ML20207F246
| ML20207F246 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 02/22/1999 |
| From: | Robert Schaaf NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20207F249 | List: |
| References | |
| GL-91-08, GL-91-8, NUDOCS 9903110204 | |
| Download: ML20207F246 (12) | |
Text
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t UNITED STATES g-j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20665 4 001
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CONSUMERS ENERGY COMPANY DOCKET NO. 50-255 PALISADES PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.184 License No. DPR-20 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Consumers Energy Company (the licensee) dated March 26,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as arnended (the Act), and 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; E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all apdcable requirements have been satisfidd.
I 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-20 is hereby amended to read as follows:
The Technical Specifications contained in Appendix A, as revised through Amendment No. 184, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. Consumers Energy Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f,
9903110204 990222 Robert G. Schaaf, Project Manager PDR ADOCK 05000255 Project Directorate Ill-1 P
PDR Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: February 22, 1999
ATTACHMENT TO LICENSE AMENDMENT NO.184 FACILITY OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by amendment number and contain verticallines indicating the areas of change.
REMOVE INSERT i
i 1-2 1-2 3-40 3-40 3-40a 3-40a 3-40b 3-40c 3-40d 3-40e 3-40f 3-40g 4-20 4-20 4-21 4-21 l
4-22 4-22 4 23 4-23 4-23a 4-23a 4-23b i
m
PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION DESCRIPTION PAGE N0 1.0 DEFINITIONS 1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 2.1 SAFETY LIMITS - REACTOR CORE 2-1 2.2 SAFETY LIMITS - PRIMARY COOLANT SYSTEM PRESSURE 2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS - RPS 2-1 Table 2.3.1 Reactor Protective System Trip Setting Limits 2-2 B2.1 Basis - Reactor Core Safety Limit B 2-1 B2.2 Basis - Primary Coolant System Safety Limit B 2-2 B2.3 Basis - Limiting Safety System Settings B 2-3 3.0 LIMITING CONDITIONS FOR OPERATION 3-1 3.0 APPLICABILITY 5-1 3.1' PRIMARY COOLANT SYSTEM 3-lb 3.1.1 Operable Components 3-lb 3.1.2 Heatup and Cooldown Rates 3-4 Figure 3-1 Pressure - Temperature Limits for Heatup 3-5 Figure 3-2 Pressure - Temperature Limits for Cooldown 3-6 3.1.3 Minimum Conditions for Criticality 3-12 3.1.4 Maximum Primary Coolant Radioactivity 3-17 3.1.5 Primary Coolant System Leakage Limits 3-20 3.1.6 Maximum PCS 0xygen and Halegen Concentration 3-23 3.1.7 Primary and Secondary Safety Valves 3-24a 3.1.8 Over Pressure Protection Systems 3-25a Figure 3-4 LTOP Limit Curve 3-25c 3.1.9 Shutdown Cooling 3-25h 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM 3-26 3.3 EMERGENCY CORE COOLING SYSTEM 3-29 3.4 CONTAINMENT COOLING 3-34 3.5 STEAM AND FEEDWATER SYSTEMS 3-38 3.6 CONTAINHENT SYSTEM 3-40
-3.7 ELECTRICAL POWER SYSTEMS 3-41 B3.7 Bases - Electrical Power Systems B 3.7.1-1 3.8 REFUELING OPERATIONS 3-46 3.9 Deleted 3-49 i
Amendment No. M9, M4, +69,184
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1.0 DEFINITIONS (continued)
CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel to verify that it is OPERABLE, including any alarm and trip initiating function.
COLD SHUTDOWN The COLD SHUTDOWN condition shall be when the primary coolant is at SHUTDOWN BORON CONCENTRATION and T., is less than 210*F.
[.QHIU NMENT INTEGRITY CONTAINMENT INTEGRITY is defined to exist when:
a.
All nonautomatic containment isolation valves and blind flanges are closed (OPERABLE).
]
b.
The equipment hatch is properly closed and sealed.
l c.
At least one door in each air lock is properly closed and sealed.
d.
All automatic containment isolation valves are OPERABLE or are locked closed.
4 e.
The uncontrolled containment leakage satisfies Specification 4.5.
l CONTROL RODS l
CONTROL RODS shall be all full-length shutdown and regulating rods.
