ML20198H547
| ML20198H547 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 09/09/1997 |
| From: | Hopkins J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20198H553 | List: |
| References | |
| NUDOCS 9709190077 | |
| Download: ML20198H547 (11) | |
Text
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t UNITED STATES I~
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NUCLEAR REGULATORY COMMISSION L
t WASHINGTON. D.C. 30666-0001
...../
THE CLEVELAND ELECTRIC ILLUMINATING COMPANY. ET AL.
DOCKET NO. 50-440 PERRY NUCLEAR POWER PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE 1
Amendment No. 86 License No. NPF-58 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by The Cleveland Electric Illuminating Company. Centerior Service Company. Duquesne Light Company. Ohio Edison Company. OES Nuclear. Inc., Pennsylvania Power Company. and Toledo Edison Company (the licensees) dated January 31. 1997, supplemented August 6,1997. complies with the standards and recuirements of the Atomic Energy Act of 1954. as amended (the Act),
anc the Commission's rules and regulations set forth in 10 CFR Chapter I:
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission:
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations:
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public: and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-58 is hereby amended to read as follows:
97o9190077 97o9o9 PDR ADOCK 05000440 s
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'C (2) Technical Snecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B. as revised through Amendr.cnt No. 86. are hereby incorocrated into this license.
The Cleveland Electric Illuminating Company ' hall operate the facility s
in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented not later than 90 days after issuance.
FOR THE NUCLEAR REGULATORY COMMISSION d) f i
on B. Hopkins. Senior Project Manager
/ Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 9 1997
ATTACHMENT TO LICENSE AMENDMENT NO. 86 FACILITY OPERATING LICENSE NO. NPF-58 DOCKET NO. 50-440 Replace the following Jages of the Appendix "A" Technical Specifications with the attached pages. T19 revised pages are identified by Amendment number and contains vertical lines indicating the ared of change.
Remove Insert 1.0-3 1.0-3 3.6-2 3.6-2 3.6-7 3.6-7 3.6-18 3.6-18 3.6-19 3.6-19 3.6-60 3.6 60 5.0-15 S.0-15 5.0-15a l
Definitions 1.1 1.1-Definitions (continued)
EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM (ECCS) RESPONSE from when the monitored )arameter exceeds-its ECCS TIME initiation setpoint at tie chanr.el sensor until the ECCS equipment is ca)able of performing.its safety function (i.e., t1e valves travel to their required positions, pum their required values, p discharge pressures reach etc.).
Times shall include diesel generator starting and sequence loading delays, where applicable.
The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
Exceptions are stated in the individual surveillance requirements.
END OF CYCLE The E0C-RPT SYSTEM RESPONSE TIME shall be that RECIRCULATION PUMP TRIP time interval from initial movement of the (E0C-RPT) SYSTEM RESPONSE associated turbine stop valve or the turbine TIME control valve to complete suppression of the electric arc between the fully open contacts of the recirculation Jump circuit breaker. The response time may )e measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at-the channel sensor until the isolation valves travel to their required positions. The response time
-may be measured by means of any series of sequential, overlapping, or total ste the entire response time is measured.ps so that Exceptions are stated in tne individual surveillance requirements.
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(continued)
PERRY - UNIT 1 1.0-3 Amendment No. W,H,06
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Primary Containment-Operating t
3.6.1.1 1
SURVEILLANCE REQUIREMENTS SURVEILLANCE.
FREQUENCY d
SR 3.6.1.1.1 Perform required visual examinations and-In accordance leakage rate testing except for primary with the l-containment air lock testing. in Primary accordance with the Primary Containment Containment Leakage Rate Testing Program.
Leakage Rate Testing Program 1
PERRY - UNIT 1 3.6-2 Amendment No.46h, 86 1
Primary Containment Air Locks 3.6.1.2 l-SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.2.1
NOTES------------------
1.
aAn inoperable air lock door does not invalidate the previous successful performance of the overall air lock
' leakage test.
2.
During MODES 1. 2. and 3. results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.1.
Perform required primary containment air In accordance lock leakage rate testing in accordance with the with the Primary Containment Leakage Rate Primary Testing Program.
Containment Leakage Rate Testing Program SR 3.6.1.2.2 Verify primary containment air lock seal 7 days air header pressure is = 90 psig.
(continued)
PERRY - UNIT.1 3.6-7 Amendment No.49,86
^
PCIVs 3.6.1.3
..s SURVEILLANCE REQUIREMENTS (continued) i i
SURVEILLANCE FREQUENCY SR 3.6.1.3.9
NOTE-------------------
Only required to be met in MODES 1. 2.
and 3.
Verify the combined leakage rate for all In accordance s
secondary containment bypass leakage with the paths is s 0.0504 L, when pressurized to Primary a P.,
Containment Leakage Rate Testing Program SR 3.6.1.3.10
NOTE-------------------
Only required to be met in MODES 1, 2.
and 3.
