TSTF-03-07, Traveler - September 17, 2003, Transmitts TSTF-459, Eliminate RHR Shutdown Cooling System TSTF-460, Control Rod Scram Time Testing; TSTF-465, Addition of Time Performance Surveillance Requirement Note to Source Range Monitor (Srm).: Difference between revisions
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
||
| Line 3: | Line 3: | ||
| issue date = 09/17/2003 | | issue date = 09/17/2003 | ||
| title = Traveler - September 17, 2003, Transmitts TSTF-459, Eliminate RHR Shutdown Cooling System TSTF-460, Control Rod Scram Time Testing; TSTF-465, Addition of Time Performance Surveillance Requirement Note to Source Range Monitor (Srm). | | title = Traveler - September 17, 2003, Transmitts TSTF-459, Eliminate RHR Shutdown Cooling System TSTF-460, Control Rod Scram Time Testing; TSTF-465, Addition of Time Performance Surveillance Requirement Note to Source Range Monitor (Srm). | ||
| author name = Furio P | | author name = Furio P, Infanger P, Silko T, Wideman S | ||
| author affiliation = B & W Owners Group, BWR Owners Group, Combustion Engineering Owners Group, Technical Specifications Task Force, Westinghouse Owners Group | | author affiliation = B & W Owners Group, BWR Owners Group, Combustion Engineering Owners Group, Technical Specifications Task Force, Westinghouse Owners Group | ||
| addressee name = Beckner W | | addressee name = Beckner W | ||
| addressee affiliation = NRC/NRR/DRIP | | addressee affiliation = NRC/NRR/DRIP | ||
| docket = | | docket = | ||
| Line 129: | Line 129: | ||
Cooling System to be in operation in MODE 3 with reactor steam dome pressure < [the RHR cut | Cooling System to be in operation in MODE 3 with reactor steam dome pressure < [the RHR cut | ||
in permissive pressure], MODE 4, and MODE 5 with irradiated fuel in the reactor pressure vessel. | in permissive pressure], MODE 4, and MODE 5 with irradiated fuel in the reactor pressure vessel. | ||
2.0 Proposed Change The Limiting Conditions for Operation (LCOs) of the following Specifications are revised to | |||
Change The Limiting Conditions for Operation (LCOs) of the following Specifications are revised to | |||
eliminate the requirement that at least one RHR shutdown cooling system must be in operation. | eliminate the requirement that at least one RHR shutdown cooling system must be in operation. | ||
| Line 196: | Line 194: | ||
in itiated system. This Frequency has been shown to be acceptable based on operating experience. | in itiated system. This Frequency has been shown to be acceptable based on operating experience. | ||
5.0 Regulatory Analysis 5.1 No Significant Hazards Consideration The TSTF has evaluated whether or not a significant hazards consideration is involved with th e proposed generic change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: | |||
Analysis 5.1 No Significant Hazards Consideration The TSTF has evaluated whether or not a significant hazards consideration is involved with th e proposed generic change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: | |||
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evalu ated? Response: No. | : 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evalu ated? Response: No. | ||
The proposed change allows the Residual Heat Removal Shutdown Cooling (RHR SDC) | The proposed change allows the Residual Heat Removal Shutdown Cooling (RHR SDC) | ||
| Line 229: | Line 225: | ||
margin of safet | margin of safet | ||
: y. B ased on the above, the TSTF concludes that the proposed change presents no significant hazards considerations under the standards set forth in 10 CFR 50.92(c), and, accordingl y , a finding of "no significant hazards consideration" is justified. | : y. B ased on the above, the TSTF concludes that the proposed change presents no significant hazards considerations under the standards set forth in 10 CFR 50.92(c), and, accordingl y , a finding of "no significant hazards consideration" is justified. | ||
5.2 Applic able Regulatory Requirements/Criteria The proposed change to the Improved Standard Technical Specifications do not change the | |||
able Regulatory Requirements/Criteria The proposed change to the Improved Standard Technical Specifications do not change the | |||
design requirements for the RHR Shutdown Cooling System and the RHR shutdown Cooling | design requirements for the RHR Shutdown Cooling System and the RHR shutdown Cooling | ||
| Line 239: | Line 233: | ||
of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to th e health and safety of the public. | of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to th e health and safety of the public. | ||
6.0 Environmental Consideration A review has determined that the proposed change would not change a requirement with respect | |||
Consideration A review has determined that the proposed change would not change a requirement with respect | |||
to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would not change an inspection or surveillance requirement. The proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or TSTF-459, Rev. 0 Page 5 of 6 significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR | to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would not change an inspection or surveillance requirement. The proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or TSTF-459, Rev. 0 Page 5 of 6 significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR | ||
| Line 246: | Line 239: | ||
51.22(c)(9). Therefore, p ursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change. | 51.22(c)(9). Therefore, p ursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change. | ||
7.0 References None. | |||
None. | |||
TSTF-459, Rev. 0 Page 6 of 6 INSERT 1 manual, power operated, and automatic valve in the flow path that is not l ocked, sealed, or otherwise secured in position, is aligned or can be aligned to its correct position. | TSTF-459, Rev. 0 Page 6 of 6 INSERT 1 manual, power operated, and automatic valve in the flow path that is not l ocked, sealed, or otherwise secured in position, is aligned or can be aligned to its correct position. | ||
INSERT 2 Verifying the correct alignment for manual, power operated, and automatic valves in the RHR | INSERT 2 Verifying the correct alignment for manual, power operated, and automatic valves in the RHR | ||
| Line 277: | Line 269: | ||
method.circulation AND recirculation pump in operation. | method.circulation AND recirculation pump in operation. | ||
AND BWRl4 STS 3.4.9-1 Rev.2, 04/30101 RHR Shutdown Cooling System-Cold Shutdown 3.4.9 ACTIONS (continued) | AND BWRl4 STS 3.4.9-1 Rev.2, 04/30101 RHR Shutdown Cooling System-Cold Shutdown 3.4.9 ACTIONS (continued) | ||
CONDITION REQUIRED ACTION COMPLETION TIME B.2coolant.1<1.--",...r hour)Jerfl'Derature.SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.9.1 FREQUENCY BWRl4STS 3.4.9-2 Rev.2, 04/30/01 RHR-High Water Level 3.9.8 3.9 REFUELING OPERATIONS | CONDITION REQUIRED ACTION COMPLETION TIME B.2coolant.1<1.--",...r hour)Jerfl'Derature.SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.9.1 FREQUENCY BWRl4STS 3.4.9-2 Rev.2, 04/30/01 RHR-High Water Level 3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)-High Water Level LCO 3.9.8 APPLICABILITY: | ||
Heat Removal (RHR)-High Water Level LCO 3.9.8 APPLICABILITY: | |||
ACTIONS One RHR shutdown cooling subsystem shall be MODE 5 with irradiated fuel in the reactor pressure vessel (RPV)and the water level[23]ft abovethetop of the[RPV flange].CONDITION REQUIRED ACTION COMPLETION TIME A, Required RHR shutdown A,1 Verify an alternate method 1 hour cooling subsystem of decay heat removal is inoperable. | ACTIONS One RHR shutdown cooling subsystem shall be MODE 5 with irradiated fuel in the reactor pressure vessel (RPV)and the water level[23]ft abovethetop of the[RPV flange].CONDITION REQUIRED ACTION COMPLETION TIME A, Required RHR shutdown A,1 Verify an alternate method 1 hour cooling subsystem of decay heat removal is inoperable. | ||
available. | available. | ||
| Line 291: | Line 280: | ||
method.circulation SURVEILLANCE REQUIREMENTS SR 3.9.8.1 SURVEILLANCE RHR shutdown cooling subsysterr(j | method.circulation SURVEILLANCE REQUIREMENTS SR 3.9.8.1 SURVEILLANCE RHR shutdown cooling subsysterr(j | ||
_....:.,.;:::::; | _....:.,.;:::::; | ||
.....FREQUENCY BWRl4 STS 3.9.8-2 Rev.2, 04/30101 RHR-Low Water Level 3.9.9 3.9 REFUELING OPERATIONS | .....FREQUENCY BWRl4 STS 3.9.8-2 Rev.2, 04/30101 RHR-Low Water Level 3.9.9 3.9 REFUELING OPERATIONS 3.9.9 Residual Heat Removal (RHR)-Low Water Level LCO 3.9.9 APPLICABI L1TY: ACTIONS RHR shutdown coo!!-ng subsystems shall be OPERABLE.*r1¥e) | ||
Heat Removal (RHR)-Low Water Level LCO 3.9.9 APPLICABI L1TY: ACTIONS RHR shutdown coo!!-ng subsystems shall be OPERABLE.*r1¥e) | |||
CB:HR_own coohAtfSuDsystemin operatiolj) | CB:HR_own coohAtfSuDsystemin operatiolj) | ||
MODE 5 with irradiated fuel in the reactor pressure vessel (RPV)and the water level<[23]ft abovethetop of the[RPV flange].CONDITION REQUIRED ACTION COMPLETION TIME A.One or two required A.1 Verify an alternate method 1 hour RHR shutdown cooling of decay heat removal is subsystem inoperable. | MODE 5 with irradiated fuel in the reactor pressure vessel (RPV)and the water level<[23]ft abovethetop of the[RPV flange].CONDITION REQUIRED ACTION COMPLETION TIME A.One or two required A.1 Verify an alternate method 1 hour RHR shutdown cooling of decay heat removal is subsystem inoperable. | ||
| Line 425: | Line 411: | ||
CONDITION REQUIRED ACTION COMPLETION TIME B.No RHR shutdown cooling subsystem in operation. | CONDITION REQUIRED ACTION COMPLETION TIME B.No RHR shutdown cooling subsystem in operation. | ||
No recir ation pump in oper n.B.2 Verify reactor coolant circulating by an alternate method.1 hour from dis ery of no reactor olant circulatio Once per hour SURVEILLANCE REQUIREMENTS FREQUENCY RHR shutdown cooling subsystem@} | No recir ation pump in oper n.B.2 Verify reactor coolant circulating by an alternate method.1 hour from dis ery of no reactor olant circulatio Once per hour SURVEILLANCE REQUIREMENTS FREQUENCY RHR shutdown cooling subsystem@} | ||
n SR 3.4.10.1 each Y6 ,,/rrJ SURVEILLANCE BWRl6STS 3.4.10-2 Rev.2, 04/30101 RHR-High Water Level.3.9.8 3.9 REFUELING OPERATIONS | n SR 3.4.10.1 each Y6 ,,/rrJ SURVEILLANCE BWRl6STS 3.4.10-2 Rev.2, 04/30101 RHR-High Water Level.3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)-High Water Level LCO 3.9.8 One RHR shutdown cooling subsystem shall be.APPLICABI L1TY: ACTIONS._-------------------- | ||
Heat Removal (RHR)-High Water Level LCO 3.9.8 One RHR shutdown cooling subsystem shall be.APPLICABI L1TY: ACTIONS._-------------------- | |||
------------------- | ------------------- | ||
.NOTE*:;t The required shutdown cooling subsyst may be not in op tion for up to 2 urs per 8 hour period.._------------------------------------ | .NOTE*:;t The required shutdown cooling subsyst may be not in op tion for up to 2 urs per 8 hour period.._------------------------------------ | ||
| Line 441: | Line 424: | ||
C.1 C.2 Ve'reactor coolant culation by an alternate method.Monitor rea temperatur | C.1 C.2 Ve'reactor coolant culation by an alternate method.Monitor rea temperatur | ||
.Once per 12 hours thereafter Once per hour SURVEILLANCE REQUIREMENTS SR 3.9.8.1 BWRl6STS SURVEILLANCERHR shutdown cooling subsystem@? | .Once per 12 hours thereafter Once per hour SURVEILLANCE REQUIREMENTS SR 3.9.8.1 BWRl6STS SURVEILLANCERHR shutdown cooling subsystem@? | ||
pe , 3.9.8-2 FREQUENCY Rev.2, 04/30/01 RHR-Low Water Level 3.9.9 3.9 REFUELING OPERATIONS | pe , 3.9.8-2 FREQUENCY Rev.2, 04/30/01 RHR-Low Water Level 3.9.9 3.9 REFUELING OPERATIONS 3.9.9 Residual Heat Removal (RHR)-Low Water Level LCO 3.9.9 APPLICABI L1TY: ACTIONS Two RHR shutdown cooling subsystems shall be OPERABLE.@) | ||
Heat Removal (RHR)-Low Water Level LCO 3.9.9 APPLICABI L1TY: ACTIONS Two RHR shutdown cooling subsystems shall be OPERABLE.@) | |||
i!§!In MODE 5 with irradiated fuel in the reactor pressure vessel and with the water level<[23]ft abovethetop of the[reactor pressure vessel flange].CONDITION REQUIRED ACTION COMPLETION TIME A.One or two RHR A.1 Verify an alternate method 1 hour shutdown cooling of decay heat removal is subsystems inoperable. | i!§!In MODE 5 with irradiated fuel in the reactor pressure vessel and with the water level<[23]ft abovethetop of the[reactor pressure vessel flange].CONDITION REQUIRED ACTION COMPLETION TIME A.One or two RHR A.1 Verify an alternate method 1 hour shutdown cooling of decay heat removal is subsystems inoperable. | ||
available for each AND inoperable RHR shutdown cooling subsystem. | available for each AND inoperable RHR shutdown cooling subsystem. | ||
| Line 527: | Line 507: | ||
Industry Contact: | Industry Contact: | ||
Tom Silko, (802) 258-4146, tsilko@entergy.com Yes Correction or Improvement: | Tom Silko, (802) 258-4146, tsilko@entergy.com Yes Correction or Improvement: | ||
Improvement | Improvement 1.0 Description The proposed Traveler changes NUREG | ||
The proposed Traveler changes NUREG | |||
-1433 (BWR/4) and NUREG | -1433 (BWR/4) and NUREG | ||
-1434 (BWR/6) by revising the Frequency of SR 3.1.4.2, control rod scram time testing, from "120 days cumulative operation in MODE 1" to | -1434 (BWR/6) by revising the Frequency of SR 3.1.4.2, control rod scram time testing, from "120 days cumulative operation in MODE 1" to | ||
"[200] days cumulative operation in MODE 1." The Bases are revised to limit the percentage of the tested rods which can be "slow" from 20% to 7.5%. | "[200] days cumulative operation in MODE 1." The Bases are revised to limit the percentage of the tested rods which can be "slow" from 20% to 7.5%. | ||
2.0 Proposed Change NUREG-1433, SR 3.1.4.2 states, " Verify, for a representative sample, each tested control rod scram time is within the limits of Table 3.1.4 | |||
Change NUREG-1433, SR 3.1.4.2 states, " Verify, for a representative sample, each tested control rod scram time is within the limits of Table 3.1.4 | |||
-1 with reactor steam dome pressure | -1 with reactor steam dome pressure | ||
> [800] psig." | > [800] psig." | ||
| Line 553: | Line 528: | ||
of the entire sample size) is satisfied, or until the total number of "slow" control rods (throughout the core, from all surveillances) exceeds the LCO limit. For planned testing, the control rods selected for the sample should be different for each test. | of the entire sample size) is satisfied, or until the total number of "slow" control rods (throughout the core, from all surveillances) exceeds the LCO limit. For planned testing, the control rods selected for the sample should be different for each test. | ||
17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. | 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. | ||
TSTF-460, Rev. 0 BWROG-90, Rev. 1 | TSTF-460, Rev. 0 BWROG-90, Rev. 1 4.0 Technical Analysis Industry operating experience has shown the control rod scram rates to be highly reliable. For example, at the Grand Gulf Nuclear Station, out of 7,660 control rod insertion tests, only 12 control rods have been slower than the insertion time limit (with the exception of test data from an anomalous cycle). The control rod drive | ||
Analysis Industry operating experience has shown the control rod scram rates to be highly reliable. For example, at the Grand Gulf Nuclear Station, out of 7,660 control rod insertion tests, only 12 control rods have been slower than the insertion time limit (with the exception of test data from an anomalous cycle). The control rod drive | |||
system has shown to be highly reliable. This high reliability supports the extension of the Surveillance | system has shown to be highly reliable. This high reliability supports the extension of the Surveillance | ||
| Line 575: | Line 547: | ||
