ML20070U293
| ML20070U293 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 01/26/1983 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20070U281 | List: |
| References | |
| TAC-8910, NUDOCS 8302100301 | |
| Download: ML20070U293 (25) | |
Text
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e 3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Test Applicability l
l Applies to containment leakage and structural integrity.
Objective To verify that the:
(1) potential leakage from containment is within acceptable limits, and (2) structural performance of all important components in the containment prestressing system is acceptable.
Specifications (1)
Containment Building Leak Rate Tests Tests shall be conducted to assure that leakage of the primary reactor containment and associated systems is main-tained within allowable leakage rate limits. Periodic surveillance shall be performed to assure proper maintenance and leak repair of the containment structure and penetrations during the plant's cperating life.
Definition of terms used in the leak rate testing specifi-cations:
Leakage rate - for test purposes is that leakage of contain-ment air which occurs in a unit of time at 60 psig that escapes to the outside atmosphere during a 24-hour test period.
Maximum allowable leakage rate (Ln) - the design bcsis leakage rate of 0.2% by weight of the containment atmosphere per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at a pressure of 60 psig; i.e., 209,850 sccm.
Overall integrated leakage rate - that leakage rate whien is obtained from a summation of leakage through all potential leakage caths including containment welds, valves, fittings, and components which penetrate containment.
Acceptance criteria - the standard against which test re-sults are to be compared for establishing the functional acceptability of the containment as a leakage limiting boundary.
(2)
Integrated Leak Rate Test (Type A Test) a.
Introduction Type A tests are intended to measure the reactor con-tainment overall integrated leakage rate at periodic in te rvals.
2100301 830126 3-37 ATTAC10!ENT A DR ADOCK 05000 a
3.0
. SURVEILLANCE REOUIREMENTS 3.5 Containment Test (Continued) b.
Pretest Requirements A general inspection of the accessible interior and exterior surfaces of the containment structures and components shall be performed prior to any Type A test-to uncover any evidence of structural deterioration which may affect either the containment structural integrity or leak-tightness.
If there is evidence of.
structural deterioration, the Type A tests shall not be performed until corrective action is taken in accordance with repair procedures, non-destructive examinations, and tests as specified in the applicable code specified in 10 CFR Part 50.55a at the commence-ment of repair work.
Such structural deterioration and corrective actions taken shall be reported as part of the Type A test report.
During the period between the initiation of the con-tainment inspection and performance of the Type A test, no repairs or adjustments shall be made so that the containment can he tested in as close to the "as is" condition as practical.
During the period between the completion of one Type A test and the initiation of the containment inspection for the subsequent Type A test, repairs or adjustments shall be made to com-ponerts whose leakage exceeds that specified in the Technical Specifications as soon as practical after 1
identification. This requirement is interpreted not to preclude performance of Type B and Type C testing and required repairs prior to initiation of the con-tainment inspection and the performance of'the Type A l
test.
I r
If during a Type A test, potentially excessive leakage paths are identified which interfere with satisfactory completion of the test, or which result in the Type A test not meeting the acceptance criteria, the Type A test shall be temporarily suspended. Thereafter, l
repairs and/or adjustments to equipment shall be made and the Type A test resumed. The corrective action taken, the change la leakage rate resulting from the l
repairs and overall integrated leakage determined from the Type A and local leak rate tests shall be included in a report submitted to the Commission.
Closure of containment isolation valves for the Type A test shall be accomplished by normal operation and I
without any preliminary exercising or adjustments (e.g., no tightening of valve after closure by valve motor).
Repairs of maloperating or leaking valves shall be made as necessary.
Information on any valve closure malfunction or valve leakage that requires corrective action before the test, shall be included in the Type A Leak Test Report submitted to the Commission.
3-38
3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Test (Continued)
The containment test conditions shall stabilize for a
. period of approximately 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s-prior to the start of a leakage rate test.
Those portions of the fluid systems that are part of the reactor coolant pressure boundary and are open -
directly to the containment atmosphere under post-accident conditions and become an extension of-the boundary of the containment shall be opened or vented to the containment atmosphere prior to and during the-test.
