ML20128P450
| ML20128P450 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/09/1996 |
| From: | Hansen A NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20128P456 | List: |
| References | |
| NUDOCS 9610170267 | |
| Download: ML20128P450 (18) | |
Text
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- t UNITED STATES g
., j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 2066H001 o
+4 * * * *
- 4 0
i WISCONSIN ELECTRIC DOWER COMPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 169 License No. DPR-24 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated May 29, 1996, as supplemented August 20, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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.
4 k
9610170267 961009 PDR ADOCK 05000266 P
. 1 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No.
DPR-24 is hereby amended to read as follows:
l B.
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.169, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective immediately upon issuance.
The Technical Specifications are to be implemented within 45 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION M A4-Allen G. Hansen, 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:
October 9, 1996
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g-UNITED STATES g
,[
NUCLEAR REGULATORY COMMISSION
't WASHINGTON, D.C. 30666-0001 49 * * * * *,o 1
WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.173 License No. DPR-27 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated May 29, 1996, as supplemented August 20, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; j
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 (1) 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 3.B of Facility Operating License No.
DPR-27 is hereby amended to read as follows:
B.
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.173, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective immediately upon issuance.
The Technical Specifications are to be implemented within 45 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
~
Allen G. Hansen, 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:
October 9, 1996
ATTACHMENT TO LICENSE AMENDMENT NOS. 169 AND 173 TO FACILITY OPERATING LICENSE NOS OPR-24 AND DPR-27 DOCKET N05. 50-266 AND 50-301 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
REMOVE INSERT 15.1-2 15.1-2 15.3.6-4 15.3.6-4 15.3.6-8 15.3.6-8 15.3.6-9 15.3.6-9 15.4.4-1 15.4.4-1 15.4.4-2 15.4.4-2 15.4.4-2a 15.4.4-3 15.4.4-2b 15.4.4-4 15.4.4-3 15.4.4-5 15.4.4-4 15.4.4-6 15.4.4-5 15.4.4-7 15.4.4-6 15.4.4-8 15.4.4-6a 15.4.4-6b 15.4.4-7 15.4.4-8 15.4.4-9 15.4.4-9a 15.4.4-9b 15.4.4-10 15.4.4-11 15.4.4-12 4
15.4.4-13 15.4.4-14 1
15.4.4-15 15.4.4-16 15.6.9-4 15.6.9-4 15.6.12-1
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l D.
Containment Integritv*
Containment integrity is defined to exist when:
1)
Penetrations required to be isolated during accident conditions are either:
l a.
Capable of being closed by an operable automatic containment isolation valve, OR b.
Closed by an operable containment isolation valve, OR c.
Closed in accordance with Specifications 15.3.6.A.1.b and 15.3.6.A.1.c.
2)
The equipment hatch is properly closed.
3)
At least one door in each personnel air lock is properly closed.
4)
The overall uncontrolled containment leakage is less than La.**
E.
Protective Instrumentation Logic 1)
Analog Channel An analog channel is an arrangement of components and modules as required to generate a single protective action signal when required by a plant condition. An analog channel loses its identity where single action signals are combined.
- Containment isolation valves are discussed in FSAR Section 5.2.
- Prior to the first startup after performing a required Containment Leakage Rate Testing Program leakage test, the applicable leakage limits specified in TS 15.6.12.D.2 must be met.
l l
l Unit 1 - Amendment No. 169 15.1-2 Unit 2 - Amendment No. 173
C.
Containment Purce Sunnly and Rvhaunt Valves The containment purge supply and exhaust valves shall be locked closed and may not be opened unless the reactor is in the cold shutdown or refueling shutdown condition.
(1)
One of the redundant valves in the purge supply and exhaust lines may be opened to perform the repairs required to conform with the Containment Leakage Rate Testing Program.
-(2)
If containment purge supply and exhaust penetration leakage results in exceeding the overall containment leakage rate acceptance criteria ( L. ), enter 15.3.6.A.1.a.
1 1
Unit 1 - Amendment No. 169 15.3.6-4 Unit 2 - Amendment No. 173
E.
corrAINMENT STRUCTURAL INTEGRTTY The structural integrity of the reactor containment shall be maintained in accordance with the surveillance criteria specified in the Containment Leakage Rate Testing Program and 15.4.4.II.