The proposed change is consistent with the amendment requests in References 1, 2, and 3 and the NRC's approvals in References 4 and 5. | The proposed change is consistent with the amendment requests in References 1, 2, and 3 and the NRC's approvals in References 4 and 5. | ||
17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. | 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. | ||
TSTF-460, Rev. 0 BWROG-90, Rev. 1 | TSTF-460, Rev. 0 BWROG-90, Rev. 1 5.0 Regulatory Analysis 5.1 No Significant Hazards Consideration The TSTF has evaluated whether or not a significant hazards consideration is involved with the proposed generic change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as | ||
Analysis 5.1 No Significant Hazards Consideration The TSTF has evaluated whether or not a significant hazards consideration is involved with the proposed generic change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as | |||
discussed below: | discussed below: | ||
| Line 597: | Line 566: | ||
protected. Therefore, the proposed change does not involve a significant reduction in a margin of safety. | protected. Therefore, the proposed change does not involve a significant reduction in a margin of safety. | ||
5.2 Applicable Regulatory Requirements / Criteria The proposed change does not affect any OPERABILITY requirements and the test Frequency being revised is not specified in regulations. As a result, no regulatory requirements or criteria are affected. | |||
Regulatory Requirements / Criteria The proposed change does not affect any OPERABILITY requirements and the test Frequency being revised is not specified in regulations. As a result, no regulatory requirements or criteria are affected. | |||
17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. | 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. | ||
TSTF-460, Rev. 0 BWROG-90, Rev. 1 | TSTF-460, Rev. 0 BWROG-90, Rev. 1 6.0 Environmental Consideration The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed | ||
Consideration The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed | |||
amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). | amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). | ||
| Line 611: | Line 575: | ||
need be prepared in connection with the proposed amendment. | need be prepared in connection with the proposed amendment. | ||
7.0 References | |||
: 1. Letter from William A. Eaton, Entergy Operations, Inc. (GNRO | : 1. Letter from William A. Eaton, Entergy Operations, Inc. (GNRO | ||
-2001/00002) to NRC, "Grand Gulf Nuclear Station, Docket No. 50 | -2001/00002) to NRC, "Grand Gulf Nuclear Station, Docket No. 50 | ||
| Line 620: | Line 583: | ||
: 2. Letter from William A. Eaton, Entergy Operations, Inc. (GNRO | : 2. Letter from William A. Eaton, Entergy Operations, Inc. (GNRO | ||
-2002/00012) to NRC, "Grand Gulf Nuclear Station, Docket No. 50 | -2002/00012) to NRC, "Grand Gulf Nuclear Station, Docket No. 50 | ||
-416, Supplement to Amendment Request Concerning Control Rod Scram Time Testing Frequency," dated February 20, 2002. | -416, Supplement to Amendment Request Concerning Control Rod Scram Time Testing Frequency," dated February 20, 2002. | ||
: 3. Letter from William R. Brian, Entergy Operations, Inc. (LAR 2001 | : 3. Letter from William R. Brian, Entergy Operations, Inc. (LAR 2001 | ||
-35) to NRC, "River Bend Station, Unit 1, Docket No. 50 | -35) to NRC, "River Bend Station, Unit 1, Docket No. 50 | ||
| Line 656: | Line 619: | ||
Control Rod Scram Times3.1.4BWR/4 STS3.1.4 - 1Rev. 2, 04/30/013.1 REACTIVITY CONTROL SYSTEMS3.1.4Control Rod Scram TimesLCO 3.1.4a.No more than [10] OPERABLE control rods shall be "slow," inaccordance with Table 3.1.4-1, andb.No more than 2 OPERABLE control rods that are "slow" shalloccupy adjacent locations.APPLICABILITY:MODES 1 and 2. | Control Rod Scram Times3.1.4BWR/4 STS3.1.4 - 1Rev. 2, 04/30/013.1 REACTIVITY CONTROL SYSTEMS3.1.4Control Rod Scram TimesLCO 3.1.4a.No more than [10] OPERABLE control rods shall be "slow," inaccordance with Table 3.1.4-1, andb.No more than 2 OPERABLE control rods that are "slow" shalloccupy adjacent locations.APPLICABILITY:MODES 1 and 2. | ||
ACTIONSCONDITIONREQUIRED ACTIONCOMPLETION TIMEA.Requirements of theLCO not met.A.1Be in MODE 3.12 hoursSURVEILLANCE REQUIREMENTS- NOTE -During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall beisolated from the associated scram accumulator.SURVEILLANCEFREQUENCYSR 3.1.4.1Verify each control rod scram time is within the limitsof Table 3.1.4-1 with reactor steam dome pressure | ACTIONSCONDITIONREQUIRED ACTIONCOMPLETION TIMEA.Requirements of theLCO not met.A.1Be in MODE 3.12 hoursSURVEILLANCE REQUIREMENTS- NOTE -During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall beisolated from the associated scram accumulator.SURVEILLANCEFREQUENCYSR 3.1.4.1Verify each control rod scram time is within the limitsof Table 3.1.4-1 with reactor steam dome pressure | ||
[800] psig.Prior to exceeding 40% RTP after each reactor shutdown 120 daysSR 3.1.4.2Verify, for a representative sample, each testedcontrol rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure | |||
[800] psig.120 days cumulative operation in MODE 1 SURVEILLANCE REQUIREMENTS (continued)Control Rod Scram TimesB 3.1.4 BASESBWR/4 STSB 3.1.4 - 4Rev. 2, 04/30/01testing can be performed. To ensure that scram time testing isperformed within a reasonable time following a shutdown 120 days orlonger, control rods are required to be tested before exceeding 40% RTP following the shutdown. This Frequency is acceptable considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by fuel movement within the associated core cell and by work on control rods or the CRD System.SR 3.1.4.2Additional testing of a sample of control rods is required to verify thecontinued 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 (e.g., 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 LCO limit. | |||
For planned testing, the control rods selected for the sample should be different for each test. 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 frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, "Control Rod Scram Accumulators."SR 3.1.4.3When work that could affect the scram insertion time is performed on acontrol 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. | For planned testing, the control rods selected for the sample should be different for each test. 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 frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, "Control Rod Scram Accumulators."SR 3.1.4.3When work that could affect the scram insertion time is performed on acontrol 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. | ||
The limits for reactor pressures < 800 psig are established based on a high probability of meeting the acceptance criteria at reactor pressures 800 psig. Limits for 800 psig are found in Table 3.1.4-1. If testingdemonstrates the affected control rod does not meet these limits, but is Control Rod Scram Times3.1.4BWR/6 STS3.1.4 - 1Rev. 2, 04/30/013.1 REACTIVITY CONTROL SYSTEMS3.1.4Control Rod Scram TimesLCO 3.1.4a.No more than [14] OPERABLE control rods shall be "slow," inaccordance with Table 3.1.4-1 andb.No more than 2 OPERABLE control rods that are "slow" shalloccupy adjacent locations.APPLICABILITY:MODES 1 and 2.ACTIONSCONDITIONREQUIRED ACTIONCOMPLETION TIMEA.Requirements of theLCO not met.A.1Be in MODE 3.12 hoursSURVEILLANCE REQUIREMENTS- NOTE -During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.SURVEILLANCEFREQUENCYSR 3.1.4.1Verify each control rod scram time is within the limitsof Table 3.1.4-1 with reactor steam dome pressure | The limits for reactor pressures < 800 psig are established based on a high probability of meeting the acceptance criteria at reactor pressures 800 psig. Limits for 800 psig are found in Table 3.1.4-1. If testingdemonstrates the affected control rod does not meet these limits, but is Control Rod Scram Times3.1.4BWR/6 STS3.1.4 - 1Rev. 2, 04/30/013.1 REACTIVITY CONTROL SYSTEMS3.1.4Control Rod Scram TimesLCO 3.1.4a.No more than [14] OPERABLE control rods shall be "slow," inaccordance with Table 3.1.4-1 andb.No more than 2 OPERABLE control rods that are "slow" shalloccupy adjacent locations.APPLICABILITY:MODES 1 and 2.ACTIONSCONDITIONREQUIRED ACTIONCOMPLETION TIMEA.Requirements of theLCO not met.A.1Be in MODE 3.12 hoursSURVEILLANCE REQUIREMENTS- NOTE -During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.SURVEILLANCEFREQUENCYSR 3.1.4.1Verify each control rod scram time is within the limitsof Table 3.1.4-1 with reactor steam dome pressure | ||
[950] psig.Prior to exceeding40% RTP after each reactor shutdown 120 daysSR 3.1.4.2Verify, for a representative sample, each testedcontrol rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure | |||
[950] psig.120 dayscumulative operation in MODE 1 SURVEILLANCE REQUIREMENTS (continued)Control Rod Scram TimesB 3.1.4 BASESBWR/6 STSB 3.1.4 - 4Rev. 2, 04/30/01at reactor steam dome pressure 950 psig ensures that the scram timeswill be within the specified limits at higher pressures. 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 scram time testing is performed within a reasonable time following a shutdown 120 days, control rods are required to be tested before exceeding40% RTP. This Frequency is acceptable, considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by fuel movement within the associated core cell and by work on control rods or the CRD System.SR 3.1.4.2Additional testing of a sample of control rods is required to verify thecontinued 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." If more than 20% of the sample is declared to be "slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 20% criterion (e.g., 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 LCO limit. For planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data were 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 frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, "Control Rod Scram Accumulators."SR 3.1.4.3When work that could affect the scram insertion time is performed on acontrol 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 that the affected control rod is still within acceptable TSTF-465, Rev. 0 BWROG-81, Rev. 1 NUREGs Affected: | |||
Addition of time performance Surveillance Requirement (SR) note to Source Range Monitor (SRM) SRs Technical Specification Task Force Improved Standard Technical Specifications Change Traveler 1430 1431 1432 1433 1434 Classification: | Addition of time performance Surveillance Requirement (SR) note to Source Range Monitor (SRM) SRs Technical Specification Task Force Improved Standard Technical Specifications Change Traveler 1430 1431 1432 1433 1434 Classification: | ||
: 3) Improve Specifications Recommended for CLIIP?: | : 3) Improve Specifications Recommended for CLIIP?: | ||
Industry Contact: | Industry Contact: | ||
Tom Silko, (802) 258-4146, tsilko@entergy.com Yes Correction or Improvement: | Tom Silko, (802) 258-4146, tsilko@entergy.com Yes Correction or Improvement: | ||
Improvement | Improvement 1.0 Description A time allowance Note is being added to the Source Range Monitor (SRM) Surveillance Requirements (SRs) 3.3.1.2.3 and 3.3.1.2.4. This change provides a time allowance to perform the subject SRs following sudden entry into MODE 3 due to a reactor scram. | ||
A time allowance Note is being added to the Source Range Monitor (SRM) Surveillance Requirements (SRs) 3.3.1.2.3 and 3.3.1.2.4. This change provides a time allowance to perform the subject SRs following sudden entry into MODE 3 due to a reactor scram. | |||
These two SRs are not routinely performed in MODE 1 and thus will likely not be in periodicity. With the two SRs out of periodicity, sudden entry into MODE 3 due to a scram results in the immediate entry into SR 3.0.3 for the SRMs, which would remain in effect until the two SRs were completed. In STS, it atypical to have a | These two SRs are not routinely performed in MODE 1 and thus will likely not be in periodicity. With the two SRs out of periodicity, sudden entry into MODE 3 due to a scram results in the immediate entry into SR 3.0.3 for the SRMs, which would remain in effect until the two SRs were completed. In STS, it atypical to have a | ||
| Line 676: | Line 636: | ||
presents a administrative distraction to Operators involved in scram recovery activities. Therefore, the | presents a administrative distraction to Operators involved in scram recovery activities. Therefore, the | ||
addition of a specific time allowance note to perform the two SRs is being proposed. | addition of a specific time allowance note to perform the two SRs is being proposed. | ||
2.0 Proposed Change A 12-hour time allowance note is added to SRs 3.3.1.2.3 (SRM CHANNEL CHECK) and 3.3.1.2.4 (SRM COUNT RATE/SIGNAL | |||
Change A 12-hour time allowance note is added to SRs 3.3.1.2.3 (SRM CHANNEL CHECK) and 3.3.1.2.4 (SRM COUNT RATE/SIGNAL | |||
-TO-NOISE). This change provides a time allowance to perform the SRs for the situation involving sudden entry into MODE 3 due to a reactor scram. The added Note is the same as that currently used for SR 3.3.1.2.6 (SRM CHANNEL FUNCTIONAL TEST/SIGNAL | -TO-NOISE). This change provides a time allowance to perform the SRs for the situation involving sudden entry into MODE 3 due to a reactor scram. The added Note is the same as that currently used for SR 3.3.1.2.6 (SRM CHANNEL FUNCTIONAL TEST/SIGNAL | ||
-TO-NOISE RATIO) and 3.3.1.2.7 (SRM CHANNEL CALIBRATION). This change is applicable to the Boiling Water Reactor (BWR) Standard Technical Specifications (STS), Revision 2 of NUREG | -TO-NOISE RATIO) and 3.3.1.2.7 (SRM CHANNEL CALIBRATION). This change is applicable to the Boiling Water Reactor (BWR) Standard Technical Specifications (STS), Revision 2 of NUREG | ||
| Line 721: | Line 679: | ||
-hour performance note. | -hour performance note. | ||
12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. | 12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. | ||
TSTF-465, Rev. 0 BWROG-81, Rev. 1 | TSTF-465, Rev. 0 BWROG-81, Rev. 1 4.0 Technical Analysis A reactor scram can result in an sudden unplanned entry into MODE 3. TS 3.3.1.2, SRM Instrumentation, requires SRM operability in MODE 3. The required SRs for MODE 3 are listed in TS Table 3.3.1.2 | ||
Analysis A reactor scram can result in an sudden unplanned entry into MODE 3. TS 3.3.1.2, SRM Instrumentation, requires SRM operability in MODE 3. The required SRs for MODE 3 are listed in TS Table 3.3.1.2 | |||
-1 and include SR 3.3.1.2.3 (SRM CHANNEL CHECK), SR 3.3.1.2.4 (SRM COUNT RATE/SIGNAL | -1 and include SR 3.3.1.2.3 (SRM CHANNEL CHECK), SR 3.3.1.2.4 (SRM COUNT RATE/SIGNAL | ||
-TO-NOISE), SR 3.3.1.2.6 (SRM CHANNEL FUNCTIONAL TEST/SIGNAL | -TO-NOISE), SR 3.3.1.2.6 (SRM CHANNEL FUNCTIONAL TEST/SIGNAL | ||
| Line 754: | Line 709: | ||
TS requirements. | TS requirements. | ||
12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. | 12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. | ||
TSTF-465, Rev. 0 BWROG-81, Rev. 1 | TSTF-465, Rev. 0 BWROG-81, Rev. 1 5.0 Regulatory Analysis A change to Boiling Water Reactor (BWR) Standard Technical Specifications (STS), Revision 2 of NUREG | ||
Analysis A change to Boiling Water Reactor (BWR) Standard Technical Specifications (STS), Revision 2 of NUREG | |||
-1433 and NUREG | -1433 and NUREG | ||
-1434 is being proposed by the Technical Specifications Task Force (TSTF) to add a 12 | -1434 is being proposed by the Technical Specifications Task Force (TSTF) to add a 12 | ||
| Line 770: | Line 722: | ||
determined that they do not represent a significant hazards consideration. The following is provided in | determined that they do not represent a significant hazards consideration. The following is provided in | ||
support of this conclusion. | support of this conclusion. | ||
