ML17262A192
| ML17262A192 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/15/1990 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| Shared Package | |
| ML17262A190 | List: |
| References | |
| NUDOCS 9010240117 | |
| Download: ML17262A192 (13) | |
Text
ATTACHMENT A Revise the Technical Specification pages as follows:
Remove Insert 3.6-2 3.6-2 3.6-4 3.6-5 3.6-6 3.6-7 3.6-7A 3.6-8 3.6-9 3.6-10 .
3.6-11 4.4-4 4. 4-4 4.4-7 4.4-7 4.4-8 4.4-8 4.4-11 4.4-11 90i0240ii7 901015 PDR ADOCK 05000244 P PDC
3.6.3 Containment Isolation Valves 3.6.3.1
~ ~ ~ With one or more of the isolation valve(s) specified in UFSAR Table 6.2-13 inoperable, maintain at least one isolation valve operable in each affected penetration that is open and either:
- a. Restore the inoperable valve(s) to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or
- b. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, or
- c. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange, or
- d. Be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Containment isolation valves are inoperable from a leakage standpoint when the demonstrated leakage of a single containment, isolation valve or cumulative total leakage of all containment isolation valves is greater than that allowed by 10 CFR 50 Appendix J.
3.6.4 Combustible Gas Control 3.6.4.1 When the reactor is critical, at least two independent containment hydrogen monitors shall be operable. One of the monitors may be the Post Accident Sampling System.
3.6.4.2 With only one hydrogen monitor operable, restore a second monitor to operable status within 30 days or be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
3.6.4.3 With no hydrogen monitors operable, r'estore at least one monitor to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
3.6.5 Containment Mini-Pur e Whenever the containment integrity is required, emphasis will be placed on limiting all purging and venting times to as low as achievable. The mini-purge isolation valves will remain closed to the maximum extent practicable but may be open for pressure control, for ALARA, for respirable air quality considerations for personnel entry, for surveillance tests that may require the valve to be open or other safety related reasons.
Amendment No. gg 3.6-2 Proposed
P 4.4.1.4 Acce tance Criteria e a. The leakage ra te at Pt (Ltm) shall be <0.75 xs defined as the containment vessel reduced pressure which is greater than or equal to 35 psig.
Lt. Pt test Ltm is defined as the total measured containment leakage rate at pressure Pt. Lt is defined as the maximum allowable leakage rate at pressure Pt.
- b. Lt shall be determined as Lt=La Sa se ~/~
which equals
.1528 percent weight per day at 35 psig. Pa is defined as the calculated peak containment internal pressure related to design basis accidents which is greater than or equal to 60 psig. La is defined as the maximum allowable leakage rate at Pa which equals
.2 percent weight per day.
- c. The leakage rate at Pa (Lam) shall be <0.75 La. Lam is defined as the total measured containment leakage rate at pressure Pa.
4.4.1.5 Test Fre uenc a ~ A set of three integrated leak rate tests shall be performed at approximately equal intervals during each 10-year service period. The third test of each set shall be conducted in the final year of the 10-year service period or one year before or after the final year of the 10-year service period provided:
1 ~ the interval between any two Type A tests does not exceed four years.
11 ~ following each in-service inspection, the containment airlocks, the steam generator inspection/maintenance penetration, and the equipment hatch are leak tested prior to returning the plant to operation, and 111 ~ any repair, replacement, or modification of a containment barrier resulting from the inservice inspections shall be followed by the appropriate leakage test.
4.4-4 Proposed
T. 1 0
shutdown and depressurized until repairs are effected and the local leakage meets the acceptance criterion.
Ce If it is determined that the leakage through a mini-purge supply and exhaust line is greater than 0.05 La an engineering evaluation shall be performed and plans for corrective action developed.
4.4.2.4 Test Fre uenc Except as specified in b. and c. below, individual penetrations and containment isolation valves shall be tested during each reactor shutdown for refueling, or other convenient intervals, but in no case at intervals greater than two years.
- b. The containment equipment hatch, fuel transfer tube, steam generator inspection/maintenance penetration, and shutdown purge, system flanges shall be tested at each refueling shutdown or after each use, if that be sooner.
Amendment No. gg 4.4-7 Proposed
~ ~
C. The containment air locks shall be tested at intervals of no more than six months by pressurizing the space between the air lock doors. In addition, following opening of the air lock door during the interval, a test shall be performed by pressurizing between the dual seals of each door opened, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the opening, unless the reactor was in the cold shutdown condition at the time of the opening or has been subsequently brought to the cold shutdown condition. A test shall also be performed by pressurizing between the dual seals of each door within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of leaving the cold shutdown condition, unless the doors have not been open since the last test performed either by pressurizing the space between the air lock doors or by pressurizing between the dual door seals.
Amendment No. gg 4.4-8 Proposed
~~
the tendon containing 6 broken wires) shall be inspected.
The accepted criterion then shall be no more than 4 broken wires in any of the additional 4 tendons. If this criterion is not satisfied, all of the tendons shall be inspected and if more than 5% of the total wires are broken, the reactor shall be shut down and depressurized.