Portions of closed svstems inside containment that penetrate containment and rupture as a result of a loss of coolant accident shall be vented to the containment atmosphere. All vented systems shall be drained of water or other fluids to the extent neces-sary to assure exposure of the system containment isolation valves-to containment air test pressure and to assure they will be subjected to the post-accident differential pressure.
Systems that are required to maintain the plant in a safe condition during the test shall be operable in their normal mode, and need not be vented.
Systems that are normally filled with water and operating under post-accident conditions, such as the containment heat removal system and the component cooling water system, need not be vented. However, the containment isolation valves in the systems defined in this section shall be tested in accordance with Section
- 3. 5 (4).
The measured leakage rate from these tests shall be reported to the Commission.
c.
Test Methods All Type A tests shall be conducted in accordance with the provisions of 10 CFP.Part 50, Appendix J.
The accuracy of any Type A test shall be verified by a supplemental test.
The supplemental test method
~
selected shall be conducted for sufficient duration to establish accuratelv the change in leakage rate between the Type a test and the supplemental Type A test.
Results from the supplemental test are acceptable provided the difference between the supplemental test data and the Type A test data is within 0.25 L i 1*"**
a 52,460 sccm.
If results are not within 52,460 sccm, the reason shall be determined, corrective action taken, and a successful supplemental test performed.
Test leakage rates shall be calculated using absolute values corrected for instrument error, d.
Acceptance Criteria The maximum allowable leakage rate shall not exceed 209,850 secm.
3-39
=
3.0 SURVEILLANCE REQUIREMENTS 3.5 containment Test (continued)
The total measured leakage rate at a pressure of 60 psig shall be less than 0.75 La; i.e., 157,380 scem.
If local leakage measurements are taken to effect repairs in order to meet 157,380 seem acceptance-criteria, these measurements shall be taken at a pressure of 60 psig.
If two consecutive Type A tests fail to meet the acceptance criteria, not withstanding the requirements of the testing frequency, a Type A test shall be per-formed at each refueling outage or approximately every 18 months, whichever occurs first, until two consecu-tive Type A tests meet the. acceptance criteria, after which time the normal testing frequency schedule may be resumed.
e.
Testing Frequency A set of three Type A tests shall be performed, at approximately equal intervals during each 10 year-service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year in-service inspections.
The performance of Type A tests shall be limited to periods when the plant facility is non-operational and secured in the shutdown condition under administrative control and in accordance with the safety procedures defined in the license.
(3)
Containment Penetrations Leak Rate Tests (Type B Tests) a.
Intreduction Type B tests are intended to detect local leaks and to measure leakage across each pressure-containing'or leakage limiting boundary for the containment pene-trations.
b.
Test Methods Type B tests shall be performed by local pneumatic pressurization of the containment penetrations, either individually or in groups, at a pressure of 60 psig.
Examination shall be performed by halide leak-detection method or by other equivalent test methods such as measurement of the rate of maken reavired to maintain the test volume at 60 psig.
)
3-40 1
3.0 SURVEILLANCE PEOUIREMENTS 3.5-Containment Test (Continued) c.
Acceptance Criteria The combined leakage rate of all penetrations and valves subject to Type B and Type C tests shall be less
^
than or equal to 0.6 L ;
i.e., 125,910 scem.
a If at a.ny time it is determined that a leakage rate is greater than 125,910 scem, repairs shall be initiated immcdiately.
If repairs are not completed and con-formance to.the acceptance criteria 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 reactor shall be shutdown and
-depressurized until repairs are completed and the local leakage meets this acceptance criteria.
The results of personnel access lock door seal tests at 5 psig shall not exceed.01 L ; i.e., 2,098 secm.
a d.
Testing Frequency Type B tests shall be performed during each refueling outage, or other convenient intervals, but in no case at intervals greater than 2-years, except the personnel' access lock (PAL) which will be tested as follows:
(i)
Every six months the-entire PAL assembly shall be leak tested at 60 psig.
(ii)
If the PAL is opened during periods when con-tainment integrity is not required, the PAL door seals shall be Icak tested at 5 psig at the end of such periods and the entire PAL assembly shall then bc leak tested at 60 psig within two weeks of achieving the required '
condition for' containment integrity.