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? | : 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? | ||
Response: No This change provides Notes to SRs 3.3.1.2.3 and 3.3.1.2.4 to avoid those Surveillances being declared not met within the required Frequency due to an expected transition into MODE 3. The Frequency of | Response: No This change provides Notes to SRs 3.3.1.2.3 and 3.3.1.2.4 to avoid those Surveillances being declared not met within the required Frequency due to an expected transition into MODE 3. The Frequency of | ||
| Line 786: | Line 738: | ||
from any accident previously evaluated | from any accident previously evaluated | ||
.3. Does the proposed change involve a significant reduction in a margin of safety? | .3. Does the proposed change involve a significant reduction in a margin of safety? | ||
Response: No This change provides Notes to SRs 3.3.1.2.3 and 3.3.1.2.4 to avoid those Surveillances being declared not met within the required Frequency due to an expected transition into MODE 3. Should the Notes not be adopted, plants would continue to invoke SR 3.0.3 until the Surveillances can be performed. SR | Response: No This change provides Notes to SRs 3.3.1.2.3 and 3.3.1.2.4 to avoid those Surveillances being declared not met within the required Frequency due to an expected transition into MODE 3. Should the Notes not be adopted, plants would continue to invoke SR 3.0.3 until the Surveillances can be performed. SR 3.0.3 would allow 24 hours to perform the missed Surveillances, while the proposed Notes allow only 12 hours. For these reasons, the proposed change does not involve a significant reduction in the | ||
allow 24 hours to perform the missed Surveillances, while the proposed Notes allow only 12 hours. For these reasons, the proposed change does not involve a significant reduction in the | |||
margin of safety. | margin of safety. | ||
| Line 796: | Line 745: | ||
significant hazards consideration" is justified. | significant hazards consideration" is justified. | ||
5.2 Applicable Regulatory Requirements/Criteria In STS, it customary to require performance of applicable SRs prior to entry into the specified condition of the Applicability, whenever feasible. In some cases, however, due to plant conditions it may not always possible to perform the SRs prior to entry. For these situations, it is typical to have a SR performance Note which | |||
Regulatory Requirements/Criteria In STS, it customary to require performance of applicable SRs prior to entry into the specified condition of the Applicability, whenever feasible. In some cases, however, due to plant conditions it may not always possible to perform the SRs prior to entry. For these situations, it is typical to have a SR performance Note which | |||
allows a reasonable time period to perform the SR. | allows a reasonable time period to perform the SR. | ||
| Line 811: | Line 758: | ||
inimical to the common defense and security or to the health and safety of the public." | inimical to the common defense and security or to the health and safety of the public." | ||
6.0 Environmental Consideration The proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant | |||
Consideration The proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant | |||
increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change. | increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change. | ||
7.0 References None Revision History OG Revision 0 Revision Status: | |||
None Revision History OG Revision 0 Revision Status: | |||
Closed Original Issue Revision | Closed Original Issue Revision | ||
Revision as of 04:00, 14 July 2019
| ML033350006 | |
| Person / Time | |
|---|---|
| Issue date: | 09/17/2003 |
| From: | Furio P, Infanger P, Silko T, Wideman S B & W Owners Group, BWR Owners Group, Combustion Engineering Owners Group, Technical Specifications Task Force, Westinghouse Owners Group |
| To: | Beckner W Division of Regulatory Improvement Programs |
| References | |
| TSTF-03-07 | |
| Download: ML033350006 (83) | |
Text
{{#Wiki_filter:1 1921 Rockville Pike, Suite 100, Rockville, MD 20852 Phone: 301 -984-4400 , Fax: 301 -984-7600 Email: tstf@excelservices.com Administered by EXCEL Services Corporation TECHNICAL SPECIFICAT IONS TASK FORCE A JOINT OWNERS GROUP ACTIVITY TSTF September 1 7 , 2003 TSTF-03-0 7
Dr. William D. Beckner, Director Operating Reactor Improvements Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 -0001 SUB JECT: TSTF -4 5 9, TSTF-460, and TSTF -465
Dear Dr. Beckner:
Enclosed for NRC consideration are the following Technical Specifica tion Task Force Traveler s: TSTF-4 5 9 , Revision 0 , "Eliminate the requirement to have one RHR Shutdown Cooling System in operation
- " TSTF-460, Revision 0, "
Control Rod Scram Time Testing Frequency
- " and TSTF-465, Revision 0, "Addition of time performance Surveillance Requirement (SR) note to Source Range Monitor (SRM) SRs
." Any NRC review fees associated with these Travelers shoul d be billed to the Boilin g Water Reactors Owners Group. Should you have any questions, please do not hesitate to contact us. Steve Wideman (WOG) Tom Silko (BWROG) Patricia Furio (CEOG) Paul Infanger (BWOG) Enclosure cc: K , Putnam, BWROG
TSTF-459, Rev. 0 BWROG-37, Rev. 1 NUREGs Affected: Eliminate the requirement to have one RHR Shutdown Cooling System in operation Technical Specification Task Force Improved Standard Technical Specifications Change Traveler 1430 1431 1432 1433 1434 Classification:
- 3) Improve Specifications Recommended for CLIIP?:
Industry Contact: Tom Silko, (802) 258-4146, tsilko@entergy.com Yes Correction or Improvement: Improvement See attached. Revision History OG Revision 0 Revision Status: Closed Original Issue Revision
Description:
Revision Proposed by: Owners Group Review Information Date Originated by OG: 16-May-97 Owners Group Comments: 2/14/2001 - discussed by TSTF. Needs Safety Evaluation quality justification and be marked on Revision 2 pages. Date: 21-Sep-99 Owners Group Resolution: Approved OG Revision 1 Revision Status: Active Remarked on Revision 2 pages and expanded justification to SE quality. Revision
Description:
Revision Proposed by: BWROG Owners Group Review Information Date Originated by OG: 21-May-03 Owners Group Comments: (No Comments) Date: 21-May-03 Owners Group Resolution: Approved TSTF Review Information TSTF Received Date: 08-Aug-03 Date Distributed for Review: 12-Aug-03 TSTF Comments: (No Comments) OG Review Completed: BWOG CEOG WOG BWROG 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-459, Rev. 0 BWROG-37, Rev. 1 Affected Technical Specifications OG Revision 1 Revision Status: Active Date: 26-Aug-03 TSTF Resolution: Approved NRC Review Information NRC Received Date: 19-Sep-03 LCO 3.9.8 RHR - High Water Level LCO 3.9.8 Bases RHR - High Water Level Action 3.9.8.C Deleted Change
Description:
RHR - High Water Level Action 3.9.8.C Bases Deleted Change
Description:
RHR - High Water Level SR 3.9.8.1 RHR - High Water Level SR 3.9.8.1 Bases RHR - High Water Level LCO 3.9.9 RHR - Low Water Level LCO 3.9.9 Bases RHR - Low Water Level Action 3.9.9.C Deleted Change
Description:
RHR - Low Water Level Action 3.9.9.C Bases Deleted Change
Description:
RHR - Low Water Level SR 3.9.9.1 RHR - Low Water Level SR 3.9.9.1 Bases RHR - Low Water Level LCO 3.4.8 NUREG(s)- 1433 Only RHR Shutdown Cooling System - Hot Shutdown LCO 3.4.8 Bases NUREG(s)- 1433 Only RHR Shutdown Cooling System - Hot Shutdown Appl. 3.4.8 Bases NUREG(s)- 1433 Only RHR Shutdown Cooling System - Hot Shutdown Action 3.4.8.A Bases NUREG(s)- 1433 Only RHR Shutdown Cooling System - Hot Shutdown Action 3.4.8.B NUREG(s)- 1433 Only Deleted Change
Description:
RHR Shutdown Cooling System - Hot Shutdown 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-459, Rev. 0 BWROG-37, Rev. 1 Action 3.4.8.B Bases NUREG(s)- 1433 Only Deleted Change
Description:
RHR Shutdown Cooling System - Hot Shutdown SR 3.4.8.1 NUREG(s)- 1433 Only RHR Shutdown Cooling System - Hot Shutdown SR 3.4.8.1 Bases NUREG(s)- 1433 Only RHR Shutdown Cooling System - Hot Shutdown LCO 3.4.9 NUREG(s)- 1433 Only RHR Shutdown Cooling System - Cold Shutdown LCO 3.4.9 Bases NUREG(s)- 1433 Only RHR Shutdown Cooling System - Cold Shutdown Appl. 3.4.9 Bases NUREG(s)- 1433 Only RHR Shutdown Cooling System - Cold Shutdown Action 3.4.9.A Bases NUREG(s)- 1433 Only RHR Shutdown Cooling System - Cold Shutdown Action 3.4.9.B NUREG(s)- 1433 Only Deleted Change
Description:
RHR Shutdown Cooling System - Cold Shutdown Action 3.4.9.B Bases NUREG(s)- 1433 Only Deleted Change
Description:
RHR Shutdown Cooling System - Cold Shutdown SR 3.4.9.1 NUREG(s)- 1433 Only RHR Shutdown Cooling System - Cold Shutdown SR 3.4.9.1 Bases NUREG(s)- 1433 Only RHR Shutdown Cooling System - Cold Shutdown Appl. 3.9.8 Bases NUREG(s)- 1433 Only RHR - High Water Level Appl. 3.9.9 Bases NUREG(s)- 1433 Only RHR - Low Water Level LCO 3.4.9 NUREG(s)- 1434 Only RHR Shutdown Cooling System - Hot Shutdown LCO 3.4.9 Bases NUREG(s)- 1434 Only RHR Shutdown Cooling System - Hot Shutdown Appl. 3.4.9 Bases NUREG(s)- 1434 Only RHR Shutdown Cooling System - Hot Shutdown Action 3.4.9.A Bases NUREG(s)- 1434 Only RHR Shutdown Cooling System - Hot Shutdown Action 3.4.9.B NUREG(s)- 1434 Only Deleted Change
Description:
RHR Shutdown Cooling System - Hot Shutdown Action 3.4.9.B Bases NUREG(s)- 1434 Only Deleted Change
Description:
RHR Shutdown Cooling System - Hot Shutdown SR 3.4.9.1 NUREG(s)- 1434 Only RHR Shutdown Cooling System - Hot Shutdown SR 3.4.9.1 Bases NUREG(s)- 1434 Only RHR Shutdown Cooling System - Hot Shutdown LCO 3.4.10 NUREG(s)- 1434 Only RHR Shutdown Cooling System - Cold Shutdown LCO 3.4.10 Bases NUREG(s)- 1434 Only RHR Shutdown Cooling System - Cold Shutdown Appl. 3.4.10 Bases NUREG(s)- 1434 Only RHR Shutdown Cooling System - Cold Shutdown 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-459, Rev. 0 BWROG-37, Rev. 1 Action 3.4.10.A Bases NUREG(s)- 1434 Only RHR Shutdown Cooling System - Cold Shutdown Action 3.4.10.B NUREG(s)- 1434 Only Deleted Change
Description:
RHR Shutdown Cooling System - Cold Shutdown Action 3.4.10.B Bases NUREG(s)- 1434 Only Deleted Change
Description:
RHR Shutdown Cooling System - Cold Shutdown SR 3.4.10.1 NUREG(s)- 1434 Only RHR Shutdown Cooling System - Cold Shutdown SR 3.4.10.1 Bases NUREG(s)- 1434 Only RHR Shutdown Cooling System - Cold Shutdown 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-459, Rev. 0 Page 1 of 6 1.0 Description This change will revise the BWR/4 and BWR/6 ISTS NUREGs to not require an RHR Shutdown
Cooling System to be in operation in MODE 3 with reactor steam dome pressure < [the RHR cut
in permissive pressure], MODE 4, and MODE 5 with irradiated fuel in the reactor pressure vessel. 2.0 Proposed Change The Limiting Conditions for Operation (LCOs) of the following Specifications are revised to
eliminate the requirement that at least one RHR shutdown cooling system must be in operation.
- BWR/4 LCO 3.4.8, RHR Shutdown Cooling System
- Hot Shutdown
- BWR/4 LCO 3.4.9, RHR Shutdown Cooling System
- Cold Shutdown
- BWR/4 LCO 3.9.8, RHR
- High Water Level
- BWR/4 LCO 3.9.9, RHR
- Low Water Level
- BWR/6 LCO 3.4.9, RHR Shutdown Cooling System
- Hot Shutdown
- BWR/6 LCO 3.4.10, RHR Shutdown Cooling System
- Cold Shutdown
- BWR/6 LCO 3.9.8, RHR
- High Water Level
- BWR/6 LCO 3.9.9, RHR
- Low Water Level LCO Notes allowing the operating RHR shutdown cooling subsystem to be stopped are removed
and ACTIONS related to no oper ating RHR shutdown cooling subsystem are eliminated. The Surveillance of each of the Specifications listed above is revised from verifying that an RHR
shutdown cooling subsystem is operating every 12 hours to verifying every 31 days that each
required RHR shutdown cooling subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is aligned or can be aligned to its correct position.
3.0 Background
The RHR Shutdown Cooling (SDC) Sy stem one mode of operation of the RHR System. This mode is associated with a U F SAR "Power Generation Objective," such that the system can "remove decay and residual heat from the reactor core to achieve and maintain a cold shutdown condition." This norma l operational mode of RHR utilizes a single suction path from one recirculation loop, which is common to both RHR divisions. Due to the inherent single failure nature of this common flow path, these valves are not required to perform an opening safety
fun ction. Also, the RHR SDC provides circulation of the reactor coolant to aid in the measurement of average reactor coolant temperature. The RHR SDC System is not required for mitigation of any event or accident evaluated in the safety analyses. The chan ge to the subject LCOs will allow RHR SDC operation to be established based on the plant conditions and will facilitate operational evolutions, such as in -vessel inspections and RHR SDC relief valve testing.
TSTF-459, Rev. 0 Page 2 of 6 4.0 Technical Analysis In the original develo pment of the ISTS NUREGs, the BWROG commented to the NRC that the requirement to have one RHR SDC subsystem in operation does not meet the criteria specified in 10CFR 50.36(c)(2)(ii). RHR SDC subsystems are only required to be operating when desired by
pl ant operations to reduce reactor coolant temperature. Its operation may also be desired on occasion to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring. Monitoring of average reactor coolant temperature may b e accomplished by continuous or intermittent operation of the subsystems, or by other systems and is associated with normal operational monitoring. Industry commitment to NUMARC 91 -06, Shutdown Risk Management, requires that plants have a conservative est imate of the time to boil for the reactor coolant system. Continuous, forced reactor coolant flow solely for the purpose of mixing to measure reactor coolant temperature is overly conservative. Natural circulation will provide sufficient mixing to obtain a reasonable estimate of average reactor coolant temperature. Periodic measurement of reactor coolant temperature or the use of temporary or alternate temperature measurement instruments, when combined with a conservatively calculated time to boil, are s ufficient to assure plant safety. Unlike Pressurized Water Reactors, Boiling Water Reactors do not use boron in the reactor
coolant for normal shutdown margin. Therefore, continuous operation of RHR SDC to ensure
mixing of a borated solution is also not required for this purpose. BWRs may use the Standby Liquid Control (SLC) System to inject boron into the reactor coolant system, but SLC is not required to be OPERABLE in the Applicability of these LCOs. The RHR SDC System is still required to be OPERABL E with this change. The system pumps can be started and stopped as dictated by plant conditions. Reactor coolant temperature can be controlled as plant conditions dictate, including maintaining adequate control to avoid
inadvertently changing MODE. Cont inuous operation of a SDC subsystem is not required to adequately perform the decay heat removal function. Establishing coolant circulation during shutdown conditions for the purpose of temperature indication of the reactor coolant is related to plant spe cific procedures for measuring reactor coolant system temperature. Allowing the stopping (and subsequent re -starting) of RHR pumps is allowed by the current RHR SDC Specifications to change operating loops or by the Notes to the various RHR -SDC LCOs. Fur thermore, the actual cooling function provided by the RHR service water system (providing cooling water to the RHR heat exchanger) is not required to be continuously operating. Operability of the RHR -SDC system, which includes the required pumps, presumes the ability to start (and re -start) any required pump. As such, these changes do not introduce any new or different failure modes nor any increased risk of loss of decay heat removal capability.
