Surve111ances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: 2 (a) for up to 2 hours for Function 3 and (b) for up to 6 hours for Functions 1 and 2 provided the associated Function maintains isolation capability. Upon completion of , the Surveillance, or expiration of the 2 hour allowance for '
, (continued) ;
O Brunswick Unit 2 B 3.3-182 Revision No. l l l
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE Function 3 or the 6 hour allowance for Functions 1 and 2, REQUIREMENTS the channel must be returned to OPERABLE status or the (continued) applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis _(Refs. 5 and 6) assumption of the average time required to perform channel surveillance. That analysis demonstrated the 2 hour I testing allowance for Function 3 and the 6 hour. testing ! allowance for Functions 1 and 2 do not significantly reduce I the probability that the SCIDs will isolate the associated penetration flow paths and that the SGT System will initiate when necessary. SR 3.3.6.2.1 , Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of O excessive instrument drift in one of the channels or something even mora serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the . instrumentation continues to operate properly between each I CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO. (continued) O Brunswick Unit 2 B 3.3-183 Revision No.
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR' 3.3.6.2.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL ~ TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment-shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of References 5 and 6. SR 3.3.6.2.3 Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.2-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be s equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability analysis of References 5 and 6. ! SR 3.3.6.2.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL _ CALIBRATION leaves the channel' adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint ; methodology. The Frequency is based on the assumption of a 24 month i calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. (continued) O Brunswick Unit 2 B 3.3-184 Revision No.
Secondary Containment Isolatten Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.5 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed on SCIDs , and the SGT System in LC0 3.6.4.2 and LCO 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the Surveillance when performed at the 24 month Frequency. REFERENCES 1. UFSAR, Section 15.6.4.
- 2. UFSAR, Section 15.7.I.
- 3. NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995. ,
- 4. 10CFR50.36(c)(2)(ii).
- 5. NEDC-31677P-A, Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation, July 1990.
- 6. NEDC-30851P-A Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
-\ Brunswick Unit 2 B 3.3-185 Revision No.
i CREV Systea Instrumentatien B 3.3.7.1 O V B 3.3 INSTRUMENTATION B 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation BASES l BACKGROUND The CREV System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. Two independent CREV subsystems are each capable of fulfilling the stated safety function. The instrumentation and controls for the CREV System automatically initiate action to pressurize the main control room (MCR) to minimize the consequences of radioactive material in the control room environment. In the event of a Control Building Air Intake Radiation-High signal, the CREV System is automatically started in the radiation / smoke protection mode. Air is then recirculated through the charcoal filter, and sufficient outside air is drawn in through the normal intake to maintain the MCR slightly pressurized with respect to outside atmosphere. The CREV System instrumentation has two tri) systems, either
-of which can initiate the CREV System. Eaci trip system receives input from the two Control Building Air Intake Radiation-High Function channels. The Control Building Air Intake Radiation-High Function is arranged in a one-out-of-two logic for each trip system. The channels include electronic equipment (e.g , trip units) that g
compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a CREV System initiation signal to the initiation logic. APPLICABLE. The ability of the CREV System to maintain the habitability SAFETY ANALYSES, of the MCR is explicitly assumed for the design basis LCO, and accident as discussed in the UFSAR safety analyses (Ref.1). APPLICABILITY CREV System operation ensures that the radiation exposure of control room personnel, through the duration of any one of the postulated accident 1, does not exceed the limits set by GDC 19 of 10 CFR 50, Appendix A. CREV System instrumentation satisfies Criterion 3 of Reference 2. (continued) O Brunswick Unit 2 B 3.3-186 Revision No.
CREV System Instrumentation B 3.1.7.1 BASES-APPLICABLE - The OPERABILITY of the CREV System instrumentation is SAFETY ANALYSES, dependent upon the OPERABILITY of the Control Building Air LCO, and Intake Radiation-High instrumentation channel Function. APPLICABILITY The Function must have a required number of OPERABLE (continued) channels, with their setpoints within the specified Allowable Value.- The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. l Allowable Values are specified for Control Building Air Intake Radiation-High Function. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. , Operation with a trip setting less conservative than the l trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., control building air intake radiation), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits O are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. ' These calibration based instrument errors are limited to instrument drift, errors associated with measurement and ) test equipment, and calibration tolerance of loop components. The trip setpoints and Allowahls Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. The control building air intake radiation monitors measure radiation levels in the control building air intake plenum. A high radiation level may pose a threat to MCR personnel; thus, automatically initiating the CREV System. (continued) O Brunswick Unit 2 B 3.3-187 Revision No.
CREV System Instrumentation B 3.3.7.1 BASES APPLICABLE The Control Building Air Intake Radiation-High Function SAFETY ANALYSES, consists of two independent monitors. Two channels per trip /A LCO, and system of Control Building Air Intake Radiation-High are fU APPLICABILITY available and are required to be OPERABLE to ensure that no (continued) single instrument failure can preclude CREV System initiation. The Allowable Value was selected to ensure protection of the control room personnel. The Control Building Air Intake Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 and during CORE ALTERATIONS, OPDRVs, and movement of irradiated fuel assemblies in the secondary containment, to ensure that-control room personnel are protected during a LOCA, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., CORE ALTERATIONS), the probability of a LOCA, main steam line break accident, control rod drop accident, or fuel damage is low; thus, the Function is not required. ACTIONS A Note has been provided to modify the ACTIONS related to CREV System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been 'O entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable CREV System instrumentation channels provide a)propriate compensatory measures for separate inoperable ciannels. As such, a Note has been provided that allows separate Condition entry for each inoperable CREV System instrumentation channel. Ad /d\ ! Because of the redundancy of sensors available to provide initiation signals and the redundancy of the CREV System design, an allowable out of service time of 7 days is provided to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the Control Building Air Intake Radiation-High function is still maintaining CREV System initiation capability (refer to Required Action B.1 Bases). g (continued) O Brunswick Unit 2 8 3.3-188 Revision No.
CREV System Instrumentation B 3.3.7.1 BASES l l ACTIONS M (continued) g If the Function is not maintaining CREV System initiation capability, Condition B must be entered. g If the inoperable channel cannot be restored to OPERABLE status within the 7 day allowable out of service time, one CREV subsystem must be placed in the radiation / smoke protection mode of operation per Required Action A.I. The method used to place the CREV subsystem in operation must provide for automatically re-initiating the subsystem upon restoration of power following a loss of power to the CREV subsystem. Placing one CREV subsystem in the radiation / smoke protection mode of operation )rovides a suitable compensatory action to ensure that tie automatic radiation protection function of the CREV System is not l e d. . g M Required Action B.1 is intended to ensure that appropriate q action is taken if multiple, inoperable, untripped channels Q result in the Control Building Air Intake Radiation-High Function not maintaining CREV System initiation capability. The Function is considered to be maintaining CREV System initiation capability when sufficient channels are OPERABLE or in trip such that one trip system will generate an initiation signal for one CREV subsystem from the Function on a valid signal. For the Control Building Air Intake Radiation-High Function, this would require one trip system to have one channel OPERABLE or in trip. blith CREV System initiation capability not maintained, one CREV subsystem must be placed in the radiation / smoke protection mode of operation per Required Action B.1 to ensure that control room personnel will be protected in the event of a Design Basis Accident. The method used to place the CREV subsystem in operation must provide for automatically re-initiating lb the subsystem upon restoration of power following a loss of power to the CREV subsystem. 8 The 1 hour Completion Time is intended to allow the operator time to place the CREV subsystem in operation. The I hour Compietion Time is acceptable because it minimizes risk g while allowing time for restoration or tripping of channels, or for placing one CREV subsystem in operation. g (continued) f O Brunswick Unit 2 B 3.3-189 Revision No.
CREV System Instrumentation-B 3.3.7.I O BASES (continued) SURVEILLANCE The Surveillances are modified by a Note tc indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours,.provided the associated Function maintains CREV System initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered t.nd Required Actions taken. This Note is based on the reliability analysis (Ref. 3) asemption of the average time required to perform channel surwillance. That analysis demonstrated that the 6 hour testing' allowance does not significantly reduce the probability that the CREV System will initiate when necessary. SR 3.3.7.1.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter O indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays y associated with channels required by the LCO. (continued) l O Brunswick Unit 2 B 3.3-190 Revision No.
CREV System Instrumentation B 3.3.7.1 BASES SURVEILLANCE SR 3.3.7.1.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended fur.ction. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analyses of Reference 3. SR 3.3.7.1.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency is based upon the assumption of a 24 month O calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
}R 3.3.7.1.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LC0 3.7.3, " Control Room Emergency Ventilation (CREV)
System," overlaps this Surveillance to provide complete testing of the assumed safety function. While this surveillance can be performed with the reactor at power, operating experience has demonstrated these components will usually pass the Surveillance when performed at the 24 month Frequency. _ Therefore, the Frequency was found to be acceptable from a reliability standpoint. (continued) I l O Brunswick Unit 2 B 3.3-191 Revision No. l l
i i CREV System Instrumentation 1 B 3.3.7.1 ) BASES (continued) REFERENCES 1. UFSAR, Section 15.6.4.5.5. i
- 2. 10 CFR 50.36(c)(2)(ii).
- 3. GENE-770-06-1-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, ;
December 1992. O O Brunswick Unit 2 B 3.3-192 Revision No.
1 Cendenser Vacuu: Pump Isolaticn Instrumentatien 8 3.3.7.2 (~ l B 3.3 ' INSTRUMENTATION B 3.3.7.2 Condenser Vacuum Pump Isolation Instrumentation l l l BASES BACKGROUND The condenser vacuum aump isolation instrumentation l initiates a trip of tte respective condenser vacuum pump and isolation of the common isolation valve following events in which main steam radiation monitors exceed a predetermined-value. Tripping and isolating the condenser vacuum pumps l limits control room doses in the event of a control rod drop j- accident (CRDA). The condenser vacuum pump isolation instrumentation includes , sensors, relays and switches that are necessary to cause ' initiation of condenser vacuum pump isolation. The channels include electronic equipment that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an isolation signal to the condenser vacuum pump , isolation logic. O The isolation logic consists of two independent trip systems, with two channels of the Main Steam Line g Radiation-High Function in each trip system. Each trip system is a one-out-of-two logic for this Function. Thus, either channel of the Main Steam Line Radiation-High Function in each trip system are needed to trip a trip system. The outputs of the channels in a trip system are ; arranged in a logic so that both trip systems must trip to i result in an isolation signal. l l There are two condenser vacuum pumps and one isolation valve associated with this function. APPLICABLE The condenser vacuum pump isolation is assumed in the safety SAFETY ANALYSES analysis for the CRDA. The condenser vacuum pump isolation instrumentation initiates an isolation of the condenser vacuum pumps to limit control room doses resulting from fuel cladding failure in a CRDA. ; The condenser vacuum pump isolation instrumentation satisfies Criterion 3 of Reference 1. l (continued) 4 Brunswick Unit 2 B 3.3-193 Revision No. l
Condenser Vacuum Pump Isolatten Instrumentation B 3.3.7.2 q h BASES (continued) LCO The OPERABILITY of the condenser vacuum pump isolation instrumentation is dependent on the OPERABILITY of the individual Main Steam Line Radiation-High Function instrumentation channels, which must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 3.3.7.2.3. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Channel 0PERABILITY also includes the condenser vacuum pump trip breakers and isolation valve. Allowable Values are specified for the condenser vacuum pump isolation Function specified in the LCO. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the actual trip settings do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its I Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. h) Trip setpoints are those 3 redetermined values of output at which an action should ta(e place. The setpoints are g v compared to the actual process parameter (i.e., main steam line radiation), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for the calibration based errors. These calibration based errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerances of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. 1 (continued) 4 O (O Br,nswick Unit 2 8 3.3-194 Revision No.
Condenser Vacuus Pump Isolaticn Instrumentation B 3.3.7.2 BASES (continued) APPLICABILITY The condenser vacuum pump isolation is required to be OPERABLE in MODES 1 and 2 when the condenser vacuum pump is in service to mitigate the consequences of a postulated CRDA. In this condition, fission products released during a CRDA could be discharged directly to the environment. Therefore, the condenser vacuum pump isolation is necessary to assure conformance with the radiological evaluation of the CRDA. In MODE 3, 4 or 5 the consequences of a control rod drop are insignificant, and are not expected to result in any fuel damage or fission product releases. When the condenser vacuum pump is not in operation in MODE 1 or 2, fission product releases via this pathway would not occur. ACTIONS A Note has been provided to modify the ACTIONS related to condenser vacuum pump isolation instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the /x Condition continue to apply for each additional failure, m 0 with Completion Times based on initial entry into the Condition. However, the Required Actions for ino)erable condenser vacuum pump. isolation instrumentation ciannels provide appropriate compensatory measures for separate I inoperable channels. As such, a Note has been provided that l allows separate Condition entry for each inoperable l condenser vacuum pump isolation instrumentation channel. ; I A.1 and A.2 With one or more channels inoperable, but with condenser vacuum pump isolation capability maintained (refer to : Required Actions B.1, B.2, and 8.3 Bases), the condenser vacuum pump isolation instrumentation is capable of performing the intended function. However, the reliability and redundancy of the condenser vacuum pump isolation instrumentation is reduced, such that a single failure in one of the remaining channels could result in the inability of the condenser vacuum pump isolation instrumentation to i perform the' intended function. Therefore, only a limited time is allowed to restore the inoperable channels to (continued) O Brunswick Unit 2 B 3.3-195 Revision No.
Condenser Vacuua Pump Isolation Instrumentation B 3.3.7.2 BASES ACTIONS A.1 and A.2 (continued) OPERABLE status. Because of the low probability of exterisive number of inoperabilities affecting multiple channels, and the low probability of an event requiring the initiation of condenser vacuum pump isolation,12 hours has been shown to be acceptable (Ref. 2) to permit restoration of any inoperable channel to OPERABLE status (Required Action A.1). Alternately, the inoperable channel, or associated trip system, may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable condenser vacuum pump trip breaker or isolation valve, since this may not adequately compensate for the inoperable condenser vacuum pump trip breaker or isolatfor, valve (e.g..' the trip breaker may be inoperable such that it will not trip). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in loss of condenser vacuum), or if the inoperable channel is the result of an inoperable condenser O vacuum pump trip breaker or isolation valve, Condition B must be entered and its Required Actions taken. b B.I. B.2. and B.3 With any Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve ',his status, the plant must be brought to at least MODE 3 with 12 hours (Required Action B.3 . Alternately, the associated condenser vacuum pum)ps may be removed from service since this performs the-intended function of the instrumentation (Required Action B.1). An additional option is provided to isolate the main steam lines (Required Action B.2), which may allow operation to continue. Isolating the main steam lines effectively provides an equivalent level of protection by precluding fission product transport to the condenser. This isolation is accomplished by isolation of all main steam lines and main steam line drains which bypass the main steam isolation valves. (continued) iO l Brunswick Unit 2 B 3.3-196 Revision No.
Condenser Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES ACTIONS B.I. B.2. and B.3 (continued) i B is also intended to ensure that appropriate Condit'on actions are taken if multiple, inoperable, untripped channels result in the Function not maintaining condenser vacuum pump isolation capability. The Function is considered to be maintaining condenser vacuum pump isolation capability when sufficient channels are OPERABLE or in trip such that the condenser vacuum pump isolation instruments will . generate a trip signal from a valid Main Steam Line Radiation-High signal, and the condenser vacuum pumps will trip. This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the condenser vacuum pump trip breakers to be OPERABLE. SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains .(% condenser vacuum pump isolation trip capability. Upon completion of the Surveillance, or expiration of the 6 hour g Q allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 2) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the condenser vacuum pumps will isolate I when necessary. SR 3.3.7.2.1 Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is nomally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of (continued) O Brunswick Unit 2 B 3.3-197 Revision No.
Condenser. Vacuum Pump Isolation Instrumentation B 3.3.7.2 BASES I SURVEILLANCE SR 3.3.7.2.1 (continued) REQUIREMENTS excessive instrument. drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. ) Agreement criteria are determined by the plant staff based on combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based on the CHANNEL CHECK Frequency requirement of other instrumentation. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO. SR 3.3.7.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended g function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 2. SR 3.3.7.2.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. The Frequency is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. (continued) O Brunswick Unit 2 B 3.3-198 Revision No.
Condenser Vacuum Pump Isolation Instrumentatien B 3.3.7.2 BASES. SURVEILLANCE SR 3.3.7.2.3 (continued) REQUIREMENTS For the purposes of this SR, background is the dose level experienced at 1005 RATED THERMAL POWER with hydrogen water chemistry at the maximum injection rate. Under these conditions, an Allowable Value of s 6 x background will ensure that General Design Criterion 19 limits of 10 CFR 50, Appendix A, will not be exceeded in the control room in the event of a CRDA. SR 3.3.7.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers and actuation of the associated isolation valve are included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a breaker is incapable of o.)erating or the isolation valve is incapable of actuating, tie instrument channel would be inoperable. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an un>1anned transient if the Surveillance were performed with tie reactor at power. REFE'RENCES- 1. 10CFR50.36(c)(2)(ii).
- 2. NEDC-30851P-A, Supplement 2, Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation, March 1989.
