ML18039A205
| ML18039A205 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/04/1997 |
| From: | Abney T TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML18039A206 | List: |
| References | |
| TAC-M96431, TAC-M96432, TAC-M96433, NUDOCS 9712180109 | |
| Download: ML18039A205 (27) | |
Text
Enclosure ITS Section 3.5 Emergency Core Cooling Systems and Reactor Core Isolation Cooling System Enclosure Contents Enclosed?
~ Response to NRC questions Yes
~ Summary Description of ITS/ITS BASES Changes. Yes
~ ITS Revised Pages Yes
~ ITS BASES Revised Pages Yes
~ CTS Mark-up Revised Pages. . Yes
~ Justifications for Changes to CTS (DOCs)
Revised Pages Yes
~ NUREG-1433 BWR/4 STS Mark-up Revised Pages. Yes
~ NUREG-1433 BWR/4 STS Bases Mark-up Revised Pages. Yes
~ Justification for Changes to NUREG-1433 (JDs)
Revised Pages Yes
~ No Significant Hazards Considerations Revised Pages Yes
~ Cross-Reference Matrix Correlating Changes Between the CTS, ITS, and NUREG-1433. Yes
ITS Section 3.5.1 Item 3.5.1-1 CTS 3.5.H/4.5.H.4 require daily monitoring of the Core Spray (CS) and Residual Heat Removal (RHR) discharge line pressure indicators (48 psig minimum) to ensure the discharge piping is full of water whenever CS and RHR are required to be operable. After further review, TVA has determined that these CTS requirements will be relocated to the Technical Requirements Manual (TRM). Changes to the TRM are controlled by 10 CFR 50.59.
Improper water fill is not considered likely since to the Pressure Suppression Chamber (PSC) head tank is alignment maintained with locked open valves. Also, under the provisions of proposed ITS SR 3.5.1.1, venting and verification of water fill for the Emergency Core Cooling Systems (ECCS) discharge piping is performed every 31 days as a formal surveillance test. The 31-day frequency is adequate to ensure that the water fill requirements are met and is based on the gradual nature of void buildup in the Emergency Core Cooling Systems (ECCS) piping, procedural controls governing syst: em operation, and industry operating experience. Relocation to the TRM is acceptable based on the criteria of 10 CFR 50.36. A new DOC R2 has been added for this relocation. DOCs LA2 and LA3 were modified appropriately.
Item 3.5.1-2 The CTS 4.5.H.l and 4.5.H.3 requirements that CS, RHR, and High Pressure Coolant Injection (HPCI) discharge piping be vented from the high point, and water flow determined or observed are considered procedural details for performing the associated surveillance tests (ITS SR 3.5.1.1). Retention of this level of detail in the ITS is not necessary to demonstrate the fill verification of ECCS discharge piping.
The Bases for SR 3.5.1.1 state that one acceptable method of ensuring that the lines are full is to vent at the high points. The function and purpose of ECCS venting is also described in FSAR Sections 6.3 and 6.4. Procedural details for venting the discharge lines will be incorporated into the surveillance test for SR 3.5.1.1. Removal of the details is consistent with the application of 10 CFR 50.36 criteria.
CTS 4.5.H.4 requires monitoring of the CS and RHR discharge line pressure on a daily, basis to ensure the lines are filled with water whenever the CS or Low Pressure Coolant Injection
(LPCI) systems are required to be operable. The CTS 4.5.H.1 monthly requirement to vent and verify water fill of the ECCS discharge piping every month is being retained as proposed ITS SR 3.5.1.1. The adequacy of the monthly surveillance is based on the gradual nature of void build-up in the ECCS piping, procedural controls governing system operation, and industry operating experience. Also, as discussed in the response to CTS 3.5.H under NRC Item 3.5.1-1, the requirements associated with the maintenance of filled discharge piping are being relocated to the TRM.
Justification for this change is provided in the referenced DOC changes in response 3.5.1-1.
The CTS 4.6.D.2 provisions for verifying Automatic Depressurization System (ADS) safety/relief valve (S/RV) operability are considered procedural details for performing the associated surveillance test (now ITS SR 3.5.1.11).
Retention of this level of detail (use of thermocouples/acoustic monitors) in the ITS is not necessary to demonstrate system operability. The. Bases for SR 3.5.1.11 state that S/RV operability may be demonstrated by any method suitable to verify steam flow. This would include the use of thermocouples or acoustic monitors. The function and purpose of the thermocouples and acoustic monitors is also described in FSAR Section 7.4. Details for the utilization of the thermocouples and acoustic monitors as satisfactory alternative methods will be incorporated into the surveillance procedure for SR 3.5.1.11. This change is consistent with the application of 10 CFR 50.36 criteria.