CORE OPERATING LIMITS REPORT (COLR) l The COLR is the document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with i
Specification 6.6.5.
Plant operation within these limits is addressed l
in individual Specifications.
j DOSE EQUIVALENT I-131 l
DOSE EQUIVALENT I-131 shall be that concentration of I 131 (yci/gm) l which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, 1-134 and I-135 actually present. The' thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Cal:ulation of Distance Factors for Power and Test Reactor Sites."
': - 2 Amendment No. M, 44, 64, W, 68, Me, 444, MS, 4W, Me, 4#,184
3.6 CONTAINMENT SYSTEM 3.6.1 CONTAINMENT INTEGRITY shall be maintained:*
a.
When the plant is above COLD SHUTDOWN, b.
When the reactor vessel head is removed (unless the PCS boron concentration is at REFUELING BORON CONCENTRATION), and c.
When positive reactivity changes are made by boron dilution or CONTROL ROD motion (except for testing one CONTROL R00 at a time).
ACTION:
With one or more containment isolation valves inoperable (including during performance of valve testing), maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
a.
Restore the inoperable valves 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 closed and deactivated automatic valve, closed manual valve, or blind flange; or c.
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 />.
3.6.2 The containment internal pressure shall not exceed:
a.
1.5 psig when above COLD SHUTDOWN and below HOT STANDBY; and b.
1.0 psig when in POWER OPERATION or HOT STANDBY.
With containment internal pressure above the limit, restore pressure to within the limit within I hour, or be in at least HOT GHUTDOWN 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 />.
3.6.3 The containment average air temperature shall not exceed 140*F when the plant is above COLD SHUTDOWN. With containment average air temperature above the limit, restore temperature to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 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 /> 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 />.
3.6.4 Two independent cor.teinment hydrogen recombiners shall be OPERABLE when the plant is in POWER OPERATION or HOT STANDBY. With one recombiner inoperable, restore the inoperable recombiner to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.6.5 The containment purge exhaust and air room supply isolation valv6s shall be locked closed whenever the plant is above COLD SHUTDOWN. With one containment purge exhaust or air room supply isolation valve not locked closed, lock the valve closed within I hour or be in at least HOT STANDBY 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 />.
Entry and exit is permissible through a " locked" air lock door to perform repairs on other air lock components. Penetration flow paths may be unisolated intermittently under administrative control.
3-40 Amendment No. H e, M 2,4 4, 184
"3. 6 CONTAINNENT SYSTEM (continued) 3.6.1 BA111 Maintaining CONTAINMENT INTEGRITY ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of
.I radioactive material to the containment atmosphere or pressurization of the i
containment. CONTAINMENT INTEGRITY also ensures that the release of radioactive material to the environment will be consistent with the assumptions used in Section 14 events of the Palisades FSAR.
COLD SHUTDOWN conditions assure that no steam will be formed and, hence, there would be no pressure butidup in the containment if the primary coolant system ruptures. REFUELING BORON CONCENTRATION provides sufficient SHUTDOWN i
MARGIN to precludes criticality under any circumstances.
A footnote to LC0 3.6.1 allows temporary deviation from the requirements of CONTAINMENT INTEGRITY. The allowance for air lock entry to perform repairs is discussed in the basis for Section 4.5.2.
The opening of locked or sealed-closed containment penetration flow paths on an intermittent basis under administrative control includes the following considerations:
(1) Stationing an operator, who is in constant communication with control room, at the valve controls, valves in an accident situation (2) Instructing this o>erator to close these
, and (3) Assuring t1at environmental conditions will not preclude access to close the valves nor preclude the valves from closing, and that this action will prevent the release of radioactivity outside the containment.
The Actions specified in LCO 3.6.1 provide time for trouble-shooting, repairs, and pressure testing of isolation valves or other components.
The containment design pressure of 55 psig would not be exceeded during a Main Steam Line Break (MSLB) or a loss of Coolant Accident (LOCA) if the average containment air temperature was $140*F and the internal containment pressure was $1.0 psig during react operation (or 51.5 psig when above COLD SHUTDOWN with the reactor shutdown)
The function of the hydrogen recombiners is to eliminate the potential breach of containment due to a sudden hydrogen-oxygen burn following a LOCA or MSLB.