O Brunswick Unit 2 B 3.3-199 Revision No.
L LOP Instrumentaticn l' B 3.3.8.1 B 3.3. INSTRUMENTATION B 3.3.8.1 Loss of Power (LOP) Instrumentation j BASES BACKGROUND Successful operation of the required safety functions of the Emergency Core Cooling Systems (ECCS) is dependent upon the availability of adequate power sources for energizing the various components such as pump motors, motor operated valves, and the associated control components. The LOP ! instrumentation monitors the 4.16 kV emergency buses. l Offsite power is the preferred source of power for the ! 4.16 kV emergency buses. If the monitors determine that i insufficient power is available, the buses are disconnected l from the offsite power sources and connected to the onsite diesel generator (DG) power sources. Each 4.16 kV emergency bus has its own independent LOP instrumentation and associated trip logic. The voltage for i each bus is monitored at two levels, which can be considered i as two different undervoltage Functions: Loss of Voltage and 4.16 kV Emergency Bus Undervoltage Degraded Voltage. Each Function causes various bus transfers and disconnects. O- The Loss of Voltage Function is monitored by one inverse time delay undervoltage relay (27/59E) and the Degraded Voltage Function is monitored by three definite time undervoltage relays (27DVA, 27DVB, and 27DVC) for each emergency bus. The Loss of Voltage function is a one-out-of-one logic configuration and the Degraded Voltage Function output is arranged as a two-out-of-three logic configuration. The channels include electronic equipment (e.g., internal relay contacts, coils, etc. measuredinputsignalswithpre-estabitshed)setpoints.that Whencompares the set >oint is exceeded, the channel output relay actuates, which tten outputs a LOP trip signal to the trip logic. APPLICABLE The LOP instrumentation is required for Engineered Safety SAFETY ANALYSES, features to function in any accident with a loss of offsite LCO, and power. The required channels of LOP instrumentation ensure APPLICABILITY that the ECCS and other assumed systems powered from the DGs, provide plant protection in the event of any of the Reference 1 and 2 analyzed accidents in which a loss of offsite power is assumed. The initiation of the DGs on loss of offsite power, and subsequent initiation of the ECCS, ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. (continued) O Brunswick Unit 2 8 3.3-200 Revision No. l i
7_ __ _ . . _ _ _ _ _ LOP Instrumentation B 3.3.8.1 BASES APPLICABLE Accident analyses credit the loading of the DG based on the SAFETY ANALYSES, loss of offsite power during a loss of coolant accident. LCO, and The diesel starting and loading times have been included in APPLICABILITY the delay time associated with each safety system component (continued) requiring DG supplied power following a loss of offsite power. The LOP instrumentation satisfies Criterion 3 of Reference 3. The OPERABILITY of the LOP instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.8.1-1. Each Function must have a required number of OPERABLE channels )er 4.16 kV emergency bus, with their setpoints within tie specified Allowable Values. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. The Allowable Values are specified for each Function in the Table. Trip setpoints are specified in the setpoint calculations. The set,90ints are selected to ensure that the trip settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less
- 3) conservative than the trip setpoint, but within the Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., degraded voltage), and when the' measured output value of the process parameter exceeds the setpoint, the associated device (e.g.,
trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for (continued) m U Brunswick Unit 2 B 3.3-201 Revision No.
LOP Instrumentation B 3.3.8.1
-G V BASES APPLICABLE channels that must function in harsh environments as defined SAFETY ANALYSES, by 10 CFR 50.49) are accounted for and appropriately amlied LCO, and for the instrumentation.
APPLICABILITY (continued) The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.
- 1. 4.16 kV Emeraency Bus Undervoltaae (Loss of Voltaael Loss of voltage on a 4.16 kV emergency bus indicates that offsite power may be com)1etely lost to the respective emergency bus and is unaal'e to supply sufficient power for proper operation of the applicable equipment. Therefore, the power supply to the bus is transferred from offsite power to DG power when the voltage on the bus drops below the Loss of Voltage Function Allowable Values (loss of voltage with a short time delay). This ensures that adequate power will be available to the required equipment.
The Bus Undervoltage Allowable Values are low enough to prevent inadvertent power supply transfer, but high enough
!eD to ensure that power is available to the required equipment.
V The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that power is available ; to the required equipment. l One channei of 4.16 kV Emergency Bus Undervoltage (l.oss of Voltage) Function per associated emergency bus is only l required to be OPERABLE when the associated DG is required to be OPERABLE to ensure that no single instrument failure !' can preclude the start of three of the four DGs. (One channel inputs to each of the four DGs.) Refer to LCO 3.8.1, "AC Sources-0perating," and 3.8.2, "AC Sources-Shutdown," for Applicability Bases for the DGs. l
- 2. 4.16 kV Emergency Bus Undervoltaae (Deoraded Voltaae_). '
A reduced voltage condition on a 4.16 kV emergency bus indicates that, while offsite power may not be completely lost to the respective emergency bus, available power may be (continued) O Brunswick Unit 2 B 3.3-202 Revision No.
LOP Instrumentaticn B 3.3.8.1
. BASES APPLICABLE 2. 4.16 kV Emeraency Bus Undervoltaae (Dearaded Voltaael SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS function. Therefore, the power supply to the bus is transferred from offsite power to onsite DG power when the voltage on the bus drops below the Degraded Voltage Function Allowable Values (degraded voltage with a time delay). This ensures that adequate power will be available to the required equipment. The Bus Undervoltage Allowable Values are low enough to prevent inadvertent power supply transfer, but high enough to ensure that sufficient power is available to the required equipment. The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that sufficient power is available to the required equipment. Three channels of 4.16 kV Emergency Bus Undervoltage (Degraded Voltage) Function per associated bus are only required to be OPERABLE when the associated DG is' required to be OPERABLE to ensure that no single instrument failure O can )reclude the DG function. (Three channels-input to each of tie four emergency buses and DGs.) Refer to LC0 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs. ACTIONS A Note has been provided to modify the ACTIONS related to LOP instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable LOP instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable LOP instrumentation channel. (continued) O Brunswick Unit 2 B 3.3-203 Revision No.
LOP Instrumentaticn B 3.3.8.1 BASES , ACTIONS M (continued) With one or more channels of a Function inoperable, the Function is not capable of performing the intended function. Therefore, only I hour is allowed to restore the inoperable channel to OPERABLE status. If the inoperable channel-cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.I. Placing the
-inoperable channel in trip would conservatively ccmpensate for the inoperability, restore capability to accommodate a single failure (within the LOP instrumentation), and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in a DG
. initiation), Condition B must be entered and its Required Action taken.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channel s. O M' If any Required Action and associated Completion Time are not met, the associated function is not capable of performing the intended function. Therefore, the associated DG(s) is declared inoperable immediately. This requires entry into applicable Conditions and Required Actions of LCO 3.8.1 and LCO 3.8.2, which provide appropriate actions for the inoperable DG(s). SURVEILLANCE As noted at the beginning of the SRs, the SRs for each LOP REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.8.1-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to (continued) O Brunswick Unit 2 B 3.3-204 Revision No. i
i m LOP Instrumentation 8 3.3.8.1 BASES SURVEILLANCE 2 hours provided: (a) for Function 1, the associated REQUIREMENTS- Function maintains initiation capability for three DGs; and (continued) (b) for Function 2, the associate runction maintains DG initiation capability. For Function 1, the loss of function for one DG for this short period is appropriate since only three of four DGs are required to start within the required times and because there is no appreciable impact on risk. Ab. upon completion of the Surveillance, or expiration of the 2 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. SR 3.3.8.1.1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 31 days is based on operating experience O with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function is a rare event. SR 3 3.8.1.2 and SR 3.3.8.1.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. ~ Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The frequencies of SR 3.3.8.1.2 and SR 3.3.8.1.3 are based upon the assumptions of 18 and 24 month calibration intervals, respectively, in the detemination of the magnitude of equipment drift in the setpoint analyses. (continued) O Brunswick Unit 2 B 3.3-205 Revision No.
LOP Instrumentaticn B 3.3.8.1 A BASES SURVEILLANCE 1R 3.3.8.1.4 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specific channel and includes simulated automatic operation of the chantiel . The system functional testing performed in LCO 3.8.1 and LC0 3.8.2 overlaps this Surveillance to provide complete testing of the assumed safety functions. The 24 month Frequency is based on the need to perform this l Surveillance under the conditions that apply during a plant ' outage and the potential for an un)lanned transient if the Surveillance were performed with tie reactor at power. Operating experience has demonstrated these components will pass the Surveillance when performed at the 24 month Frequency. REFERENCES 1. UFSAR, Section 6.3.
- 2. UFSAR, Chapter 15.
- 3. 10CFR50.36(c)(2)(ii).
i l l O Brunswick Unit 2 B 3.3-206 Revision No.
)
RPS Electric Power Monitoring B 3.3.8.2 B 3.3 INSTRUNENTATION B 3.3.8.2 Reactor Protection System (RPS) Electric Pow'r e Monitoring BASES BACKGROUND RPS Electric Power Monitoring System is provided to isolate the RPS bus from the motor generator (MG) set or an alternate power supply in the event of overvoltage, undervoltage, or underfrequency. This system protects the loads connected to the RPS bus against unacceptable voltage and frequency conditions (Ref. 1) and forms an important part.of the primary success path of the essential safety circuits. Some of the essential , equipment powered from the RPS buses includes the RPS logic and scram solenoids. RPS electric power monitoring assembly will detect any abnormal high or low voltage or low frequency condition in the outputs of the two MG sets or the alternate power supply and will de-energize its respective RPS bus, thereby causing i all safety functions normally powered by this bus to 1 de-energize. In the event of failure of an RPS Electric Power Monitoring O System (e.g., both in series electric power monitoring assemblies), the RPS loads may experience significant 1 i l effects from the unregulated power supply. Deviation from ; the nominal conditions can potentially cause damage to the scram solenoids and other Class IE devices. In the event of a low voltage condition for an extended period of time, the scram solenoids can chatter and potentially lose their pneumatic control capability, resulting in a loss of primary scram action. In the event of an overvoltage condition, the RPS logic , relays and scram solenoids may experience a voltage higher ' than their design voltage. If the overvoltage condition persists for an extended time period..it may cause equipment degradation and the loss of plant safety function. Two redundant Class IE circuit breakers are connected in series between each RPS bus and its MG set, and between each RPS bus and the alternate power supply. Each of these circuit breakers has an associated independent set of Class IE overvoltage, undervoltage, and underfrequency sensing logic. Together, a circuit breaker and its sensing logic constitute an electric power monitoring assembly. If (continued) O Brunswick Unit 2 8 3.3-207 Revision No.
l 1 RPS Electric Power Monitoring 8 3.3.8.2 O (V BASES BACKGROUND the output of the MG set or the alternate power supply , (continued) exceeds predetermined limits of overvoltage, undervoltage, I or underfrequency, a trip coil driven by this logic circuitry opens the circuit breaker, which removes the associated power supply from service. l APPLICABLE The RPS electric power monitoring is necessary to meet the SAFETY ANALYSES assumptions of the safety analyses by ensuring that the RPS equipment powered from the RPS buses can perform its intended function. RPS electric power monitoring provides protection to the RPS components, by acting to disconnect I the RPS from the power supply under specified conditions that could damage the RPS equipment. RPS electric power monitoring satisfies Criterion 3 of Reference 2. i LCO The OPERABILITY of each RPS electric power monitoring assembly is dependent on the OPERABILITY of the overvoltage, undervoltage, and underfrequency logic, as well as the j OPERABILITY of the associated circuit breaker. Two electric n) 4
" power monitoring assemblies are required to be OPERABLE for each inservice power supply. This provides redundant j
protection against any abnormal voltage or frequency conditions to ensure that no single RPS electric power monitoring assembly failure can preclude the function of RPS components. Each of the inservice electric power monitoring assembly trip logic setpoints is required to be within the specified Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Allowable Values are specified for each RPS electric power monitoring assembly trip logic (refer to SR 3.3.8.2.2 and SR 3.3.8.2.3). Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure i that the trip settings do not exceed the Allowable Value l between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Tri) setpoints are those predetermined values of output at w11ch an action should take place. The setpoints are compared to the actual process parameter (e.g., overvoltage), and when the measured output value of (continued) ( Brunswick Unit 2 B 3.3-208 Revision No. - ~ A
l l- RPS Electric Power Monitoring l B 3.3.8.2 BASES LCO the process parameter exceeds the setpoint, the associated (continued) device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation. The Allowable Values for the instrument settings of the normal power supply (RPS MG set) electric power monitoring assembly are based on the RPS MG sets providing a 57 Hz and O 117 Y i 10%. The Allowable Values for the instrument settings of the alternate power su) ply electric power monitoring assembly are based on tle alternate power supply providing a 57 Hz and 120 V i 10%. The most limiting voltage requirement and associated line losses determine'the settings of the electric power monitoring instrument channels. The settings are calculated based on the line : resistance losses at the downstream locations of the solenoids and relays. APPLICABILITY The o)eration of the RPS electric power monitoring assem)1ies is essential to disconnect the RPS components from the MG set or alternate power supply during abnormal voltage or frequency conditions. Since the degradation of a nonclass IE source supplying power to the RPS bus can occur as a result of any random single failure, the OPERABILITY of the RPS electric power monitoring ess.emblies is required when the RPS components are required to be OPERABLE. This results in the RPS Electric Power Monitoring System OPERABILITY being required in MODES 1 and 2; and in MODES 3, 4, and 5 with any control rod withdrawn from a core cell g containing one or more fuel assemblies. (continued) O Brunswick Unit 2 B 3.3-209 Revision No.
'l RPS Electric Power Monitoring B 3.3.8.2 BASES (continued)
ACTIONS- M If one RPS electric power monitoring assembly for an inservice power supply (MG set or alternate) is inoperable, or one RPS electric power monitoring assembly on each inservice power supply is inoperable, the OPERABLE assembly will still provide protection to the RPS components under degraded voltage or frequency conditions. However, the reliability and redundancy of the RPS Electric Power Monitoring System is reduced, and only a limited time (72 hours) is allowed to_ restore the inoperable assembly to OPERABLE status. If the inoperable assembly cannot be restored to OPERABLE status, the associated power supply (s) must be removed from service (Required Action A.1). This places the RPS bus in.a safe condition. An alternate power supply with OPERABLE powering monitoring assemblies may then be used to power the RPS bus.
]
The 72 hour completion Time takes into account the remaining OPERABLE electric power monitoring assembly and the low probability of an event requiring RPS electric power monitoring protection occurring during this period. It l allows time for plant operations personnel to take ' Os corrective actions or to place the plant in the required condition in an orderly manner and without challenging plant systems. Alternately, if it is not desired to remove the power supply l from service (e.g., as in the case where removing the power ' supply (s) from service would result in a scram or isolation), Condition C or D, as applicable, must be entered and its Required Actions taken. M If both power monitoring assemblies for an inservice power supply (MG set or alternate) are inoperable or both power monitoring assemblies in each inservice power supply are inoperable, the system protective function is lost. In this : condition, I hour is allowed to restore one assembly to ! OPERABLE status for each inservice power supply. If one inoperable assembly for each inservice power supply cannot be restored to OPERABLE status, the associated power (continuedl i O Brunswick Unit 2 B 3.3-210 Revision No.
l 4 RPS Electric Power Monitoring ! B 3.3.8.2 l VO) BASES ACTIONS M (continued) supply (s) must be removed from service within I hour (Required Action B.1). An alternate power supply with OPERABLE assemblies may then be used to power one RPS bus. The I hour Completion Time is sufficient for the plant i operations )ersonnel to take corrective actions and is acceptable secause it minimizes risk while allowing time for restoration or removal from service of the electric power I monitoring assemblies.
]
Alternately, if it is not desired to remove the power ! supply (s) from service (e.g., as in the case where removing i the power supply (s) from service would result in a scram or isolation), Condition C or D, as applicable, must be entered and its Required Actions taken. C.1 and C.2 If any Required Action and associated Completion Time of Condition A or B are not met in MODE 1 or 2,' a plant p shutdown must be performed. This places the plant in a 4.,"J . condition where minimal equipment, powered through the inoperable RPS electric power monitoring assembly (s), is required and ensures that the safety function of the RPS (e.g., scram of control rods) is not required. The plant I shutdown is accomplished by placing the plant in MODE 3 ! within 12 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. M If any Required Action and associated Completion Time of Condition A or B are not met in MODE 3, 4, or 5 with any control rod withdrawn from a core cell containing one or d more fuel assemblies, the operator must immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Required Action D.1 results in the least reactive condition for the reactor core and ensures that the safety function of the RPS (e.g., scram of control rods) is not required. (continued) ! Brunswick Unit 2 B 3.3-211 Revision No.