1.
If more than one tendon is observed with a prestressing force between the predicted lower limit (PLL) and 90% of the PLL or if one tendon is observed with prestressing force less than 90% of the PLL, the tendon (s) shall be restored to the required level of integrity within 15 days or the reactor shall be in hot standby within the next six hours 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 />. An engi-1 neering evaluation of the situation shall be conducted and a special report submitted in accordance with Specification 15.4.4.II.D within 30 days.
i 2.
With an abnormal degradation of the containment structural integrity in excess of that specified in 15.3.6.E.1, and at a level below the acceptance criteria of Specification 15.4.4.II, restore the contain-ment structural integrity to the required level within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot shutdown within the next six hours an ' dn 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 />.
Perform an engineerit.
valuation of the containment structural integrity and provide a special report in accordance with Specification 15.4.4.II.D within 30 days.
i Unit 1 - Amendment No. 169 15.3.6-8 Unit 2 - Amendment No. 173
l Basis Specification 15.3.6.A.1 The Reactor Coolant System conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the Reactor Coolant System ruptures.
Specification 15.3.6.A.1.a.
i The safety design basis for the containment is that the containment must withstand the pressures and temperatures of the design basis LOCA without exceeding the design leakage rate. The design allowable leakage rate (L.) is 0.4% of containment air weight per day at 60 psig (P ). "'
containment operability is maintained by limiting the overall containment leakage rate to within the design allowable leakage rate (L.).
Prior to the first startup after performing a required Containment Leakage Rate Testing Program leakage test, however, the applicable leakage limits specified in TS 15.6.12.D.2 must be met.
Compliance with Specification 15.3.6.A.1.a. will ensure a containment configuration that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis.
If penetration or air lock leakage results in exceeding L.,
Specification 15.3.6.A.1.a. shall be entered simultaneously with the LCO applicable to the penetration or air lock with the excessive leakage. Once the overall containment leakage rate is restored to less than L.,
Specification 15.3.6. A.1.a. may be exited and operation continued in accordance with the applicable LCo.
Specification 15.3.6. A.1.a. (1)
In the event the containment is inoperable, containment must be restored to operable status within one hour. The one hour completion time provides a period of time to correct the problem commensurate with the importance of maintaining containment integrity during plant operation. This time period also ensures that the probability of an accident (requiring containment integrity) occurring during periods when containment is inoperable is minimal.
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Unit 1 - Amendment No. 169 15.3.6-9 Unit 2 - Amendment No.173
15.4.4 CONTAINMENT TESTS Applicability 1
Applies to containment leakage and structural integrity.
4 Obiective i
To verify that potential leakage from the containment and the pre-stressing tendon loads are maintained within acceptable valuer Specification I.
Perform required visual examinations and leakage rate testing in accordance with the Containment Leakage Rate Testing Program.
l II.
TENDON SURVEILLANCE A.
Obiect In order to insure containment structural integrity, selected tendons shall be periodically inspected for symptoms of material deterioration or lift-off force reduction. The tendons for inspection shall be randomly but representative 1y selected from each group for each inspec-tion; however, to develop a history and to correlate the observed data, one tendon from each grcup shall be kept unchanged af ter initial selec-tion. Tendons selected for inspection will consist of five hoop tendons, three vertical tendons located approximately 120' apart, and three dome tendons, one from each of the three dome tendon groups.
B.
Freguency Tendon surveillance shall be conducted at five-year intervals in accordance with the following schedule:*
M Xgar surveillance Reemired 1
1984 Physical 2
1984 Visual 1
1989 Visual 2
1989 Physical
- Subsequent five-year interval inspections repeat this pattern.
Unit 1 - Amendment No.169 15.4.4-1 Unit 2 - Amendment No.173
C.
Yn=pectinna Tendon surveillance in accordance with 15.4.4.II.B shall consist of either a visual or physical inspection.
(1) Visual Inspection Tendon anchorage assembly hardware of the randomly selected a.
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tendons shall be visually examined to the extent practicable without dismantling load bearing components of the anchorage.
The immediate concrete area shall be checked visually for indications of abnormal material behavior.
(2)
Phvnieml Inspection j
a.
Tendons which are physically inspected shall first be visua) inspected in accordance with C. (1).
b.