Changes to the Bases will be controlled by the Bases Control Process in BFN ITS Section 5.5.10. Changes to the FSAR and the TRM are controlled by provisions of 10 CFR 50.59.
Changes to plant SRs are controlled by site administrative procedures which include a review for 10 CFR 50.59 applicability.
Item 3.5.1-3 The LA5 reference for the CTS 3.5.B.14 mark-up is incorrect and new DOC LAS has been added to justify this change. The CTS mark-up has been revised to reference LA8.
Operability of the recirculation pump discharge valves is required to support the LPCI ECCS function as indicated in the Background Bases for ITS 3.5.1 and the Bases for ITS SR 3.5.1.5. Hence, it is not necessary to relocate CTS LCO 3.5.B.14 as a separate ITS LCO since operability of the recirculation pump discharge valves is required to meet the ITS LCO 3.5.1 for ECCS LPCI operability. Proposed ITS SR 3.5.1.5 will ensure recirculation pump discharge valves are operable when required to support the LPCI ECCS function.
This change is consistent with the application of 10 CFR 50.36 criteria.
Item 3.5.1-4 The CTS 3.5.E option for use of an auxiliary (boiler) as a low pressure steam source is relocated to the ITS Bases Section, SR 3.5.1.6, SR 3.5.1.7, and SR 3.5.1.8. This provision is considered an operational detail for conducting the surveillance and, hence, more appropriately stated in the Bases. The procedure which implements SR 3.5.1.8 will have additional details, precautions, and limitations concerning the use of the auxiliary boiler as the steam supply source for the low pressure HPCI surveillance. This change is consistent with the application of 10 CFR 50.36 criteria.
Item 3. 5. 1-5 This proposed change eliminates the existing requirement in CTS 4.5.H.1 to vent and verify water discharge piping "prior to testing".
fill These of the RHR and CS requirements are considered operational details for operating the RHR and CS systems to ensure the systems are completely filled with water prior to operating or testing. Under the provisions of proposed ITS SR 3.5.1.1, the venting and verification of water fill for the ECCS discharge piping will be performed every 31 days as a formal surveillance. Also, as discussed in the response to NRC comment 3.S.1-1, CTS requirements for daily monitoring of discharge line pressure for= RHR and Core Spray are being relocated to the TRM.
The requirement to vent and verify water fill "prior to testing" will be 'controlled by plant procedures. Operating Instructions (OIs) will require the systems to be filled and vented prior to manual initiation. Similarly, test instructions which require manual initiation will likewise require the system to be filled and vented. Changes to these plant instructions are controlled by site administrative procedures which include a review for 10 CFR 50.S9 applicability. Since testing other than that explicitly required by ITS may be performed, we are re-categorizing this aspect of this change as less restrictive. A new DOC L14 has been added to address this change.
Item 3.5.1-6 In response to the NRC letter dated October 22, 1997, concerning ITS 3.5.1 Condition A, the second provision in Condition A has been deleted (in this submittal). This also results in a revision to Condition H to match the removal of the second Condition in A. As a result of these changes, 3
this NRC comment is no longer applicable. DOC L7 has also been revised to address this change. JD P24 is no longer applicable and has been deleted.
Item 3.5.1-7 DOC LS has been revised to properly reference CTS SR 4.6.D.2 and ITS SR 3.5.1.11.
Item 3.5.1-8 The HPCI system is designed to supply 5000 gpm to the reactor at'essel pressures ranging from 150 psig to 1120 psig. CTS 4.5.E.l.d requires that HPCI deliver 5000 gpm at normal reactor operating pressures. This requirement is similar to the ITS SR 3.5.1.7 requirement that HPCI supply 5000 gpm between 920 psig and 1010 psig reactor pressure, since the nominal no load reactor pressure is 920 psig and full load reactor pressure is 1005 psig (this allows a 5 psig margin at full load). Thus, the ITS SR would allow HPCI testing over the full range of potential expected operating pressures.
This change is also consistent with NUREG-1433 which provides bracketed ranges for demonstration. of HPCI flow at operating pressures.