RPS Electric Power Monitoring 8 3.3.8.2 O O BASES (continued) SURVEILLANCE SR 3.3.8.2.1 REQUIREMENTS . A CHANNEL FUNCTIONAL TEST is performed on each overvoltage, undervoltage, and underfrequency channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. As noted in the Surveillance, the CHANNEL FUNCTIONAL TEST is only required to be )erformed while the plant is in a condition in which tie loss of the RPS bus will not jeopardize steady state power operation (the design of the system is such that the power source must be removed from service to conduct the Surveillance). The 24 hours is intended to indicate an outage of sufficient duration to allow for scheduling and proper performance of the Surveillance. The 184 day Frequency and the Note in the Surveillance are based on guidance provided in Generic Letter 91-09 (Ref. 3). SR 3.3.8.2.2 and SR 3.3.8.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. , 1 The frequencies are based on the assumption of a 24 month calibration interval in the determination of the magnitude ' of equipment drift in the setpoint analysis. SR 3.3.8.2.4 Performance of a system functional test demonstrates that, with a required system actuation (simulated or actual) signal, the logic of the system will automatically trip open the associated power monitoring assembly. Only one signal (continued) O 'U Brunswick Unit 2 B 3.3-212 Revision No.,
l I RPS Electric Power Monitoring B 3.3.8.2 BASES SURVEILLANCE SR 3.3.8.2.4 (continued) REQUIREMENTS per power monitoring assembly is required to be tested. This Surveillance overlaps with the CHANNEL CALIBRATION to provide complete testing of the safety function. The system functional test of the Class IE circuit breakers is included as part of this test to provide complete testing of the safety function. If the breakers are incapable of operating, the associated electric power monitoring assembly would be . inoperable. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for.an unplanned transient if the 1 Surveillance were performed with the reactor at power. l Operating experience has demonstrated that these components l will usually pass the Surveillance when performed at the ! 24 month Frequency. REFERENCES 1. UFSAR, Section 7.2.1.1.1.3.
- 2. 10 CFR 50.36(c)(2)(ii).
- 3. NRC Generic Letter 91-09, Modification of Surveillance Interval for the Electrical Protective Assemblies in Power Supplies for the Reactor Protection System.
O Brunswick Unit 2 B 3.3-213 Revision No.
c+. . . ;;.,+.-/ .. r \ INSTRUMENTATION
,3 SQWhCE RANCE H0gr%RS ION OR . 01'en og LIMI ilG~i ,
/4 0 7. 7./.2. Q Three source. range monitmrs shall be OPERABLE [dri % hree ^
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, and 4. l APPLICABILITY OPERATI0llAL CONDITIONS 2
. ACTIOWs l l IaOPERATIONALCONDITION2R,wjione'of the above required source range '
S monitors inoperable, restore Woource ranse monitors to CPERABLE status ( g g,4
}}g,a k within 4 hours, @r ce to at Least BOT SuuiLOWW within tne next 12 hour g 0 ACT4u C.
In OPERATIONAL CONDITIONS 3 or 4, with-4we e more o he ebeve required h! '
$. source range monitors inoperable, y 1 all p ontrol rods 6's'Ea) fully 4'l de swate a the Shutdown fcTho O A in the core, and oet the reacto pqsition within one hour. y
} hde l7 SURVEILLANCE REQUIREMENTS l
" Y.2.'.4-Eachoftheaboverequiredsource se monitors shall be I
b M ~ b demonstrated OPERABLE bys f, ,y gg [ i 24 hours sur w O I'I *I A*0 h Perin === of a Nam runuuvnAL TET witch f g shutdown y not perf within reificus[. j the SEM count rate . . Verifying,$pfor to4tehdraudT of cos(rol e f g 3 ,3, 3,7, s.l h is at lease. 3 cps drith>hi detector fvKr inspeed. $ '. t'AI em
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t h.\ h CHANNEL CHECE at least once pers
- @ 12 hours in OPERATIONAL CONDITION P , and g g 7,T,/.2.t f(( 3.3. l.2. 3 h24hoursinOPERATIONALCONDITION3or'4. months.
y,T./.E.7 { CHANNEL CALIBRATION @at least once per L6.1 f) ele.2 in . l) q S E 9.s.t.i.7 , f/f MI (a)PWith IBMs on range 2 or below. Nofc.14o ff.7.J.lf.'} My exclude neutron detectors. 3/4 3-43 Amendment No.130 BRUNSWICr - UNIT 1 pge,Iof b
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C C M 3. 3. l. 2. /% b hkLIME ILadiREMEld h Each of the above required SRM channels shall be demonstrated OPERABLE bys j h At least once per 12 hours; Performance of a CHANNEL CHECK, g,,h N fg 3,,3, g,7, g -h
, Ve fying tg detectors a\e inserted to thA norma hoperating) c %,u t s u.1 su.u.zt W.y .
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- Ort,RABLE SRM channel is located i The core cuadrant wher g 9).g,j,f,g
> 9- (, , . ti. MTERATIONS are being performe e is located in the } ,
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- 69. 7 *3.l.*2. 4g During CORE ALTERATIOWss verifying that the channel count rate (s at least 3 cps excepj;ss notedAn Speg+ncattogO.9.2.b@ Migg' M Tf,$ P.; Na d**3
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- 5. Du' g a corm si unwnu or brpL RELOAD, verpg that the 1 movemen e is beine fokreved- -
g g l N*k I b h VerifyingGrion:Atf been raised to the ntset of a SPJJWCL RELCAD a count rate of at least a cps by the e SRMs have insertion of up ! St a 3.1. 7 tj ,, g' j,,p to four fuel assemblies around each of the four SRMs. h Performance of a CHANNEL FUNCTIONAL TESTS
$l' 3.3.1.2..r -d. Withi M ours prior start of C0 % L'TERATIONS, and
[ At least once per seven days. 7 M.3 i i i 1 BRUNSWICK - UNIT 1 3/4 9-4 Amendment No. 89 3 r v t
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3 3.l. L l f TEUNENTATION N j
@TRAucr/MONITpas 1
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g( t.ll OPERABIM [(o 3. "i (t h64 Three source range sionitors shall "Y *" *
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APPLICABILITTs OPERATIONAL CONDITIONS % . (ed c vV*$ ew v. s
- ACTIONS he above required source range l ne
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@ In OPERATIONAL monitors CONDITION 2 , witsource range monitors to g, inoperable, restore CPERABL within 4 houtspeTa at n least nur suum 0WN within the next g g w pf p o f the @ required g,,
ActioeJ C ~g In OPERATIONAL CONDITIONS 3a or allo 4ontrol rods 6 fully 4, vich . source range monitors inoperable, in the corc, and the reacto mode switch in the Shutdown Auwab e tion ylthin one hour. (% p ,( ,ng y ,
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SURVEILLANCE REQUIREMENTS O A.S ~ M 8 O 4* .. Each of the above required source range monitors shall be A(, one pe 3 e .($ g, demonserated CPERAB ^ j [
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(**'# jf[g'd3 .i a pryr 6v -7 wit the previousa de ( @ (Performance of a 2^? vet FUNCTIONAL TESTEEe' mod < SR h t.2.4 { he SRM count race is g
@ VerifyingE$ene cyvatnesswat cycor rv roasfthat at least 3 cps drt% t,me,4etpccorAupy spas SR 13f.?.4
[prforydace t9y CHANNEL CHECK at least once pers 60 12 hours in CPERATIONAL CONDITION 2F, and 9 13 ai -
$p p i,7,3 h 24 hours in OPERATICHAL CONDITION months.
3 or 4. l l
@ CHANNEL CALIBRATI0tM at least once per .
Se 3.' i.2 71 I'I Ll". I MD lasic L y . b sg 34.2, App l.'<A f. (s.) dPWith IRMs on range 2 or below. b y exclude neutron detectors.
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5A % 51.2.7 Amendment No. 160 3/4 3-63 BEUNSWICE - UNIT 2 4 7'g 4 ef k
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Spe4A sesi,z i-lo (A.1 o ILL IRSMI M Each of the above required SRM channels shall be demonstrated OPERABLE bys h At 1esst once per 12 hoursi Performance of a CHANNEL CHECK, "JR %~5, Q l @ eri g the,6etectJds are,dnsert[d to t/nordi q/arathfg] a v % .m e m. u ,a y. , ' 3
$ 8p,,, p io m r itratATIONh ver Mifying that OPE m ti m channet 1. toca thecuseant ee core detector
.,,ere of CoKan s,. u. , ., ,, b. (Ittraatt0MS are neang performe @ one is located in cne ;
- c. (adjacent quadrant w g %g , ., ,,
During CORE ALTERATIONS 4 verifying that the channel count rate sg 3,3. s.c. 4g is at least 3 ep ( 4ece nc =s
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a h L g , L.O h Verifying 666r D6 tMe sta/t of,4 SP)(AL RE@ that the SRMs have c,g 94 g 4, @ been raised to a count rate of at least J cps by the insertion of u g 3,3,o,y to four fuel assemblies around each of the four SRMs. cwt'Ryu.J M+-Cl ,
$ Performance of a CHANNEL FUNCTIONAL TEST
- MithM 24 bdurs prig to t)d stat c/f COREpTEpIONSg an At least once per seven days.
/.4,3 A48' $# 3.3.12 I
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- 3/4 9-4 Amendment No. 114 BRUNSWICK - UNIT 2 p
m., v c
l DISCUSSION OF CHANGES ITS: 3.3.1.2 - SRM INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS 4.3.5.4.b requires verification that SRM count rate is k 3 cps prior to withdrawal of control rods. ITS SR 3.3.1.2.4 requires' this verification to be performed once per 24 hours.. The requirement to perfom the Surveillance periodically during the A applicable MODES and condition, instead of only prior to LO withdrawal of control rods, represents an additional restriction on plant operation necessary to achieve consistency with NUREG-1433, Revision 1. M.2 CTS 3.9.2.e allows up to four fuel assemblies to be loaded into different control cells containing control blades around each SRM to obtain 3 cps. Under the same condition, Note I to ITS SR 3.3.1.2.4 allows SRM count rate drop below 3 cps provided no b other fuel assemblies are in.the associated quadrant. The added requirement, that no other fuel assemblies be located in the core quadrant where the SRM is located, ensures that even with a control rod withdrawn, the core configuration will not be g critical. This requirement represents an additional restriction on plant operation since the allowance for SRM count rate to drop below 3 cps, and the SRM still be considered OPERABLE, will now only apply if no other fuel assemblies are in the core quadrant. A Therefore, in ITS, if other fuel assemblies are in the same core e quadrant as the SRM, other than the fuel assemblies adjacent to , the SRM, the SRM count rate must be a 3 cps. ) M.3 A requirement to perform a CHANNEL CALIBRATION every 24 months on SRM instrumentation (ITS SR 3.3.2.1.7) is added for SRMs during MODE 5. This Surveillance verifies the performance of the SRM detectors and asseciated circuitry. This Surveillance Requirement represents an additional restriction on plant operation necessary to ensure the OPERABILITY of the SRMs during MODE 5. M.4 CTS 4.9.2.a.3 requires verifying that the detector of an OPERABLE-SRM channel is located in the core quadrant where CORE ALTERATIONS are being performed and one is located in the adjacent quadrant. ITS SR 3.3.1.2.2 requires verifying that an OPERABLE SRM detector is located in the fueled region; the core quadrant where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region; and in a core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region. - As a result of providing the additional criteria on where the OPERABLE SRMs must be A relocated, Note 2 to ITS SR 3.3.1.2.2 is also added to clarify a that more than one of the three requirements of ITS SR 3.3.1.2.2 can be n tisfied by the same SRM since only two SRMs are required to be OPERABLE. Providing additional criteria on where the SRMs must be located to satisfy the Surveillance represents an additional restriction on plant operation necessary to provide adequate coverage of potential reactivity changes in the core and to achieve consistency with NUREG-1433, Revision 1. O BNP UNITS 1 & 2 2 Revision 0 y
DISCUSSION OF CHANGES ITS: 3.3.1.2 - SRM INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE (continued) M.5 In MODE 5, CTS 4.g.2.a.4 requires verification that SRM count rate l 1s a: 3 cps once per 12 hours during CORE ALTERATIONS. ITS SR 3.3.1.2.4 requires this verification to be performed 12 hours 1 during CORE ALTERATIONS and once per 24 hours during other times j in MODE 5. The requirement to perform the Surveillance l , periodically in MODE 5 when CORE ALTERATIONS are not being ; I performed, instead of only in MODE 5 during CORE ALTERATIONS, i represents an additional restriction on plant operation necessary i to achieve consistency with NUREG-1433, Revision 1. ' t M.6 In MODE 2 (with IRMs on Range 2 or below) and in MODES 3 and 4, L CTS 4.3.5.4.a effectively requires performance of a CHANNEL
- FUNCTIONAL TEST of the SRMs once prior to entering MODE 2 from l MODE 3 or 4 (i.e., prior to moving the mode switch from Shutdown).
In addition, CTS 4.3.5.4.a does not require performance of a CHANNEL FUNCTIONAL TEST of the SRMs at other times (i.e., prior to entering MODE 2 (with the IRMs on Range 2 or below) from MODE 1, prior to entering NODE 3 or 4, or periodically during the applicable MODES or condition). ITS SR 3.3.1.2.6 requires performance of a CHANNEL FUNCTIONAL TEST of the SRMs once per ! 31 days in the same applicable MODES and condition. As a result, l ITS SR 3.3.1.2.6 will require a CHANNEL FUNCTIONAL TEST of the l' SRMs to be performed periodically and prior to entering MODE 2 from MODE 3 or 4 during a startup instead of only once during a startup. The requirement to perform the Surveillance periodically b ' during the applicable MODES and condition, instead of only prior to moving the mode switch from Shutdown, represents an additional i l restriction on plant operation necessary to help ensure that the i SRMs function properly when required and to achieve consistency l with NUREG-1433, Revision 1. l Requiring performance of the CHANNEL FUNCTIONAL TEST of-the SRMs prior to entering MODE 2 (with the IRMs on Range 2 or below) from i MODE 1 does.not provide adequate time in order to establish the necessary conditions to perform the test and complete performance of the CHANNEL FUNCTIONAL TEST of the SRMs. Therefore, a Note is i added to ITS SR 3.3.1.2.6 which allows the Surveillance to be l delayed until 12 hours after the IRMs on Range 2 or below. As a result. ITS SR 3.3.1.2.6 requires performance of a CHANNEL i FUNCTIONAL TEST of the SRMs to be completed 12 hours after the-IRMs are on Range 2 or below during a plant shutdown. Since the i CTS does not require performance of the CHANNEL FUNCTIONAL TEST of the SRMs during a plant shutdown, this change represents an additional restriction on plant o)eration necessary to help ensure ! that the SRMs function properly w1en required and to achieve consistency with NUREG-1433, Revision 1. ; r i l O BNP UNITS 1 & 2 3 Revision 0 i-l
i DISCUSSION OF CHANGES ITS: 3.3.1.2 - SRM INSTRUMENTATION I TECHNICAL CHANGES - LESS RESTRICTIVE
" Generic" !
LA.1 The detail of the method for performing the CTS Surveillance 4.3.5.4.b is to be relocated to the ITS 3.3.1.2 Bases. The detail ; to be relocated is a procedural detail that is not necessary for. ; assuring SRM OPERABILITY. ITS SR 3.3.1.2.4, along with the other Surveillance Requirements of ITS 3.3.1.2, provide adequate assurance the SRMs are maintained OPERABLE. Therefore, the relocated detail of the method of performing this Surveillance is not required to be in the Technical Specifications to provide adequate protection of the public health and safety. Changes to the ITS Bases will be controlled by the provisions of the ITS Bases Control Program described in Chapter 5 of the.ITS. LA.2 The details of CTS 3/4.9.2 relating to SRM OPERABILITY (in this case that the SRMs shall be inserted to the normal operating level with continuous indication in the control room) are to be relocated to the ITS Bases. These details for system OPERABILITY are not necessary in the LCO. The definition of OPERABILITY l suffices. Therefore, the relocated details are not required to be ' in the Technical Specifications to provide adequate protection of i the public health and safety. Changes to the ITS Bases will be i controlled by the provisions cf the ITS Bases Control Program described in Chapter 5 of the ITS. -l LA.3 The requirement of CTS 3.9.2.c associated with the removal of RPS shorting links is to be relocated from the Technical l Specifications. The shorting links are required to be removed i with any control rod withdrawn from a core cell containing one or l more fuel assemblies when SHUTDOWN MARGIN has not been ! demonstrated. The primary reactivity control functions during refueling are the refueling interlocks and the SHUTDOWN MARGIN. The refueling interlocks are required to be OPERABLE by ITS LCO 3.9.1 and ITS LCO 3.9.2. Although SHUTDOWN MARGIN may not be ; demonstrated until after CORE ALTERATIONS are complete in MODE 5, SHUTDOWN MARGIN calculations performed prior to altering the core, , along with procedural compliance for any CORE ALTERATIONS, provides assurance that adequate SHUTDOWN MARGIN is available. i When the SRM shorting links are removed, IRM instrumentation continues to provide backup for the credited functions for any significant reactivity excursions. Since the SRN channel high flux scram (with shorting links removed) provides only an uncredited backup in MODE 5, the relocation of the shorting link removal requirement does not significantly affect safety. As such, the relocated requirement is not required to be in the Technical Specifications to provide adequate protection of the i public health and safety. Details for control of shorting link removal are controlled by the UFSAR. Changes to the UFSAR are controlled by the provisions of 10 CFR 50.59. h b O BNP UNITS 1 & 2 4 Revision 0
l DISCUSSION OF CHANGES ITS: 3.3.1.2 - SRM INSTRUMENTATION g !q) TECHNICAL CHANGES - LESS RESTRICTIVE (continued) LE.1 CTS 4.3.5.4.c.2 specifies the Frequency for SRM CHANNEL CALIBRATION as at least once every 18 months. ITS SR 3.3.1.2.7 will extend the required Frequency for this SR to 24 months. Therefore, the surveillance test interval of this SR is being increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25% grace period. Extending the SRM calibration interval from 18 months to 24 months is acceptable for the following reasons: SRMs are not required to measure neutron level just changes in neutron level; SRMs satisfy their design function when shutdown if calibration is sufficient to ensure changes in neutron level are observable when the reactor is shutdown. This capability is verified once within 24 hours prior to control rod withdrawal in MODE 2, 3, or 4 when the reactor is shutdown; SRMs satisfy their design function in Mode 2 if calibration is sufficient to ensure overlap with ths IRMs and IRWSRM overlap is verified whenever power level is changed in the overlap region; and, SRMs have no safety function and are not assumed to function during any UFSAR design basis accident or transient analysis. Additionally, SRM response to reactivity changes is distinctive and well known to plant operators and SRM response is closely monitored during these reactivity changes. Therefore, any substantial degradation of the SRMs will be evident prior to the scheduled performance of these tests. Based on the above discussion, the impact, if any, from the surveillance test (] G frequency increase on system availability will be minimal. A review of the surveillance test history for each of these Surveillance Requirements was performed to validate the above conclusion. This historical review of the surveillance test history demonstrates that there are no failures that would invalidate the conclusion that the impact, if any, of this change on system reliability is minimal.