4.4.4.2 Pre-Stress Confirmation Test a ~ Lift-off tests shall be performed on the 14 tendons identified in 4.4.4.1a above, at the intervals specified in 4.4.4.1b. If the average stress in the 14 tendons checked is less than 144,000 psi (60% of ultimate stress), all tendons shall be checked for stress and retensioned, 144,000 psi.
if necessary, to a stress of
- b. Before reseating a tendon, additional stress (6%)
shall be imposed to verify the ability of the tendon to sustain the added stress applied during accident conditions.
4.4.5 Containment Isolation Valves 4.4.5.1
~ ~ ~ Each isolation valve specified in UFSAR Table 6.2-13 shall be demonstrated to be operable in accordance with the Ginna Station Pump and Valve Test Program submitted in accordance with 10 CFR 50.55a.
4.4.6 Containment Isolation Res onse 4.4.6.1 Each containment isolation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.1-1.
4.4.6.2 The RESPONSE TIME of the containment isolation valves, as listed in UFSAR Table 6.2-13, shall be demonstrated to be within the limit at least once per 18 months. This response time includes only the valve travel times for all valves that change position.
Amendment No. gg 4.4-11 Proposed
I~ XI ATTACHMENT B The purpose of this amendment is to remove Table 3.6-1, "Containment Isolation Valves", from the R.E. Ginna Technical Specifications. The reference to Table 3.6-1 in Technical Specification sections 3.6.3.1, 4.4.5.1, and 4.4.6.2 will be deleted and replaced by a reference to UFSAR Table 6.2-13. In addition, the inoperability definition for Technical Specification 3.6.3.1 will be clarified. Technical Specifications 4.4.1.5, section a (ii) and 4.4.2.4, section b, will be revised to include the modified steam generator inspection/maintenance penetration.
Technical Specification 4.4.1.5, section a (ii) will also be clarified. The temporary notes associated with the purge system and mini-purge valves (Technical Specifications 3.6.5, 4.4.2.4 section a, and 4.4.2.4 section d) will be removed since the valves have been installed. Also, the acceptance criteria for containment leakage criteria in Technical Specification 4.4.1.4 will be clarified.
The 1988 Inservice Test (IST) Program provided a complete review of the Ginna Containment Isolation Valves and their testing requirements. The information obtained during this review was submitted to the NRC to define the IST requirements for the third ten-year interval at Ginna. This submittal was subsequently approved by the NRC. As a result of this submittal and approval, numerous clarifications were required of Technical Specification Table 3.6-1 and UFSAR Table 6.2-13. However, this amendment will remove Technical Specification Table 3.6-1. The necessary changes to UFSAR Table 6.2-13 have been completed. Attachment C contains the safety evaluation related to these changes for your information.
The removal of Table 3.6-1 from the Technical Specifications and the incorporation of the required information into Ginna UFSAR Table 6.2-13 will keep the listing of the Containment Isolation Valves within a licensee controlled document. Changes to this document can only be performed under the criteria of 10CFR50.59 to ensure that no unreviewed safety questions are related to the change. Any additional changes to UFSAR Table 6.2-13 will be submitted as part of the annual UFSAR update. In addition, a report summary of the changes to the Ginna UFSAR are furnished to the NRC on an annual basis.
The addition of the steam generator inspection/maintenance penetration to both the UFSAR Table and the necessary Technical Specification surveillance requirements is the result of a modification to enhance containment closure during mid-loop operation (Generic Letter 88-17) . No new containment isolation valves were added as a result of this modification. The addition of this penetration to the UFSAR Table and Technical Specifications 4.4.1.5, section a (ii) and 4.4.2.4, section b, causes the new penetration to be treated consistent with respect to the Personnel and Equipment Hatches, and the fuel transfer tube (see letter from R.C. Mecredy, RG6E, to A.R. Johnson, NRC, dated March 13, 1990) .
1 ~ ~ 1 The first line of Technical Specification 4.4.1.5, section a (ii) will also be modified to state "following each in-service inspection..." The hyphenation of "in.-service" is to correct a typographical error only. The replacement of "one" with "each" provides greater understanding of the test frequency requirements.
These changes are a minor clarification only and do not involve a technical change.
The changes related to the inoperability definition for containment isolation valves in Technical Specification 3.6.3.1 do not involve any technical changes. Instead, these clarifications will provide consistency with 10CFR50 Appendix J requirements. In addition, the changes with respect to containment leakage criteria in Technical Specification 4.4.1.4 are clarifications only. All terms contained in the definition for Lt will be specified in the Technical Specifications consistent with 10CFR50 Appendix J.
The temporary notes associated with the purge and mini-purge valves in Technical Specifications 3.6.5, 4.4.2.4 section a, and 4.4.2.4 section d will be removed since these valves have been installed.
This is not a technical change since the notes were only intended to be applicable until the completion of the necessary modifications.
In accordance with 10CFR50.91, these changes to the Technical Specifications have been evaluated to determine if the operation of the facility in accordance with the proposed amendment would:
involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2. create the possibility of a new or different kind of accident previously evaluated; or
- 3. involve a significant reduction in a margin of safety.
These proposed changes do not increase the probability or consequences of a previously evaluated accident or create a new or different type of accident. Furthermore, there is no reduction in the margin of safety for any particular Technical Specification.
The detailed changes are described in Table 1.
Therefore, Rochester Gas,and Electric submits that the issues associated with this Amendment request are outside the criteria of 10CFR50.91; and a no significant hazards finding is warranted.
kg ~ l 0
Attachment C (UFSAR Changes Safety Evaluation)