TSTF-459, Rev. 0 Page 3 of 6 The revised Specifications are similar to the Specificatio ns governing other required modes of RHR operation. Specification 3.6.2.3, "RHR Suppression Pool Cooling," require s two RHR subsystems to be OPERABLE, but does not require a system to be in operation. It is assumed that the pumps can and will be started as required for plant safety. T he RHR-SDC Surveillance Requirement is also revised to require periodic verification that the system is aligned, or can be aligned, for operation. This is consistent with the Surveillance for Specification 3.6.2.3. The Fre quency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually
in itiated system. This Frequency has been shown to be acceptable based on operating experience. 5.0 Regulatory Analysis 5.1 No Significant Hazards Consideration The TSTF has evaluated whether or not a significant hazards consideration is involved with th e proposed generic change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evalu ated? Response: No.
The proposed change allows the Residual Heat Removal Shutdown Cooling (RHR SDC) System to not be in continuous operation. The RHR SDC System is not a precursor to
any accident previously evaluated. The RHR SDC System is not required for mitigation of any accident previously evaluated. The proposed changes do not adversely affect accident the design assumptions, conditions, or configuration of the facility. The
proposed changes do not alter or prevent the ability of structures, syst ems, and components (SSCs) from performing their intended function. Therefore, it is concluded that this change does not significantly increase the probability
or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possi bility of a new or different kind of accident from any accident previously evaluated?
Response: No. The proposed change allows the Residual Heat Removal Shutdown Cooling (RHR SDC)
System to not be in continuous operation. This revision will not impact t he accident analysis. The changes will not alter the methods of operation of the RHR SDC System. No new or different accidents result. The changes do not involve a physical alteration of TSTF-459, Rev. 0 Page 4 of 6 the plant (i.e., no new or different type of equipment will be ins talled) or a change in the methods governing normal plant operation. The changes do not alter assumptions made in the safety analysis. Therefore, the possibility of a new or different kind of accident from any accident
previously evaluated is not created . 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed changes do not alter the manner in which safety limits, limiting safety
system settings or limiting conditions for operation are determined. Th e safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis. The level of
redundancy required for the RHR SDC system is unaffected. The prop osed changes do not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. Therefore, it is concluded that this change does not involve a significant reduction in the
margin of safet
- y. B ased on the above, the TSTF concludes that the proposed change presents no significant hazards considerations under the standards set forth in 10 CFR 50.92(c), and, accordingl y , a finding of "no significant hazards consideration" is justified.
5.2 Applic able Regulatory Requirements/Criteria The proposed change to the Improved Standard Technical Specifications do not change the
design requirements for the RHR Shutdown Cooling System and the RHR shutdown Cooling
System will continue to comply with applicab le regulatory requirements and criteria. The system design will still be consistent with GDC 34, Residual heat removal. In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety
of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to th e health and safety of the public.
6.0 Environmental Consideration A review has determined that the proposed change would not change a requirement with respect
to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would not change an inspection or surveillance requirement. The proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or TSTF-459, Rev. 0 Page 5 of 6 significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR
51.22(c)(9). Therefore, p ursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
7.0 References None. TSTF-459, Rev. 0 Page 6 of 6 INSERT 1 manual, power operated, and automatic valve in the flow path that is not l ocked, sealed, or otherwise secured in position, is aligned or can be aligned to its correct position. INSERT 2 Verifying the correct alignment for manual, power operated, and automatic valves in the RHR -shutdown cooling flow path provides assurance that the proper flow paths will exist for RHR operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or
securing. A valve that c an be manually (locally or remotely) aligned is allowed to be in a non-RHR shutdown cooling position provided the valve can be repositioned. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves cap able of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The Frequency of 31 days is justified because the valves are operated under procedural co ntrol, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This
Frequency has been shown to be acceptable based on operating experience.
RHR Shutdown Cooling System:.Hot$hlltdoWh 3.4.8 3.4 REACTOR COOlANT SYSTEM (RCS)3.4.8 Residual Heat Removal (RHR)Shutdown Cooling System-HotShutGiOwn LCO 3.4.8 ACTIONS One RHR shutdown cooling*subsystem may be inoperable for up to ,2 hours for the performance of Surveillanqas. MODE 3, with reactor steam dome pressure<[the RHRcutln PEltmlssive pressure]. .NOTES*1.LCO 3.0.4 is not applicable. 2.Separate CQnditicm entry is allowed for each RHR shutdowni cooling subsystElm; ._---------------------------------------------------------------------,------------------------- CONDITION A.One or two RHR A.1 shutdown cooling subsystems inoperable. REQUIRED ACTION Initiate action to restore RHR shutdown cooling subsystem(s) to OPERABLE.status.COMPLETION TIME Immediately BWRl4STS A.2 Verify an alternate method 1 hour of decay heat removal is available for each inoperable RHR shutdown cooling subsystem. 3.4.8-1 Rev.2, 04/30/01 RHR Shutdown Cooling System-Hot Shutdown 3.4.8 B.3 Monit reactor coolant Once per hour te rature and pressure.ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A.3 Be in MODE 4.24 hours B.No RHR shutdown I nitiate action to restore Immediately cooling subsystem in one RHR shutdown operation. cooling subsystem or one recirculation pump to AND operation. AND B.2 Verify reactor coolant 1 hour from discovery circulation by an alternate of no reactor cool method.circulation SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1-NOTE-Not required to be met until 2 hours after reactor steam dome pressure is<[the RHR cut in permissive pressure]. 0c.h h71 L...:,o......--Veri Q e RHR shutdown cooling subsystem V re.Ionoperating BWRl4STS 3.4.8-2 Rev.2, 04/30/01 RHR Shutdown Cooling System-Cold Shutdown 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.9 Residual Heat Removal (RHR)Shutdown Cooling System-Cold Shutdown LCO 3.4.9 1.Bo ay be not in oper.t./[rJne RHR shutdown cooling subsystem may be inoperable for up to M?hours for the performance of Surveillances ..------------------------------------------------------------------------- APPLICABILITY: MODE 4.ACTIONS.NOTE*Separate Condition entry is allowed for each shutdown cooling subsystem. CONDITION REQUIRED ACTION COMPLETION TIME A.One or two RHR A.1 Verify an alternate method 1 hour shutdown cooling of decay heat removal is subsystems inoperable. available for each AND inoperable RHR shutdown cooling subsystem. Once per 24 hours thereafer B.No RHR shutdown B.1 Verify reactor coolant 1 hour from discovery cooling subsyste circulating by an alter of no reactor coolant operation. method.circulation AND recirculation pump in operation. AND BWRl4 STS 3.4.9-1 Rev.2, 04/30101 RHR Shutdown Cooling System-Cold Shutdown 3.4.9 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.2coolant.1<1.--",...r hour)Jerfl'Derature.SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.9.1 FREQUENCY BWRl4STS 3.4.9-2 Rev.2, 04/30/01 RHR-High Water Level 3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)-High Water Level LCO 3.9.8 APPLICABILITY: ACTIONS One RHR shutdown cooling subsystem shall be MODE 5 with irradiated fuel in the reactor pressure vessel (RPV)and the water level[23]ft abovethetop of the[RPV flange].CONDITION REQUIRED ACTION COMPLETION TIME A, Required RHR shutdown A,1 Verify an alternate method 1 hour cooling subsystem of decay heat removal is inoperable. available. AND Once per 24 hours thereafter B.Required Action and B.1 Suspend loading irradiated Immediately associated Completion fuel assemblies into the Time of Condition A not RPV.met.AND B.2 Initiate action to restore Immediately [secondary] containment to OPERABLE status.AND BWRl4 STS 3.9.8-1 Rev.2, 04/30/01 Once per 12 hours thereafter itor reactor coolant emperature. C.2 RHR-High Water Level 3.9.8 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.3 Initiate action to restore Immediately one standby gas treatment subsystem to OPERABLE status.AND B.4 Initiate action to restore Immediately isolation capability in each required[secondary] containment penetration flow path not isolated.C.No RHR shutdown C.1 Verify reactor coolant1hr from discovery cooling subsystem in circulation by an alternate o reactor coolant operation. method.circulation SURVEILLANCE REQUIREMENTS SR 3.9.8.1 SURVEILLANCE RHR shutdown cooling subsysterr(j _....:.,.;:::::; .....FREQUENCY BWRl4 STS 3.9.8-2 Rev.2, 04/30101 RHR-Low Water Level 3.9.9 3.9 REFUELING OPERATIONS 3.9.9 Residual Heat Removal (RHR)-Low Water Level LCO 3.9.9 APPLICABI L1TY: ACTIONS RHR shutdown coo!!-ng subsystems shall be OPERABLE.*r1¥e) CB:HR_own coohAtfSuDsystemin operatiolj) MODE 5 with irradiated fuel in the reactor pressure vessel (RPV)and the water level<[23]ft abovethetop of the[RPV flange].CONDITION REQUIRED ACTION COMPLETION TIME A.One or two required A.1 Verify an alternate method 1 hour RHR shutdown cooling of decay heat removal is subsystem inoperable. available for each AND inoperable required RHR shutdown cooling Once per 24 hours subsystem. thereafter B.Required Action and B.1 Initiate action to restore Immediately associated Completion [secondary] containment to Time of Condition A not OPERABLE status.met.AND B.2 Initiate action to restore Immediately one standby gas treatment subsystem to OPERABLE status.AND BWRl4 STS 3.9.9-1 Rev.2, 04/30/01 RHR-Low Water Level 3.9.9 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.3 Initiate action to restore Immediately isolation capability in each required[secondary] containment penetration flow path not isolated.C.No RHR shutdown cooling subsyste In operation. C.1 Verify reactor coolant circulation by an alternat method.Monitor reactor coolant temperature. 1 hour from discovery of no reactor coolant circulation Once per 12 hours thereafter SURVEILLANCE REQUIREMENTS SR 3.9.9.1 BWRl4 STS SURVEILLANCE shutdown cooling ing 3.9.9-2 FREQUENCY Rev.2, 04/30101 RHR Shutdown Cooling System-Hot Shutdown B 3.4.8 B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.8 BASES Residual Heat Removal (RHR)Shutdown Cooling System-Hot Shutdown BACKGROUND APPLICABLE SAFETY ANALYSES LCO Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant.This decay heat must be removed to reduce the temperature of the reactor coolant to::;200°F.This decay heat removal is in preparation for performing refueling or maintenance operations, or for keeping the reactor in the Hot Shutdown condition. The two redundant, manually controlled shutdown cooling subsystems of the RHR System provide decay heat removal.Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves.Both loops have a common suction from the same recirculation loop.Each pumpdischargesthe reactor coolant, after circulation through the respective heat exchanger, to the reactor via the associated recirculation loop.The RHR heat exchangers transfer heat to the RHR Service Water System (LCO 3.7.1,"Residual Heat Removal Service Water (RHRSW)System").Decay heat removal by operation of the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses.Decay heat removal is, however, an important safety function that must be accomplished or core damage could result.The RHR shutdown cooling subsystem satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). Two RHR shutdown coolin subsystems r be OPERABLE.and w no pump IS in 0 Ion 0 shutdow.coo In s stem must bwfO ooeration. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, and the associated pipingandvalves. The two subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping.Thus, to meet the LCO, both pumps in one loop or one pump in each of the two loops must be OPERABLE.Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems. Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local)in the shutdown cooling mode for removal of decay heat.In MODE 3, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. BWRl4 STS B 3.4.8-1 Rev.2, 04/30/01 BASES LCO (continued) APPLICABILITY +0 k;.e S:,.,+RHR Shutdown Cooling System-Hot Shutdown B 3.4.8 oth RHR shutdowJ).Qe01ing subsyste s kJ1iDt be i oper or.an 8 hour eri ote all ws one R shutdown cooling subsystem to be inoperable for up to 2 hours for the performance of Surveillance tests.These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiringLffi2'lZ'!Ow Inter'@tlon) of redundancy. In MODE 3 with reactor steam dome pressure below[the RHR cut in ermissive ressure.., the actual pressure at which the interlock resets)the RHR Syste ay be operated in the shutdown cooling mode to remove decay heat to reduce or maintain coolant reclrcu a Ionrequire 0 e ra Ion. In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to[the RHR cut in permissive pressure], this LCO is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping.Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS)(LCO 3.5.1,"ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation. The requirements for decay heat removal in MODES 4 and 5 are discussed in LCO 3.4.9,"Residual Heat Removal (RHR)Shutdown Cooling System-Cold Shutdown," LCO 3.9.8,"Residual Heat Removal (RHR)-High Water Level," and LCO 3.9.9,"Residual Heat Removal (RHR)-Low Water Level." BWRl4 STS B 3.4.8-2 Rev.2, 04/30/01 BASES ACTIONS RHR ShutdownCoolingSystem -Hot Shutdown B 3.4.8 A Note to the ACTIONS excludes the MODE change restriction of LCO 3.0.4.This exception allows entry into the applicable MODE(S)while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown.This exception is acceptable due to the redundancy of the OPERABLE subsystems, the low pressure at which the plant is operating, the low probability of an event occurring during operation in this condition, and the availability of alternate methods of decay heat removal capability. A second Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems. Section 1.3, Completion Times, specifies 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 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 shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdowncooling subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem.A.1,A.2, and A.3With one required RHR shutdown coolin bsystem inoperable for decay heat removal, except as permitted by CO NoteCi the inoperable subsystem must be restored to OPERABLE status without delay.In this condition, the remaining OPERABLE subsystem can necessary decay heat removal.The overall reliability is reduceci,'-..LJ however, because a single failure in the OPERABLE subsystem could result in reduced RHR shutdown cooling capability. Therefore, an alternate method of decay heat removal must be provided.With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO.The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Decay heat removal by ambient losses can be BWRl4 STS B 3.4.8-3 Rev.2, 04/30/01 RHR Shutdown Cooling System-Hot Shutdown B 3.4.8 BASES ACTIONS (continued) considered as, or contributing to, the alternate method capability. Alternate methods that can be used include (but are not limited to)the Spent Fuel Pool Cooling System and the Reactor Water Cleanup System.However, due to the potentially reduced reliability of the alternate methods of decay heat removal, it is also required to reduce the reactor coolant temperature to the point where MODE 4 is entered.With no RHR shutd n cooling subsystem and no recirculation pump in operation, exceR s permitted by LCO Note 1, reactor coolant circulation by the RHR s down cooling subsystem or recirculation pump must be restored w'out delay.Until R or recirculation pump operation is re-established, an alternate m od of reactor coolant circulation must be placed into service.This II provide the necessary circulation for monitoring coolant temperature. The 1 hour Completion Time is based on the coolant circulation function and is modified such that the 1 hour is applicable separately for each occurrenceinvolvinga loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours thereafter. This will provide a rance of continued temperature monitoring capability. During the period when the reactor coolant is be'circulated by an alternate method (other than by the require R shutdown cooling subsystem or recirculation pump), the re or coolant temperature and pressure must be periodically monitor to ensure proper function of the alternate method.The once per ho Completion Time is deemed appropriate. SURVEILLANCE REQUIREMENTS SR 3.4.8.1 This Surveillance v les that one RHR shutdown cooling bsystem or recirculation pu is in operation and circulating reacto olant.The required flo te is determined by the flow rate ne sary to provide sufficient cay heat removal capability. The Fre ency of 12 hours is suffici in view of other visual and audible in'tions available to the op tor for monitoring the RHR subsystem' the control room.BWR/4 STS B 3.4.8-4 Rev.2, 04/30/01 RHR Shutdown Cooling System-Hot Shutdown B 3.4.8 BASES SURVEILLANCE REQUIREMENTS (continued) REFERENCES None.BWRl4 STS B 3.4.8-5 Rev.2, 04/30/01 ... RHR Shutdown Cooling System-Cold Shutdown B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.9 BASES Residual Heat Removal (RHR)Shutdown Cooling System-Cold Shutdown BACKGROUND APPLICABLE SAFETY ANALYSES LCO Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant.This decay heat must be removed to maintain the temperature of the reactor coolant s 200°F.This decay heat removal is in preparation for performing refueling or maintenance operations, or for keeping the reactor in the Cold Shutdown condition. The two redundant, manually controlled shutdown cooling subsystems of the RHR System provide decay heat removal.Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves.Both loops have a common suction from the same recirculation loop.Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via the associated recirculation loop.The RHR heat exchangers transfer heat to the RHR Service Water System.Decay heat removal by operation of the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses.Decay heat removal is, however, an important safety function that must be accomplished or core damage could result.The RHR Shutdown Cooling System satisfies Criterion 4 of 10 CFR 50.36(c}(2}(ii}. Two RHR shutdown coolin subs stems are re uired to be OPERABLE.an w no is in oper'one RH own In subs stem rn.etSfbe in operation. An OPERABLE RHR shutdown cooling subsystem consists 0 one ERABLE RHR pump, one heat exchanger, and the associated piping and valves.The two subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping.Thus, to meet the LCO, both pumps in one loop or one pump in each of the two loops must be OPERABLE.Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems. In MODE 4, the RHR cross tie valve (2E11-F010) may be opened to allow pumps in one loop to discharge through the opposite recirculation loop to make a complete subsystem. Additionally, each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local)in the shutdown cooling mode for removal of decay heat.In MODE 4, one RHR shutdown cooling BWR/4 STS B 3.4.9-1 Rev.2, 04/30/01 BASES LCO (continued) APPLICABI L1TY/s to-fh.+r'of RHR Shutdown Cooling System-Cold Shutdown B 3.4.9 RHR shutdown cooling subsystem to be inoperable for up to 2 hours for the performance of Surveillance tests.These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations ) of redundancy. In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive pressure, this LCO is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping.Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS)(LCO 3.5.1,"ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation. The requirements for decay heat removal in MODE 3 below the cut in permissive pressure and in MODE 5 are discussed in LCO 3.4.8,"Residual Heat Removal (RHR)Shutdown Cooling System-Hot Shutdown," LCO 3.9.8,"Residual Heat Removal (RHR)-High Water Level," and LCO 3.9.9,"Residual Heat Removal (RHR)-Low Water Level." BWRl4 STS B 3.4.9-2 Rev.2, 04/30/01 RHR Shutdown Cooling System-Cold Shutdown B 3.4.9 BASES ACTIONS A Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems. Section 1.3, Completion Times, specifies 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 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 shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem. A.1With one of the two required RH hutdown co ling subsystems inoperable, except by CO Note£(he remaining subsystem is capable of provldln 15required decay heat removal.However, the overall reliability is reduced.Therefore, an alternate method of decay heat removal must be provided.With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO.The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these alternate method(s}must be reconfirmed every 24 hoursthereafter.This will provide assurance of continued heat removal capability. The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability. Alternate methods that can be used include (but are not limited to)the Spent Fuel Pool Cooling System and the Reactor Water Cleanup System.B.1 and B.2 With no RH utdown cooling subsystem and no re culation pump in operation xcept as permitted by LCO Note 1, an ntil RHR or recirc ion pump operation is re-established, alternate method of rea r coolant circulation must be placed in service.This will provide necessary circulation for monitoring c ant temperature. The 1 hour BWRl4 STS B 3.4.9-3 Rev.2, 04/30/01-------------------------------_._-------------------------------- RHRShutdownCooling System-Cold Shutdown B 3.4.9 BASES ACTIONS (continued) .------------------- During the period when the reactor coolant i eing circulated by an alternate method (other than by the requi f:l RHR Shutdown Cooling System or recirculation pump), the re or coolant temperature and pressure must be periodically mon'red to ensure proper function alternate method.The once pe our Completion Time is dee appropriate. Completion Time'ased on the coolant circulation function and is modified such at the 1 hour is applicable separately for each occurrenc.volving a 1055 of coolant circulation. Furthermore verifica.of the functioning of the alternate method must reco rmed every 12 hours thereafter. This will provid ssurance of c inued temperature monitoring capability, SURVEILLANCE REQUIREMENTS SR 3.4.9.1 This Surveillance verifie one RHR shutdown cooling 5 bsystem or recirculation pumR.operation and circulating reacto olant.The required flow is determined by the flow rate ne sary to provide sufficient cay heat removal capability. The Fr uencyof 12 hours is suffic'in view of other visual and audible i lcations available to the o ator for monitoring the RHR subs ste in the control room.REFERENCES None.BWRl4 STS B 3.4.9-4 Rev.2, 04/30/01 RHR-High Water Level B 3.9.8 B 3.9 REFUELING OPERATIONS B 3.9.8 BASES Residual Heat Removal (RHR)-High Water Level BACKGROUND APPLICABLE SAFETY ANALYSES LCO The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34.Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal.Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves.Both loops have a common suction from the same recirculation loop.Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop or to the reactor via the low pressure coolant injection path.The RHR heat exchangers transfer heat to the RHR Service Water System.The RHR shutdown cooling mode is manually controlled. In addition to the RHR subsystems, the volume of water above the reactor pressure vessel (RPV)flange provides a heat sink for decay heat removal.With the unit in MODE 5, the RHR System is not required to mitigate any events or accidents evaluated in the safety analyses.The RHR System is required for removing decay heat to maintain the temperature of the reactor coolant.The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). Onl one RHR shutdown cooling subsystem is required to be OPERABLE an e'on'n MODE 5 with irradiated fuel in the RPV and the water evel[23]above the RPV flange.Only one subsystem is required because the volume of water above the RPV flange provides backup decay heatremovalcapability. An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.In MODE 5, the RHR cross tie valve is not required to be closed;thus, the valve may be opened to allow pumps in one loop to discharge through the opposite loop's heat exchanger to make a complete subsystem. Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local)in the shutdown cooling mode for removal of decay heat.Operation (either BWRl4 STS B 3.9.8-1 Rev.2, 04/30101 BASES LCO (continued) APPLICABILITY ACTIONS RHR-High Water Level B 3.9.8 One RHR shutdown cooling subsystem must be MODE 5, with irradiated fuel in the reactor pressure vessel and with the water level[23]feet abovethetop of the RPV flange, to provide decay heat removal.RHR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS);Section 3.5, Emergency Core Cooling Systems (ECCS)and Reactor Core Isolation Cooling (RCIC)System;and Section 3.6, Containment Systems.RHR Shutdown Cooling System requirements in MODE 5 with irradiated fuel in the reactor pressure vessel and with the water level<[23]ft above the RPV flange are given in LCO 3.9.9.With no RHR shutdown cooling subsystem OPERABLE, an alternate method of decay heat removal must be established within 1 hour.In this condition, the volume of water above the RPV flange provides adequate capability to remove decay heat from the reactor core.However, the overall reliability is reduced because loss of water level could result in reduced decay heat removal capability. The 1 hour Completion Time is based on decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these alternate method(s)must be reconfirmed every 24 hours thereafter. This will ensure continued heat removal capability. Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. For example, this may include the use of the Reactor Water Cleanup System, operating with the regenerative heat exchanger bypassed.The method used to remove the decay heat should be the most prudent choice based on unit conditions. BWRl4 STS B 3.9.8-2 Rev.2, 04/30101 RHR-High Water Level B 3.9.8 BASES ACTIONS (continued) B.1.B.2.B.3.and B.4 If no RHR shutdown cooling subsystem is OPERABLE and an alternate method of decay heat removal is not available in accordance with Required Action A.1, actions shall be taken immediately to suspend operations involving an increase in reactor decay heat load by suspending loading of irradiated fuel assemblies into the RPV.Additional actions are required to minimize any potential fission product release to the environment. This includes ensuring secondary containment is OPERABLE;one standby gas treatment subsystem is OPERABLE;and secondary containment isolation capability (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each associated penetration not isolated that is assumed to be isolated to mitigate radioactive releases.This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons.It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status.In this case, a surveillance may need to be performed to restore the component to OPERABLE status.Actions must continue until all required components are OPERABLE.Ing the period when the reactor coolant is being c'ulated by an Iternate method (other than by the required RHR utdown Cooling System), the reactor coolant temperature must periodically monitored to ensure proper functioning of the alternate thod.The once per hour Completion Time is deemed appropriate. C.1 and C.2 If no RHR Shutd Cooling System is in operation, an alternate of coolant circ tion is required to be established within 1 hour.e Completio ime is modified such that the 1 hour is applicabl eparately for eac ccurrenceinvolvinga loss of coolant circulation. BWRl4 STS B 3.9.8-3 Rev.2, 04/30/01 tion BASES SURVEILLANCE REQUIREMENTS RHR-High Water Level B 3.9.8 SR 3.9.8.1 This Surveillance demonstrat and circulating reactor nt.The required fl rate is determined by the flow rate cessary to provide sufficient y heat removal capability. The Fre ency of 12 hours is suffici In view of other visual and audible in'tions available to the tor for monitoring the RHR subsystem' the control room.REFERENCES None.BWRl4 STS B 3.9.8-4 Rev.2, 04/30/01 RHR-Low Water Level B 3.9.9 B 3.9 REFUELING OPERATIONS B 3.9.9 BASES Residual Heat Removal (RHR)-Low Water Level BACKGROUND APPLICABLE SAFETY ANALYSES LCO The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34.Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal.Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves.Both loops have a common suction from the same recirculation loop.Each pumpdischargesthe reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop or to the reactor via the low pressure coolant injection path.The RHR heat exchangers transfer heat to the RHR Service Water System.The RHR shutdown cooling mode is manually controlled. With the unit in MODE 5, the RHR System is not required to mitigate any events or accidents evaluated in the safety analyses.The RHR System is required for removing decay heat to maintain the temperature of the reactor coolant.The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV)and the water level<23 ft above the reactor pressure vessel (RPV)flange both RHR shutdown cooling subsystems must be OPERABLE.An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.To meet the LCO, both pumps in one loop or one pump in each of the two loops must be OPERABLE.In MODE 5, the RHR cross tie valve is not required to be closed;thus, the valve may be opened to allow pumps in one loop to discharge through the opposite loop's heat exchanger to make a complete subsystem. Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local)in the shutdown cooling mode for removal of decay heat.Operation (either continuous or intermittent) of one subsystem n maintain and reduce the reactor coolant temperature as required.However, to ensure qua e core flo ow or accurate e reactor coolant t eraturem.ring, nearly continu 0 eration is re uired.ote is provided to BWRl4 STS B 3.9.9-1 Rev.2, 04/30/01 BASES LCO (continued) APPLICABILITY ACTIONS RHR-Low Water Level B 3.9.9 be in subsystems are required to be and QAt'fmust be.in MODE 5, with irradiated fuel in the RPV an WIte water level<[23]ft above the top of the RPV flange, to provide decay heat removal.RHR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS);Section 3.5, Emergency Core Cooling Systems (ECCS)and Reactor Core Isolation Cooling (RCIC)System;and Section 3.6, Containment Systems.RHR Shutdown Cooling System requirements in MODE 5 with irradiated fuel in the RPV and with the water level[23]ft above the RPV flange are given in LCO 3.9.8,"Residual Heat Removal (RHR)-High Water Level." With one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem is capable of providing the required decay heat removal.However, theoverallreliability is reduced.Therefore an alternate method of decay heat removal must be provided.With both required RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO.The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of this alternate method(s)must be reconfirmed every 24 hours thereafter. This will ensure continued heat removal capability. Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. For example, this may include the use of the Reactor Water Cleanup System, operating with the regenerative heat exchanger bypassed.The method used to remove decay heat should be the most prudent choice based on unit conditions. BWRJ4 STS B 3.9.9-2 Rev.2, 04/30/01 RHR-Low Water Level B 3.9.9 BASES ACTIONS (continued)B.1.B.2.and B.3 With the required decay heat removal subsystem(s} inoperable and the required alternate method(s}of decay heat removal not available in accordance with Required Action A.1, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring secondary containment is OPERABLE;one standby gas treatment subsystem is OPERABLE;and secondary containment isolation capability (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each associated penetration not isolated that is assumed to be isolated to mitigate radioactive releases.This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons.It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status.In this case,thesurveillance may need to be performed to restore the component to OPERABLE status.Actions must continue until all required components are OPERABLE.C.1 and C.2 If no RHR subsystemi.operation, an alternate method of co ant circulation is requir 0 be established within 1 hour.The pletion Time is modified ch that the 1 hour is applicable separ Iy for each Iving a loss of coolant circulation. During e period when the reactor coolant is bein circulated by an alter te method (other than by the required R Shutdown Cooling S tern), the reactor coolant temperature mu e periodically monitored o ensure proper functioning of the alternate ethod.The once per hour Completion Time is deemed appropriate. SR 3.9.9.1 This Surveillance de rates that one RHR shutdown subsystem is in ration and circulating reactor coola.The required flow rate is ermined by the flow rate necessary t rovide sufficient t removal capabili.BWRl4 STS B 3.9.9-3 Rev.2, 04/30/01 RHR-Low Water Level B 3.9.9 BASES SURVEILLANCE REQUIREMENTS (continued) The Frequency 0 ours is sufficient in view of oth ual and audible indications able to the operator for monitori e RHR subsystems in the rol room.REFERENCES None.BWRl4 STS B 3.9.9-4 Rev.2, 04/30/01 RHR Shutdown Cooling System-Hot Shutdown 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.9 Residual Heat Removal (RHR)Shutdown Cooling System-Hot Shutdown LCO 3.4.9 APPLICABILITY: ACTIONS Two RHR shu n coolin su s shall be nd no pumP)l'r operation, at lea ne RHR shutdown shalloperation. One RHR shutdown cooling subsystem may be inoperable for up to hours for performance of Surveillances. MODE 3 with reactor steam dome pressure<[the RHR cut in permissive pressure]. .NOTES*1.LCO 3.0.4 is not applicable. 2.Separate Condition entry is allowed for each RHR shutdown cooling subsystem. CONDITION REQUIRED ACTION COMPLETION TIME A.One or two RHR A.1 shutdown cooling subsystems inoperable. Initiate action to restore Immediately RHR shutdown cooling subsystem to OPERABLE status.BWRl6STS 3.4.9-1 Rev.2, 04/30101 RHR Shutdown Cooling System-Hot Shutdown 3.4.9 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A.2 Verify an alternate method 1 hour of decay heat removal is available for each inoperable RHR shutdown cooling subsystem. AND A.3 Be in MODE 4.24 hours B.No RHR shutdown Initiate action to restore mmediately cooling subsystem in one RHR shutdown operation. cooling subsystem or one recirculation pump to AND operation. AND B.2 Verify reactor coolant circulation by an alternate method.Once per 12 hours thereafter Monito eactor coolant Once per hour tem rature and pressure.BWRl6STS 3.4.9-2 Rev.2, 04/30/01 RHR Shutdown Cooling System-Hot Shutdown 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.9.1-NOTE-Not required to be met until 2 hours after reactor steam dome pressure is<[the RHR cut in permissive pressure]. eacltl ("erujfe!J-- __FREQUENCY BWRl6STS 3.4.9-3 Rev.2, 04/30/01 RHR Shutdown Cooling System-Cold Shutdown 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.10 Residual Heat Removal (RHR)Shutdown Cooling System-Cold Shutdown LCO 3.4.10 1.One RHR shutdown cooling subsystem may be inoperable for up to 2 hours for the performance of Surveillances. APPLICABILITY: ACTIONS MODE 4..NOTE*Separate Condition entry is allowed for each RHR shutdown cooling subsystem. CONDITION REQUIRED ACTION COMPLETION TIME A, One or two RHR A,1 shutdown cooling subsystems inoperable. Verify an alternate method 1 hour of decay heat removal is available for each AND inoperable RHR shutdown cooling subsystem. Once per 24 hours thereafter BWRl6 STS 3.4.10-1 Rev.2, 04/30/01 RHR Shutdown Cooling System-Cold Shutdown 3.4.10 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.No RHR shutdown cooling subsystem in operation. No recir ation pump in oper n.B.2 Verify reactor coolant circulating by an alternate method.1 hour from dis ery of no reactor olant circulatio Once per hour SURVEILLANCE REQUIREMENTS FREQUENCY RHR shutdown cooling subsystem@} n SR 3.4.10.1 each Y6 ,,/rrJ SURVEILLANCE BWRl6STS 3.4.10-2 Rev.2, 04/30101 RHR-High Water Level.3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)-High Water Level LCO 3.9.8 One RHR shutdown cooling subsystem shall be.APPLICABI L1TY: ACTIONS._--------------------
.NOTE*:;t The required shutdown cooling subsyst may be not in op tion for up to 2 urs per 8 hour period.._------------------------------------