(iii)
If the PAL is opened during the intervci between the six uonth tests when containment integrity is required, the PAL door seats shall be leak tested at a pressure not less than 5 psig within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
If the PAL is opened more frequently than once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the door-seals shall be leak tested at a pressure of 5-psig at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during the period of frequent openings.
Penetrations to be Tested (I) c.
(i)
Equipment Hatch (ii)
Personnel Access Lock (iii)
Mechanical Penetrations M-1 through M-100 t
s (iv)
Electrical Penetrations 3-41
.-~
3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Test (Continued) 4 A-1 B-9 D-6 F-2 E-HCV-383-3A A-2 B-10 D-7 F-4 E-HCV-383-3B A-4 B-11 D-8 F-5 E-HCV-383-4A A-5 C-1 D-9 F-6 E-HCV-383-4B
~A-6 C-2 D-10 F-7 A-7 C-4 D-ll F-8 A-8 C-5 E-1 F-9 A-9 C-6 E-2 F-10 A-10 0-7 E-4 F-11 A-ll C-8 E-5 G-1 B-1 C-9 E-6 G-2 B-2 C-10 E-7 G-3 B-4 C-11 E-8 G-4 B-5 D-1 E-9 H-1 B-6 D-2 E-10 H-2 B-7 D-4 E-]l H-3 B-8 D-5 F-1 H-4 (4)
Containment Isolation Valves Leak Rate Tests (Type C Tests) a.
Introduction Type C tests are intended to measure containment iso-lation valve leakage rates.
b.
Test Methods.
Type C tests shall be performed by local pressurization with air or nitrogen at a pressure of 60 psig. The pressure shall be applied in the same direction as that when the valve would be required to perform its safety function, unless it can be determined that the results from the tests for a pressure applied in a different direction will provide equivalent or more conservative results. Each valve to be tested shall be closed by normal operation and without any preliminary exercising or adjustments (e.g., no tightening of valve af ter closure by valve motor).
A c.
Acceptance Criteria l
The combined Icakage rate of all penetrations and valves subject to Type B and Type C tests shall be less than or equal to 125,910 secm.
[
If at any time it is determined that a leakage rate is greater than 125,910 scem, repairs shall be initiated immediately.
If repairs are not completed and con-formance to the acceptance criteria 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 reactor shall be shutdown and depressurized until repairs are completed and the local leakage meets this acceptance criteria.
3-42 m,__
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3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Test (Continued) d.
Testing Frequency Type C rests shall be performed during each refueling.
outage, or other convenient intervals, but in no case at intervals greater than 2 years.
Penetrations to be Tested ( )
e.
M-2 M-31 M-52 M-7 M-38 M-53 M-8 M-39 M-57 M-11 M-40 M-58 M-14 M-42 M-69 M-15 M-43 M-73 M-18 M-44 M-74 M-19 M-45 M-79 M-20 M-46 M-80 M M-47 M-87 M-24 M-48 M-88 M-25
-M-50 M-HCV-383-3 M-30 M-51 M-HCV-383-4 (5)
Special Testing Requirements Any major modification or replacement of a component which is part of the containment boundary shall be followed by either Type A, Type B, or Type C tests as applicable for the area affected by the modification and shall meet the applicable acceptance criteria. Minor modifications, or replacements, performed directly prior to the conduct of a scheduled Type A test do not require a separate test.
(6)
Report on Test Results Leak rate tests shall be the uubject of a summary technical report submitted to the Commission approximately three months after the conduct of each test.
The report shall be titled " Reactor Containment Building Integrated Leak Rate Test".
The report chall contain an analysis and interpretation of the Type A test results and a summary analysis of periodic Type B and Type C tests that were performed since the last Type A test.
Amendment No. 24 3-43
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3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Test (Continued)
Leakage test results from Type _A, B, and C tests that failed to meet the applicable acceptance criteria shall be reported in a separate summary report approximately three months after the conduct of these tests.
The Type A test report shall include an analysis and interpretation of the test data, the least-squares fit analysis of the test data (Type A test only), the instrumentation error analysis (Type _A test only), and the structural conditions of the containment-or components, if any, which contributed to the failure in meeting the acceptance criteria.