The recombiners accomplish this by recombining hydrogen and oxygen in a slow continuous manner, to form water vapor. Operation of the hydrogen ru ombiners is manually initiated. Two 100% capacity, independent hydrogen recombiners are provided. A single recombiner is capable of maintaining the containment hydrogen concentration in containment below the hydrogen flammability limit.
The containment purge exhaust and air room supply isolation valves are required to be locked closed above COLD SHUTD0g because they are not assured to be capable of closing during DBA conditions To ensure that the valves are closed and that the seals have not degraded, a between the valves leak rate test is periodically performed. Maintaining these valves locked closed during plant operatiun ensures that excessive quantities of radioactive materials will not be released via the containment purge exhaust or air room sup>1y ventilation systems. The valves may be locked closed electrically, mecianically, or by other physical means.
References (1) FSAR, Section 14.18.
l (2) Standard Review Plan 6.2.4 and Branch Technical Position CSB 6-4.
3-40a Amendment No. E, 90, H8, 184
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l 4.5 CONTAINMENT TESTS
. 4.5.2 -
Local Leak Qatection Tests (continued)
I b.
Acceptance Critedd i
(1)
The total leakage from all penetrations and isolation valves shall not exceed O.60 L.
(2)
The leakage for a Personnel air lock door seal test shall not exceed 0.023 le-
- (3)
An acceptable Emergency Escape Airlock door seal contact check consists of a verification of continuous contact between the seals and the sealing surfaces.
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c.
Corrective Action (1)
If at any time it is determined that 0.60 L, is exceeded, repairs shall be initiated immediately. If repairs are not completed and conformance to the
+
acceptance criterion of 4.5.2.b(1) is not demonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the plant shall be placed 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 />.
1 (2)
If at any time it is determined that total containment leakage exceeds L _,
within one hour action shall be initiated to place the plant in at least HO,T 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 />.
l (3)
If the Personnel air lock door seal leakage is greater than 0.023 L,, or if the Emergency Escape Lock door seal contact check falls to meet its acceptance criterion, repairs shall be initiated immediately to restore the door seal to the acceptance criteria of specification 4.5.2.b(2) or 4.5.2.b(3).
In the event repairs cannot be completed within 7 days, the plant shall be placed 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 /> snd 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 />.
-l (4)
If air lock door seal leakage results in one door causing total containment leakage to exceed 0.60 L., the door shall be declared inoperable and the remaining OPERABLE door shall be immediately locked closed
- and tested within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. As long as the remaining door is found to be OPERABLE, the provisions of 4.5.2.c(2) do not apply. Repairs shall be initiated immediately to establish conformance with specification 4.5.2.b(1). In the event conformance to this specification cannot be established within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the plant shall be placed 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 />.
L r
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Entry and exit is permissible through a " locked" air lock door to perform repalm on the affected air lock components.
4-20 Amendment No. +e6,94,4i9,99,184
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4.5 CONTAINMENT TESTS i
4.5.2 Local Leak Detection Tests (continued) l d.
Test Freauency (1)
Individual penetrations and containment isolation valves shall be leak rate tested at a frequency of at least every refueling, not exceeding a two-year interval, except as specified in (a) and (b) below:
(a) The containment equipment hatch and the fuel transfer l
tube shall be tested at each refueling outage or after l
l each time used, if that be sooner.
3 (b) A full air lock penetration test shall be performed at six-month intervals. During the period between the six-month tests when CONTAINMENT INTEGRITY is required, a l
reduced pressure test for the door seals or a full air lock penetration test shall be performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after either each air lock door opening or the first of a series of opetings.
M 4.5.3 containment isolation Valvet l
a.
The isolation valves shall be demonstrated OPERABLE by performance l
of a cycling test and verification of isolation time for auto isolation valves price to declaring the valve to be OPERABLE after maintenance, repcir, or replacement work is performed on the valve or its associated actuator, control, or power circuit.
b.
Each isolation valvt4 shall be demonstrated OPERABLE by verifying l
that on each contaiement isolation right channel or left channel test signal, applicable isolation valves actuate to their required position during C010 SHUTDOWN or at least once per refueling cycle.
l c.
The isolation time of each power operated or automatic valve shall be verified in accordance with Section XI of the ASME Boiler and l
Pressure Vessel CtAe.
d.