MODE 5 with irradiated fuel in the reactor pressure vessel (RPV)and with the water level[22 ft 8 inches]abovethetop of the[reactor pressure vessel (RPV)flange].CONDITION REQUIRED ACTION COMPLETION TIME A, Required RHR shutdown A,1 Verify an alternate method 1 hour cooling subsystem of decay heat removal is inoperable. available. AND Once per 24 hours thereafter B.Required Action and B.1 Suspend loading irradiated Immediately associated Completion fuel assemblies into the Time of Condition A not RPV.met.AND B.2 Initiate action to restore Immediately [primary or secondary] containment to OPERABLE status.AND BWRl6STS 3.9.8-1 Rev.2, 04/30/01 ACTIONS (continued) CONDITION RHR-High Water Level 3.9.8 REQUIRED ACTION COMPLETION TIME B.3 Initiate action to restore Immediately one standby gas treatment subsystem to OPERABLE status.B.4 Initiate action to restore Immediately isolation capability in each required secondary containment penetration flow path not isolated.C.No RHR shutdown cooling subsystem in operation. C.1 C.2 Ve'reactor coolant culation by an alternate method.Monitor rea temperatur .Once per 12 hours thereafter Once per hour SURVEILLANCE REQUIREMENTS SR 3.9.8.1 BWRl6STS SURVEILLANCERHR shutdown cooling subsystem@? pe , 3.9.8-2 FREQUENCY Rev.2, 04/30/01 RHR-Low Water Level 3.9.9 3.9 REFUELING OPERATIONS 3.9.9 Residual Heat Removal (RHR)-Low Water Level LCO 3.9.9 APPLICABI L1TY: ACTIONS Two RHR shutdown cooling subsystems shall be OPERABLE.@) i!§!In MODE 5 with irradiated fuel in the reactor pressure vessel and with the water level<[23]ft abovethetop of the[reactor pressure vessel flange].CONDITION REQUIRED ACTION COMPLETION TIME A.One or two RHR A.1 Verify an alternate method 1 hour shutdown cooling of decay heat removal is subsystems inoperable. available for each AND inoperable RHR shutdown cooling subsystem. Once per 24 hours thereafter B.Required Action and B.1 Initiate action to restore Immediately associated Completion [primary or secondary] Time of Condition A not containment to met.OPERABLE status.AND B.2 I nitiate action to restore Immediately one standby gas treatment subsystem to OPERABLE status.AND BWRl6 STS 3.9.9-1 Rev.2, 04/30/01 RHR-Low Water Level 3.9.9 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.3 Initiate action to restore Immediately isolation capability in each required secondary containment penetration flow path not isolated.C.No RHR shutdown cooling subsystem in operation. C.1 C.2 rify reactor coolant circulation by an alternate method.Once per hour SR 3.9.9.1 BWRl6STS SURVEILLANCE RHR shutdown cooling 3.9.9-2 FREQUENCY Rev.2, 04/30/01 RHR Shutdown Cooling System-Hot Shutdown B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.9 BASES Residual Heat Removal (RHR)ShutdownCoolingSystem -Hot Shutdown BACKGROUND APPLICABLE SAFETY ANALYSES LCO Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant.This decay heat must be removed to reduce the temperature of the reactor coolant to s 200°F.This decay heat removal is in preparation for performing refueling or maintenance operations, or for keeping the reactor in the Hot Shutdown condition. The two redundant, manually controlled shutdown cooling subsystems of the RHR System provide decay heat removal.Each loop consists of a motor driven pump, two heat exchangers in series, and associated piping and valves.Both loops have a common suction from the same recirculation loop.Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via separate feedwater lines or to the reactor via the LPCI injection path.The RHR heat exchangers transfer heat to the Standby Service Water System (LCO 3.7.1, U[Standby Service Water (SSW)]System and[Ultimate Heat Sink (UHS)]U).Decay heat removal by the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses.Decay heat removal is, however, an important safety function that must be accomplished or core damage could result.The RHR Shutdown Cooling System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). BWRl6STS B 3.4.9-1 Rev.2, 04/30/01 RHR Shutdown Cooling System-Hot Shutdown B 3.4.9 BASES LCO (continued) Note 1 permit RHR shutdown cooli ystems a irculation urnst be in operation for a'i i d. allows oneRs utdown cooling subsystem to be inoperable for up to 2 hours for performance of surveillance tests.These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations analloss of redundancy. APPLICABI L1TY I'.,) 50I}In MODE 3 with reactor steam dome pressure below the RHR cut in permissive pressure (Le., the actual pressure at which the interlock reset wn Coolin S ste may be operated in the shutdown cooling mode to remove decay heat to reduce or maintain coolant a PUillp Is to be i'9 In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive pressure, this LCO is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping.Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS)(LCO 3.5.1,"ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation. The requirements for decay heat removal in MODES 4 and 5 are discussed in LCO 3.4.10,"Residual Heat Removal (RHR)Shutdown Cooling System-Cold Shutdown," LCO 3.9.8,"Residual Heat Removal (RHR)-High Water Level," and LCO 3.9.9,"Residual Heat Removal (RHR)-Low Water Level." ACTIONS A Note to the ACTIONS excludes the MODE change restriction of LeO 3.0.4.This exception allows entry into the applicable MODE(S)while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown.This exception is acceptable due to the redundancy of the OPERABLE subsystems, the low pressure at which the plant is operating, the low probability of an event occurring during BWRl6STS B 3.4.9-2 Rev.2, 04/30101 RHR Shutdown Cooling System-Hot Shutdown B 3.4.9 BASES ACTIONS (continued) operation in this condition, and the availability of alternate methods of decay heat removal capability. A second Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems. Section 1.3, Completion Times, specifies 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 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 shutdown cooling subsystems provide appropriate compensatory measures forseparateinoperable shutdown cooling subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem. A.1 , A.2, and A.3With one required RHR shutdown cooling subsystem ina rable for decay heat removal, except as permitted by CO inoperable subsystem must be restored to OPERABLE status without delay.In this condition, the remaining OPERABLE subsystem can necessary decay heat removal.The overall reliability is reduced, CV however, because a single failure in the OPERABLE subsystem could result in reduced RHR shutdown cooling capability. Therefore an alternate method of decay heat removal must be provided.With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO.The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability. Alternate methods that can be used include (but are not limited to)the Spent Fuel Pool Cooling System or the Reactor Water Cleanup System.BWRl6STS B 3.4.9-3 Rev.2, 04/30101 RHR Shutdown Cooling System-Hot Shutdown B 3.4.9 BASES ACTIONS (continued) However, due to the potentially reduced reliability of the alternate methods of decay heat removal, it is also required to reduce the reactor coolant temperature to the point where MODE 4 is entered.B.1.B.2, and B.3 With no RHR shutdown Ing subsystem and no recirculation pump in operation, except as i ermitted by LCO Note 1, reactor coolant circulation by the shutdown cooling subsystem or one recirculation pump must be r: to red without delay.Until RH r recirculation pump operation is re-established, an alternate meth of reactor coolant circulation must be placed into service.This wil rovide the necessary circulation for monitoring coolant temperature.e1 hour Completion Time is based on the coolant circulation function and is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours thereafter. This will provide ass continued temperature monitoring capability. During the period when the reactor coolant is bein irculated by an alternate method (other than by the required R shutdown cooling subsystem or recirculation pump), the react coolant temperature and pressure must be periodically monitored ensure proper function of the alternate method.The once per hour ompletion Time is deemed appropriate. SURVEILLANCE SR 3.4.9.1 REQUIREMENTS BWRJ6STS B 3.4.9-4 Rev.2, 04/30/01 RHR Shutdown Cooling System-Cold Shutdown B 3.4.10 B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.10 Residual Heat Removal (RHR)Shutdown Cooling System-Cold Shutdown BASES BACKGROUND APPLICABLE SAFETY ANALYSES LCO Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant.This decay heat must be removed to maintain the temperature of the reactor coolant at200°F.This decay heat removal is in preparation for performing refueling or maintenance operations, or for keeping the reactor in the Cold Shutdown condition. The two redundant, manually controlled shutdown cooling subsystems of the RHR System provide decay heat removal.Each loop consists of a motor driven pump, two heat exchangers in series, and associated piping and valves.Both loops have a common suction from the same recirculation loop.Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via separate feedwater lines or to the reactor via the LPCI injection path.The RHR heat exchangers transfer heat to the Standby Service Water System.Decay heat removal by the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses.Decay heat removal is, however, an important safety function that must be accomplished or core damage could result.The RHR Shutdown Cooling System satisfies Criterion 4 of 10 CFR 50.36(c}(2}(ii}. BWRl6STS B 3.4.10-1 Rev.2, 04/30101 RHR Shutdown Cooling System-Cold Shutdown B 3.4.10 BASES LCO (continued) Note 1 P R shut own g su systems andr.ulation fi\;"...our eriod.&-Note allows one RHR shutdown cooling subsystem to be inoperable for.up to 2 hours for performance of surveillance tests.These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations ar}SDloss of redundancy. APPLICABILITYfeet u.,I'ftd-b bL ope;e A 8LE.-'t>-{-hA-it-ACTIONS In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive pressure, this LCO is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping.Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS)(LCO 3.5.1,"ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation. The requirements for decay heat removal in MODE 3 below the cut in permissive pressure and in MODE 5 are discussed in LCO 3.4.9,"Residual Heat Removal (RHR)Shutdown Cooling System-Hot Shutdown," LCO 3.9.8,"Residual Heat Removal (RHR)-High Water Level," and LCO 3.9.9,"Residual Heat Removal (RHR)-Low Water Level." A Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems. Section 1.3, CompletionTimes,specifies 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 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 BWRl6 STS B 3.4.10-2 Rev.2, 04/30/01 RHR Shutdown Cooling System-Cold Shutdown B 3.4.10 BASES ACTIONS (continued) for inoperable shutdown cooling subsystems provided appropriate compensatory measures for separate inoperable shutdown cooling subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem. A.1-TY'e With one of the two required RHR shutdown cooling subsystems inoperable except as permitted b CO Note;t'the remaining subsystem is capable of providing the required decay heat removal.However, the overall reliability is reduced.Therefore, an alternate method of decay heat removal must be provided.With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO.The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these alternate method(s)must be reconfirmed every 24 hours thereafter. This will provide assurance of continued heat removal capability. The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Decay heat removal by ambient losses can be considered as, or contributing to the alternate method capability. Alternate methods that can be used include (but are not limited to)the Spent Fuel Pool Cooling System or the Reactor Water Cleanup System.B.1 and B.2 With no RHR shutdown coolin ubsystem and no recirculation ump in operation, except as is per ed by LCO Note 1, and until RH or recirculation pump ope Ion is re-established, an alternate ethod of reactor coolant circ ion must be placed into service.Ts will provide the necessary ci lation for monitoring coolant tempe ture.The 1 hour Completion r e is based on the coolant circulation nction and is modifiedsh that the 1 hour is applicable separa Iy for each occurr e involving a loss of coolant circulatio .Furthermore, veri*ation of the functioning of the alternate ethod must be r nfirmed every 12 hoursthereafter.Thi ill provide assurance of continued temperature monitoring capabili.BWRl6 STS B 3.4.10-3 Rev.2, 04/30101 RHR Shutdown Cooling System-Cold Shutdown B 3.4.10 BASES ACTIONS (continued) During the period when eactor coolant is being circul alternate method(r than by the required RHR shut n cooling system or recir ation pump), the reactor coolant te erature and pressure be periodically monitored to ensure oper function of the alterna ethod.The once per hour Completia ime is deemed ap priate.SURVEILLANCE REQUIREMENTS SR 3.4.10.1 This Surveillance verifies tha e RHR shutdown cooling su ystem or r IA Le,f:2 recirculation pump is in 0 ation and circulating reactor lant.The , ,--required flow rate is d rmined by the flow rate necess to provide sufficient decay he removal capability. The Frequen of 12 hours is sufficient in vie f other visual and audible indicatio s available to the operator for onitorin the RHR subsystem in the ntrol room.REFERENCES None.BWRl6STS B 3.4.10-4 Rev.2, 04/30/01 BASES LCO (continued) APPLICABILITY ACTIONS RHR-High Water Level B 3.9.8 One RHR shutdown cooling subsystem must be OPERABLE in MODE 5, with irradiated fuel in the RPV and with the water level[22 ft 8 inches]above the top of the RPV flange, to provide decay heat removal.RHR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS);Section 3.5, Emergency Core Cooling Systems (ECCS)and Reactor Core Isolation Cooling (RCIC)System;and Section 3.6, Containment Systems.RHR ShutdownCoolingSystem requirements in MODE 5, with irradiated fuel in the reactor pressure vessel and with the water level<[22 ft 8 inches]above the RPV flange, are given in LCO 3.9.9,"Residual Heat Removal (RHR)Low Water Level." With no RHR shutdown cooling subsystem OPERABLE, an alternate method of decay heat removal must be established within 1 hour.In this condition, the volume of water above the RPV flange provides adequate capability to remove decay heat from the reactor core.However, the overall reliability is reduced because loss of water level could result in reduced decay heat removal capability. The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these alternate method(s)must be reconfirmed every 24 hours thereafter. This will ensure continued heat removal capability. Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. For example, this may include the use of the Reactor Water Cleanup System, operating with the regenerative heat exchanger bypassed.The method used to remove the decay heat should be the most prudent choice based on unit conditions. BWRl6 STS B 3.9.8-2 Rev.2, 04/30/01 RHR-High Water Level B 3.9.8 BASES ACTIONS (continued) B.1!B.2.B.3.and B.4 If no RHR shutdown cooling subsystem is OPERABLE and an alternate method of decay heat removal is not available in accordance with Required Action A.1, actions shall be taken immediately to suspend operations involving an increase in reactor decay heat load by suspending the loading of irradiated fuel assemblies into the RPV.Additional actions are required to minimize any potential fission product release to the environment. This includes ensuring secondary containment is OPERABLE;one standby gas treatment subsystem is OPERABLE;and secondary containment isolation capability (i.e., one secondary containment isolation valve and associated instrumentation are operable or other acceptable administrative controls to assure isolation capability) in each associated penetration not isolated that is assumed to be isolated to mitigate radioactivity releases.This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons.It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component isinoperable,then it must be restored to OPERABLE status.In this case, a surveillance may need to be performed to restore the component to OPERABLE status.Actions must continue until all required components are OPERABLE.C.1 and C.2 If no RHR shutdown co g subsystem is in operation, an alte e method of coolant c'lation is required to be established w*In 1 hour.The Completion.e is modified such that 1 hour is appl" ble separately for ch occurrenceinvolvinga loss of cool circulation. During th eriod when the reactor coolant is bei alterna method (other than by the required R Shutdown Cooling Syst), the reactor coolant temperature mu be periodically monitored to sure proper functioning of the alternat ethod.The once per hour mpletion Time is deemed appropriate. BWRl6 STS B 3.9.8-3 Rev.2, 04/30/01 BASES SURVEILLANCE REQUIREMENTS (S:n5 v i]REFERENCES RHR-High Water Level B 3.9.8 SR 3.9.8.1 at the RHR subsystem is'operation and circulating reactor co t.The required flow rate is termined by the flow rate necessa 0 provide sufficient decay he emoval capability. The Fr ency of 12 hours is sufficient i iew of other visual and audible in'tions available to the operator f monitoring the RHR subsystem' he control room.None.BWRl6 STS B 3.9.8-4 Rev.2, 04/30101 RHR-Low Water Level B 3.9.9 B 3.9 REFUELING OPERATIONS B 3.9.9 BASES Residual Heat Removal (RHR)-Low Water Level BACKGROUND APPLICABLE SAFETY ANALYSES LCO The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34.Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal.Each loop consists of one motor driven pump, a heat exchanger, and associated piping and valves.Both loops have a common suction from the same recirculation loop.Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via separate feedwater lines, to the upper containment pool via a common single flow distribution sparger, or to the reactor via the low pressure coolant injection path.The RHR heat exchangers transfer heat to the Standby Service Water System.The RHR shutdown cooling mode is manually controlled. With the unit in MODE 5, the RHR System is not required to mitigate any events or accidents evaluated in the safety analyses.The RHR System is required for removing decay heat to maintain the temperature of the reactor coolant.The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV)and with the water level<22 ft 8 inches above the RPV flange both RHR shutdown cooling subsystems must be OPERABLE.An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.BWRl6STS B 3.9.9-1 Rev.2, 04/30/01 RHR-Low Water Level B 3.9.9 BASES ACTIONS (continued) associated instrumentation are operable or other acceptable administrative controls to assure isolation capability) in each associated penetration not isolated that is assumed to be isolated to mitigate radioactivity releases.This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons.It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status.In this case, a surveillance may need to be performed to restore the component to OPERABLE status.Actions must continue until all required components are OPERABLE.C.1 and C.2 During the per*when the reactor coolant is being circulate y an od (other than by the required RHR Shutdo Cooling System e reactor coolant temperature must be peri Ically monitored to en re proper function of the alternate method.Te once per hour Co pletion Time is deemed appropriate. If no RHR shutdown cooling su stem is in operation, an alternate method of coolant circulatio .required to be established within 1 hour.The Completion Time is dified such that the 1 hour is applicable separately for each urrence involving a loss of coolant circulatio SURVEILLANCE SR 3.9.9.1 REQUIREMENTS __ REFERENCES None.BWRl6STS B 3.9.9-3 Rev.2, 04/30101 TSTF-460, Rev. 0 BWROG-90, Rev. 1 NUREGs Affected: Control Rod Scram Time Testing Frequency Technical Specification Task Force Improved Standard Technical Specifications Change Traveler 1430 1431 1432 1433 1434 Classification:
- 1) Technical Change Recommended for CLIIP?:
Industry Contact: Tom Silko, (802) 258-4146, tsilko@entergy.com Yes Correction or Improvement: Improvement 1.0 Description The proposed Traveler changes NUREG -1433 (BWR/4) and NUREG -1434 (BWR/6) by revising the Frequency of SR 3.1.4.2, control rod scram time testing, from "120 days cumulative operation in MODE 1" to "[200] days cumulative operation in MODE 1." The Bases are revised to limit the percentage of the tested rods which can be "slow" from 20% to 7.5%. 2.0 Proposed Change NUREG-1433, SR 3.1.4.2 states, " Verify, for a representative sample, each tested control rod scram time is within the limits of Table 3.1.4 -1 with reactor steam dome pressure > [800] psig." NUREG-1434, SR 3.1.4.2 states, " Verify, for a representative sample, each tested control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure > [950] psig." Both SRs have a Frequency of "120 days cumulative operation in MODE 1." The proposed change revises the Frequency to "[200] days cumulative operation in MODE 1." The Bases are revised to reference the new Frequency and to reduce the percentage of the tested rods which can be "slow" from 20% to 7.5%.