Regarding accident/transient analyses, HPCI is assumed capable of delivering 5000 gpm over the specified design pressure range (150 psig to 1120 psig) . Actual reactor pressure in the transient and accident analyses is, of course, dependent on the specifics of the event. Performance of the HPCI SR at operating pressure ranges confirm that the HPCI pump and turbine are functioning in accordance with design speci.fications in the higher pressure domain. A small decrease in pressure to as low as 920 psig at which the performance to design specifications is verified will not affect the validity of the HPCI test to demonstrate proper operation of the pump and turbine. Furthermore, the reactor is expected to be at the lower pressure only on a very limited basis (during startup or shutdown evolutions).
Hence, the HPCI SR will normally be performed at rated reactor pressure. DOC L9 has been augmented with the above additional information.
The ITS SR 3.5.1.8 value of less than or equal to 165 psig was chosen to be consistent with the NUREG-1433 suggested value for low pressure HPCI performance testing. CTS 4.5.E.l.e requires reactor pressure above 150 psig. As noted in DOC L10, this small change in pressure will not significantly delay or affect the validity of the test to determine that the pump and turbine are still operating at the design specifications.
As..noted.,above, in the accident/transient analyses, HPCI is assumed capable of delivering flow over the specified design range. Actual reactor pressure depends on the specific event mechanics. Performance of ITS SR 3.5.1.8 at 165 psig or slightly below confirms that the HPCI pump and turbine are functioning in accordance with design specifications at the low end of the pressure range. A small increase to 165 psig (from 150 psig) at which the design specification is verified will not invalidate the validity of the HPCI test. DOC L10 has been augmented with the above additional justification.
Item 3.5.1-9 As noted in the NRC comments, the'taff is currently reviewing the acceptability of the Unit 3 LPCI subsystem
,cross-tie alternate verification proposed for SR 3.5.1.4 (which is classified beyond scope). No additional action or response from BFN is required at this time.
Item 3.5.1-10 ITS SR 3.5.1.5 Notes and Frequencies have been revised in the manner suggested in this NRC comment. The SR 3.5.1.5 Bases discussion were similarly modified. NUREG Justification (JD)
P21 has also been modified to reflect this change. This should address NRC's comment on the Bases.
Item 3.5.1-11 The wording of ITS SR 3.5.1.12 has been revised in the manner suggested by the NRC comment.
Item 3.5.1-12 I
As suggested in the NRC comment, the ITS Applicable Safety Analysis Bases was revised to include a discussion of the limiting single failures as follows: For a large or small pipe break LOCA and events requiring ADS operation, selected battery failure is considered the most severe single failure.
Item 3.5.1-13 NRC 93-102 is the NRC press release (issued July 23, 1993) which announced the issuance of the Policy Statement on Technical Specification Improvements for Nuclear Power Plants. The July 22, 1993, reference provided in the NRC question is the federal register notice of the subject policy
t statement. We consider that NRC 93-102 provides sufficient linkage to the corresponding Policy Statement.
0 ITS Section 3.5.2 Item 3.5.2-1 We agree with the NRC comments regarding the CTS mark-ups.
The CTS mark-up DOC M3 reference has been corrected from SR 3.5.2.4 to SR 3.5.2.3 and DOC M4 has been corrected from SR 3.5.2.5 to SR 3.5.2.4. The DOC M3 text has been corrected from SR 3.5. 2. 4 to SR 3. 5.2. 3. Also, CTS mark-up was corrected by changing Note for SR 3.5.2.4 to Note for SR 3.5.2.3 and CTS 4.5.B.9 mark-up changing SR 3.5.2.5 to SR 3.5.2.4.
A 'new DOC M5 and justification for the addition of ITS SR 3.5.2.5 has been included. CTS DOC MS addresses that SR 3.5.2.5 requires verification that each required ECCS subsystem actuates on an actual or simulated automatic initiation signal every 18 months. Addition of this SR is considered to be more restrictive since CTS only requires actuation testing during ECCS-Operating (i.e., MODES 1, 2, and 3) .
Item 3.5.2-2 This proposed change eliminates the existing requirement in CTS 4.5.H.l to vent and verify water discharge piping "prior fill of the RHR and CS to testing". These requirements are considered operational details for operating the RHR and CS systems to ensure the systems are completely filled with water prior to operating or testing. Under the provisions of proposed ITS SR 3.5.2.2, the venting and verification of water fill for the ECCS discharge piping is performed every 31 days as a formal surveillance. Also, as discussed in the response to NRC comment 3.5.1-1, CTS requirements for daily monitoring of discharge line pressure for RHR and Core Spray are being relocated to the TRM.
