ML15261A518
| ML15261A518 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/28/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML15261A517 | List: |
| References | |
| NUDOCS 9902050347 | |
| Download: ML15261A518 (5) | |
Text
-13 that unit conditions would exist that would require the information that is provided by this instrumentation. The remedial action changes are consistent with the NUREG if any of this instrumentation is inoperable. For this, and as explained above, the proposed changes are acceptable.
L33 The CTS SR 3.7.5.1 requirement to perform a manual Keowee start (SR 3.7.1.11) during operation above cold shutdown was not retained. The manual start function is only required to be operable when in shutdown (Modes 5 and 6) and during movement of irradiated fuel assemblies. Consequently, the SRs should only be applicable under these conditions. In addition, the accident analyses do not take credit for a manual Keowee start during operation above cold shutdown. Therefore, this change is acceptable.
Section 3.4 - Less Restrictive L1 For measured leakage > 1.0 gpm (gallons per minute) and 5.0 gpm, CTS 3.1.6.10.a.2 and 3.1.6.10.a.3 require that the increase in the measured rate of PIV leakage since the previous test does not reduce the margin between the previous leakage and the 5.0 gpm limit by 50 percent. This requirement was not retained in the ITS. The provisions removed were intended to retain a margin of safety based upon changes from past leakage rate measurements. Since past performance has not been shown to be a reliable means for predicting future performance, based on a long history of leakage determinations, the additional limitations of the CTS over those established by the STS is not warranted. Overall, leakage rates must be within acceptable limits, without regard to changes based on past performance. Therefore, this change is acceptable.
L2 CTS 3.1.6.8 permits the reactor coolant system (RCS) leak detection system sensitive to radiation to be inoperable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This was changed to permit the containment atmosphere radiation monitor to be inoperable for 30 days. This change was based upon recognition that at least one other form of leakage detection is available. This is consistent with the STS and is based on the fact that a loss of RCS leak detection capability using the radiation monitors would not warrant more restrictive remedial actions. Therefore, this change is acceptable.
L8 A Note was added which states that the E determination is not required to be performed until 31 days after operating for a minimum of 2 effective full power days and 20 days of Mode 1 since the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. These provisions are consistent with the STS and ensures that radioactive materials are at equilibrium so that the analysis is representative and not skewed by a crud burst or similar event.
Therefore, this change is acceptable.
L15 Note 2 to CTS Table 4.1-3, Item 1.c requires E determination to be started when gross gamma activity indicates > 10p Cilml and be determined for each 10,bCi/ml increase thereafter. This requirement was deleted. An E determination is required to be performed every 184 days consistent with the STS and under the conditions noted under L8 above. Therefore, this change is acceptable.
9902050347 990128 PDR ADOCK 05000269 P
-20 To the extent that requirements and information have been relocated to licensee-controlled documents, such information and requirements are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.
Further, where such information and requirements are contained in LCOs and associated requirements in the CTS, the NRC staff has concluded that they do not fall within any of the four criteria in 10 CFR 50.36 (discussed in Part II of this safety evaluation). Accordingly, existing detailed information and specific requirements, such as generally described above, may be deleted from the CTS.
E. Relocated Specifications Section 50.36 states that LCOs and associated requirements that do not satisfy or fall within any of the four criteria specified in 50.36(c)(2)(ii) may be relocated from existing TS (an NRC controlled document) to appropriate licensee-controlled documents. These requirements include the LCOs, Action Statements (Actions), and associated SRs. In its application, the licensee proposed relocating such specifications to the Chapter 16, "Selected Licensee Commitments" of the UFSAR. The staff has reviewed the licensee's submittals, and finds that relocation of these requirements to the UFSAR (Chapter 16) is acceptable, in that changes to these documents will be adequately controlled by 10 CFR 50.59. The licensee, in electing to implement the specifications of STS, also proposed, in accordance with the criteria in 10 CFR 50.36, to entirely remove certain TS from the CTS and place them in licensee-controlled documents noted in Table R. Table R lists all specifications and specific CTS details that are relocated, based on 10 CFR 50.36, to licensee-controlled documents in ITS. Table R provides:
a CTS reference; a summary description of the requirement; the name of the document that retains the CTS requirements; and the method for controlling future changes to relocated requirements. The NRC staff evaluation of each relocated specification and specific CTS detail presented in Table R is provided below.