Results and analyses af the supplemental verification test employed to demonstrate the validity of the leakage rate test measurements shall also be included.
(7)
Surveillance for Prestressing System a.
Surveillance Requirements 210 dome tendons and 616 wall tendons shall be inspected for symptoms of material deterioration or force reduction.
Inspections will be performed on three dome tendons, one from each Jayer, and on three wall tendons of each orientation.
The tendons shall be inspected as follows:
(i)
Lift-off readings shall be taken on each of the tendons selected to determine the load existing in the tendon at the time of inspection. At each surveillance period, readings may also bc taken on the load cells of the special instru-mented tendons.
Force reductions on the sur-veillance tendons and on the instrumented tendons will be compared.
If good correlation exists between these two groups of tendons through several surveillance periods, con-sideration will be given to eliminating some lift-off readings and monitoring of the load cells as an citernative. Each selected tendon shall be completely detensioned and examined for broken wires and any evidence of damage or deterioration of anchorage hardware.
(ii)
One wire from each of three helical tendons and
[
one wire of a deme tendon shall be removed.
l Each removed wire shall be carefully examined over its entire length for evidence of cor-resion or other deleterious effects.
Tensile I
tests shall be made on at least three samples cut from each of the four wires removed, one at each end end ene et midlength, the samples i
being of a maximum length practical for test-ing.
In special cases, the use of fatigue tests and accelerated corrosion tests may be considered.
3-44
e 3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Test (Continued)
(iii)
Comparisons shall be made between the quality control records and each of the surveillance inspection records for each of the surveillance tendons.
Af ter completion of the tendon surveillance the indi-vidual detensioned tendons shall be retensioned to a fcrce commensurate with the average wire stress indi-cated by the last lif t-off reading for that tendon.
b.
Acceptante Criteria (1)
The tendon force determined by the lift-off test shall be considered adequate if it is not less than the force shown on the appropriate lower limit curve of USAR Figure 5.10-4, as l_
adjusted for wire removal, for the elapsed time between the original prestressing and the particular surveillance period. These lower limit curves have been generated by calculating the dif ference between the anticipated tendon force at end of plant life and the minimum tendon force to meet the design requirements.
Que half of this difference has been added to the anticipated total loss of prestress at the.
end of plant life and the curves have been drawn to meet this limit.
Since the lock-off force on individual tendons is varied to com-pensate for elastic shortening of the structure, the tendon force at 707. of ultimate strength, rather taan the actual lock-off force shall be taken as the initial prestress force. An allowable limit of not more than one defective tendon out of the total. sample population is acceptable, provided an adjacent tendon on each side of the defective tendon is tested and is found to meet criteria.
Should one of the adjacent tendons be also found defective, the Commission shall be notified in accordance with Regulatory Guide 1.16, " Reporting of Operating Information".
(ii)
No unexpected change in corrosion conditions or grease properties.
(iii)
All three tensile tests on any one wire indi-cate an ultimate strength at least equal to the specified minimum ultimate strcngth of the wire.
If a single test on any wire shows an ultimate strength less than the specified mini-mum, the Commission will be notified in accord-ance with Regulatory Guide 1.16, " Reporting of Operating Information".
3-45
P d
3.0 SURVEILLANCE REQUIREMENTS:
3.5 Containment Test (Continued) i c.
Corrective Action i.
~If~the above acceptance criteria are not met,'an im-mediate investigation shall be made to determine the cause(s) for the non-conformance to the criteria, and results will be reported to the Commission within'90 i
days.
d.
Testing Frequency The tendons in the prestressing system shall be in -
spected once every five years.
Basis The containment is designed for an accident pressure of 60 psig. (2).
i While the reactor is operating, the internal environment of the-containment will be air at approximately atmospheric pressure. and a maximum temperature of about 120 F.
With these initial condi-tions the temperature of the steam-air mixture at the peak accident-pressure of 60 psig is 288 F. -
j' Prior to initial operation, the containment was strength-tested at l
69 psig and then was leak-tested. The design objective of the pre-operational leakage rate' test has been established as 0.2% by
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weight for 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 eith the construction of the containment, which'is equipped with l
independent leak-testable penetrations and contains channels over all inaccessible containment liner welds, which were independently l
1eak-tested during construction.