All tendons which are physically inspected shall be subjected to a lift-off test to monitor their prestressing force.
(i)
If the prestressing force of a selected tendon in a group lies above the predicted lower limit, the tendon is considered to be acceptable.
(ii)
If the prestressing force of a selected tendon liea between the predicted lower limit and 90% of the predicted lower limit, two tendons, one on each side of.the test tendon, shall be checked for their prestressing forces.
If the prestressing forces for these tendons are above the predicted lower limit for the tendons, all three tendons j
shall be restored to the required level of integrity. A single deficiency shall be considered unique and accept-able.
If the prestressing force of either of the adjacent tendons falls below the predicted lower limit of the tendon, additional lift-off testing should be done if necessary, so that the cause and extent of such occurrence can be determined and the condition shall be considered an ab-normal degradation of the containment structure and the l
provisions of Specification 15.3.6.E are applicable.
(iii) If the prestressing force of the selected test tendon falls below 90% of the predicted lower limit, the tendon shall be completely detensioned and a determination shall be made as to the cause of the condition. Such a condition shall be considered an abnormal degradation of the containment l
structure and the provisions of Specification 15.3.6.E are applicable.
Unit 1 - Amendment No. 169 15.4.4-2 Unit 2 - Amendment No. 173
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(iv)
If the average of all measured tendon forces for each group (corrected for average condition) is found to be less than the minimum required prestress level at Anchorage location for that grcup, the condition should be considered i
as abnormal degradation of the containment structure and l.
the provisions of 15.3.6.E are applicable. The average i
minimum design values adjusted for elastic losses are as j
follows:"'
Hoop.
124.s kai Vertical 140.s kal Dome 137.4 kai J
f' c.
One randomly selected tendon from each group of tendons shall be subjected to complete detensioning in order to identify broken or damaged wires. During the retensioning of the detensioned tendon, simultaneous measurements of elongation and jacking force shall be made at a minimum of two levels of force between the required seating force and zero. During the detensioning and retensioning of the tendons tested, if the elongation corresponding to a specific load differs by more than 50 from that recorded during installation of the tendons, an investiga-tion shall be made to ensure that such discrepancies are not related to wire failures or slippage of wires in anchorages.
d.
A tendon wire shall be removed from the one tendon from each group which has been completely detensioned. The wire shall be inspected over its entire length to determine if evidence of corrosion or other deleterious effects are present. Tensile tests shall be made on three samples cut from each removed wire. The samples will be cut from the midsection and each end of the removed wire.
Failure of the material to demonstrate the minimum required tensile strength of 240,000 psi shall be considered an abnormal condition of the containment structure and the engineering evaluation provisions of specification l
15.3.6.E.1 are applicable.
If an acceptable justification for continued operation cannot be concluded from this evaluation, j'
then the shutdown requirements of Specification 15.3.6.E.1 are applicable.
e.
The sheathing filler grease will be sampled and inspected on each physically inspected tendon. The operability of the sheathing filler grease shall be verified by assuring:
1)
There are no voids in the filler material in excess of 5% of net duct volume.
Unit 1 - Amendment No. 169 15.4.4-3 Unit 2 - Amendment No. 173
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2)
Complete grease coverage exists for the different parts of the Anchorage system, and 3)
The chemical properties of the filler material are within the tolerance limits specified by the manufacturer.
D.
Renerte A final report documenting the results of each tendon surveillance will be prepared and maintained as a permanent plant record.
Abnormal conditions observed during testing will be evaluated to de-termine the effect of such conditions on containment structural integrity.
This evaluation should be completed within 30 days of the identification of the condition. Any condition which is determined in this evaluation to have a significant adverse effect on containment structural integrity j
will be considered an abnormal degradation of the containment structure.
Any abnormal degradation of the containment structure identified during the engineering evaluation of abnormal conditions shall be reported to the Nuclear Regulatory Commission pursuant to the requirements of 10 CFR 50.4 within thirty days of that determination. Other conditions that indicate possible effects on the integrity of two or more tendons f
shall be reportable in the same manner.
Such reports shall include a description of the tendon condition, the condition of the concrete 1
(especially at tendon anchorages), the inspection procedure and the corrective action taken.