CTS 3.5.2.2.b.6 and CTS 4.7.2, Control Rod Drive Patch Panel Operation The control rod drive patch panels are a feature of the control rod drive mechanism (CRDM) power supplies that provide the capability to patch (i.e., "program") any rod into any group with the exception of.Group 8. This feature provides flexibility in establishing the value of rod worth between rod groups. The panels are located in two locked cabinets. The program ensures that the control rods are programmed to operate in the core position and rod group that is consistent with the licensing basis.
The control rod program verification test (CTS 4.7.2) verifies that the control rods are programmed in the proper sequence, in accordance with the installed program. The program also requires an independent verification of power or instrumentation cables when they are disconnected and reconnected at the bulkhead or on the reactor vessel head. If the provisions of the verification program are not met, a control rod is inoperable (CTS 3.5.2.2.b.6), which is carried forward as a selected licensee commitment. These provisions are prudent post maintenance testing checks that assure that the control rods are properly programmed so they can perform their intended control function following maintenance operations that could impact their normal control function. It is inherent in any post maintenance procedure to assure that adequate checks and tests are performed prior to returning equipment to service. (Note that ITS 3.1.4.2 will verify that each individual rod moves in response to a command for a change in its position and is not affected by this change).
-24 Therefore, these requirements do not satisfy the criteria of 10 CFR 50.36 for inclusion in the ITS and their relocation to Chapter 16 of the UFSAR is acceptable.
CTS Table 4.1-2, Item 8, High Pressure Service Water (HPSW) Pumps and Power Supplies CTS Table 4.1-2, Item 8, contains an SR for which there is no corresponding LCO in CTS Section 3. The HPSW pumps are used primarily for fire protection throughout the Oconee station. In the event of a loss of the normal Low Pressure Service Water (LPSW) supply, the HPSW system automatically supplies cooling water to the HPI pump motor coolers. For loss of ac power, HPSW via the elevated water storage tank automatically supplies cooling water to the turbine driven emergency feedwater pump and its associated oil cooler, and maintains condenser circulating water (CCW) pump bearing cooling water and cooling water for the CCW pump motors. This SR is not associated with detection of abnormal degradation of the RCS boundary, it is not an initial condition of a design-basis accident or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier, it is not a structure, system, or component that is part of the primary success path for design-basis accident mitigation, and is a noncontributor to the CDF and plant risk. Therefore, the requirement does not satisfy the criteria of 10 CFR 50.36 for inclusion in the ITS and its relocation to Chapter 16 of the UFSAR is acceptable.
CTS 3.12, RCS Polar Crane and Auxiliary Hoist CTS 3.12 contains the requirements for the RCS Polar Crane and Auxiliary Hoist. The
.specification applies to the use of the RCS polar crane over the SG compartments and the fuel transfer canal and the auxiliary hoist over the fuel transfer canal. These restrictions preclude the dropping of materials or equipment into the reactor vessel and possibly damaging the fuel to the extent that an escape of fission products would result.
The fuel transfer canal is delineated by readily visible markers at an elevation above which the RCS polar crane does not normally handle loads. Restriction in the use of the RCS polar crane over the SG compartments is administratively controlled to preclude damage to the SGs and the RCS system.
The limits are not associated with detection of abnormal degradation of the RCS boundary, they are not an initial condition of a design-basis accident or transient analysis that either assumes the failure of or challenge to the integrity of a fission product barrier, they are not a structure, system, or component that is part of the primary success path for design-basis accident mitigation, and they are a noncontributor to CDF and plant risk. Therefore, these requirements do not satisfy the criteria of 10 CFR 50.36 for inclusion in the ITS and their relocation to Chapter 16 of the UFSAR is acceptable.
CTS 4.16, Radioactive Material Sources CTS 4.16 imposes a SR which implies an LCO exists. However, there is no corresponding LCO for the Radioactive Material Sources in CTS Section 3. The CTS specification requires leakage testing for sealed sources containing radioactive material in nongaseous form, other than tritium with a half life greater than 30 days. This specification ensures that leakage from byproduct, source and special nuclear material seal sources do not exceed allowable limits.
Sealed sources are exempt when the source contains < 100 micro Curies of beta and/or
- 29 6.