" Specific" L.1 Since CTS 3.3.5.4 only specifies an action for one required SRM inoperable during MODE 2, CTS 3.3.5.4 requires a plant shutdown if two or more required SRMs become inoperable (in accordance with CTS 3.0.3). This requirement is unnecessarily restrictive and does not allow concentration of the efforts on repair. The words "or more" are added (ITS 3.3.1.2 Condition A) to allow the action to apply to two or three inoperable SRMs. This is acceptable based on the limited risk of an event occurring during the time the SRMs are inoperable and the desire to concentrate efforts on repair, rather than an immediate shutdown which is currently required by CTS 3.0.3, with one or no SRMs OPERABLE.
Additionally, with no OPERABLE SRMs, the ability to monitor positive reactivity changes is significantly restricted, thus ITS 3.3.1.2 Required Action B.1 is added to ensure that no further rT U BNP UNITS 1 & 2 5 Revision 0
' DISCUSSION OF CHANGES ITS: 3.3.1.2 - SRM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE
' L.1 control rod withdrawal is allowed. Further, requiring an (cont'd) immediate plant shutdown could, with no SRMs OPERABLE, pose a greater risk since the APRMs and IRNs are inadequate for monitoring neutron flux in the source range.
L.2 CTS 3.3.5.4 Action b requires the mode switch to be locked in the shutdown position. Under the same conditions, ITS 3.3.1.2 Required Action D.2 requires the mode switch to be placed in the shutdown position.' The required position of the reactor mode d switch in MODE 3 or 4 is adequately controlled by theiMODES ! definition Table (ITS Table 1.1-1). Movement of the reactor mode i switch from the Shutdown position (and therefore any requirement to " lock") is adequately controlled by ITS Table 1.1-1. Reactor l' mode switch positions other than Shutdown result in the unit entering some other MODE; with the associated Technical Specification compliance requirements of that MODE and of ITS ! LC0 3.0.4 becoming applicable. Therefore, the requirement of CTS 3.3.5.4 Action b to " lock" the mode switch in shutdown is i deleted from the Technical Specifications and replaced with the requirement to " place" the mode switch in shutdown. d. L L.3 The CTS 3.9.2 Action "... fully insert all insertabic control rods" is revised to "immediately initiate action to insert all insertable control rods..." During MODE 5, it may not be possible i to immediately insert all insertable control rods. In this 5 situation, the CTS do not provide direction as to the action to take if control rods cannot be inserted immediately. As a result, the ITS provide a Required Action (ITS 3.3.1.2 Required l Action to insert E.2)ll a insertable control rods.to immediately This changeinitiate ensuresaction that and continue a l actions are taken to insert'all insertable control rods in a timely manner while continuing to provide direction if attempts fail to immediately insert all insertable control rods. This change is considered to be acceptable since ITS 3.3.1.2 Required Action E.1 ensures the probability 'of occurrence of postulated events involving changes in reactivity in the MODE 5 is minimized by suspension of CORE ALTERATIONS. L.4 A Note is added (Note 2 to ITS SR 3.3.1.2.7) which allows entry into the MODES and conditions where the SRMs are required to be OPERABLE, prior to satisfactory completion of the required CHANNEL CALIBRATION. This is an exception to CTS 4.0.4. The SRMs are required in MODES 2, 3, 4, and 5 but not in MODE 1, and the required Surveillance cannot be performed at high power (prior to L entry into the applicable condition) without utilizing jumpers or lifted leads. Use of these devices is not recommended since minor A errors in their use may significantly increase the probability of LM_ a reactor transient or event which is a precursor to a previously analyzed accident. Additionally, performance of the Surveillance at low power, prior to entry into the applicable condition, poses a greater risk of error and may significantly increase the
, probability of an event due to the plant shutdown which is a BNP UNITS 1 & 2 6 Revision 0 L
i l l DISCUSSION OF CHANGES i ITS: 3.3.1.2 - SRM INSTRUMENTATION l TECHNICAL CHANGES - LESS RESTRICTIVE l L.4 precursor to a previously analyzed accident. Therefore, time is (cont'd) allowed to conduct the SR after entering the applicable MODE. This Note is consistent with the BWR Standard Technical Specifications, NUREG-1433. L.5 CTS 3.9.2.e and CTS 4.9.2.b require verifying SRM ccunt rate is at least 3 cps prior to a core Spiral Reload. A specific performance of ITS SR 3.3.1.2.4 prior to a core Spiral Reload is not required. ITS SR 3.3.1.2.4 Note 1 allows SRM count rate to be below 3 c as with less than or equal to four fuel assemblies adjacent to tse SRM regardless of the type of core reload that will be conducted provided no other fuel assemblies are located in the associated core quadrant (See Comment M.2). This change is acceptable since in this condition, even with a control rod withdrawn, the configuration will not be critical. L.6 The CTS 3.9.2 Action which requires sus)ending operations involving positive reactivity changes wien the minimum required SRMs are inoperable is deleted. This change is acceptable because most positive reactivity changes are already included in the definition of CORE t.LTERATION, which is also required to be suspended by the CTS 3.9.2 Action (ITS 3.3.1.2 Required Action E.1). These requirements and the requirements of ITS LCO 3.1.1, SHUTDOWN MARGIN, are adequate to ensure the core is O maintained subcritical. Movement of a control rod while in MODE 5 (Refueling) is allowed by this change provided there are no fuel assemblies in the associated core cell. Also, movement of source range monitors, local power range monitors, intemediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement) is allowed by this change. The reactivity change associated with movement of these components is insignificant. L.7 CTS 4.9.2.c.1 requires an SRM CHANNEL FUNCTIONAL TEST be performed periodically while in MODE 5. The required periodic Frequency (every 7 days of CTS 4.9.2.c.2 and ITS SR 3.3.1.2.5) has been determined to be sufficient verification that the source range monitors are properly functioning. A review of historical test data for the past 3 years shows there have been no functional failures of the SRMs. Perforaing CORE ALTERATIONS does not impact d the ability of the monitors to perform their required function. Additionally, the requirements of ITS 3.3.1.2 Required Action E.1 preclude beginning CORE ALTERATIONS unless the required equipment is OPERABLE. Therefore, an additional Surveillance required to be performed "within 24 hours prior to CORE ALTERATIONS" is an extraneous and unnecessary performance of a Surveillance and is , deleted. In addition, it is concluded that the impact, if any, on SRM availability is minimal as a result of this change. O BNP UNITS 1 & 2 7 Revision 0 t
DISCUSSION OF CHANGES ITS: 3.3.1.2 - SRM INSTRUNENTATION , TECHNICAL CHANGES - LESS RESTRICTIVE (continued) l L.8 CTS 3.3.5.4 Action b requires verification that all control rods are fully inserted in the core within one hour. ~ ITS 3.3.1.2 Required Action 0.1 requires fully inserting all insertable control rods within I hour. In MODES 3 and 4 with the reactor mode switch in the shutdown position, the ITS Required Action is consistent with the CTS Action in that the control rods will be fully inserted. In MODE 3 or 4 with the reactor modeiswitch in the refuel position (per ITS 3.10.3, " Single Control Rod Withdrawal-Hot Shutdown," or ITS 3.10.4, " Single Control Rod Withdrawal-Cold Shutdown," respectively), one rod may be withdrawn and it may not be ca>able of being inserted. This
.10.3 or change ITS 3.10.4is acceptable and the SDMbecause tie requirements requirements of ITS ybeen of ITS 3.1.1 have determined to be adequate to maintain the reactor subcritical.
L.9 In MODE 5, the CTS 3.9.2 Action requires fully inserting all insertable control rods if one or more required SRMs are inoperable. In'this condition, ITS 3.3.1.2 only requires inserting all insertable control rods in core cell containing one or more fuel assemblies (ITS 3.3.1.2 Required Action E.2). Control rods withdrawn from or inserted into a core cell containing no fuel assemblies have a negligible impact on the reactivity of the core and therefore are not required to be inserted to maintain the reactor subcritical. L.10 CTS 4.9.2.5 requires verification that the fuel movement sheet is being followed during a core spiral unload and reload. This requirement is not required to verify the OPERABILITY of the SRM instrumentation. The Surveillance Requirements of ITS 3.3.1.2 are considered adequate to maintain OPERABILITY of the SRMs. Therefore, CTS 4.9.2.5 is to be removed from Technical Specifications. In addition, ITS 3.1.1, SHUTDOWN MARGIN (SDM), requires SDN to be maintained in MODE 5. During spiral unload and ! reload sequences, SDN is maintained and ITS SR 3.1.1.1 (which requires verification that SDN limits are met) is satisfied. Therefore, CTS 4.9.2.5 requires periodic verification that the specified applicable conditions for satisfying the requirements of ! ITS SR 3.1.1.1 (associated with spiral unload and reload) are met. A l
^
In general, this type of requirement is addressed by plant 2 specific processes which continuously monitor plant conditions to l ensure that changes in the status of the plant that require entry into ACTIONS (as a result of failure to satisfy a Surveillance Requirement) are identified in a timely manner. This verification is an implicit part of using Technical Specifications and determining the appropriate Conditions to enter and Required Actions to take in the event of a failure to meet a Surveillance Requirement. In addition, plant status is continuously monitored by control room personnel. The results of this monitoring process ; are documented in records / logs maintained by control room personnel. The continuous monitoring process includes re-evaluating the status of compliance with Technical Specification requirements when the plant conditions change (in this case BNP UNITS 1 & 2 8 - Revision 0 l l
DISCUSSION OF CHANGES ITS: 3.3.1.2 - SRM INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE L.10 compliance with the requirements of fuel movement sheets). (cont'd) Therefore, the explicit requirement to periodically verify that the fuel movement sheet is being followed during a core spiral unload and reload is considered to be unnecessary for ensuring ds compliance with the applicable Technical Specification requirements. L.11 CTS 3.3.5.4 requires three SRMs to be OPERABLE when in MODES 2, 3, and 4 and CTS 3.3.5.4 Action b requires actions to be performed in MODES 3 and 4 when two or more of the three SRMs are ino)erable. CTS 3.3.5.4 does not provide direction when one of the tiree SRMs is inoperable in these MODES. During startup in MODE 2, control rods are capable of being withdrawn or are being withdrawn to bring the reactor to a critical state. Therefore, three SRMs are required to be OPERABLE to monitor the reactor flux level prior to and during control rod withdrawal to ensure that the approach to criticality and the achievement of criticality occurs as expected. The three required SRMs ensure a representation of overall core response during periods when reactivity changes are occurring throughout the reactor core. In MODES 3 and 4, the reactor mode switch is in the shutdown position and, as a result, all control rods are inserted. In this condition, ITS 3.3.1.2 is revised to only require two SRMs to be OPERABLE. This reduction in the number of SRMs required to be OPERABLE is considered to be 8 O acceptable since the reactor is shutdown and reactivity changes that may result in criticality are not expected since the ITS 3.1.1, SHUTDOWN MARGIN (SDM), requirements must still be met. Should a reactivity change occur, redundant monitoring capability of flux levels of the reactor core will continue to be provided by the two required SRMs. In addition, ITS Table 1.1-1 requires the reactor mode switch to be in the shutdown position in MODES 3 and 4, which ensures that all control rods are inserted. A corresponding change to CTS 3.3.5.4 Action b is also made to require actions to be taken when one or more required SRMs are inoperable in MODES 3 and 4 (the same level of degradation addressed in CTS 3.3.5.4 Action b). L.12 In MODE 2 (with IRMs on Range 2 or below) and in MODES 2 and 3, CTS 4.3.5.4.a requires performance of a CHANNEL FUNCTIONAL TEST of the SRMs once within 24 hours prior to moving the mode switch from Shutdown (i.e., prior to entering MODE 2 from MODE 3 or 4) if not performed within the previous 7 days. The required periodic Frequency (every 31 days of ITS SR 3.3.1.2.6) has been determined to be sufficient verification that the SRMs are properly functioning. A review of historical test data for the past b 3 years shows there have been no functional failure of the SRMs. Moving the mode switch from Shutdown does not impact the ability of the SRMs to perform their required function. Additionally, the requirements of ITS 3.3.1.2 Required Action B.1 preclude beginning control rod withdrawal unless the required equipment is OPERABLE. O BNP UNITS 1 & 2 9 Revision 0
I
-DISCUSSION OF CHANGES l ITS: 3.3.1.2 - SRM INSTRUMENTATION l TECHNICAL CHANGES - LESS RESTRICTIVE L.12 Therefore, an additional Surveillance required to be performed (cont'd) "within 24 hours )rior to moving the mode switch from. Shutdown if not performed witiin the previous 7 days" is an extraneous and unnecessary performance of a Surveillance and is deleted. In b, addition, it is concluded that the impact, if any, on SRM availability is minimal as a result of this change.
RELOCATED SPECIFICATIONS None i BNP UNITS 1 & 2 10 Revision 0
,- =
/3 iJ A.i
.br:4<.4w 9 s.i 1
3.3.1 - j Table
~~
NMotinue) , 1 ACCIDEWT MONITORINC INSTRUMENTATION ACTIONS 1 1 i .s With the number of OPERABLE chainnels les, than required by the l
' minimum channels OPERABLE requiremeRLs /in late A e pc
- anaf Qg j (altynatemtho# of sp6itorig thq/appro latesfarame r(s)/ l k fa Agc i N- ' T L. ,
7' gQ [d either restore at astTlWaddip*AueverMI operable 1-. I channel (s) tio CPERABLE status within 7 days of the event, L, or 3To [2)__ prepare and submit a Special Report to the Consission pursuant to Specification 6.9.2 within the next 14 days f"T * p f
Aulod f pew sb i outlining the action taken, the cause of the inoperability
\ and the plans and schedule for restoring the system to
)L,f 4' OPERABLE status.
eA.# 4 % ris C.6.h
@ With the number of OPERABLE channels one less than the required number of chancels shown in Table 3.3.5.3-1, g AcncpJ A restore the inoperable eh=aael to OPr8 Anff armena withi rye 17 at teap HOT J,buiugim within/be pga _
t" h 50 ' *r %Js pa.p adbu" _b ) - h With the number of OPERABLE channels less than the minimum channels OPERABLE requirements of Table 3.3.5.3-1, restore ' A C '04 C- at least the .mintave number of operable channel (s) to fat.f T A m oas E
;- OPERABLE status within 7 days or be in at least. HOT SPUTDOWN
[withinthenext12 hours, f .g.s.P ! pcuou b l I d 4
)
l BaUNSWICE - UNIT 1 3/4 3-61 Amendment No. 130 \ N d b
* ~a .,;. .
.n ~
Ct EC4NA ~ 3. 3. *3, l l g] .( () . . - _. .. - .. I l coMNMENT/ SYSTEMS / ER $Y
/ / L h 'b kLIMITINC wanTION OP 10h Llo *5.% \ gas analystr.; systems for the dryvell and suppression o h ) Two [f pep / 4 with each systes' consisting of an oxygen analyser
{ .
*Y chamber shall es vr end a hydrogen analyser.
- APPLICABII.ITY: OPERATIONA!. CONDITION 1 ACTION:
p With one oxygen and/or one hydrogen analyzer inoperable,, restore at Acnou A s se two oxygen and two en anairzers to CPERABLE, status within j ifhi/the'nes( 8 . b6uts The L,1 sys or (e Ah ara =* . provisions of SpecificationT0.4 are not applicable. gg With no gas analyzer CPERABLE for oxygen and/or hydrogen, be in at (f) leaot @Tilrwi)A Acnaj 6 h e'*Pr.S'O ^ * *" c/4 p.pwp Aos. CS A,Q suuYEID AMCE REQUIREMENTS i Each gas analyser system shall be demonstrated OPERABLE at least once 9 9 I l' '2 - per 92 days by performing a CHANNEL CALIBRATION,415g stenaars gas sampis. I n a Ri' inal! I lume decent hy dgen, b 1 nce ni en. a Zero
- b. Se to a volume eccent drogen, alance ni ogen.