3.0 Background
Control rod scram time is verified following each refueling. 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 (e.g., 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 LCO limit. For planned testing, the control rods selected for the sample should be different for each test. 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-460, Rev. 0 BWROG-90, Rev. 1 4.0 Technical Analysis Industry operating experience has shown the control rod scram rates to be highly reliable. For example, at the Grand Gulf Nuclear Station, out of 7,660 control rod insertion tests, only 12 control rods have been slower than the insertion time limit (with the exception of test data from an anomalous cycle). The control rod drive
system has shown to be highly reliable. This high reliability supports the extension of the Surveillance
Frequency from 120 days of cumulative operation in MODE 1 to 200 days. The current TS Bases states that the acceptance criteria have been met if 20 percent or fewer of the random sample control rods that are tested within the 120 day surveillance period are found to be slow. The Bases are
revised to change the control rod insertion time acceptance criterion for percentage of slow rods allowed, be
reduced to 7.5 percent of the random at -power surveillance sample when the surveillance period is extended to 200 cumulative days of operation in MODE 1. The more restrictive 7.5 percent acceptance criterion for testing the random sample is consistent with the TS 3.1.4 objective of ensuring that no more than 14
OPERABLE control rods are slow at any given time. Plants submitting amendments to extend the Surveillance Frequency should demonstrate the reliability of the control rod insertion system, based on historical control rod scram time test data, and by the more restrictive
acceptance criterion for the number of slow rods allowed during at -power surveillance testing. The justification provided should be comparable to that used in References 1 and 2. The proposed change is consistent with the amendment requests in References 1, 2, and 3 and the NRC's approvals in References 4 and 5. 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-460, Rev. 0 BWROG-90, Rev. 1 5.0 Regulatory Analysis 5.1 No Significant Hazards Consideration The TSTF has evaluated whether or not a significant hazards consideration is involved with the proposed generic change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as
discussed below:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No. The proposed change extends the Frequency for testing control rod scram time testing from every 120 days of cumulative MODE 1 operation to [200] days of cumulative MODE 1 operation. The Frequency of Surveillance testing is not an initiator of any accident previously evaluated. The Frequency of Surveillance testing does not affect the ability to mitigate any accident previously evaluated, as the tested component is still required to be OPERABLE. Therefore, the proposed change does not involve a significant increase in the probability or 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?
Response: No. The proposed change extends the Frequency for testing control rod scram time testing from every 120 days of cumulative MODE 1 operation to [200] days of cumulative MODE 1 operation. The proposed change does not result in any new or different modes of plant operation. Therefore, the proposed change
does not create the possibility of a new or different kind of accident from any previously evaluated. 3.Does the proposed change involve a significant reduction in a margin of safety? Response: No.
The proposed change extends the Frequency for testing control rod scram time testing from every 120 days of cumulative MODE 1 operation to [200] days of cumulative MODE 1 operation. The proposed
change continues to test the control rod scram time to ensure the assumptions in the safety analysis are
protected. Therefore, the proposed change does not involve a significant reduction in a margin of safety. 5.2 Applicable Regulatory Requirements / Criteria The proposed change does not affect any OPERABILITY requirements and the test Frequency being revised is not specified in regulations. As a result, no regulatory requirements or criteria are affected. 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-460, Rev. 0 BWROG-90, Rev. 1 6.0 Environmental Consideration The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed
amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment
need be prepared in connection with the proposed amendment. 7.0 References
- 1. Letter from William A. Eaton, Entergy Operations, Inc. (GNRO
-2001/00002) to NRC, "Grand Gulf Nuclear Station, Docket No. 50 -416, License No. NPF -29, Control Rod Scram Time Testing Frequency, Proposed Amendment to the Operating License, LDC 2001 -001," dated January 25, 2001.
- 2. Letter from William A. Eaton, Entergy Operations, Inc. (GNRO
-2002/00012) to NRC, "Grand Gulf Nuclear Station, Docket No. 50 -416, Supplement to Amendment Request Concerning Control Rod Scram Time Testing Frequency," dated February 20, 2002.
- 3. Letter from William R. Brian, Entergy Operations, Inc. (LAR 2001
-35) to NRC, "River Bend Station, Unit 1, Docket No. 50 -458, License Amendment Request, Control Rod Scram Time Testing Frequency," dated July 10, 2002.4. Letter from S. Patrick Sekerak, NRC, to Mr. William A. Eaton, Entergy Operations, Inc., Grand Gulf Nuclear Station, Unit 1 - Issuance of License Amendment re: Control Rod Scram Time Testing Frequency, dated May 14, 2002.
- 5. Letter from Michael Webb, NRC, to Mr. Paul D. Hinnenkamp, Engergy Operations, Inc., River Bend Station Unit 1
- Issuance of Amendment Re: Control Rood Testing Frequency, dated December 12, 2002 Revision History OG Revision 0 Revision Status: Closed Original Issue Revision
Description:
Revision Proposed by: Grand Gulf Owners Group Review Information Date Originated by OG: 21-Aug-02 Owners Group Comments: (No Comments) Date: 21-May-03 Owners Group Resolution: Superceeded OG Revision 1 Revision Status: Active Bracketed the 200 day Frequency, added additional information on plant-specific justification of new Frequency. Revision
Description:
Revision Proposed by: Grand Gulf 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-460, Rev. 0 BWROG-90, Rev. 1 Affected Technical Specifications OG Revision 1 Revision Status: Active Owners Group Review Information Date Originated by OG: 21-May-03 Owners Group Comments: (No Comments) Date: 21-May-03 Owners Group Resolution: Approved TSTF Review Information TSTF Received Date: 11-Aug-03 Date Distributed for Review: 12-Aug-03 TSTF Comments: (No Comments) Date: 26-Aug-03 TSTF Resolution: Approved OG Review Completed: BWOG CEOG WOG BWROG NRC Review Information NRC Received Date: 07-Sep-03 SR 3.1.4.2 Control Rod Scram Times SR 3.1.4.2 Bases Control Rod Scram Times 17-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. Control Rod Scram Times3.1.4BWR/4 STS3.1.4 - 1Rev. 2, 04/30/013.1 REACTIVITY CONTROL SYSTEMS3.1.4Control Rod Scram TimesLCO 3.1.4a.No more than [10] OPERABLE control rods shall be "slow," inaccordance with Table 3.1.4-1, andb.No more than 2 OPERABLE control rods that are "slow" shalloccupy adjacent locations.APPLICABILITY:MODES 1 and 2. ACTIONSCONDITIONREQUIRED ACTIONCOMPLETION TIMEA.Requirements of theLCO not met.A.1Be in MODE 3.12 hoursSURVEILLANCE REQUIREMENTS- NOTE -During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall beisolated from the associated scram accumulator.SURVEILLANCEFREQUENCYSR 3.1.4.1Verify each control rod scram time is within the limitsof Table 3.1.4-1 with reactor steam dome pressure [800] psig.Prior to exceeding 40% RTP after each reactor shutdown 120 daysSR 3.1.4.2Verify, for a representative sample, each testedcontrol rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure [800] psig.120 days cumulative operation in MODE 1 SURVEILLANCE REQUIREMENTS (continued)Control Rod Scram TimesB 3.1.4 BASESBWR/4 STSB 3.1.4 - 4Rev. 2, 04/30/01testing can be performed. To ensure that scram time testing isperformed within a reasonable time following a shutdown 120 days orlonger, control rods are required to be tested before exceeding 40% RTP following the shutdown. This Frequency is acceptable considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by fuel movement within the associated core cell and by work on control rods or the CRD System.SR 3.1.4.2Additional testing of a sample of control rods is required to verify thecontinued 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 (e.g., 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 LCO limit. For planned testing, the control rods selected for the sample should be different for each test. 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 frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, "Control Rod Scram Accumulators."SR 3.1.4.3When work that could affect the scram insertion time is performed on acontrol 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. The limits for reactor pressures < 800 psig are established based on a high probability of meeting the acceptance criteria at reactor pressures 800 psig. Limits for 800 psig are found in Table 3.1.4-1. If testingdemonstrates the affected control rod does not meet these limits, but is Control Rod Scram Times3.1.4BWR/6 STS3.1.4 - 1Rev. 2, 04/30/013.1 REACTIVITY CONTROL SYSTEMS3.1.4Control Rod Scram TimesLCO 3.1.4a.No more than [14] OPERABLE control rods shall be "slow," inaccordance with Table 3.1.4-1 andb.No more than 2 OPERABLE control rods that are "slow" shalloccupy adjacent locations.APPLICABILITY:MODES 1 and 2.ACTIONSCONDITIONREQUIRED ACTIONCOMPLETION TIMEA.Requirements of theLCO not met.A.1Be in MODE 3.12 hoursSURVEILLANCE REQUIREMENTS- NOTE -During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.SURVEILLANCEFREQUENCYSR 3.1.4.1Verify each control rod scram time is within the limitsof Table 3.1.4-1 with reactor steam dome pressure [950] psig.Prior to exceeding40% RTP after each reactor shutdown 120 daysSR 3.1.4.2Verify, for a representative sample, each testedcontrol rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure [950] psig.120 dayscumulative operation in MODE 1 SURVEILLANCE REQUIREMENTS (continued)Control Rod Scram TimesB 3.1.4 BASESBWR/6 STSB 3.1.4 - 4Rev. 2, 04/30/01at reactor steam dome pressure 950 psig ensures that the scram timeswill be within the specified limits at higher pressures. 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 scram time testing is performed within a reasonable time following a shutdown 120 days, control rods are required to be tested before exceeding40% RTP. This Frequency is acceptable, considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by fuel movement within the associated core cell and by work on control rods or the CRD System.SR 3.1.4.2Additional testing of a sample of control rods is required to verify thecontinued 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." If more than 20% of the sample is declared to be "slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 20% criterion (e.g., 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 LCO limit. For planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data were 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 frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, "Control Rod Scram Accumulators."SR 3.1.4.3When work that could affect the scram insertion time is performed on acontrol 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 that the affected control rod is still within acceptable TSTF-465, Rev. 0 BWROG-81, Rev. 1 NUREGs Affected: Addition of time performance Surveillance Requirement (SR) note to Source Range Monitor (SRM) SRs Technical Specification Task Force Improved Standard Technical Specifications Change Traveler 1430 1431 1432 1433 1434 Classification:
- 3) Improve Specifications Recommended for CLIIP?:
Industry Contact: Tom Silko, (802) 258-4146, tsilko@entergy.com Yes Correction or Improvement: Improvement 1.0 Description A time allowance Note is being added to the Source Range Monitor (SRM) Surveillance Requirements (SRs) 3.3.1.2.3 and 3.3.1.2.4. This change provides a time allowance to perform the subject SRs following sudden entry into MODE 3 due to a reactor scram. These two SRs are not routinely performed in MODE 1 and thus will likely not be in periodicity. With the two SRs out of periodicity, sudden entry into MODE 3 due to a scram results in the immediate entry into SR 3.0.3 for the SRMs, which would remain in effect until the two SRs were completed. In STS, it atypical to have a
forced entry into SR 3.0.3 due to an anticipated operational occurrence (in this case, a scram) and the situation
presents a administrative distraction to Operators involved in scram recovery activities. Therefore, the
addition of a specific time allowance note to perform the two SRs is being proposed. 2.0 Proposed Change A 12-hour time allowance note is added to SRs 3.3.1.2.3 (SRM CHANNEL CHECK) and 3.3.1.2.4 (SRM COUNT RATE/SIGNAL -TO-NOISE). This change provides a time allowance to perform the SRs for the situation involving sudden entry into MODE 3 due to a reactor scram. The added Note is the same as that currently used for SR 3.3.1.2.6 (SRM CHANNEL FUNCTIONAL TEST/SIGNAL -TO-NOISE RATIO) and 3.3.1.2.7 (SRM CHANNEL CALIBRATION). This change is applicable to the Boiling Water Reactor (BWR) Standard Technical Specifications (STS), Revision 2 of NUREG -1433 and NUREG -1434. See the attached mark -ups for the specific changes. 12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-465, Rev. 0 BWROG-81, Rev. 1
3.0 Background
The primary use of the SRMs is during plant start -up. During start -up, the SRMs provide the operator with information relative to the neutron flux level at very low flux levels in the core. As such, the SRM indication is used by the operator to monitor the approach to criticality and determine when criticality is achieved. The
SRMs are maintained fully inserted until the count rate is greater than a minimum allowed count rate (a
control rod block is set at this condition), at which time they are partially withdrawn. After the Intermediate
Range Monitors (IRMs) are on range 3 or above, the SRMs are fully withdrawn from the core, where they
remain during normal power operation. The SRMs are required to be OPERABLE in MODES 2, 3, 4, and 5 prior to the IRMs being on scale on Range 3 to provide for neutron monitoring. In MODE 1, the APRMs provide adequate monitoring of
reactivity changes in the core; therefore, the SRMs are not required. In MODE 2, with IRMs on Range 3 or
above, the IRMs provide adequate monitoring and the SRMs are not required. The SRMs have no safety
function and are not assumed to function during any FSAR design basis accident or transient analysis. However, the SRMs do provide the only onscale monitoring of neutron flux levels during startup and refueling. As noted above, the SRMs are fully withdrawn from the reactor during startup. Accordingly, SRs 3.3.1.2.3 (SRM CHANNEL CHECK) and SR 3.3.1.2.4 (SRM COUNT RATE/SIGNAL -TO-NOISE) are not performed at power and thus will routinely be out of periodicity during MODE 1 power operation. A reactor scram results in a sudden entry into MODE 3 from MODE 1, which reestablishes TS requirements for SRM
operability. However, with the two SRs out of periodicity, the entry into MODE 3 results in the immediate
entry into SR 3.0.3 for the SRMs, which would remain in effect until the SRs were completed. In STS, it atypical to have a forced entry into SR 3.0.3 due to an anticipated operational occurrence (in this case, a scram) and the situation presents a administrative distraction to Operators involved in scram recovery activities. Hence, a time allowance to perform the SR is needed to avoid the unnecessary invocation of SR 3.0.3 for surveillance tests not met within the required Frequency. To address this situation, this TSTF