The requirement to vent and verify water fill "prior to testing" will be controlled by plant procedures. Operating Instructions (OIs) require the systems to be filled and vented prior to manual initiation. Similarly, any test instruction which requires manual initiation will require the system to be filled and vented. Changes to these plant instructions are controlled by site administrative procedures for which a review for 10 CFR 50.59 applicability is required. Since other testing other, than that explicitly required by TS may be performed, we are categorizing this aspect of this change as less restrictive. A new DOC L3 has been added to reflect this change.
Item 3.5.2-3 The CTS Specification 4.5.H.1 requirements that CS and RHR discharge pipe venting be vented from the high point and water flow determined are considered procedural details for performing the associated surveillance test (now ITS SR 3.5.2.2). Retention 'of this level of detail in the ITS is not necessary to demonstrate the fill verification of ECCS discharge piping. The Bases for SR 3.5.2.2 state that one acceptable method of ensuring that the lines are full is to vent at the high points. The function and purpose of ECCS venting is also described in FSAR Sections 6.3 and 6.4.
Procedural details for venting the discharge lines will be incorporated into the surveillance tests for SR 3.5.2.2.
These changes are consistent with the application of 10 CFR 50.36 criteria.
System monitoring required per CTS 4.5.H.4 is discussed in responses 3.5.1-1 and 3.5.1-2.
Item 3.5.2-4 To address the NRC comment, BFN has added the CTS 3.5.A.5 footnote "*" requirements into the ITS 3.5.2 APPLICABILITY statement. This will require two low pressure ECCS injection/spray subsystems in MODE 5 if the spent fuel
,storage gates are removed and water level ~ 22 feet over the top of the reactor pressure vessel flange while operations with the potential for draining the reactor vessel (OPDRVs) are in progress. This will retain the existing CTS requirement and is more restrictive than the CTS 3.5.A.5 note since ITS will require two ECCS subsystems be available instead of either CS or RHR as is required in the CTS requirement. Accordingly, DOC M6 has been added for this change. DOC L1 has been deleted.
Item 3.5.2-5 Under CTS 3.5.A.5, Core Spray is allowed to be removed from service during refueling operations if a Residual Heat Removal Service Water System (RHRSW) pump is available through the cross-connection, provided the fuel pool gates are removed and level normal. The RHR/RHRSW cross-connection provides a redundant source of make-up water for fuel pool as discussed in Section 10.5.5 of the Updated Final Safety Analysis Report. It is considered a back-up source since if all it is raw water (river water) that would be used only other normal sources were unavailable.
The need for the availability of a RHRSW pump through the cross-connect as a prerequisite for the allowing the core spray system to be inoperable is not included in ITS 3.5.2 8
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for the same refueling conditions. This is consistent with the application of 10 CFR 50.36 criteria in that this feature of the RHRSW system is not credited as a primary system for mitigation of transients or accidents. We also consider that the provisions of proposed BFN ITS 3.5.2, which are consistent with NUREG-1433, provide adequate requirements for ensuring adequate water inventory is maintained during refueling activities. Accordingly, this change has been recategorized as Less Restrictive and a new DOC L4 generated as justification. As noted above, this design feature of the RHR/RHRSW will continue to be described in the UFSAR.
Changes to the UFSAR are controlled in accordance with 10 CFR 50.59.
Item 3.5.2-6 The current BFN design and licensing basis does not postulate a LOCA in either MODE 4 or 5 since the primary system is not pressurized. Hence, no specific analysis for a MODE 4 or 5 LOCA is available for reference. For this reason, the standard NUREG-1433 wording for the APPLICABLE SAFETY ANALYSIS in the Bases of Section 3.5.2 has not been incorporated. BFN-specific wording has been included to support the establishment of LCO 3.5.2.
Item 3.5.2-7 ITS 3.7.2 defines the operability requirements for Emergency Equipment Cooling Water (EECW) in MODES 1, 2, and 3. In MODES 4 and S, therefore, it is appropriate to require the necessary portions of the EECW be available in support ITS LCO 3.S.2 system operability requirements. This same statement is not required in ITS LCO 3.5.1 since EECW requirements in Modes 1, 2, and 3 are as specified in ITS 3.7.2.
ITS Section 3.5.3 Item 3.5.3.1 Removal of these details described by DOC LA1 from the CTS is in accordance with the "Application of Screening Criteria to BFN TS" and the criteria of 10 CFR 50.36. Proposed SR 3.5.3.1 ensures Reactor Core Isolation Cooling (RCIC) discharge piping is filled with water and proposed SR 3.5.3.2 ensures the RCIC valve lineup is correct.