Verify leakage rate through each main In accordance steam line is s 25 scfh when tested at with the a P., leakage rate through perating Cycle 6.
Until the end of O Primary the one main steam Containment o,V
-line is limited to s 35 scfh when tested Leakage Rate through. as long as the total leakage rate-Testing at a P all four main steam lines is Program s 100 scfh.
(continued)
PERRY - UNIT 1 3.6-18 Amendment No.6,86
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PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.1.3.11
NOTE---
Only required to be met in MODES 1. 2, l
and 3.
I Verify combined leakage rate of 1 gpm In accordance times the total number of PCIVs through with the hydrostatically tested lines that Primary penetrate the primary containment is not Containment exceeded when these isolation valves are Leakage Rate tested at = 1.1 P.
Testing Program SR 3.6.1.3.12
NOTE------------------
Only required to be met in MODES 1.
2, and 3.
Verify each outboard 42 inch primary 18 months containment purge valve is blocked to restrict the valve from opening > 50.
SR 3.6.1.3.13
NOTE-------------------
Not required to be met when the Backup Hydrogen Purge System isolation valves are open for pressure control, ALARA or air quality considerations for personnel entry, or Surveillances or special testing of th. Backup Hydrogen Purge System that require the valves to be open.
Verify each 2 inch Backup Hydrogen Purge 31 days System isolation valve is closed.
PERRY - UNIT 1 3.6-19 Amendment No. iME, 86
Drywell 3.6.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1.1 Verify bypass leakage is less than or
NOTE-----
equal to the bypass leakage limit.
The 3erformance However, during the first unit startup of tie drywell following bypass leakage testing bypass leakage performed in accordance with this SR, the test is acceptance criterion is s 10% of the extended to the drywell bypass leakage limit, sixth refueling outage and need not be performed during the fifth refueling outage.
18 months SR 3.6.5.1.2 Visually inspect the exposed accessible Three times interior and exterior surfaces of the durina each 10-drywell.
year service period, at approximately equal intervals.
(continued)
PERRY - UNIT 1 3.6-60 Amendment No.e;e2, 86
Programs an' Manuals 5.5 5.5 Programs and Manuals 5.5.10 Safety Function Determination Proaram (SFDP)
(continued)
The SFDP identifies where a loss of safety function exists.
If a
)
loss of safety function is determined to exist by this )rogram, the appropriate Conditions and Required Actions of the _C0 in which the loss of safety function exists are required to be entered.
5.5.11 Technical Soecifications (TS) Bases Control Procram This 3rogram provides a means for processing changes to the Bases for t1ese TS.
a.
Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
b.
Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
1.
a change in the TS incorporated in the license: or 2.
a change to the USAR or Bases that involves an unreviewed s5fety question as defined in 10 CFR 50.59.
c.
The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.
d.
Proposed changes that meet the criteria of S)ecification 5.5.11.b.1 or Specification 5.5.11.b.2 above siall be reviewed and approved by the NRC prior to implementation.
Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
5.5.12 Primary Containment Leakaae Rate Testina Proaram A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50. Appendix J. Option B as modified by a> proved exemptions, This program shall be in accordance wit 1 the guidelines contained in Regulatory Guide 1.163. " Performance Based Containment Leak-Test Program." dated September 1995, as modified by the following exceptions:
(continued)
PERRY - UNIT.1 5.0-15 Amendment No. e, 86 i
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment leakaae Rate Testina ProararD (continued)
BN-TOP-1 methodology may be used for Type A tests.
The corrections to NEl 94 01 which are identified on the Errata Sheet attached to the NEl letter. "A>pendix J Workshop Questions and Answers," dated Marc 1 19, 1996,are considered to be an integral part of NEI 94 01.
The peak calculated primary containment internal pressure for the design basis loss of coolant accident. P., is 7.80 psig.
The maximum allowable primary containment leakage rate. L., shall be 0.20% of primary containment air weight per day at the calculated peak containment pressure (P,).
Leakage rate acceptance criteria are:
a.
Primary containment leakage rate acceptance criterion is
< 1. 0 L,.
However, during the first unit startup following testing performed in accordance with this Program, the leakage rate acceptance criteria are < 0.6 L for'the Type B and Type C tests, and s 0.75 L, for the Type,A tests:
b.
Air lock testing acceptance criteria are:
1)
Overall air lock leakage rate is s 2.5 scfh when tested at a P,.
2)
For each door, leakage rate is s 2.5 scfh when the gap between the door seals is pressurized to a P,.
The provisions of SR 3.0.2 do not apply to tha test frequencies specified in the Primary Containment Leakage Rate Testing Program.
The provisions of SR 3.0.3 are ap)licable to the Primary Containment Leakage Rate Testing 3rogram.
PERRY - UNIT 1 5.0-15a-Amendment No. 86
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