proposes the addition of a 12 -hour time allowance note to perform SR 3.3.1.2.3 and SR 3.3.1.2.4. This change also promotes consistency with existing SRs 3.3.1.2.6 (SRM CHANNEL FUNCTIONAL TEST/SIGNAL -TO-NOISE RATIO) and 3.3.1.2.7 (SRM CHANNEL CALIBRATION), which both already have a 12 -hour performance note. 12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-465, Rev. 0 BWROG-81, Rev. 1 4.0 Technical Analysis A reactor scram can result in an sudden unplanned entry into MODE 3. TS 3.3.1.2, SRM Instrumentation, requires SRM operability in MODE 3. The required SRs for MODE 3 are listed in TS Table 3.3.1.2 -1 and include SR 3.3.1.2.3 (SRM CHANNEL CHECK), SR 3.3.1.2.4 (SRM COUNT RATE/SIGNAL -TO-NOISE), SR 3.3.1.2.6 (SRM CHANNEL FUNCTIONAL TEST/SIGNAL -TO-NOISE RATIO) and 3.3.1.2.7 (SRM CHANNEL CALIBRATION). Since the SRMs are fully withdrawn from the reactor during startup and there are no operability requirements for the SRMs in MODE 1, none of the above four SRM SRs are required to be performed during normal
power operation. So, on a reactor scram, it would not be unusual for all four of the SRs to be out of
periodicity. SRs 3.3.1.2.6 and 3.3.1.2.7 both currently have 12 -hour performance Notes, which provide a nominal time period to perform the SRs. In current STS, however, SR 3.3.1.2.3 and SR 3.3.1.2.4, do not have a similar performance Note, which would result in the immediate entry into SR 3.0.3 for the SRMs until the
SRs were completed. The current SRM TS are primarily constructed with start -up activities in mind. In a shutdown condition with the SRMs fully inserted, all of the SRM SRs can be readily performed and maintained in periodicity. Therefore, it is simple to maintain MODE 3 SRs in periodicity and, during startup, transition into Mode 2 and
subsequently MODE 1. After the IRMs are on Range 3, SRM operability is no longer required and the SRMs
are withdrawn. A scram results in sudden entry into Mode 3, which reestablishes TS requirements for SRM operability. With SR 3.3.1.2.3 or SR 3.3.1.2.4 out of periodicity, this situation results in the immediate entry into SR 3.0.3 for the SRMs due to SRs not being within the required Frequency. The invocation of SR 3.0.3 allows an
additional 24 hours to perform SRs, which are discovered out of frequency. Therefore, the addition of a 12
hour time allowance note is conservative with respect to the 24 -hour time allowance provided by SR 3.0.3. In this regard, the proposed TSTF change is administrative in that it simply establishes TS provisions to avoid to a forced entry into SR 3.0.3. The proposed 12 -hour allowance to perform the SRs is reasonable based on the small safety significance of the delay in completing the SR, the inability to perform the SR prior to entering the Applicability, and the recognition that the most probable result of the SR being performed is verification of conformance with the
TS requirements. 12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-465, Rev. 0 BWROG-81, Rev. 1 5.0 Regulatory Analysis A change to Boiling Water Reactor (BWR) Standard Technical Specifications (STS), Revision 2 of NUREG -1433 and NUREG -1434 is being proposed by the Technical Specifications Task Force (TSTF) to add a 12 -hour time allowance note to SRs 3.3.1.2.3 (SRM CHANNEL CHECK) and 3.3.1.2.4 (SRM COUNT RATE/SIGNAL -TO-NOISE). 5.1 No Significant Hazards Consideration The Technical Specifications Task Force (TSTF) has evaluated whether or not a significant hazards consideration is involved with the proposed generic change by focusing on the three standards set forth in
10 CFR 50.92, "Issuance of amendment" as discussed below. In accordance with the criteria set forth in
10 CFR 50.92, the TSTF has evaluated these proposed Standard Technical Specifications changes and
determined that they do not represent a significant hazards consideration. The following is provided in
support of this conclusion.
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No This change provides Notes to SRs 3.3.1.2.3 and 3.3.1.2.4 to avoid those Surveillances being declared not met within the required Frequency due to an expected transition into MODE 3. The Frequency of
Surveillances is not an initiator of any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The Frequency of Surveillances has no effect on the consequences of an accident as the most likely outcome of any Surveillance is verification that the equipment is OPERABLE. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No
The proposed change does not involve a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a new or different kind of accident
from any accident previously evaluated .3. Does the proposed change involve a significant reduction in a margin of safety? Response: No This change provides Notes to SRs 3.3.1.2.3 and 3.3.1.2.4 to avoid those Surveillances being declared not met within the required Frequency due to an expected transition into MODE 3. Should the Notes not be adopted, plants would continue to invoke SR 3.0.3 until the Surveillances can be performed. SR 3.0.3 would allow 24 hours to perform the missed Surveillances, while the proposed Notes allow only 12 hours. For these reasons, the proposed change does not involve a significant reduction in the
margin of safety. 12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-465, Rev. 0 BWROG-81, Rev. 1 Based on the above, the TSTF concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no
significant hazards consideration" is justified. 5.2 Applicable Regulatory Requirements/Criteria In STS, it customary to require performance of applicable SRs prior to entry into the specified condition of the Applicability, whenever feasible. In some cases, however, due to plant conditions it may not always possible to perform the SRs prior to entry. For these situations, it is typical to have a SR performance Note which
allows a reasonable time period to perform the SR. For the situation described in this TSTF, a reactor scram results in the sudden entry into a plant condition (MODE 3) that requires the operability of the SRMs. The required SRs will be out of periodicity, which
results in a forced entry into SR 3.0.3. In using STS, it is atypical to have a forced entry into SR 3.0.3 due to
an anticipated operational occurrence (in this case, a scram) and the situation presents a administrative distraction to Operators involved in scram recovery activities. Therefore, this TSTF proposes the addition of a time allowance note to allow performance of the SRs. This is consistent with STS general practice and
meets regulatory objectives. Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the approval of the proposed change will not be
inimical to the common defense and security or to the health and safety of the public." 6.0 Environmental Consideration The proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant
increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change. 7.0 References None Revision History OG Revision 0 Revision Status: Closed Original Issue Revision
Description:
Revision Proposed by: BWROG 12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-465, Rev. 0 BWROG-81, Rev. 1 Affected Technical Specifications OG Revision 0 Revision Status: Closed Owners Group Review Information Date Originated by OG: 27-Jun-00 Owners Group Comments: Traveler must be marked on Revision 2 pages and have an SE quality justification prior to TSTF review. Discussed at 8 21/02 BWROG meeting. BWROG prioritized change and wanted BWROG-81 to address the BF proposed change. Date: 08-Nov-00 Owners Group Resolution: Superceeded OG Revision 1 Revision Status: Active Revised to mark on ISTS Revision 2 pages and upgraded justification to Safety Evaluation quality. Revision
Description:
Revision Proposed by: Browns Ferry Owners Group Review Information Date Originated by OG: 21-Aug-02 Owners Group Comments: (No Comments) Date: 21-Aug-02 Owners Group Resolution: Approved TSTF Review Information TSTF Received Date: 25-Nov-02 Date Distributed for Review: 12-Aug-03 TSTF Comments: WOG chairman pointed out that change is applicable to PWRs. Will consider PWR-specific change if beneficial. Date: 12-Sep-03 TSTF Resolution: Approved OG Review Completed: BWOG CEOG WOG BWROG NRC Review Information NRC Received Date: 18-Sep-03 SR 3.3.1.2.3 SRM Instrumentation SR 3.3.1.2.3 Bases SRM Instrumentation SR 3.3.1.2.4 SRM Instrumentation 12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-465, Rev. 0 BWROG-81, Rev. 1 SR 3.3.1.2.4 Bases SRM Instrumentation 12-Sep-03 Traveler Rev. 3. Copyright (C) 2003, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited. TSTF-465, Rev. 0
Insert 1 The Note to SR 3.3.1.2.3 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 allowance to enter the Applicability with t he 24 hour Frequency not met is reasonable based on the limited time of 12 hours and the most probable result of performing the Surveillance being the verification of conformance with the requirements. Insert 2 Note 2 to the surveillance allows the Sur veillance to be delayed until entry into the specified condition of the Applicability (THERMAL POWER decreased to IRM Range 2 or below). The allowance to enter the Applicability with the 24 hour Frequency not met is
reasonable based on the limited time of 12 hours and the most probable result of performing the Surveillance being the verification of conformance with the requirements.
SRM Instrumentation 3.3.1.2 ACTIONS (continued) CONDITION E.2 SURVEILLANCE REQUIREMENTS REQUIRED ACTION COMPLETION TIME Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies. .NOTE-Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified conditions. SR 3.3.1.2.1 SR 3.3.1.2.2 SURVEILLANCE Perform CHANNEL CHECK.-NOTES-1.Only required to be met during CORE ALTERATIONS. 2.One SRM may be used to satisfy more than one of the folloWing. Verify an OPERABLE SRM detector is 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 c.A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region.FREQUENCY 12 hours 12 hours SR 3.3.1.2.3 CHANNEL CHECK.BWR/4 STS\3.3.1.2-2-NOTE-(Not required to be performed until 12 hours after.(IRMs on Range 2 or below..) -.,.----24 hours Rev.2, 04f30/0 1 SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE.A) FREQUENCY SR 3.3.1.2.4----------------------.-NC)TEf---------------------
- 't:>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.
Verify count rate is: 12 hours during CORE a.?:[3.0]cps with a signal to noise ratio;::[2:1]or ALTERATIONS b.?:[0.7]cps with a signal to noise ratio?:[20:1].AND 24 hours SR 3.3.1.2.5 SR 3.3.1.2.6 SR 3.3.1.2.7 BWR/4 STS Perform CHANNEL FUNCTIONAL TEST[and determination of signal to noise ratio]..NOTE*Not required to be performed until 12 hours after IRMs on Range 2 or below.Perform CHANNEL FUNCTIONAL TEST[and determination of signal to noise ratio]..NOTES-1, Neutron detectors are excluded.2.Not required to be performed until 12 hours after I RMs on Range 2 or below.Perform CHANNEL CALIBRATION. 3.3.1.2-3 7 days 31 days[18]months Rev.2, 04/30/01 SRM Instrumentation B 3.3.1.2 BASES ACTIONS (continued) Action (once required to be initiated) to insert control rods must continue until all insertable rods in core cells containing one or more fuel assemblies are inserted.SURVEILLANCE REQUIREMENTS The SRs for each SRM Applicable MODE or other specified conditions are found in the SRs column of Table 3.3.1.2-1. SR3.3.1.2.1and SR 3.3.1.2.3 Performance of the CHANNEL CHECK 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 another channel.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 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 0\ the channels required by the LCO..l.('---\"'0'"*SR 3.3.1.2.2.D'.,.'.'..'-t-./\""\.: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 anadjacentquadrant containing fuel.Note 1 states that the SR is required to be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring.ThisSurveillance consists of a review of plant BWR/4 STS Rev.2.04/30101 SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE REQUIREMENTS (continued) logs to ensure that SRMs required to be OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE.In the event that only one SRM is required to be OPERABLE, per Table 3.3.1.2-1, footnote (b), only the a.portion of this SR is required.Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRM.The 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, 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.To accomplish this, the SR is modified by a Note 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.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 changes are in progress, the Frequency is relaxed from 12 hours to 24 hours. SR 3.3.1.2.5 and SR 3.3.12.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly.A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay.This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay.This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical BWR/4 STS Rev.2, 04130/01 SRM Instrumentation 3.3.1.2 ACTIONS (continued) CONDITION E.2 SURVEILLANCE REQUIREMENTS REQUIRED ACTION Initiate action to insert all insertable control rods in core cells containing one or more fuel assemblies. COMPLETION TIME Immediately -NOTE-Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified conditions. SR 3.3.1.2.1 SR 3,3.1.2.2 SURVEILLANCE Perform CHANNEL CHECK.-NOTES*1.Only required to be met during CORE ALTERATIONS. 2, One SRM may be used to satisfy more than one of the following. Verify an OPERABLE SRM detector is 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 c.A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region.FREQUENCY 12 hours 12 hours SR 3.3.1.2.3 f1'erform CHANNEL CHECK./ Not required to be performed until 12 hours after 1 IRMs on Range 2 or below.I 24 hours Rev.2, 04/30/01 .".:1"Not required to be performed until 12 hours after IRMs on Range 2 or below.__. ....__ ..----.----,------- SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE8R 3.3.1.2.4 cD Not required to be met with less than or equal to four fuel assemblies adjacent to the 8RM and no other fuel assemblies in the associated core quadrant.Verify count rate is either: a.:<:[3.0]cpswitha signal to noise ratio:<:[2:1]or SRM Instrumentation 3.3.1.2 FREQUENCY 12 hours during CORE ALTERATIONS b.:<:[0.7]cps with a signal to noise ratio:::[20:1].AND 24 hours SR 3.3.1.2.5 SR 3.3.1.2.6 SR 3.3.1.2.7 BWR/6 STS Perform CHANNEL FUNCTIONAL TEST[and determination of signal to noise ratio]..NOTE*Not required to be performed until 12 hours after IRMs on Range 2 or below.Perform CHANNEL FUNCTIONAL TEST[and determination of signal to noise ratio]..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. 3.3.1.2-3 7 days 31 days[18]months Rev.2, 04/30/01 SRM Instrumentation B 3.3.1.2 BASES ACTIONS (continued) Action (once required to be initiated) to insert control rods must continue until all insertable rods in core cells containing one or more fuel assemblies are inserted.SURVEILLANCE REQUIREMENTS The SRs for each SRM Applicable MODE or other specified condition are found in the SRs column of Table 3.3.1.2-1. SR 3.3.1.2.1 and SR 3.3.1.2.3 Performance of the CHANNEL CHECK 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 the same parameter indicated 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 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 the Leo.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 this SR is required to be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant BWR/6 STS B 3.3.1.2-5 Rev.2, 04/30/01 SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE REQUIREMENTS (continued) logs to ensure that SRMs required to be OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE.In the event that only one SRM is required to be OPERABLE, per Table 3.3.1.2-1, footnote (b), only the a.portion of this SR is required.Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRM.The 12 hour Frequency is based upon operating experience and supplements operational controls over refueling activities, which include steps to ensure that the SRMs required by the LCO are in the proper quadrant.SR 3.3.12.4 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.This 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.To accomplish this, the SR is modified by a Note 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 assembliesloadedaround each SRM and no other fuel assemblies in the associated quadrant, even with a control rod withdrawn the configuration 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 changes are in progress, the Frequency is relaxed from________-12 hours to 24 hours.//\\2 r----:..-""'.7
--\...........----- 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_A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay.This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay.This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical BWR/6 STS B 3.3.1.2-6 Rev.2, 04/30/01}}