Safety analyses have been performed on the basis of a leakage ~ rate of 0.2% of the free volume.per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' follow-ing the maxinum hypothetical accident. With this' leakage rate,'a i
I reactor power level of 1500 MWt, and with minimum containment engineered safety systems for iodine removal in operation (one air cooling and filtering unit), the public exposure would be well below 10 CFR Part 100 values in the event of the maximum hypo-thetical accident.(3) The performance of a periodic integrated leakage rate test during plant life provides a current assessment of potential leakage from the containment.
1 The reduced pressure (5 psig) test on the PAL is a conservative i.
method of testing and provides adequate indication of any potential j
containment Icakage path. The test is conducted by pressurizing between two resilient seals on each door.
The test pressure tends to unseat the resilient seals which is opposite to the accident pressure thet tends to seat the resilient seals.
The six month test ensures the overall PAL integrity at 60 psig, i
3-46
v
' 3.0 SURVEIL'ANCE REQUIREMENTS 3.5 Coi.tainuent Iest (Continued)
The frequency of the periodic integrated leakage rate test (Type
' A test) is keyed to the refueling schedule for the reactor, because this test can only be performed during refueling shut-downs.
The specified frequency of. periodic integrated leakage rate tests is based on three major considerations.
First is the low prob-ability of leaks in the liner because of the test of-the leak--
tightness of the welds during erection and conformance of the -
complete containment to a low leak rate at.60 psig during pre-
)
operational testing, which is consistent with 0.2% leakage at
- design basis accident conditions and absence of any significant stresses in the liner during reactor operation.
Second is the more frequent testing, at the full accident pressure, of.those portions of the containment envelope that are most likely. to develop leaks during reactor operation'(penetrations and isolation valves) and the low value (0.60).of the total leakage that is specified as acceptable from penetrations and isolation valves.
Third is the tendon stress surveillance program,-which provides assurance that an important part of the structural integrity of the containment is maintained. -
A reduction in prestressing force and changes in physical con-
~
ditions are expected for the prestressing system. Allowances have.
been made in the reactor building design for the reduction and chenges. The inspection results for each tendon shall be recorded on the forms provided for.that purpese and comparison will be made with the previous test results and the initial quality control records. Force-time trend lines will also be established and maintained for each of the surveillance tendoas.
If the force-time trend line, as extrapolated, falls below-the predicted force-time curve for cne cn more surveillance tendons, then before the next scheduled surveillance inspection, an in-l vestigation shall be made to determine whether the rate of force i
reduction 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 ex-tension of the investigation shall determine the ~needed changes in the surveillance inspection schedule and the criteria and initial planc.ing for corrective action.
If the force-time trend lines of the surveillance tendons at any time exceed the upper bound curve of the band on the force-time graph, an investigation shall be made to determine the cause.
l 3-47
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3.0 SURVEILLANCE REQUIREMENTS 3.5 containment Test (Continued)
If the comparison of the corrosion conditions, including chemical tests of the corrosion plotection material, indicates larger than expecte
- change in the conditions from the time of installation or last surveillance inspection, an 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 tot being exceeded. The technique chosen for surveillance is based upon the rate of change of force and physical conditiens 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 have concluded that the design is adequate and this has been confirmed by tests. The prestreesing sequence has shown that the end anchorage concrete can withstand loads in excess of those which result when the tendons are anchored.
Further, the containment building was pressure tested to 1.15 times the maximum design pressure.
References J
i (1)
USAR, Section 5.9.1 l
(2)
USAR, Section 5.1.1 (3)
USAR, Section 14.15 3-48
3.0 SURVEILLANCE REQUIREMENTS a
3.5 containment Test (continued) 1
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. 3.0 SURVEILLANCE REQUIREMENTS
. 3.5 containment Test (Continued) i
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3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Test (Continued)
DELETED 3-52
3.0 SURVEILLANCE REQUIREMENTS 3.5 containment Test (Continued)
DELETED 9
4 3-53
i
'i Reference 3, entitled " Omaha Public Power Distcict, Fort Calhoun i
Station Unit No. 1, Reactor Containment Building Leak Rate Test _,
Supplementary Report, June 1973", has been removed from the. Technical Specifications. This deletion encomnassed the cover sheet, Table l
of Contents, and the six (6) Reference 3 pages.