- c. vent Ive to t rty voi perce hydrogen balance a rosen.
cent es en, bal ce nitrog d Ze volume
- e. ven to e volume reent o en, bat a nitros .
Twenty t twenty- ive volume percent o on, balance nitro en. f. t - ta ,3 a[4 yqod se ns I.g . G 3/4 6-30 amendment so. no Q stuMSutcr - uwIT 1
%. c .no
.. ; . , ... . . +
- t. < i eg, .fi ,Jf 33*3.I )
~
~3 ~5.5.t - '
Table f.'3dft/Cear(ingsdI) IDENT MONITORINC INSTRUlGDITATION
*'^
ACTIONS
, l f With the number of OPERABt.E channels less than required by the minimum channels OPERABLE requirementefl ^ isse cas ag# t h y - - em ponspbrio ea to f
3s 4 - m-s v ey og
' "r 4D operable
, L. I ; h either rectore at least the_==+
{ channel (s) to "JPERABLE status within 7 days of the event,-
.. L or ,
1 l Ms. B ) prepare and submit a special Report to the Commiss !
"g p O. j 1r pursuant to specification 6.9.2 within the next 14 days i WH D grep, p outlining the action taken, the cause of the l inoperability,,and the plans and schedule for restoring / I A[6 the system to OPEF.ABLE status. j ym.a 4 n s. @_- I g, p With the number of OPERABLE channels one less than the required number of channels shown in Table 3.3.5.3-1, hcrieJ A -tore the inoperable channel to OPERABLE status within y atfeast grf SagrauGipf witpfn thg4e
= au ..,..., n.g h ith the number of OPERABLE channels less than the minimum channels OPERABLE requirements of
' Table 3.3.5.3-1, restore at least the minimum number of 83 # ff *** - operable channel (s) to OPERABLE status within 7 days or Acnouj g {beinatleastHOTSHUTDOWNwithinthenext12 hours.
3/4 3-61 Amendment No. 160 p BtuusWICK - UNIT 2 a, o w
.3-
.a . , ~ .:}K ;,::~ *-
y.n u.6.i (G) . . . - - . . _
,Y I lelENT 5%87 EMS, LYZ SYS S LLI TI
~
CC TIow ot OP TION Miyzer systems for the drywell and suppression
^
Two
' W '3.'5.3.; .
LE with each system consisting of an oxygen analyzer g,g., g c r shall and a hydrogen analyzer.
*- APPLICABILITY OPERATIONAL CONDITION 1 ACTION ht.rskJ 14 p With one oxygen and/or one hydrogen analyzer inoperable, restore atle
' * ' " " ML So days or m( Wa nV WASfUP/wiphinf fhe 9/zt g hpdrs) The [,
,5, u,4m i 4. 6Nes provisions o specification 3'.0.4 are not applicante. Acne b
Aerio.J 6 $ With aus analyzer OPERABI.E for orygen and/or hydrogen,/ e in at least traitTv vi,xtii.n's gu3 tvoN h f~f*"O b * *" SURVEILLANCE REQUIREMENTS M ._Each gas analyzer system shall be demonstrated OPERABLE at least once O Sg' s =,,.5.s.7 y r M1 days by performing a CHANNEL CALI singstandardgassamples'l (v . .I c aaK tra. .
' ' s. Zero vol percent drcgen, ance nitr n. A,q
- b. Se to ten une perc hydrogen alance nit n.
volume pe cut hydrogen alance ni gen. d *c. Twenty-f a to thir
- d. Zer volume per at oxygen alance nitro n.
- e. even to t percent a gen, balance trogen.
Twenty t twenty-fi percent oxy n, balan nitrogen [J. k
/A.S &
. c.es p,.,pos,p sr_ ' .u.i.9 ,
I e 3/4 6-30 Amendment No. 160 BRUNSWICK - UNIT 2 (
'~Q& loef to 3
, DISCUSSION OF CHANGES ITS: 3.3.3.1 - POST ACCIDENT MONITORING INSTRUMENTATION TECHNICAL CHANGES MORE RESTRICTIVE (continued)
M.4 The Reactor Vessel Water Level instrumentation consists of instruments with different ranges to satisfy RG 1.97 requirements. For the Reactor Vessel Water Level instrumentation, the different ranges are: a. -150 inches to +150 inches; b. 0 inches to
+210 inches; and c. +150 inches to +550 inches. Currently, CTS 3.3.5.3-1 only specifies requirements for two channels but does not specify the required ranges. Using the ITS fomat, these instruments are delineated in ITS Table 3.3.3.1-1 as separate line items under Function 2, with each channel now consisting of only one instrument. This change represents an additional restriction on plant operation.
M.5 CTS 4.6.6.4 requires CHANNEL CALIBRATIONS of the drywell and suppression chamber hydrogen and oxygen analyzers once per 92 days. CTS Table 4.3.5.3-1 requires CHANNEL CALIBRATIONS of the drywell oxygen concentration and drywell hydrogen concentration analyzer and monitor once per 18 months. These two CTS Functions are satisfied with the same instrumentation. As a result the CHANNEL CALIBRATION requirement (SR 3.3.3.1.2) for the drywell and suppression chamber hydrogen and oxygen analyzers is specified as once per 92 days. This change represents an additional restriction on plant operations and achieves consistency with vendor recommendations. M.6 CTS Table 3.3.5.3-1 ACTION 82 item a provides 31 days to restore \ inoperable accident monitoring instrumentation channels to OPERABLE status and CTS 3.3.6.4 ACTION a provides 31 days to restore inoperable hydrogen and oxygen analyzers to OPERABLE status. Under the same conditions, ITS 3.3.3.1 Required ! Action A.1 provides a Completion Time of 30 days for restoration of these post accident monitoring instruments. The reduction of b the time period allowed to restore inoperable post accident monitoring instrumentation channels represents an additional ; restriction on plant operation necessary to achieve consistency 1 with NUREG-1433, Revision 1. IECHNICAL CHANGES - LESS RESTRICTIVE
' Generic" LA.1 The use of alternate methods of monitoring in CTS Table 3.3.5.3-1 ACTION 81 are to be relocated to the Bases. These detatis are not necessary to be included in Technical Specifications to ensure actions are taken to initiate the preplanned alternate method of monitoring since ITS 3.3.3.1 Condition F requires action to be immediately initiated in accordance with ITS 5.6.6. ITS 5.6.6 requires a report to be submitted to the NRC within the following 14 days and that the report outline the preplanned alternate method of monitoring. As such, the relocated details are not O BNP UNITS 1 & 2 3 Revision 0
DISCUSSION OF CHANGES ITS: 3.3.3.1 POST ACCIDENT MONITORING INSTRUNENTATION TECHNICAL CHANGES - LESS RESTRICTIVE LA.1 required to be in Technical Specifications to provide. adequate (cont'd) protection of the public health and safety. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications.' LA.2 The details in CTS Table 4.3.5.3-1 Note (a) of the method for performing the Surveillance (in this case the CHANNEL CALIBRATION of the drywell area radiation monitors) are to be relocated to the Bases. These requirements are procedural details that are not necessary for assuring the OPERABILITY of the drywell area radiation monitors. The Surveillance Requirements of ITS 3.3.3.1 provide adequate assurance the drywell area radiation monitors are maintained OPERABLE. As such, the relocated details are not required to be in Technical Specifications to arovide adequate protection of the public health and safety. Caanges to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications. LA.3 The details relating to system design in CTS 3.6.6.4 (that tb2 two drywell and suppression chamber hydrogen and oxygen analyzu systems are independent) are to be relocated to the Bases. These are design details that are not necessary to be included in the Technical Specifications to ensure the OPERABILITY of the drywell and suppression chamber hydrogen and oxygen analyzers, since O OPERABILITY requirements are adequately addressed in BNP ITS 3.3.3.1. As a result, the relocated details are not required to be included in the BNP ITS to provide adequate protection of the pubite health and safety. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications. LA.4 The details in CTS 4.6.6.4 of the method for performing the Surveillance (in this case the CHANNEL CALIBRATION of the drywell and suppression chamber hydrogen and oxygen analyzers) are to be relocated to the Bases. These requirements are procedural details that are not necessary for assuring the OPERABILITY of the drywell and suppression chamber hydrogen and oxygen analyzers. The Surveillance Requirements of ITS 3.3.3.1 provide adequate assurance the drywell and suppression chamber hydrogen and oxygen analyzers are maintained OPERABLE. As such, the relocated details are not required to be in Technical Specifications to provide adequate protection of the public health and safety. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications. O BNP UNITS 1 & 2 4 Revision 0
q_ DISCUSSION OF CHANGES ITS: 3.3.3.1 - POST ACCIDENT MONITORING INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIE (continued) LE.1 CTS Table 4.3.5.3-1 establishes 18 months as the required Frequency for performance of CHANNEL CALIBRATION of the Post Accident Monitoring (PAN) instruments. 1TS SR 3.3.3.1.3 will extend the required frequency for these SRs to 24 months. Therefore, the surveillance test interval of these SRs is being increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25% grace period. The subject SR ensures that the PAM instruments will function as designed during an analyzed event. Extending the SR Frequency is acceptable because the PAM instruments are designed to be single failure proof and therefore, their function is highly reliable. Furthermore, a CHANNEL CHECK is performed on a more frequent basis (BNP ITS SR 3.3.3.1.1). The CHANNEL CHECK provides a qualitative demonstration of the OPERABILITY of the instrument. A separate drift evaluation has not been performed for the PAM Instruments based on the design of the PAM Instruments and equipment history. The following discussion supports this conclusion. The BNP PAM Instruments are designed with a high degree of 1 accuracy and reliability. The BNP Units 1 and 2 UFSAR Section 7.5 defines the design accuracy of the various monitoring O instrumentation indicators. ANSI /ANS Standard 4.5-1980, " Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors" endorsed by the NRC through Regulatory Guide 1.97 provides a recommendation for the design accuracy of these devices. A review and comparison of the design accuracy of the BilP PAM Instruments with the accuracy recommended for the same or similar devices in ANSI /ANSStandard 4.5-1980 indicates that the design accuracy of the BNP PAM Instruments are an order of magnitude greater than the ANSI recommended accuracy for the same or similar functions. Furthermore, it should be noted that these instruments are calibrated to the tighter tolerances of their design accuracy. Based on the fact that these instruments have a significantly greater accuracy than recommended in the industry standard and are calibrated to a tighter tolerance, it has been determined that a drift calculation for these instruments is not necessary and a review of the surveillance test history provides an acceptable
- method to determine if the instrument calibration frequency can be !
extended to a 24 month operating cycle. ' Based on the design of the instrumentation, it is concluded that the impact, if any, on system availability is minimal as a result of the change in the surveillance test interval. 1 O BNP UNITS 1 & 2' 5 Revision 0
h DISCUSSION OF CHANGES ITS: 3.3.3.1 - POST ACCIDENT MONITORING INSTRUMENTATION L TECHNICAL CHANGES - LESS RESTRICTIVE (continued) ! LE.1 A review of the surveillance test history was performed to ! (cont'd) validate the above conclusion. This review of the surveillance I test history, demonstrates that there are no failures that would invalidate the conclusion that the impact, if any, on system availability is minimal from a change to a 24 month operating cycle. i
" Specific" L.1 CTS Table 3.3.5.3-1 ACTION 81 is changed for one or-two drywell area radiation monitors inoperable. With one monitor inoperable, ITS 3.3.3.1 Required Action A.1 provides 30 days for the restoration of the monitor prior to initiating the alternate method of monitoring. With two monitors inoperable, ITS 3.3.3.1 Required Action C.1 provides 7 days for restoration of one monitor prior to initiating the alternate method of monttcring. With one or two monitors inoperable CTS Table 3.3.5.3-1 ACTION 81 requires initiation of the alternate method of monitoring within 72 hours and restoration of both channels to OPERABLE status within 7 days.
The Completion Times (30 days when one monitor is inoperable or /A 7 days when two monitors are inoperable) for restoration of one W channel or initiation of the alternate method of monitoring is considered acceptable based on the relatively low probability of an event requiring PAH instrumentation, the passive function of O- the instruments, the availability of the redundant monitor (for the condition of one monitor inoperable), and the availability of alternate means to obtain the information. L.2 CTS Table 3.3.5.3-1 ACTION 82, item a, for one channel inoperable l in one or more Functions for more than the allowable outage time j is revised from requiring a shutdown to requiring a Special Report 1 (ITS 3.3.3.1 Required Action B.1) in accoraance with the Administrative Control section of the Technical Specifications. Due to the passive function of these instruments and the operator's ability to respond to an accident utilizing alternate instruments and methods for monitoring, it is not appropriate to impose stringent shutdown requirements for out of service ! instrumentation. The change is considered acceptable since i another OPERABLE channel is monitoring the Function and the i probability of an event, requiring the operator to utilize this instrumentation to respond to the event, is low. This change is g , i consistent with the BWR Standard Technical Specifications, l NUREG-1433. L.3 CTS Table 4.3.5.3-1 requires a CHANNEL CHECK to be performed once-per day for the Suppression Chamber Water Temperature Function. For this Function, ITS SR 3.3.3.1.1 requires a CHANNEL CHECK to be performed only once per 31 days. Operating experience has shown that these instruments normally pass the Surveillance at the current frequency. The change is made to conform to NUREG-1433 i and is acceptable given the passive nature of these devices and BNP UNITS 1 & 2 6 Revision 0
DISCUSSION OF CHANGES ITS: 3.3.3.1 - POST ACCIDENT MONITORING INSTRUMENTATION .(D U TECHNICAL CHANGES - LESS RESTRICTIVE L.3 the fact that the most common outcome of the performance of a (cont'd) surveillance is demonstrating the acceptance criteria are satisfied. In addition, the instrumentation for the Suppression Chamber Water Temperature Function, required to be OPERABLE in i MODES 1 and 2 in accordance with ITS 3.3.3.1, is also used to satisfy ITS SR 3.6.2.1.1. In MODES 1, 2, and 3. ITS SR 3.6.2.1.1 requires the suppression pool temperature to be verified to be 1 within the applicable limit at a maximum Frequency of once per ! 24 hours. The use of these instruments, by the operators, to i monitor suppression chamber water temperature at least once per l 24 hours will allow failure of this instrumentation to be ) identified. 'ITS SR 3.0.1 states that SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. ITS SR 3.0.1 n also states that failure to meet a Surveillance even if E experienced between performances of the Surveillance, shall be failure to meet the LCO. Therefore, the requirement to perform a daily test to ensure that a gross failure of the Suppression Chamber Water Temperature instrumentation has not occurred is implicitly maintained through the requirements of ITS SR 3.6.2.1.1 and ITS SR 3.0.1. Therefore, the deletion of the explicit requirement to perform a CHANNEL CHECK, which will detect a gross failure, of the Suppression Chamber Water Temperature instrumentation is considered to be acceptable. L.4 CTS 3.6.6.4 ACTION b requires a shutdown when the two required drywell and suppression chamber hydrogen and oxygen analyzers are inoperable. ITS 3.3.3.1 ACTION C allows 7 days for restoration of the required analyzers when two analyzers are inoperable prior to requiring a shutdown. Due to the passive function (no required automatic action) of these instruments and the operator's ability to respond to an accident utilizing alternate instruments and methods for monitoring, it is not appropriate to impose stringent shutdown requirements for out of service instrumentation. The change to the allowed outage time is considered acceptable based on the passive nature of the instruments and the low probability of an event requiring PAM instrumentation during the intervals. l, RELOCATED SPECIFICATIONS l R.1 3/4.3.5.3 ACCIDENT MONITORING INSTRUMENTATION 4 LCO Statement: The accident monitoring instrumentation channels shown in l Table 3.3.5.3-1 shall be OPERABLE. ! O BNP UNITS 1 & 2 7 Revision 0
4 DISCUSSION OF CHANGES' ITS: 3.3.3.1 - POST ACCIDENT MONITORING INSTRUMENTATION i l b RELOCATED SPECIFICATIONS R.1 Discussion: i
'(cont'd).
Each individual accident monitoring parameter has a specific purpose; however, the general purpose for. accident monitoring instrumentation is to provide sufficient information to confirm an ) accident is proceeding per prediction, i.e. automatic, safety systems are performing properly, and deviations from expected accident course are minimal. Comparison to Deterministic Screenina Criteria: r The NRC position on application of the deterministic screening i~ criteria to post-accident monitoring instrumentation is documented
.in letter dated May 7,1988 from T.E. Murley (NRC) to R.F._ Janecek
.(BWROG). The position was that the post-accident monitoring instrumentation table list should contain, on a plant specific n basis, all Regulatory Guide 1.97 Type A instruments spcified in the plant's Safety Evaluation Report'(SER) on Regulatory ,
Guide 1.97, and all Regulatory Guide 1.97 Category 1 instruments. 1 Accordingly, this position has been applied to the BNP Regulatory Guide 1.97 instruments. Those instruments meeting these criteria have remained in Technical S)ecifications. The instruments not meeting these criteria have 3een relocated from the Technical Specifications to plant controlled documents. O V The following summarizes the BNP position for those instruments currently in Technical Specifications. From NRC SER dated 5/14/85,
Subject:
Emergency-Response Capability: Conformance to R.G. 1.97, l Revision 2. l l TvDe A Variables I
- 1. Reactor Pressure Vessel (RPV) Pressure
- 2. RPV Water Level l Suppression Pool Water Temperature 3.