Description:
The applicability of ITS 3.3.14 would be expanded to include Mode 4 when the SG is relied upon for heat removal, which then would be consistent with the applicability of ITS LCO 3.7.5 for the emergency feedwater (EFW) system. The NUREG specification combines the EFW system initiation, main steamline isolation, and main feedwater isolation functions into one specification. The specification titles, LCOs, Actions, and SRs would be modified to reflect Oconee-specific terminology and design requirements. Where appropriate, ITS-required actions would be based on similar NUREG-required actions. Requiring the EFW pump initiation circuitry to be OPERABLE in MODE 4 when relied upon for decay heat removal is a more restrictive requirement with no comparable NUREG requirement.
Evaluation:
This proposed ITS is more restrictive than the CTS and is consistent with STS 3.7.5 for the requirement of the EFW system. This beyond-scope item is acceptable.
7.
Description:
ITS 3.3.15 Action A.1 would be added to allow 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to declare the turbine stop valves (TSVs) inoperable prior to requiring that the unit shut down when one or more TSV closure channels is inoperable. The NUREG specification combines the EFW system initiation, main steamline isolation, and main feedwater isolation functions into one specification. The
.specification titles, LCOs, Actions, and SRs would be modified to reflect Oconee-specific terminology and design requirements. Where appropriate, ITS-required actions would be based on similar NUREG-required actions. Action A added to require declaring the TSVs inoperable when one or more TSV closure channels is inoperable, which although not directly comparable to NUREG required actions, is comparable to similar NUREG required actions for main steam isolation valves at other facilities.
Evaluation:
This Required Action is consistent with STS 3.3.7, Required Action A.2 and is more restrictive than the CTS relative to the required shutdown time. This beyond-scope item is acceptable.
8.
Description:
CTS 3.8.10 and 4.4.4.5 frequency would be changed from "...immediately prior to refueling operation" to "Once each refueling outage prior to Core Alterations or movement of irradiated fuel assemblies within containment" in ITS SR 3.3.16.2 for testing frequency of the radiation monitor associated with the purge system valve isolation and ITS SR 3.9.3.2 for testing isolation function of the reactor building purge supply and exhaust valves. These changes are less restrictive.
Evaluation:
The licensee indicated that the proposed change permitting the specified testing to be conducted once each refueling outage prior to core alterations or movement of irradiated
-32 12.
Description:
ITS 3.5.3 LCO Note 3 would be added to explicitly require that the low pressure injection (LPI) discharge header crossover valves be operable and capable of being opened manually when in Modes 1, 2, and 3. ITS 3.5.3 Action B would require that the LPI discharge header crossover valves be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of being discovered incapable of being manually opened when in Modes 1, 2, and 3. These changes are more restrictive.
Evaluation:
The Oconee CTS does not include an explicit requirement for manual operability of the LPI discharge header crossover valves. The proposed ITS 3.5.3 requires the manual operability of the LPI discharge crossover valves during Modes 1, 2 and 3. Also, it requires that the LPI discharge header crossover valves to be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of being discovered manually inoperable in Modes 1, 2 and 3. This proposed addition to the TS will support the current safety analysis assumption in the event of a core flood line break concurrent with a single failure of the unaffected LPI train. The LPI discharge header crossover valves must be capable of being manually opened for necessary accident mitigation. The staff finds this proposed addition to Oconee TS acceptable.
13.
Description:
ITS 3.5.3 would require the LPI system to be operable in Modes 1, 2, 3, and 4. LCO Note 1 would be added to specify that only one LPI train is required to be operable in Mode 4. LCO Note 2 would be added to allow an LPI train to be considered operable during alignment, when aligned, or when operating if capable of being manually realigned to the LPI mode of operation. Action D would be added to require action be initiated immediately to restore the required LPI train to operable status and to require the reactor to be placed in Mode 5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required LPI train cannot be restored to operable status (provided an RCS loop is available). These changes are more restrictive.
Evaluation:
The ITS requirements more closely conform to the purpose and design of the LPI system.
The actions have been reviewed by the staff and found to be acceptable since they provide the desired requirements for system operability.
The note that is added in the proposed ITS for not entering Mode 5 when a DHR loop is not operable is necessary to ensure that a method of DHR is available prior to entering the mode when it will be needed. This note and action is appropriate since, in this condition, the unit is not prepared to continue cooling down using the full complement of LPI pumps and LPI heat exchangers. The proposed changes are acceptable.