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3.0 SURVEILLANCE REQUIREMENTS 3.16 Recirculation Heat Removal System Integrity Testing Applicability Applies to determination of the integrity of the shutdown cooling system and associated components.
Objective To verify that the leakage from the recirculation heat removal.
system components is within acceptable limits, 4
Specifications (1) a.
The portion of the shutdown cooling system that is outside the containment shall be tested at 250 psig at each refueling outage, or other convenient intervals, but in no case at intervals greater than 2 years.
b.
Piping from valves HCV-383-3 and HCV-383-4 to the dis-charge isolation valves of the safety injection pumps and containment spray pumps shall be hydrostatically tested at no less than 100 psig at the testing fre-quency specified in (1)a. above.
c.
Visual inspection of the system's components shall be performed at the frequency specified in (1)a. above to uncover any significant leakage.
The leakage shall be measured by collection and weighing or by any other equivalent method.
(2) n.
The maximum allowable leakage from the recirculation heat removal system's components (which include valve stems, flanges, and pump seals) shall not exceed one galJon per minute, under the normal hydrostatic head from the SIRW tank.
b.
Repairs shall be made as required to maintain leakage within the acceptabic limite.
Basis The limiting leakage rates from the shutdown cooling system are judgment values based primarily ot assuring that the components could operate without mechanical failure for a period on the order of 200 days after a design basis accident. The test pressure (250 psig) achieved citiier by normal ystem operation or by hydrostatic testing gives an adequate margin over the highest pressure within the system after a design basis accident.(1) Similarly, the hydrostatic test pressure for tot return lines from the contain-ment to the shutdov, cooling sys am (100 psig) gives an adequate margin ever the highest pressure within the lines af ter a design basis accident.
3-84
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3.0 SURVEILLANCE REQUIREMENTS 3.16 Recirculation Heat Removal System Integrity Testing (Continued)
A shutdown cooling system leakage of one gpm will limit off-site 4
exposures due to leakage to insignificant levels relative to those calculated for direct leakage from the.contaiament in the. design basis accident. The safety injection system pump rooms are equipped with individual charcoal filters which are placed"into operation by means of switches in the control rocm. The radiation detectors in the auxiliary building exhaust duct are used to detect high radiation level.
The one gpm leak rate is suffi-ciently high to permit prompt detection and to allow for reason-able leakage through the pump seals and valve packings, and yet
)'
small enough to be readily handled by the pumps and radioactive-waste system. Leakage to the safety injection system sumps will be returned to _the spent regenerant tanks. (gump room
) Addi-tional makeup water to the containment sump inventory can be readily accommodated via the charging pumps from either the SIRW tank or the concentrated boric
- acid storage tanks.
In case of failure to meet the ac:,
nce criteria for leakage from the shutdown cooling system c; the associated components,-it may be possible to effect repairs within a short time.
If so, it is considered unnecessary and unjustified to shutdown the reactor.
The times allowed for repairs are consistent with the-times developed for other engineered safeguards components.
References (1)
USAR, Section 9.3 (2)
USAR, Section 6.2 4
3-85
DISCUSSION The proposed changes to Technical Specification Section 3.5 are to ensure that reactor containment building leak rate testing (Type A integrated test, Type B and Type C local tests) is performed in ac-cordance with and as specified by 10 CFR Part 50, Appendix J, as amended, given the existing design limitationc of the Fort Calhoun Station Unit No. 1.
The Type A test will be conducted as specified in 10 CFR Part 50, Appendix J.
This integrated test will be conducted on the containment building and those systems which are considered part of the containment boundary and potential leakage paths.
The performance of the Type A test with the plant in a refueling shutdown condition may result in leakage paths in certain safety related systems which normally would-not exist during power operation. These paths may exist in the SI system which will be at a pressure greater than the test pressure if required to operate during a DBA. These leakage paths are repairable without any local leak testing and do not contribute to the failure of the Type A test.
The design basis leakage rate of 0.2% by weight of the containment atmosphere per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has been revised from the 0.1% value which was identified in the District's previous amendment application revision.