Prior to the reattor going critical after a refueling outage, a visual check will be made to confirm that all " locked-closed" manual containnut isolation valves are closed and locked (except for valves that are open under administrative control as permitted by LC0 3.6.1).
e.
Each three months the isolation valves must be stroked to the position required to fulfill their safety function unless it is established tliat such operation is not practical during plant operation. T'he latter valves shall be full-stroked during each COLD SHUTDOW.
4-21 Amendment No. M6, M8, W4,184 m
4.5 CONTAINMENT TESTS (continued)
HAlil The containment 's designed for an accident pressure of 55 psig."'
While the remer is operating, the internal environment of the containment will not exceed a pressure of 1.0 psig or a temperature of 140*F. With these initial conditions, following a design basis LOCA, the steam-air mixture will not exceed 55 psig.
Prior to initial operation, the containment was strength-tested at 63 psig
. and then leak rate tested. The design objective of this preoperational leak rate test was established as 0.1% by weight per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 55 psig.
leakage rate is consistent with the construction of the containment,'"This which is equipped with independent leak-testable penetrations and contains channels over all unaccessible containment liner welds. which were independently leak-tested during construction.
Accident analyses have been performed on the basis of a leakage rate of 0.1% by weight per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With this leakage rate and with a reactor power level of 2530 MWt, the potential public exposure would be below 10 CFR 100 guideline values in the event of the Maximum Hypothetical Accident.'"
The performance of a periodic integrated leak rate test during plant life provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment.-
In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic leak rate test is to be performed without preliminary repairs or adjustments unless those repairs or adjustments are preceded and followed by local leak rate tests and the integrated leak rate results are adjusted to reflect the as found condition of the containment.
This nomal manner is a coincident two-of-four high radiation or two-of-four high containment pressure signals which will close all containment isolation valves not required for engineered safety features except the component cooling lines' valves which are closed by CHP only. The control system is designed on a two-channel (right and left) concept with redundancy and physicals,egaration.
Each channel is capable of initiating containment isolation The Type A test requirements including pretest test methods, test pressure, acceptancecriteria,andreportingreggementsareinaccordancewiththe Containment Leak Rate Testi g Program The frequency of the periodic integrated leak rate test is keyed to the refueling set.edule for the reactor because these tests can best be performed during refueling shutdowns. The specified frequency is based on three major l-considerations:
First is the low probability of leaks in the liner because of (a) the test of the leak tightness of the welds during erection; b) conformance of the complete containment to a low leak rate at 55 psig(during l
preoperational testing which is consistent with 0.1% leakage at design basis accident (DBA) conditions: and (c) absence of any significant stresses in the liner during reactor operation.
4-22 Amendment No. 499, 446, W4,184 '
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- 4.5. CONTAINMENTTESTS Basis (continued)
Second is the more frequent testing, at the full accident pressure, of those portions of the containment envelope Miat are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value (0.60L.) of the total leakage that is l
specified as acceptable from penstrations and isolation valves.
1 L
Third is the Containment Structural Integrity Surveillance Program which provides assurance that an important part, of the structural integrity of the containment is maintained.
' The basis for specification of a total leakage rate of 0.60 L, leak rate would remain from penetrations and isolation l
valves is specified to provide assurance that the integrated j
'specified limits during the intervals between integrated leak rate tests. This value allows for i
possible deterioration in the intervals between tests.'
The basis for specification of a Personnel air lock door seal leakage rate of 0.023 L,is to l.
3rovide assurance that the failure of a single air lock door will not result in the total containment eakage exceeding 0.60 L,. Due to its design, a seal contact check is used on the Emergency Escape air lock. The seal contact check is intended to provide assurance that the Emergency Escape air lock doors will not leak excessively. The 7 day period specified for restoring the air lock door leakage to within limits is acceptable since it requires that the total containment i
leakage limit is not exceeded, i
Action 4.5.2c(4) is modified by a footnote that allows entry and exit to perform repairs on the affected air lock component. After each entry and exit, the OPERABLE door must be immediately closed. If the outer door is inoperable, then it may be easily accessed i
for most repairs. However, if the inner door is inoperable, or if repairs on the outer door i
must be performed from the barrel side, then it is permissible to enter the air lock through the OPERABLE door, which means there is a short time during which the containment boundary is not intact (during access through the OPERABLE door). The ability to open the OPERABLE door, even if it means the containment boundary is temporarily not intact, is acceptable because of the low probability of an event that could pressunze the containment during the short time in which the OPERABLE door is i
p expected to be open.