III. End Anchorage cencrete surveillance A.
Specific locations for surveillance will be determined by information I
obtained from design calculations, as-built end anchorage concrete and prestressing records, observations of the end anchorage concrete during and after prestressing, and results of deformation measurements made i
during prestressing and the initial structural test.
B.
The inspection intervals will be approximately one-half year and one year after the initial structural test and shall be chosen such that the inspection occurs during the warmest and coldest part of the year following the initial structural test.
Unit 1 - Amendment No. 169 15.4.4-4 Unit 2 - Amendment No. 173
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C.
The inspections made shall include (1)
Visual inspection of the end anchorage concrete exterior surfaces.
(2)
A determination of the temperatures of the liner plate area or containment interior surface in locations near the end anchorage concrete under surveillance.
(3)
Measurement of concrete temperatures at specific end anchorage concrete surfaces being inspected.
(4)
The mapping of the predominant visible corerete crack patterns.
(5)
The measurement of the crack widths, by use of optical compara-tors or wire feeler gauges.
(6)
The measurement of movements, if any, by use of demountable mechanical extensometers.
D.
The measurements and observations shall be compared with those to which prestressed structures have been subjected in normal and abnormal load conditions and with those of preceding measurements and observations at the same location on the reacto containment.
E.
The acceptance criteria shall be as follows:
If the inspections determine that the conditions are favorable in comparison with experience and predictions, the close inspections will be terminated by the last of the inspections stated in the schedule and a report will be prepared which documents the findings and recommende the schedule for future inspections, if any.
If the inspections detect symptoms of greater than normal cracking or movements, an immediate investigation will be made to determine the cause.
lIV.
Idner Plate A.
The liner plate will be examined before the initial pressure test to determine the following:
(1) Locate areas which have inward deformations. The magnitude of the inward deformations will be measured and recorded.
The areas will be permanently marked for future reference.
The inward deformations will be measured between the angle stiffeners which are on 15-inch centers. The measurements will be accurate to 2.01 inch.
Unit 1 - Amendment No.169 15.4.4-5 Unit 2 - Amendment No.173
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)
l (2)
Try to locate areas having strain concentrations by visual examination paying particular at:ention to the condition of
{
the liner surface.
Record the location of any areas having strain concentrations.
B.
Shortly after the initial pressure test and at about one year after initial start-up, reexamine the areas located in section (A).
Measure and record inward deformations. Record observations per-taining to strain concentrations.
C.
If the difference in the measured inward deformations exceeds 0.25 inch (for a particular location) and/or changes in strai-9ancentra-tion exist, then an investigation will be made. The investigation will determine the cause and any necessary corrective action.
D.
The surveillance program will only be continued beyond the one year after initial start-up inspection if some corrective action was needed.
If required, the frequency of inspection for a continued surveillance program will be determined shortly af ter the "one year after initial start-up inspection".
E.
In addition to the preceding requirements, temperature readings will ba obtained at the locations where inward deformations were measured.
Temperature measurements will also be obtained on the outcide of the containment building wall.
BAR!.a The containment is designed for an accident pressure of 60 psig."'
While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a temperature of about 105'F.
With these initial conditions, the temperature of the steam-air mixture at the peak accident pressure of 60 psig is 286*F.
Prior to initial operation, the containment was strength tested at 69 psig and then leak-tested.
The design objective of this preoperational leakage rate test was established as 0.4% by weight per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig. This leakage rate is consistent with the construction of the containment,"' which is equipped with independent leak-testable penetrations and contains channels over all containment liner welds, which were independently leak-tested during construction.
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Unit 1 - Amendment No. 169 15.4.4-6 Unit 2 - Amendment No. t73
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Safety analyses have been performed on the basis of a leakage rate of 0.40% by weight per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig. With this leakage rate and with minimum containment engineered safety systems for iodine removal in operation, i.e. one cpray pump with sodium hydroxide addition, the public exposure would be well below 10 CFR 100 values in the event of the design basis accident.*
j The safety analyses indicate that the containment leakage rates could be slightly in excess of 0.75% per day before a two-hour thyroid dose of 300R could be j
received at the site boundary, i
l The performance of periodic integrated leakage rate tests during plant life provide a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment. These tests cre perforumut in accordance with the Containment Leakage Rate Testing Program, i
j Periodic visual and physical inspection of the containment tendons is the l
method to be used to determine loss of load-carrying capability because of j
wire breakage or deterioration. The tendon surveillance program specified in l 15.4.4.II is based on the recommendation of Regulatory Guide 1.35 Rev. 3.