Item 3.5.3-2 The CTS Specification 4.5.H.3 requirements that RCIC discharge pipe venting be vented from the high point and water flow observed are considered procedural details for performing the associated surveillance test (ITS SR 3.5.3.1).
Retention of this level of detail in the ITS is not necessary to demonstrate the fill verification of ECCS discharge piping. The Bases for SR 3.5.3.1 states that one acceptable method of ensuring that the lines are full is to vent at the high points. Details for venting the discharge lines will be incorporated into the surveillance procedures for SR 3.5.3.1.
Relocation of these details is consistent with the application of 10 CFR 50.36 criteria.
Item 3.5.3-3 The CTS 3.5.F.1 option for use of an auxiliary (boiler) as a low pressure steam source is relocated to the ITS Bases Section for SR 3.5.3.3 and SR 3.5.3.4. This provision is considered an operational detail for conducting the surveillance and, hence, more appropriately stated in the Bases. The procedure which implements SR 3.5.3.4 will have additional details, precautions, and limitations concerning the use of the auxiliary boiler as the steam supply source for the low pressure RCIC test. This change is consistent with the application of 10 CFR 50.36 criteria.
Item 3.5.3-4 The RCIC system is designed to supply 600 gpm to the reactor vessel at reactor vessel pressures ranging from 150 psig to 1120 psig. CTS 4.5.F.1.d requires that RCIC deliver 600 gpm at normal reactor op'crating pressures. This is similar to the proposed ITS SR 3.5.3.3 requirement that RCIC supply 600 gpm between 920 psig and 1010 psig reactor pressure, since the nominal no load reactor pressure is 920 psig and full 10
0 load reactor pressure is 1005 psig (this allows a 5 psig margin at full load). Thus, the ITS SR would allow RCIC testing over the full range of expected operating pressures.
This change is also consistent with NUREG-1433 which provides a bracketed ranges for demonstration of RCIC flow at operating pressures.
Regarding transient analyses, RCIC is assumed capable of delivering 600 gpm over the specified design pressure range (150 psig to 1120 psig). Actual reactor pressure in-the transient analyses is, of course, dependent on the specifics of the event. Performance of the RCIC SR at operating pressure ranges confirm that the RCIC pump and turbine are functioning in accordance with design specifications in the higher pressure domain. A small decrease in pressure to as low as 920 psig at which the performance to design specifications is verified will not affect the validity of the RCIC test to demonstrate proper operation of the pump and turbine. Furthermore, the reactor is expected to be at the lower pressure only on a very limited basis (during startup or shutdown evolutions). Hence, the RCIC SR will normally be performed at rated reactor pressure. DOC L4 will be augmented with the above additional information.
The ITS SR 3.5.3.4 value of less than or equal to 165 psig was chosen to be consistent with the NUREG-1433 suggested value for low end RCIC pressure testing. The surveillance test which implements CTS 4.5.F.1.e requires reactor pressure above 150 psig. As noted in DOC L5, this small change in pressure will not significantly delay or affect the validity of the test to determine that the pump and turbine are still operating at the design specifications.
As noted above, in the transient analyses, RCIC is assumed capable of delivering flow over the specified design range.
Actual reactor pressure depends on the specific event mechanics. Performance of ITS SR 3.5.3.4 at 165 psig or below confirms that the RCIC pump and turbine are functioning in accordance with design specifications at the low end of the pressure range. A small increase to 165 psig (from 150 psig) at which the design specification is verified will not invalidate the validity of the RCIC test. DOC L5 will be augmented with the above additional justification.
Item 3.5.3-5 The RCIC system requires approximately 40,000 lb/hr of steam flow to deliver 600 gpm pump flow at rated reactor pressure.
This is equivalent to a reactor power of 0.3%. The equivalent steam flow from a bypass valve (BPV) would require one bypass valve to be open approximately 10%. Therefore, operation of RCIC has little effect on the reactor. However, to address the NRC concern, the Bases will be revised to 11
require the reactor" be critical for performance of SR 3.5.3.3 and 1 BPV greater than 50% open for performance of SR 3.5.3.4.