CONTAINMENT INTEGRITY will be assured if a visual check is made of all manual containmeni isolation valves which are required to be locked closed, to verify they are actually closed and locked, prior to plant start up after a refueling outage where one or more valves could inadvertently be left open (except for valves that are open under administrative control as permitted by LCO 3.6.1).
i Containment isolation valves which are required to be locked closed are discussed in the FSARA. These valves are those manual containment isolation valves which are not opened during operation except as allowed by LCO 3.6.1.
l 4-23 l
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Amendment No. +2, +4,469,496,495,494,477,184 4
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i-4b CONTAINMENT TESTS Basis (continued)
A reduction in prestressing force and change in physical conditions are expected for the prestressing system. Allowances have been made in the J
reactor building design for the reduction and changes. The inspection i
results for each tendon inspected shall be recorded on the forms provided for that purpose and comparison will be made with previous test results and the j
initial quality control records.
I Force-time records will'be established and maintained for each of the tendon.
l groups, dome, hoop and vertical.
If the force measured for a tendon is less 1
than the lower bound curve of the force-time graph, two adjacent tendons will J
be tested.
If either of the adjacent or more than one of the original sample population falls below the lower bound of the force-time graph, an
)
investigation will be conducted before the next scheduled surveillance. The investigation shall be made to determine whether the rate of force reduction 1
I is indeed occurring for other tendons.
If the rate of reduction is confirmed, the investigation shall be extended so as to identify the cause of the rate of force reduction. The extension of the investigation shall determine the needed changes in the surveillance inspection schedule and the criteria and initial planning for corrective action.
l If the force measured for a tendon at any time exceeds the upper bound curve 3
of the band on the force-time graph, an investigation shall be made to determine the cause, j
l If the comparison of corrosion conditions, including chemical tests of the corrosion protection material, indicate a larger than expected change in the conditions from the time of installation or last surveillance inspection, an j
investigation shall be made to detect and correct the causes.
The prestressing system is a necessary strength element of the plant safeguards and it is considered desirable to confirm that the allowances are not being exceeded. The technique chosen for surveillance is based upon the rate of change of force and physical conditions so that the surveillance can either confirm that the allowances are sufficient, or require maintenance before minimum levels of force or physical conditions are reached.
The end anchorage concrete is needed to maintain the prestressing forces.
The design investigations concluded that the design is adequate. The prestressing sequence has shown that the end anchorage concrete can withstand loads in excess of those which result when the tendons are anchored. At the time of initial pressure testing, the containment building had been subjected to temperature gradients equivalent to those for normal operating conditions while the prestressing tendon loads are at their maximum.
However, after the initial pressure test both concrete creep and prestressing losses increase with the greatest rapidity and result in a redistribution of the stresses and a reduction in end anchor force.
Because of the importance of the containment and the fact that the design was new, it was considered prudent to continue the surveillance after the initial period.
4-23a Amendment No. 44, M9, W4,-tM,184
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l 4.5 CONTAINMENT TESTS g.
L-Basis (continued) l Containment dome delamination inspections performed in 1970 and 1982 have I
confirmed that no concrete delamination has occurred, The possibility that delamination might occur in the future is remote because dome tendon prestress forces gradually diminish through normal tendon relaxation and concrete strength normally increases over time. To account for this remote possibility, however, an additional delamination inspection will be performed in the event that 5% or more of the installed tendons must be retensioned to compensate for excessive loss of prestress. This inspection would be to confirm that any systematic excessive prestress loss did not result from delanination and that the retensioning process did not result in delamination.
References
-(1) Updated FSAR Section 5.8.1.
l j
(2) Updated FSAR Section 5.8.8 (3) Updated FSAR Section 14.22 l
(4) Updated FSAR Section 6.7.2.3 l
(5) 10 CFR Part 50, Appendix J.
(6) Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program", September 1995.
(7) Updated FSAR Section 5.1.
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4-23b Amendment No. 44, 409, 4M,184
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