As addressed in the original Fort Calhoun Station FSAR Section 14.21,
" Maximum Hypothetical Accident," a containment leak rate of 0.2% of the containment free volume was utilized and no resultant safety hazard to the public was identified (i.e., with regard to 10 CFR 100 radiological limits).
Tl - District's amendment application and safety analysis for increasing our maximum thermal power level to 1500 megawatts also utilized this value and the conclusion was the same.
Therefore, the District believes the incorporation of this value into the Technical Specifications (TS) fer determining the maximum allowable leakage rate is justified.
The District believes the 0.1%/ day leakage rate presently in the TS was incorporated as a conservative limiting criterion for the initial pressure test of the Fort Calhoun Station containment and subsequently was never revised to the less limiting, although allowable, value of C.2%/ day.
The local leakage Type B tests (electrical penetrations, mechanical per.etration sleeve welds, equipment hatch, fuel transfer tube, and personnel access lock) are conducted as specified in 10 CFR Part 50, Appendix J, with the exception of the testing of the personnel access lock (PAL).
i The existing PAL doors are so designed that an accident pressure (60 psig) test can only be performed after strong backs (structural bracing) have been installed on the inner door. The air lock requires strong backs to be installed prior to testing since the test pressure on the inner door is opposite that of the door's designed pressure direction.
The strong backs are extremely difficult to install and the outer door must be opened to remove the strong backs. Therefore, the District will l
leak test the PAL assembly at 60 psig every six months and perform a reduced pressure test (i.e., 5 psig) of the door seals within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ATTACHMENT B
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- if the PAL is opened during periods when containment integrity is re-quired.
However, the District is requesting an exemption from the Appendix J requirement applicable to the PAL that specifies " Air locks opened during periods when containment integrity is not required by the plant's Technical Specifications shall be tested at the end of such periods at not less than Pa."
The District proposes to leak test the door seals at 5 psig prior to achieving the containment integrity required condition and within two weeks test the entire PAL assembly at 60 psig.
The District believes this proposal meets the intent of the requirement in regard to identifying and correcting significant leakage prior to achieving a containment integrity required condition.
This exemption is requested because, as indicated above, testing at Pa re-quires approximatel; 18-24 hours to complete and is a significant and unnecessary delay in the return to power operation.
The PAL is equipped with a reduced pressure testing system which applies 5 psig between the two seals of each of the two locked doors.
Although the te t is a reduced pressure test, it is capable of identi-fying a potenti.1 leak path in the air lock.
The test pressure which is applied between the two seals of each of the locked doors tends to unseat the seals which will aid in identifying seal leaks (normal DBA pressure helps to ensure the seals are seated).
In addition, the test pressure will identify leaks in any single door seal. The PAL reduced test acceptance criterion of less than or equal to 0.01 La is conservative and was agreed upon during a telephone conference call on November 24, 1982 between the District and tne Commission's staff.
The testing frequency of the reduced pressure test provides a sufficient frequency (as a result of the conservative test method and acceptance criteria) to ensure any potential lesk path is identified, before any leakage limits are exceeded.
The valves associated with the penetrations listed in the proposed Technical Specification change are those on which the Type C test will be performed in accordance with 10 CFR Part 50, Appendix J.
Penetration M-44 has been added to this list of penetrations in accordance with the Distric t's let ter dated November 5, 1982.
Valves associated with penetrations M-16 (shutdown cooling), M-86 (containment spray), and M-89 (containment spray) are part of the core cooling / safety injection system.
In the event of a DBA (LOCA) these systems will operate.
In addition, the pressure seen by the isolation valves is in the direction of flow toward the containment building and will always be greater than the maximum containment pressure (60 psig).
The District is also requesting an exemption from Type C testing for the isolation valves associated with penetration M-3, " Charging Pump Discharge Line".
The justification for not testing these valves is that the pressure seen by the valves in the direction of flow toward con-tainment is always much greater than the maximum containment accident pressure (2100 vs. 60 psig).
The charging pump (s) remain operational and the subject isolation valves remain open during receipt of an accident signal.