- 4. Suppression Pool Water Level
- 5. Drywell Pressure
- 6. Drywell Temperature l
- 7. Suppression Pool Pressure ~i
- 8. Drywell and Suppression Pool Hydrogen and Oxygen Concentration I
from R.G.1.97 and CP&l. submittal to the NRC dated 2/01/84,
" Emergency Response Capability, Regulatory Guide 1.97, ;
Revision l'. ' Other Type. Cateaory 1 Variables Primary Containment Area Radiation - High Range (Drywell l AreaRadiationMonitors) BNP UNITS 1 & 2 8 Revision 0 ,
DISCUSSION OF CHANGES ITS: 3.3.3.1 - POST ACCIDENT MONITORING INSTRUMENTATION RELOCATED SPECIFICATIONS R.1 For other post-accident monitoring instrumentation currently in (cont'd) Technical Specifications, their loss is not risk-significant since the variable they monitored did not qualify as a Type A or Category I variable (one that is important to safety and needed by j the operator, so that the operator can perform necessary normal actions). Conclusion Since the screening criteria have not been satisfied for non-Regulatory Guide 1.97 Type A or Category 1 variable instruments, their associated LC0 and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications (a Technical Requirements Manual). The i instruments to be relocated are as follows: l
- 1. Suppression Chamber Atmosphere Temperature
- 2. Drywell Radiation (Airborne radiation monitors)
- 3. Safety / Relief Valve Position Indication
- 4. Turbine Building Ventilation Monitor
- 5. Offgas Stack Ventilation Monitor 4
I I O I O BNP UNITS 1 & 2 9 Revision 0
i C$ C C A 3,], J, Z (m) w A. \ . 3,'3 STRUMENTATION
$RUTD@fh HMITOUlic INSdtUMQftATIQN MI WC WDIT N OPr ION The remote shutdown sopitoring instrumentation channels va a t.co 3 5 3.2 I)shall be OPERABLEfwity reedoutsAtisytaye#' enstra t h/ U-l
- APPLICABILITY OPERATIONAL CONDITIONS 1 I l ACTIOWt p p,p,[ [Cr10eas Nok ,
@ With the number of OPERABLE remotershutdown monitoring channels .less than ACTWJ A the requirements (f/Ts54V5/@, either restore the inoperable channel to CPERABLE status with' et1M) days or be in at least HOT SHUTDOWN within ACTM3 B the ours ylfr wynuayaWN wipa thyIollying 2r'hpr{
y,/c / /. @ The av ons of Specifica' tion 3.0.4 are not applicable. Actious SURVEI1f HCE REQUIREMENTS O C Each o'f the above required remote shutdown monitoring instrumentation irs channels shall be demonstrated OPERABLE by performance of the CHANNEL CHE and CHANNEL CALIBRATION operations at the frequenciesfstjewp I,a 3.'532I j92 G& M391-Y-e c ueL kh
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A stlo n)$ . suRvtit.t.ANCE REQUIREMENT 5 Q Each of the above required remote shutdown monitoring EL CHECKp' instrumentatio g5 @hannelsshallbedemonstratedOPERABLEbyperformanceofthe c vg Tpb and GANNEL CALI5fATION ope' rations at the frequencies a - (f.
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DISCUSSION OF CHANGES ITS: 3.3.3.2 - REMOTE SHUTDOWN MONITORING INSTRUMENTATION ADMINISTRATIVE
-A.1- In the conversion of the Brunswick Nuclear Plant (BNP) current Technical Specifications (CTS) to the plant specific Improved .
Technical Specifications (ITS), certain wording preferences or I conventions are adopted which do.not result in technical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the ITS ; consistent with the Boiling Water Reactor Standard Technical ! Specifications, NUREG-1433, Rev. 1. ; A.2 The change to CTS 3.3.5.2 ACTION provides more explicit instructions for proper application of the ACTIONS for Technical Specification compliance. In conjunction with ITS 1.3,
" Completion Times," Note 2 to the ITS 3.3.3.2 ACTIONS (" Separate Condition entry is allowed for each....") and the wording for ITS 3.3.3.2 ACTION A ("one or more required Functions") provides direction consistent with the intent of the existing ACTION for an inoperable remote shutdown monitoring' instrumentation channel.
Since this change only provides more explicit direction of the i current interpretation of the existing specifications, this change ' is considered administrative. A.3 CTS Table 4.3.5.2-1 requires a CHANNEL CHECK for the remote shutdown monitoring instrumentation. Some instrumentation associated with CTS Table 4.3.5.2-1 Functional Unit 2, Reactor Yessel Water Level, are deenergized during normal operation. No specific acceptance criteria would apply to the CHANNEL CHECK (since the instruments would not be indicatir.9). Therefore, this Surveillance Requirement in ITS SR 3.3.3.2.1 is modified to i exclude the CHANNEL CHECK requirement on these deenergized j channels. This change is considered administrative (since the l channels are normally deenergized and any CHANNEL CHECK requirement would be essentially equivalent to no requirement). TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS 3.3.5.2 ACTION a provides 31 days to restore inoperable remote shutdown monitoring instrumentation channels to OPERABLE status. Under the same conditions, ITS 3.3.3.2 Required Action A.1 provides a Completion Time of 30 days for restoration of these remote shutdown monitoring instruments. The reduction of the time period allowed to restore inoperable remote shutdown monitoring b instrumentation channels represents an additional restriction on plant operation necessary to achieve consistency with NUREG-1433, Revision 1. O BNP UNITS 1 & 2 1 Revision 0
DISCUSSION OF CHANGES ITS: 3.3.3.2 - REMOTE SHUTDOWN MONITORING INSTRUNENTATION TECHNICAL CHANGES - LESS RESTRICTIVE
" Generic" LA.1 Details relating to system design and operation (specific instrument listings, the location of the instruments and associated readouts, and which portion of the Function is excluded from the CHANNEL CHECK since the design does not include displayed indication) in CTS 3/4.3.5.2 are to be relocated to the Bases.
These are design details that are not necessary to be included in the Technical Specifications to ensure the OPERABILITY of the remote shutdown monitoring instrumentation, since OPERABILITY requirements are adequately addressed in ITS 3.3.3.2. As a result, the relocated details are not required to be included in the BNP ITS to provide adequate protection of the public health and safety. Changes to the Bases will be controlled by the provisions of the Bases Control Program described in Chapter 5 of the ITS. LE.1 CTS Table 4.3.5.2-1 establishes 92 days and 18 months as the required frequency for performance of CHANNEL CALIBRATION for Remote Shutdown Monitoring Instrumentation. ITS SR 3.3.3.2.2 will extend the required Frequency for these SRs to 24 months. Therefore, the survelliance test interval of these SRs is being increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25% grace period. The subject SR ensures that the Remote Shutdown Monitoring Instrumentation will function as designed during an analyzed event. Extending the SR Frequency is acceptable because the instrumentation and systems are designed to be highly reliable. A drift evaluation has not been performed on the Remote Shutdown Instrumentation because of the design and reliability of the instrumentation. The purpose of this monitoring instrumentation is to provide a means to monitor the s&fe shutdown the reactor following an event where operation from the control room cannot be accomplished. Several of the instruments which are being extended are only energized when the Remote Shutdown Panel is placed in service. For these instruments, the change in the CHANNEL CALIBRATION Frequency will have no impact. This conclusion is based on the fact that in the deenergized state there are no time dependent factors which would cause a change in the input / output relationship of the deenergized device. For those instruments which remain energized, a CHANNEL CHECK is required by BNP ITS. This CHANNEL CHECK provides an effective means to demonstrate the OPERABILITY of the monitoring instrumentation used at the Remote Shutdown Panel during the operating cycle. Any gross failure or excessive drift of these instruments would be detected by this surveillance test. Therefore, based on the above discussion, it ,O .BNP UNITS 1 & 2 [ 2 Revision 0
DISCUSSION OF CHANGES ITS: 3.3.3.2 - REMOTE SHUTDOWN MONITORING INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE LE.1 has been determined that a drift calculation for these instruments (cont'd) is not necessary and a review of the surveillance test history provides an acceptable method to determine if the instrument calibration frequency can be extended to a 24 month operating cycle. Based on the design of the instrumentation and the drift evaluations, it is concluded that the im>act, if any, on system availability is minimal as a result of tie change in the surveillance test interval. A review of the surveillance test history was performed to validate the above conclusion. This review of the surveillance test history demonstrates that there are no failures that would invalidate the conclusion that the impact, if any, on system availability is minimal from a change to a 24 month operating ~ cycle. '
" Specific" L.1 CTS 3.3.5.2 requires the remote shutdown monitoring instrumentation to be OPERABLE in Operational Conditions 1, 2, and 3. ITS 3.3.3.2 requires the remote shutdown monitoring O instrumentation to be OPERABLE in MODES I and 2. The remote shutdown monitoring instrumentation is provided to sup) ort ) lacing and maintaining the plant in MODE 3 from a location otier tian the control room. In MODE 3 the plant is already subcritical and in a condition of reduced Reactor Coolant System energy. Under this condition, considerable time is available to restore the necessary instrument functions if the control room instruments and controls become unavailable and the probability of an event that would require evacuation of the control room is low. Therefore, the remote shutdown monitoring instrumentation is not required to be OPERABLE in MODE 3. A commensurate change to the shutdown requirements associated with placing the unit in a non-a)plicable MODE (in this case MODE 3 within 12 hours) when remote stutdown monitoring instrumentation is not restored within the required time period is also made in ITS 3.3.3.2 Required Action B.1. This change is consistent with NUREG-1433.
RELOCATED SPECIFICATIONS None O BNP UNITS 1 & 2 3 Revision 0
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a ektdio-s UA. l bI ha-) c.&,Q (j] A L- N I TABLE- J M (Continued) 1 L.L.x Sol.ATION ACTUATION INSTRUMENTATION ACTIONS L. b , ( 2. Actrotis "I) wI b ACTION 20 - e in at least HOT SHLrJDOW withi fr' ours and in COLD SHUTDOW within the following# hou s . -
- . 14- L. .
MOWD 0 e% atAeas/STJdTUfwi che in steam line isolation L4 1 within closed withMhours valves or be in at least HOT SHUTDOW in COLD SHUTDOW thin the next hours. G A CTlaag ACTION 22 - Be in at WP withi hours. FI L*5 ACTION 23 - In OPERATIONAL CONDITIONS 1, 2, or 3, establish SECONDARY CONTAINMENT INTECRITY with the standby gas treatment system opersting within one hour. ., Ire,* 3,h, In OPERATIONAL CONDITION 5 or when handling irradiated fuel in the secondary containmentt , I
!) Establish SECONDARY CONTAINMENT INTEGRITY with the standby' f gas treatment system operating within one hcurt j 1
- 2) Otherwise, suspend handling of irradiated fuel in the l f~T 4 secondary containment, CORE ALTERATICES, or activities that t
d ggg could reduce the SHUTDOW MARCIN. F W '4 Actron 2' - tsota.c. th d.' etar - c r et aua =
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. p ,og ACTION 25 - Close the af fected s]rstem isolatic Iv gy ecpr pa p dTfsi:ted'sysr'em isbper6ble/_ - < ' N JdN 2[ - Ve[fy pode availadlity todhe bus at east e pe r hou . .!
ON 27 ,pptt hrice shutdown cooling supply Ent reattor bessel'%ea ] gUgg @ cay stolatwn vaWn tne closee position /uht the rpa p" M.4 pp djunega U 'Viadhe' specified Almit 9 ACTION 28 - Close the affected isolation valves within de or be to NOT Af 4 ) F SHUTDOW within 12 hours and in COLD SHtrT within the following 24 hours. l.E L&[ 1.gw C Fwon L. BRUNSWICK - UNIT 1 3/4 3-17 Amendment No.148 A G P.g. P/ h
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SOLATION ACTUATION INSTRUMENTATION ACTIONS l 7, dCT1c43$ 'Be in.at'Least HOT SH withi A" ours and in COLD SHUTDOWN g gg ACTION 20 within the following ours. L,f ACT 21 - p/ip/atjiesst,/$TA47UJ with the sa n stes's line isolation valves l Mou D .UTDOWN within y' I 2-Q __jgclosee withMhours or be in at least HOT ours. hours and in COLD SHUTDOWN within the next gy L.5 ggg ACTION 22 - SeinatleastSTARTUPwithinhours. 6 ACTION 23 - In OPERATIONAL CONDITIONS 1, 2, or 3 establish SECONDARY CONTAINMENT INTECRITY with the standby gas treatment system operating within one hour. b' *- In OPbEATIONAL CONDITION 5 or when handling irradiated fuel in ,
-r g r3,33 ,the secondary containments
- 1) Establish SECONDARY CONTAINMENT INTECRITY with the standby gas treatment system operating within one hourl }
Otherwise, suspend handling of irradiated fuel in the l
- 2) i secondary containment, CORE ALTERATIONS, or activities tha l could reduce the SHUTDOWW MARCIN.
( the reactor water cleanup system
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3/4 3-17 Amendment No. 178 SRUNSWICK - UNIT 2 V i TL, u .c %
DISCUSSION OF CHANGES ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION i v ADMINISTRATIVE A.13 HPCI Steam Line Area Delta Temperature-High (CTS Function 4.a.8) (cont'd) revised to HPCI Steam Line Tunnel Differential Temperature-High (ITS Function 3.h); RCIC Steam Line Tunnel Temperature-High (CTS Function 4.b.4) revised to RCIC Steam Line Area Temperature-High (ITS Function 4.f); RCIC Steam Line Ambient Temperature-High (CTS Function 4.b.7) revised to RCIC Steam Line Tunnel Ambient Temperature-High (ITS Function 4.g); RCIC Steam Line Area Delta Temperature-High (CTS Function 4.b.8) revised to RCIC Steam Line Tunnel Differential Temperature-High (ITS Function 4.1); RCIC Equipment Room Ambient Temperature-High (CTS Function 4.b.9) revised to RCIC Equipment Area Ambient Temperature-High (ITS Function 4.j); RCIC Equipment Room Delta Temperature-High (CTS Function 4.b.10) revised to RCIC Equipment Area Differential Temperature-High (ITS Function 4.k); and OU RCIC Steam Line Tunnel Temperature-High Time Delay Relay (CTS Function 4.b.ll) revised to RCIC Steam Line Tunnel and Area Temperature-High Time Delay (ITS Function 4.h). The design and location of the actual instrumentation (including the sensors that monitor the affected areas) is unchanged. Therefore, this change is considered administrative. TECHNICAL CHANGES - MORE RESTRICTIVE M.1 The Applicability for the Reactor Vessel Water Level-Low Level 1 Function (CTS Table 3.3.2-1, Functions 1.a.1 (valve group 8) and C 5.a; ITS Table 3.3.6.1-1 Function 6.b) is changed to include d MODES 4 and 5. These new Applicabilities will protect against potential draining of the reactor vessel through the RHR suction line during shutdown conditions, which is when the RHR Shutdown Cooling System is normally operated. Appropriate ACTIONS (ITS 3.3.6.1 Required Actions 1.1 and I.2) are also added for when the Function is inoperable in MODES 4 and 5. This is an additional restriction on plant operations and is consistent with the BWR Standard Technical Specifications, NUREG-1433. M.2 The name of the Main Steam Line Tunnel Temperature-High Function (CTS Table 3.3.2-1 Function 1.d) is revised to reflect the location of the specific main steam line temperature monitoring instruments that were credited in the high energy line break [s analysis for Function 1.e in ITS Table 3.3.6.1-1. The design of BNP UNITS 1 & 2 5 Revision 0
DISCUSSION OF CHANGES ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE LA.4- - primary containment isolation instruments are maintained OPERABLE. (cont'd) As such, the relocated details are not required to be in Technical e Specifications to provide adequate protection of the >ublic health and safety. Changes to the Bases will be controlled >y the provisions.of the Bases Control Program described in Chapter 5 of the Technical Specifications. LA.5 The additional requirement for bypassing the Condenser Vacuum-Low Function (with reactor steam pressure < 500 asig) for Function 1.e of CTS Table 3.3.2-1 is to be relocated to tte UFSAR. The mode switch position contacts and the turbine stop valve position % contacts are arranged in a series logic scheme that is parallel to the Condenser Vacuum-Low instrumentation. Per this logic scheme, the Condenser Vacuum-Low Function is automatically unbypassed when either the reactor mode switch is placed in the run position b or any one of the turbine stop valves is not closed regardless of reactor steam dome pressure. In ITS the Condenser Vacuum-Low Function (ITS Table 3.3.6.1-1 Function 1.d) is required to be OPERABLE in MODE I and also in MODES 2 and 3 when any turbine stop d valve is not closed. As such, the Condenser Vacuum-Low Function I is automatically enabled at the point when the function is required to be OPERABLE and cannot be otherwise bypassed without a plant modification. Therefore, the ITS Applicability of the Condenser Vacuum-Low Function is adequate to ensure the function b O is not bypassed when required and the relocated requirement does not need to be included in the Technical Specifications to ensure the Condenser Vacuum-Low Function is not bypassed during a MODE or specified condition when it is required to be OPERABLE or to provide adequate protection of the public health and safety. Changes to the UFSAR are controlled by the provisions of 10 CFR 50.59. g LA.6 Not used. b O BNP UNITS 1 & 2 9 Revision 0
DISCUSSION OF CHANGES ITS:'3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE L.10 the'RWCU System. The resulting time period allowed for continued (cont'd) - operation by this change is acceptable due to the low probability of an event requiring the SLC System Initiation isolation during l- this time period when the capability to isolate the RWCU System is lost. L.11 CTSTable3.3.2-1 Action 27requiresdeactivatingtheaffeded system isolation valves closed when CTS Table 3.3.2-1 Shutdown Cooling System Isolation Trip Functions 1.a.1 and 5.a'(Reactor Vessel Water Level-Low, Level 1) or Trip Function 5.b (Reactor Steam Dome Pressure-High) are inoperable. ITS 3.3.6.1 Required Action F.1 only requires closure of the affected valve; deactivating the affected valve is not required. The requirement to deactivate the affected valve is an additional administrative requirement to assist.in ensuring the affected valve remains closed. ITS LCO 3.0.2 states that when an LCO is not met, the Required Actions must be met. Thus, when the valve is closed (to isolate the' affected penetration flow path), the valve must remain g closed to comply with the Required Action. In addition, inadvertent movement of a closed valve is an unlikely occurrence since plant administrative controls are in place to govern operation of these valves. Plant personnel would only operate a closed valve using a plant procedure, and these procedures are controlled by the requirements of ITS 5.4.1.a. Therefore, these procedures will also hel) ensure a closed valve is not O inadvertently opened. T1e ITS ACTION is also consistent with CTS Table 3.3.2-1 Actions 24, 25, and 28 which require affected I
. isolation valves to only be closed, not deactivated closed, when the associated Trip Functions are inoperable. Therefore, the requirement to " deactivate" the affected system isolation valves is removed from the Technical Specifications.