12
0
SUMMARY
DESCRIPTION of ITS/BASES CHANGES ITS SECTION 3.5 TVA is submitting a proposed supplement to TS-362 for ITS Section 3.5, EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM. This supplement makes several changes associated with NRC comments on Section 3.5 (
Reference:
NRC Request for Additional Information Regarding Improved Standard Technical Specifications, dated September 17, 1997, -(TAC NOS. M96431, M96432,,M96433) and.the NRC letter October 22, 1997, regarding LCO 3.5.1 dated. This submittal also incorporates changes resulting from internal TVA reviews. A synopsis of the ITS and ITS BASES changes is provided below.
LCO 3.5.1 CONDITION A AND H, AND ASSOCIATED BASES In a letter dated October 22, 1997, NRC indicated that the proposed BFN ITS change regarding one inoperable Low Pressure Coolant Injection (LPCI) pump in each loop be reviewed on a generic rather than a plant specific basis, and requested TVA withdraw the change. In response to NRC, the attached ITS revision removes the seven day allowable out-of-service provision proposed in ITS Revision 0 for one inoperable LPCI pump in each subsystem (by deleting the second condition under CONDITION A). CONDITION H is also modified to accommodate the change to CONDITION A.
Appropriate BASES sections are likewise revised.
A generic change on this item was separately initiated by TVA and was approved by the BWR Owners Group Technical Specifications Task Force on November 5, 1997. If approved by NRC, the originally requested ITS changes will be resubmitted at a later date.
SR 3.5.1.3 AND BASES In response to a TVA internal review, changed ADS minimum air supply header pressure to plant specific value of 81 psig established by calculation MD-Q0032-870288 R6.
'I a
SR 3.5.1.5 AND BASES In response to an NRC corn'ment regarding cycl'i."ng of recirculation pump discharge valves, revised SR NOTE and FREQUENCY, and corresponding BASES"description to provide -a--
more concise description of the TS requirements. This change retains the current licensing'b'ases while being more consistent with NUREG-1433.
SR 3.5.1.6 In response to a TVA internal review, changed LPCI pump flow rate values and pressures to agree with current licensing basis values.
SR 3.5.1.11 In response to a TVA internal review, the word "pressure" is being added to the SR NOTE. This is consistent with NUREG-1433.
SR 3.5.1.12 In response to an NRC comment, removed excess wording "the 480 V Reactor MOV Board supplying power to" to be consistent with wording previously approved for Peach Bottom ITS for the same change.
LCO 3 5 2 APPLICABILITY~ REQUIRED ACT ION B 1 g AND ASSOCIATED BASES In response to an NRC comment, added "no operations with the potential to drain down the reactor vessel (OPDRVs) are in progress" to existing MODE 5 applicability statement (with the spent fuel storage pool gates removed and water level
> 22 feet over the top of the reactor pressure flange). In Required Action B.l, an associated editorial change was made. This change is consistent with current licensing basis.
SR 3.5.2.4 In response to a TVA internal review, changed LPCI pump flow rate values and pressures to agree with current licensing basis values.
ITS 3.5.1 APPLICABLE SAFETY ANALYSIS BASES In response to an NRC comment, included discussion of "selected battery failure" as the most severe limiting
single failure for a design basis accident. The inclusion of this plant specific detail is consistent with NUREG-1433.
SR 3 5 1 6/ SR 3 5 1 7~ AND SR 3 5 1 8 BASES In response to a TVA internal review, added reference 15 to an ECCS syst: em ana'lysis. This is the companion analysis report (non-LOCA events analyses) to Reference 13 (LOCA analyses). Also, revised Bases to indicate that two turbine bypass valves are required to be open to conduct the High Pressure Coolant Injection (HPCI) SR 3.5.1.8.
SR 3.5.1.11 BASES In response to a TVA internal review, changed the required minimum reactor steam dome pressure value needed for ADS valve testing from 250 psig to 920 psig to be consistent with manufacturers recommendations. In the BASES added wording "pressure recommended by the valve manufacturer". This change is consistent with NUREG-1433.
ITS LCO 3.5.2 BASES (Unit 1 Onl )
In response to a TVA internal review, changed "RHR System LPCI valve" to "LPCI crosstie valve" to correct a typographical error. This change is consistent with the previously submitted NUREG Bases mark-up and the typed version of the ITS Bases for Units 2 and 3; SR 3.5.3.3 & SR 3.5.3.4 BASES In response to an NRC comment, added description of what is meant by adequate steam flow for RCIC pump flow testing.
The inclusion of the parameters that define adequate steam flow for the purpose of the SRs is consistent with NUREG-1433.