Thus, the charging pump flow provides a seal barrier against escape of the containment atmosphere. Maintaining this barrier
O
. 1 during a loss of coolant accident is assured since upon receipt of a safety injection actuation signal, the charging pumps are automatically aligned to the boric acid storage tanks.
The volume held by these tanks provide a source of supply to the pumps for approximately 80 minutes and as demonstrated in USAR Section 14.16, the increased containment pres-sure would be reduced to near atmospheric levels within 50 minutes.
Thus, the District believes exempting the penetration M-3 valves is justified.
Valves associated with penetrations M-91 and M-97 are within the emergency feedwater system. These valves open on containment isolation and are not required to be tested in 10 CFR Part 50, Appendix J.
Valves associated with penetrations M-94 and M-95 are on the main steam line. These valves receive a close signal from low steam generator pressure or containment pressure high and not the containment isolation signal. Their function is to prevent an uncontrolled heat extraction.
They are not required to be tested under 10 CFR Part 50, Appendix J.
Lue remaining valves associated with penetrations M-9, M-10, M-12, M-13, M-49, M-63, M-93, and M-96 are associated with the secondary side of the steam generators.
Although the penetration isolation valves are closed on a containment isolation signal as an added precaution (with the exception of M-9 and M-12 whose valves are manually operated, locked closed) the steem generator shell is a containment boundary and no testing is required for such closed systems.
Due to the method of operation of the emergency feedwater and main steam penetration, the valves associated with the secondary side of the steam generators were not designed for external leak tightness, rather the steam generator shell is the containment boundary. The steam generators are normally pressurized and would be pressurized for every accident except MSLB or feedwater line break.
For these two accidents, the reactor coolant porti.a of the generator would remain intact and would, therefore, not contribute to containment leakage to the atmosphere.
The testing requirements for the recirculation heat removal (RHR' system, presently addressed in Specification 3.5(6), have been moved to a new section.
Section 3.16 (pages 3-84 and 3-85) now addresses the RHR system leckage testing specifications. This administrative change was com teted because the subject RHR system components are located outside containment and Section 3.5 addresses containment tests.
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IDENTIFICATION OF SIGNIFICANT CHANGES TO PRESENT TECHNICAL SPECIFICATIONS RESULTING FROM THIS AMENDMENT APPLICATION REVISION Technical Specification Page No.
Remarks 3-37 1.
Inclusion of several definitions related to containment leak rate testing.
2.
Definition of a." Type A" test in Section 3.5(2)(a) is added.
3.
Maximum allowable leakage rate cal-culated using 0.2% by weight of the containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> vice 0.1%.
3-38 1.
Discussion of Type A pretest require-ments which are consistent with Appendix J.
3-39 1.
Discussion of test methods and ac-ceptance criteria which are con-sistent with Appendix J.
3-40 1.
No significant or technical changes.
3-41 1.
Identification of Type B testing fre-quency, including testing of the personnel access lock (PAL). The reduced test pressure for the PAL will be 5 psig, with an acceptance criterion of less than or equal to 0.01 L, as agreed upon during the a
November 24, 1982 telecon. Addi-tionally, the test frequency for the PAL is consistent with Section III.D.2(b) of Appendix J.
Includes our exemption request for reduced pressure test of PAL prior to reach-ing a containment integrity required condition.
3-42 1.
Discussion of Type C resting, special 3-43 testing, and reportability require-ments which are consistent with Appendix J.
2.
Section 3.5(6), "Facirculation Heat Removal Systems", of the present TS's has been moved to a new section, Section 3.16, as detailed in the discussion section of this amendment application.
Technical Specification Page No.
Remarks 3-44 1.
No significant or technical changes.
3-45 1.
No significant or technical changes.
3-46, 1.
Section 3.5(8), "End Anchorage Con-3-46a, crete Surveillance", Section 3.5(9),
3-47,
" Liner Plate Surveillance", and 3-48, Section 3.5(10), " Penetrations 3-49 Surveillance", of'the present TS's have been deleted because these tests were only required during the initial years of plant operation and are no longer performed.
3-49, 1.
The " Basis" section of the present 3-50, TS's has been revised in accordance 3-51, with the changes detailed above and 3-52 with the requirements of Appendix J.
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