L.12 CTS Table 3.3.2-1 Action 27 requires deactivating the reactor vessel head spray isolation valves closed when CTS Table 3.3.2-1 Shutdown Cooling System Isolation frip Functions 1.a.1 and 5.a (Reactor Vessel Water Level-Low, level 1) or Trip Function 5.b (Reactor Steam Dome Pressure-High) are inoperable. ITS 3.3.6.1 Required Action F.2 does not require closure or deactivation of the reactor vessel head spray isolation valves. The reactor vessel head spray mode of the Residual Heat Removal System has been deactivated. As described in UFSAR Table 6.2.4-1, the reactor vessel head spray line has been abandoned in place in both units. The primary containment penetrations associated with these d lines, in both units, are isolated using welded pipe caps. With the reactor vessel head spray line isolated, the intended functiori of the affected isolation instrumentation is: satisfied since the function of the affected instrumentation is to isolate the reactor i vessel head spray line penetration. Unisolating the affected ; l' penetrations (i.e., removal of the welded pipe caps) requires a j plant modification. As a result, the function of the instrumentation associated with the closure of the reactor vessel BNP UNITS 1 & 2 29 Revision 0 1 } l
l l l DISCUSSION OF CHANGES ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL' CHANGES - LESS RESTRICTIVE L.12 head spray isolation valves is not needed to ensure the associated 1 (cont'd) penetrations are isolated. Therefore, the requirement to I deactivate the reactor vessel head spray isolation valves closed I when CTS Table 3.3.2-1 Shutdown Cooling System Isolation Trip A fcnctions 1.a.1 and 5.a (Reactor Vessel Water Level-Low, Level 1) m , or Trip Function 5.b (Reactor Steam Dome Pressure-High) are inoperable is deleted. RELOCATED SPECIFICATIONS R.1 TABLE 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION LCO Statement: The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2. 3/4.3.2.4.a.5 HPCI Bus Power Monitor 3/4.3.2.4.b.5 RCIC Bus Power Monitor Discussion: The Bus Power Monitors for HPCI and RCIC trip systems alarm if a fault is detected in the power system to the appropriate system's logic. No design basis accident (DBA) or transient analyses take. credit for the Bus Power Monitors. This instrumentation provides a motitoring/ alarm function only. Comparison to Screenino Criteria:
- 1. The HPCI and RCIC Bus Power Monitors are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The HPCI and RCIC Bus Power Monitors are not process 1 variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. j
- 3. The HPCI and RCIC Bus Power Monitors are not part of the ,
primary success path that function or actuate to mitigate a ; DBA or transient that either assumes the failure of or i presents a challenge to the integrity of a fission product ; barrier. j O BNP UNITS 1 & 2 30 Revision 0
l i l' DISCUSSION OF CHANGES ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION R& 0CATED SPECIFICATIONS R.1 4. As discussed in Sections 3.5 and 6 of NED0-31466 and (cont'd) summarized in Table 4-1 (item 106) of NED0-31466, Supplement 1, the loss of the HPCI and RCIC Bus Power Monitorr was found to be a non-significant risk contributor ' to core damage frequency and offsite releases. CP&L has , reviewed this evaluation, considers it applicable to BNP, { and concurs with the assessment.
Conclusion:
Since the screening criteria have not been satisfied, the ! Isolation Actuation Instrumentation LCO and Surve111ances associated with the HPCI and RCIC Bus Power Monitors may be relocated to other plant controlled documents outside the Technical Specifications (a Technical Requirements Manual). R.2 TABLE 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION LC0 Statement: The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set i consistent with the values shown in the Trip Setpoint column of i Table 3.3.2-2. ] i 3/4.3.2.1.d Main' Steam Line Tunnel Temperature - High ; 3/4.3.2.1.f Turbine Building Area Temperature - High Discussion: ! The Turbine Building Area Temperature Function and instruments of the Main Steam Line Tunnel Temperature Function that sense main steam line temperature outside the main steam isolation valve (MSIV) pit are provided to detect and initiate an MSIV isolation. l However, these primary containment isolation instruments constitute only one method of determining steam leakage in their respective areas. In addition to temperature monitoring, excess coolant inventory loss is detected by the reactor vessel low water level functions and main steam line high flow functions which continue to be required by Technical Specifications. The turbine butiding area temperature instrumentation is not assumed to mitigate any accident described in the BNP UFSAR and the only main steam line tunnel temperature instruments assumed in the mitigation of analyzed events are the MSIV instruments that sense temperature in the MSIV pit. O BNP UNITS 1 & 2 31 Revision 0 , J
DISCUSSION OF CHANGES ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION l
- V RELOCATED SPECIFICATION 1 R.2 Comparison to Screenina Criteria
'(cont'd)
- 1. The Main Steam Line Tunnel Temperature and Turbine Building l Area Temperature Functions are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Main Steam Line Tunnel Temperature and Turbine Building ,
Area Temperature Instrumentation are not process variables ' that are initial conditions of a DBA cc transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. l
- 3. The Turbine Building Area Temperature Instrumentation is not part of the primary success path that function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The main steam line tunnel temperature (
instruments that sense main steam line temperature outside l the MSIV pit are not assumed as part of the primary success i path that function or actuate to mitigate a DBA or transient ' that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Credit is not taken in pressure-temperature analyses, radiation dose O calculations, or equipment qualification for the operation
'V of the Turbine Building Area Temperature Function or instruments of the Main Steam Line Tunnel Temperature Function that sense main steam line temperature outside the MSIV pit. In addition, adequate redundancy is available by other TS required instruments to detect a steam leak in the main steam tunnel or the turbine building and perform the associated primary containment isolation function.
- 4. As discussed in Appendix B of the Application of Selection 4 Criteria to the BNP Technical Specifications, CP&L found the requirements of the Turbine Building Area Temperature function and instruments of the Main Steam Line Tunnel Temperature Function that sense main steam line temperature j outside the MSIV pit not being met to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the l Isolation Actuation Instrumentation LCO and Surveillances l associated with the Turbine Building Area Temperature Function and , instruments of the Main Steam Line Tunnel Temperature Function i that sense main steam line temperature outside the MSIV pit may be relocated to other plant controlled documents outside the Technical Specifications (a Technical Requirements Manual). K BNP UNITS 1 & 2 32 Revision 0 l l l
DISCUSSION OF CHANGES ITS: 3.3.6.2 - SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE A.1 In the conversion of the Brunswick Nuclear Plant (BNP) current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain wording preferences or conventions are adopted which do not result in technical changes (either actual .or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the ITS consistent with the Bolling Water Reactor Standard Technical Specifications, NUREG-1433, Rev. 1. A.2 The change to the CTS 3.3.2 ACTIONS provides more explicit instructions for proper application of the ACTIONS for Technical Specification compliance. In conjunction with the ITS 1.3,
" Completion Times," the ITS 3.3.6.2 ACTIONS Note (" Separate Condition entry is allowed for each....") and the wording for l ITS 3.3.6.2 ACTION B ("One or more functions") provides direction consistent with the intent of the CTS 3.3.2 ACTION for an inoperable isolation instrumentation channel. Since this change only provides more explicit direction of the current interpretation of the existing specifications, this change is considered administrative.
A.3 In MODE 1, 2, or 3. CTS Table 3.3.2-1 ACTION 23 requires Secondary Containment Integrity to be estabitshed with the standby gas hs j treatment system operating within I hour. This change to CTS Table 3.3.2-1 ACTION 23 replaces the use of the defined term s SECONDARY CONTAINMENT INTEGRITY with the elements of that definition and clarifies the need to start the associated SGT subsystem (s). 'ITS 3.3.6.2 Required Action C.l.1 requires the associated penetration flow path to be isolated within I hour and ITS 3.3.6.2 Required Action C.2.1 requires the associated standby gas treatment subsystem (s) to be placed in operation within I hour. These requirements are the same as those required by CTS , Table 3.3.2-1 ACTION 23 since the CTS attributes of Secondary i Containment Integrity are addressed in ITS 3.6.4.1, Secondary Containment OPERABILITY, ITS 3.6.4.2, Secondary Containment Isolation Dampers, and ITS 3.6.4.3, Standby Gas Treatment System. The Applicabilities of ITS 3.6.4.1, ITS 3.6.4.2, and ITS 3.6.4.3 l are the same as the Applicabilities of the Secondary Containment n Isolation Instrumentation. Whenever the ACTIONS of ITS 3.3.6.2 CO are applicable, the requirements of ITS 3.6.4.1, ITS 3.6.4.2, and ITS 3.6.4.3 are also applicable; i.e., secondary containment integrity must be already established, including requirements for doors to be closed and penetration seals to be in place or the applicable actions must be taken. As such, the CTS Table 3.3.2-1 ACTION 23 requirements associated with establishing secondary containment integrity only require actions be taken to ensure the automatic functions of the inoperable instrumentation (i.e., isolation of the associated secondary containment penetration flow path (s) and starting of the associated standby gas treatment subsystem (s)) are satisfied since the non-automatic functions must already be satisfied in accordance with the secondary containment
~
BNP UNITS 1 & 2 1 Revision 0
. DISCUSSION OF CHANGES ITS: 3.3.6.2 - SECONDARY CONTAllMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE A.3 Technical Specifications. Therefore, the change from CTS (cont'd) . Table 3.3.2-1 ACTION 23 to ITS 3.3.6.1 Required Actions C.1.1 and C.2.1 is administrative. Refer also to the Discussion of b
Changes associated with the Definitions Section in ITS Chapter 1.0 which addresses deletion of the Secondary Containment Integrity
-definition.
A.4 For Secondary Containment Isolation Instrumentation Functions with trip units, CTS Table 4.3.2-1 specifies Surveillance Requirements for the transmitter and the trip logic. CTS Table 4.3.2-1 Note (a) states that the transmitter channel check is satisfied by the trip unit channel check. A separate transmitter check is not required. The purpose of an instrument CHANNEL CHECK per the CTS definition and the ITS <lefinition is to essentially verify gross error in an instrument channel by observation. Observation of control room indication, output from the associated channels, would detect a gross error in any part of the instrument channel and satisfies the CTS and ITS definition of a CHANNEL CHECK. ITS Table 3.3.6.2-1 specifies the Secondary Containment _ Isolation Instrumentation functions, but does not distinguish'between the transmitter and the trip unit. Therefore, a statement that specifies that the transmitter CHANNEL CHECK is satisfied by the trip unit CHANNEL CHECK is not necessary. Additionally, the ITS Surveillances are considered adequate to verify OPERABILITY of the O secondary containment isolation instrumentation transmitters. As such, the deletion of the note is considered administrative in nature. A.5 Note (b) of CTS Table 4.3.2-1 exempts secondary containment isolation instrumentation transmitters from quarterly channel calibration which are required for the secondary containment i isolation instrumentation trip units. Since these trip units do not include the secondary containment isolation instrumentation , transmitters, Note (b) is unnecessary and is deleted from ' ITS 3.3.6.2. The requirement to calibrate the trip units is included in ITS SR 3.3.6.2.3. This change is considered j administrative in nature. 1 1 A.6 The technical content of CTS 4.6.6.1.d.2 is divided into two ; Surveillances. The majority of this Surveillance is performed as ITS SR 3.3.6.2.5, a LOGIC SYSTEM FUNCTIONAL TEST (LSFT). The LSFT i verifies that each signal functions properly. The actual system functional test portion is performed in LCO 3.6.4.3 Surveillance Requirements. This will ensure that the entire system is tested l with proper overlap. Therefore, since actual requirements are not 1 changed, this change is considered administrative. A.7 Note (g) to CTS Table 4.3.2-1 (which identifies which functions do not require response time testing) is not included in ITS 3 Table 3.3.6.2-1. Response time testing is not required for any of /_A the functions in ITS Table 3.3.6.2-1 and ITS 3.3.6.2 does not include Surveillance Requirements for response time testing. Each BNP UNITS 1 & 2 2 Revision 0
DISCUSSION OF CHANGES ITS: 3.3.6.2 - SECONDARY CONTAINMENT ISOLATION INSTRUNENTATION
.I ADMINISTRATIVE A.7 of the functions in CTS Table 4.3.2-1, included in ITS (cont'd) Table 3.3.6.2-1, do not currently require response time testing to /A be performed. Therefore, the deletion of Note (g) to CTS E Table 4.3.2-1 is considered to be an administrative presentation preference.