Containment tendon structural integrity was demonstrated for both units at 4
i the end of one, three and eight years following the initial containment struc-I tural integrity test.
The pre-stress lif t-off test provides a direct measure of the load-carring j
capability of the tendon. A deterioration of the corrosion preventive proper-ties of the sheathing filler will be indicated by a change in the physical cppearance of the filler.
If the surveillance program indicates, by extensive wire breakage, tendon stress-strain relations, or other abnormal conditions, that the pre-stressing tendons are not behaving as expected, the abnormal conditions will be subjected to an engineering analysis and evaluation in accordance with specification 15.4.4.II.D to determine whether the condition could result in a significant adverse impact on the containment structural integrity.
The specified acceptance criteria are such as to alert attention to the situation well before the tendon load-carrying capability would deteriorate to a point that failure during a design basis accident might be possible.
Thus, the cause of the incipient deterioration could be evaluated and corrective action studied without need to shut down the reactor. If the engineering evaluation determines that the abnormal condition could result in a significant adverse impact on the containment structural integrity,.am abnormal degradation situation will be declared and a report submitted"to the i
NRC in accordance with the specifications.
The purpose of the leakage tests of the isolation valves in the containment purge supply and exhaust lines is to identify excessive degradation of the resilient seals for these valves.
Unit 1 - Amendment No.169 15.4.4-7 Unit 2 - Amendment No.173
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References I
(1)
PSAR Section 5.1.2.3 (2)
FSAR Section 5.1.2 (3)
FSAR Section 14.3.5 (4)
FSAR Section 14.3.4 4
(5)
FSAR Section 6.2.3 (6)
PSAR pages 5.1-86 and 5.1-87 e
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2
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Unit 1 - Amendment No.169 15.4.4-8 Unit 2 - Amendment No.173
15.6.9.2 Uniaue Reportino Raouir nts The following written reports shall be submitted to the Director, Office of Nuclear Reactor Regulation, USNRC f
l A.
Deleted B.
Poison Ass-hly h val From Soent Fuel Storaae Racka Plans for removal of any poison assemblies from the spent fuel i
storage racks shall be reported and described at least 14 days j
prior to the planned activity. Such report shall describe
'l
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neutron attenuation testing for any replacement poison assemblies, if r.pplicable, to confirm the presence of boron 5
)
- material, t
C.
Overoressure Mitiaatino System operation 11 j
In the event the overpressure mitigating system (power operated 1
i relief valves in the low temperature overpressure protection i
j mode) or residual heat removal system relief valves are operated i
to relieve a pressure transient which, by licensee's evaluation,
{
could have resulted in an overpressurization incident had the i
system not been operable, a special report shall be prepared and submitted to the commission within 30 days. The report shall describe the circumstances initiating the transient, the effect of the system on the transient and any corrective action necessary to prevent recurrence.
l Unit 1 - Amendment No. 169 15.6.9-4 Unit 2 - Amendment No. 173
.-.- -.=
15.6.12 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A.
A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J.
Option B, as modified by approved exemptions. This program shall be in
]
accordance with the guidelines contained in Regulatory Guide 1.163,
" Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:
)
1.
The interval between the 1992 Unit 2 Type A test and the next Unit j
2 Type A test shall be 60 months.
i B.
The peak calculated containment internal pressure for the design i> asis loss i
of coolant accident, P, is 53 psig.
C.
The maximum allowable primary containment leakage rate, L., at P., shall be 0.4% of containment air weight per day.
1
}
D.
Leakage rate acceptance criteria are:
i l
1.
The containment leakage rate acceptance criterion is $1.0 L,.
]
2.
During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are so.6 L, for the combined Type B and Type C tests and 50.75 L, for Type A
- tests, i
E.
The provisions of Specification 15.4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testi g Program.
l F.
The provisions of Specification 15.4.0.3 are applicable to the Containment l
Leakage Rate Testing Program.
I Unit 1 - Amendment No.
169 15.6.12-1 Unit 2 - Amendment No.
173
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