1 TECHNICAL CHANGES - MORE RESTRICTIVE None TECHNICAL CHANGES - LESS RESTRICTIVE
" Generic" LA.I CTS 3/4.3.2 includes trip setpoints for the associated instrumentation. Trip setpoints are an operational detail that is not directly related to the OPERABILITY of the instrumentation.
i These details are to be relocated to the Technical Requirements Manual (TRM). The Allowable Value is the required limitation for g i the parameter and this value is retained in the Technical ! Specifications. As such, trip setpoints are not required to be in ! Technical Specifications to )rovide adequate protection of the i O public health and safety. 01anges to the relocated trip setpoints in the TRM will be controlled by the provisions of 10 CFR 50.59. ^ 1 l LA.2 The details in CTS 4.3.2.2 relating to methods for performing the LOGIC SYSTEM FUNCTIONAL TEST are to be relocated to the Bases. j These details are not necessary to ensure the OPERABILITY of the i secondary containment isolation instrumentation. The requirements j of ITS 3.3.6.2 and the associated Surveillance Requirements are adequate to ensure the secondary containment isolation instruments are maintained OPERABLE. As such, the relocated details are not required to be in Technical Specifications to provide adequate l protection of the pubile health and safety. Changes to the Bases ; will be controlled by the provisions of the Bases Control Program described in Chapter 5 of the Technical Specifications. LA.3 Details of the methods for performing Required Actions in the "*" l Note to CTS 3.3.2 ACTIONS (which trip system to trip) are to be relocated to the Bases. These details represent operational considerations and are not required in the associated action to assure equipment is placed in a safe condition in the event a . secondary containment isolation instrument channel becomes A inoperable. As such, these details do not represent limits, d3 conditions for establishing equipment OPERABILITY, or remedial actions or instructions necessary to establish limits, conditions, or remedial actions. These details are not necessary to be included'in Technical Specifications to ensure actions are taken to restore isolation capability. The ACTIONS of ITS 3.3.6.2 are adequate to ensure action is taken to restore isolation capability BNP UNITS I & 2 3 Revision 0 l
l 1 DISCUSSION OF CHANCES ITS: 3.3.6.2 - SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION
/"'i j V TECHNICAL CHANGES - LESS RESTRICTIVE LA.3 (including tripping one of the affected trip systems). As such, (cont'd) the relocated details are not required to be in Technical ,
Specifications to provide adequate protection of the >ublic health ! and safety. Changes to the Bases will be controlled )y the i provisions of the Bases Control Program described in Chapter 5 of j the Technical Specifications. LA.4 System design and operational details in CTS Table 3.3.2-1 are to be relocated to the Bases. Details relating to system design and i operation (e.g., specific valves or valve groups operated by a ' specific signal) are unnecessary in the LCO. These details are not necessary to ensure the OPERABILITY of the secondary i containment isolation instrumentation. The requirements of ITS 3.3.6.2 and the associated Surveillance Requirements are adequate to ensure the secondary containment isolation instruments are maintained OPERABLE. As such, the relocated details are not required to be in Technical Specifications to provide adequate protection of the public health and safety. Changes to the Bases . will be controlled by the provisions of the Bases Control Program ! described in Chapter 5 of the Technical Specifications. For the Allowable Value associated with reactor vessel water level, CTS Table 3.3.2-2 Note (a) states that vessel water levels are referenced to REFERENCE LEVEL ZERO. This detail (the ( reference to REFERENCE LEVEL ZERO) is to be relocated to the N Bases. This reference is not necessary to be included in the . Technical Specifications to ensure the OPERABILITY of the i secondary containment isolation instrumentation. OPERABILITY requirements are adequately addressed in ITS 3.3.6.2 and the specified Allowable Values. As such, this relocated reference is not required to be in the Technical Specifications to provide adequate protection of the public health and safety. Changes to the ITS Bases are controlled by the provisions of the Bases Control Program described in Chapter 5 of the Technical Specifications. LD.1 CTS 4.3.2.2 specify the frequency for the Isolation Actuation Instrumentation LOGIC SYSTEM FUNCTIONAL TEST (LSFT) as once every 18 months. In ITS SR 3.3.6.2.5, the frequency for the LSFT is specified as once every 24 months. The surveillance test interval of this SR is being increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25% grace period. , This SR ensures that Secondary Containment Isolation Instrumentation logic will function as designed to ensure proper l response during an analyzed event. The secondary containment I isolation dampers and the Standby Gas Treatment System including the actuating logic are designed to be single failure proof and therefore, are highly reliable. Furthermore, as stated in the NRC O G BNP UNITS 1 & 2 4 Revision 0
1 DISCUSSION OF CHANGES L ITS: - 3.3.6.2 - SECONDARY CONTAINNENT ISOLATION INSTRUNENTATION TECifflCAL CHANGES - LESS RESTRICTIVE LD.1 Safety Evaluation Report (dated August 2,1993) relating to j (cont'd) extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3 surveillance intervals from 18 to 24 months:
. " Industry reliability studies for boiling water reactors (BWRs), prepared by the BWR Owners Group (NEDC-30936P) show i that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical-components, (e.g., pumps and valves),
which are consequently tested on a more frequent basis. Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, ! increasing the logic system functional test interval represents no significant change in the overall safety system unavailability." Based on the above discussion, the impact, if any, of this change on system availability is minimal. A review of the surveillance test history was performed to validate the above conclusion. This historical review of the surveillance test history demonstrates that there are no failures that would invalidate the conclusion that the impact, it any, on system availability is minimal from a change to a 24 month operating cycle. LE.1 CTS Table 4.3.2-1 establishes 18 months as the required Frequency for performance of CHANNEL CALIBRATION. ITS SR 3.3.6.2.4 will extend the required Frequency for these SRs to 24 months. i
, Therefore, the surveillance test interval of these SRs is being i increased from once every 18 months to once every 24 months for a maximum interval of 30 months including the 25% grace period.
The subject SR ensures that the Secondary Containment Isolation i Instrumentation will function as designed during an analyzed event. Extending the SR Frequency is acceptable because the i isolation initiation logic is designed to be single failure proof . and therefore is highly reliable. Furthermore, the impacted i isolation instrumentation has been evaluated based on make, ' manufacturer and model number to determine that the l instrumentation's actual drift falls within the assumed drift in the associated setpoint calculation. The following paragraphs, listed by ITS Table 3.3.6.2-1 function number, identify by make, l manufacturer and model number the drift evaluations performed: l l- Function 1, Reactor Vessel Water Level - Low Level 2 l l l This function is performed by a Rosemount 1153DB5 Transmitter and 71000 Trip Units. The Rosemount Trip Units are functionally checked and setpoint verified more frequently, and if necessary, , recalibrated. These more frequent testing requirements remain unchanged. Therefore, an increase in the surveillance interval to BNP UNITS 1 & 2 5 Revi:lon 0 0 4
e i i DISCUSSION OF CHANGES ITS: 3.3.6.2 - SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION 1[CHNICAL CHANGES - LESS RESTRICTIVE i LE.1 accommodate a 24 month fuel cycle does not affect the Rosemount l ' '(cont'd)- Trip Units with respect to drift. The Rosemount 1153DB5 Transmitter drift was evaluated using the GE methodology. The results of the analysis indicated that the projected 30 month drift values for the instruments do not exceed the drift allowance provided in the setpoint calculation for this instrument. ! I i Function 2, Drywell Pressure - High This function is performed by a Rosemount 1153GB5 Transmitter and 510DU Trip Units. The Rosemount Trip Units are functionally 4 checked and setpoint verified more frequently, and if necessary, j recalibrated. These more frequent testing requirements remain unchanged. Therefore, an increase in the surveillance interval to accommodate a 24 month fuei cycle does not affect the Rosemount Trip Units with respect to drift. The Rosemount 1153GB5 l Transmitter drift was evaluated using the GE methodology. The ! results of the analysis indicated that the projected 30 month i drift values for the instruments do not exceed the drift allowance provided in the setpoint calculation for this instrument. Function 3, Reactor Building Exhaust Radiation - High This function is performed by General Electric radiation t^ detectors. Based on the equipment design, method of calibration, and more frequent testing, a drift calculation could not be readily performed and was not necessary to demonstrate the acceptability of extending this instrument to a 24 month CHANNEL CALIBRATION Frequency. This conclusion is based on the following discussion. The General Electric Radiation detector sensor-converter component is a Geiger Mueller (GM) Tube detector which works on the principle of detecting ionization events caused by radiation. This makes the sensor a radiation counter which will not be subject to any time based change in its input / output relationship. The indicator-trip unit is electronically checked on a more frequent basis'during the CHANNEL FUNCTIONAL TEST (BNP ITS SR3.3.6.2.2). The CHANNEL FUNCTIONAL TEST requires the performer to check the trip unit setpoint by ut'ilizing the potentiometer and verifying the trip occurs between two settings. - The General Electric radiation detector is calibrated using a calibrated source as an input signal to the detector. The source check is perfomed by exposing the sensor-converter to a known source in a constant geometry. Source checks of radiation monitors are subject to far more uncertainties than electronic calibration checks because of source decay problems, positioning of the sources, signal strength, and the sensor response curves of i O BNP UNITS 1 & 2 6 Revision 0
L l DISCUSSION OF CHANGES ITS: 3.3.6.2 - SECONDARY CONTAINNENT ISOLATION INSTI.UNENTATION TEClGIICAL CHANGES - LESS RESTRICTIVE LE.1 that particular monitoring system. Because of the uncertainties (cont'd) associated with the calibration methods for these devices, any drift evaluation would provide no true indication of the instruments performance over time. l Based on the design of the instrumentation and the drift evaluations, it is concluded that the impact, if any, on system availability is minimal as a result of the change in the surveillance test interval. A review of the surveillance test history was performed to validate the above conclusion. This review of the surveillance test history, demonstrates that there are no failures that would invalidate the conclusion that the impact, if any, on system
-availability is minimal from a change to a 24 month operating cycle. 1 L.F .1 This change revises the Current Technical Specifications (CTS)
Allowable Values for the Improved Technical Specifications (ITS). The BNP ITS Section 3.3 reflects Allowable Values consistent with i the philosophy of NUREG-1433. These Allowable Values have been i established consistent with the methods described in CP&L's Instrument Setpoint Methodology (Design Guide DG-VIII.0050
" Instrument Setpoints" Rev. 5). The Allowable Value l O determinations were done using vendor documented performance specifications, where available and applicable. Where vendor documented performance specifications for drift were not'available 1
l or applicable, the Allowable Value was determined using plant l specific operating and surveillance trend data or an allowance as l provided for by the CP&L's Instrument Setpoint Methodology. The Allowable Value verification used actual BNP operating and surveillance trend information to ensure the validity of the developed Allowable Value. All changes to safety analysis limits applied in the methodologies were evaluated and confirmed as ensuring safety analysis licensing acceptance limits are maintained. All design limits applied in the methodologies were confirmed as ensuring that applicable design requirements of the associated systems and equi > ment are maintained. The methodologies used have been compared witi the guidance of ISA Recommended Practice ISA-RP67.04 Part II, " Methodologies for the Determination of Set >oints for Nuclear Safety-Related Instrumentation," . Septem>er 1994. Plant calibration procedures will ensure that the assumptions regarding calibration accuracy, measurement and test equipment accuracy, and setting tolerance are maintained. Setpoints for each design or safety analynis limit have been established by accounting for the applicable instrument accuracy, calibration and drift uncertainties, environmental effects, power supply fluctuations, as well as uncertainties related to process ! and primary element measurement accuracy using the CP&L Instrument !' Setpoint Methodology. The Allowable Values have been established l from each design or safety analysis limit by combining the errors O' BNP UNITS I & 2 7 Revision 0
L ~ l } l~ DISCUSSION OF CHANGES I l ITS: 3.3.6.2 - SECONDARY CONTAINNENT ISOLATION INSTRUMENTATION d V TECHNICAL CHANGES - LESS RESTRICTIVE L LF.1 associated with channel / instrument calibration (e.g., device (cont'd) accuracy, setting tolerance, and drift) with the calculated Nominal Trip Setpoint also using the.CP&L Instrument Setpoint Methodology. Additionally, each channel / instrument has been evaluated and analyzed to support a fuel cycle extension to a 24 month interval. These evaluations and analyses has been performed utilizing the guidance provided in General Electric's NE00-32160P " Calibration Interval Extension" Licensing Topical Report. These evaluations and analyses were performed using the GE Instrument Trending and Analysis (GEITAS) Software tool. The GEITAS analyses are used to demonstrate that the data collected by the operating plant-(from surveillance testing) has remained acceptable and reasonable with regard to the manufacturers design specifications. Use of the previously discussed methodologies for determining Allowable Values, instrument setpoints and analyzing channel / instrument performance ensure that the design basis and associated safety limits will not be exceeded during plant operation. These evaluations, determinations and analyses now form a portion of the plants design bases.
" Specific" L.1 The Applicability of the Reactor Building Exhaust Radiation-High-Function of CTS Table 3.3.2-1 and Table 4.3.2-1 is revised from "0PERATIONAL CONDITIONS 1, 2, 3, 5, and when irradiated fuel is being handled in secondary containment" to " MODES 1, 2, 3, during movement of irradiated fuel assemblies in secondary containment, during CORE ALTERATIONS, and during operations with the potential for draining the reactor vessel (OPDRVs)" in ITS 3.3.6.2, Secondary Containment Isolation Instrumentation. The Reactor Building Exhaust Radiation-High function is required to support the OPERABILITY of secondary containment isolation dampers and the Standby Gas Treatment System to ensure fission products entrapped within secondary containment are treated prior to discharge to the environment. When the plant is in MODE 4 or 5, the probability and consequences of a design basis accident that is postulated to leak fission products into secondary containment are reduced due to the temperature and pressure limitations in these MODES.
However, in MODE 4 or 5, activities are conducted for which significant releases of radioactivity are postulated. Therefore, Reactor Building Exhaust Radiation-High Function is only required i to be OPERABLE in MODE 4 or 5, when activities are in progress which could, if an event occurs, result in significant releases of radioactivity (during movement of irradiated fuel assemblies in secondarycontainment,duringCOREALTERATIONS,orduringOPDRVs). This change alters the Applicability of the Reactor Building L Exhaust Radiation-High Function of CTS Table 3.3.2-1 and Table 4.3.2-1 to only include these activities. This is BNP UNITS 1 & 2 8 Revision 0 l
DISCUSSION OF CHANGES ITS: 3.3.6.2 - SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION i TECHNICAL CHANGES - LESS RESTRICTIJE L.1 . considered acceptable since the BNP ITS 3.3.6.2 requires the (cont'd) Reactor Building Exhaust Radiation-High Function to be OPERABLE l when it is required to mitigate postulated events in MODE 4 or 5. The BNP ITS 3.3.6.2 Applicability for the Reactor Building Exhaust Radiation-High Function maintains and adds situations for which significant releases of radioactivity are postulated while the plant is in MODE 4 or 5. This change allows operations that do not have a potential for a significant radioactive release to be performed without requiring the Reactor Building Exhaust Radiation-High Function to be OPERABLE and provides additional scheduling flexibility during plant refueling outages. L.2 New Required Actions are added (ITS 3.3.6.2 Required Actions C.I.2 and C.2.2) to require declaring the affected components inoperable and taking the' appropriate actions in the associated Secondary Containment Isolation Damper (SCID) or Standby Gas Treatment (SGT) System Specifications if the associated penetrations and SGT subsystems are not placed in the proper condition within I hour. Currently, CTS Table 3.3.2-1 ACTION 23 requires either a CTS 3.0.3 entry, which would result in an immediate shutdown, or an d immediate suspension of activities which have the potential for a radioactive release. By declaring the affected supported equipment inoperable (i.e., the SCIDs and the SGT subsystems), and as a result taking the Technical Specifications actions of the d O affected supported equipment, unit operation is maintained within the' bounds of the Technical Specifications and approved ACTIONS. Since this instrumentation provides a signal for the SCIDs and SGT System (i.e.,itsupportsSCIDsandSGTSystemOPERABILITY), itis appropriate that the proper action would be to declare the associated SCIDs and SGT subsystems inoperable. The CTS instrumentation requirements are overly restrictive, in that if the associated SCIDs and SGT subsystems were inoperable for other , reasons, a much longer restoration time is provided in the . 1 associated CTS system specifications. Currently if an instrument is inoperable but the associated SCIDs and SGT subsystems are ; otherwise fully OPERABLE, an immediate shutdown or an immediate i suspension of activities which have the potential for a radioactive release is required. The time period provided for declaring the associated SCIDs and SGT subsystem inoperable, and the resulting time >eriod allowed for continued operation, is ! acceptable due to tie low probability of an event requiring the n secondary containment isolation instrumentation to function to CD isolate the SCIDs and start the SGT System during the short time ; in which continued operation is allowed and the capability to automatically isolate the secondary containment may be lost. ! RELOCATED SPECIFICATIONS None O BNP UNITS 1 & 2 9 Revision 0
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ONTROL ROOK EMERGEEY VENTJLATION SYSTEM L ITI ONDIT FOR RATION Leo 317lM The Control Room Emergency Ventilation Systeg inst shown in Table 3.3.5.51 shall be OPERABLEF . ation 1
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APPLICABILITY: As.shown in Table 3.3.5.5-1. . bC11%; : h6 propod Acxtods _ Hole. With one or more detectors inoperable, take the ACTION required by pa 4 ($ Table 3.3.5.5-1. [ /e provipfons of Sp .ficationf 0.4 are no),arpplicab" g,,
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SURVEILLANCE REQUIREMEN$i. 6-Each of the above required control room emergency ventilation Sum &,M . instruments shall be demonstrated OPERABLE by performance of the
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,( ( l {* e Contr sideredOPEbencVentilationSystem(CREVS)instrumen Room Emer l . consistent wi the conditio a specified in
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footno *** to Technic Specification .7.2. durin from bruary 6. 1998 o May 1. 1998 In this conf uration. the sys m is not consi red to be in ap CTION stat t for.the purp es . , of echnical Spect ication 3.0.4. ( i t I 5 j k,' BRUNSWICK - UNIT 1 3/4 3-64 Anendment No. 191 i f
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. , ,, .c A k W:b*3.'bl NSTRUMENTATION, ONTROL ROOM EMERGENCY VENTIkATION SMTEIi) l L TI. ONDITI FOR 0 TION g 3,g,f@ The Control Room Emergency Ventilation System inst ntation shown in Table 3.3.5.5-1 shall be OPERABLE g 1 APPLICABILITY: . As shown in Table 3.3.5.5-1.
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/4fF) Each of the above required control room emergency ventilation instruments shall be demonstrated OPERABLE by performance of the g
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- Th Control R Emergency Ve ilation Sys fn (CREVS) instrupentation y be consi red OPEPABLE. onsistent wi the conditions /pecified in ootnote ** to Technical pecification .7.2. durhg the/ me ti period - M f from Feb ary 6. 1998. Hay 1. 1998 In this configu tion. the CP S inst .n at tion is no considered to e in an ACTION atement for e
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DISCUSSION OF CHANGES ITS: 3.3.7.1 - CREV SYSTEM INSTRUMENTATION G ADMINISTRATIVE-(continued) ! A.4 The ACTIONS of CTS 3.3.5.5, Control Room Emergency Ventilation System Instrumentation, provides the option to restore the inoperable channel (detector) to OPERABLE status or isolate the Control Room and operate in the Radiation / Smoke Protection Mode. l l The revised presentation of ITS 3.3.7.1 Required Action A.1 (based on the BWR Standard Technical Specifications, NUREG-1433) is to k not explicitly detail options of " restore...to OPERABLE status." This action is always an option, and is implied in all ACTIONS. Omitting this action is editorial.}}