ML20032B425
| ML20032B425 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/04/1981 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20032B417 | List: |
| References | |
| NUDOCS 8111050536 | |
| Download: ML20032B425 (6) | |
Text
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'30 SURVEILLANCE REQUIREMENTS 3.5 Containment Test (Continued) conservative results.
Type B test methods may be substituted where appropri-ate.
Each valve to be tested shall be closed by normal operation and without any preliminary exercising or adjustments (e.g...no tightenink of valve after closure by valve motor).
Detailed supplementary criteria that' establish specific test requirements to fulfill G.ese Specifications are provided in reference 3 vhich is included as an en-closure to t.hls Specification.
f b.
Acceptance Criteria The combined leakage rate for all components subject to Type B and C tests and subject to the 0.6 L leakage a
limit shall not exceed 60 percent of L. For the six a
month purge isolation valve tests, the measured purge valve leakage rate shall be substituted for the purge valve leakage rate from the last complete Type B and C test and the total leak rate recomputed.
Leakage of the containment air purge isolation valves shall not exceed 18,000 standard cubic centimeters per minute (SCCM).
If the leakage rate is determined to be greater than 18,000 SCCM, repairs shall be initiated immediately in order to meet this acceptance criteria, c.
Frequency Type C tests shall be performed during each reactor shutdown for major refueling but in no case at inter-vals greater than two years, except that containment purge isolation valves shall be Icakage tested at intervals not to excccd six months + 25%.
(4)
Specific Testing Requirements Any major modification or replacement of components of the primary reactor containment performed af ter the initial preoperational leakage rate test shall be followed by either a Type A test, Type B test, or a Type C test of the area affected by the modification and shall meet the applicable acceptance criteria.
3-42 ATTACEMENT A e111050536 5
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e 3.0 SURVEILLAMCE REQUIREMENTS 3.5 Containment Test (Continued)
(5)
Inspection and Reporting of Tests f
a.
Containment Inspection A detailed visual examination of critical creas and general inspection of the accessible interter and exterior surfaces of the containment structures and components shall be performed at each reactor shutdown for a refueling outage and prior to any Type A test to uncover any evidence of structural deterioration whicii may affect the containment's structural integrity leaktightness.
If there is evidence of significant structural deterioration, Type A tests shall_not be performad until corrective. action is taken in accordance with repair procedures, nondestructive examinations, and tests as specified in the construction code under which rules the containment was built. Such structural deterioration and corrective actions taken shall b2 reported as part of the Type A test report.
b.
Report of Test Results-The > f eial Type A test shall be the subject of a summary technical report submitted to the Commission after the conduct of the test.
This report shall include a schematic arrangement of the leakage rate measurement system, the. instrumentation used, the-supplemental test method, and the test program selected as applicable to the initial test, and all subsequent periodic tests. The report shall contain an analysis and interpretation of the leakage rate test data to the extent necessary to demonstrate the acceptability of the containment's leakage rate in meeting the accept-ance criteria.
Periodic test leakage rate results of Type A, B, and C tests that meet the acceptance criteria shall be reported in the licensee's operating report. Leakage test results of Type A, B, and C tests that fail to meet the acceptance criteria shall be reported in a separate summary that includes an analysis and inter-pretation of the test data, the least-squares fit analysis, and the structural conditions of the contain-ment or components, if any, which contributed to the failure in meeting the acceptance criteria. Results and analyses of the supplemental verification test employed to demonstrate the validity of the leakage rate test measurements shall also be included.
(6)
Recirculaticn Heat Removal Systems a.
Testing Requirements The portion of the shutdown cooling system that is out-side the containment shall be tested either by use in normal operation or hydrostatically tested at 250 psig at the interval specified in the 3.5(3)a(iii).
' Amendment No. 24 3-43
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3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (Continued) leakage rate and duplicates the pre-operational leak rate test of g
30 psig. The specification provides relationships for relating in a conservative manner the measured leakage of air at 30 psig to the potential leakage of a steam-air mixture at 60 psig and 188 F.
The specification also allows for possible deterioration of the leakage rate between tests, by requiring that only 70% or 80% of the allowable leakage rates actually be measured. The basis for the deterioration allowances is arbitrary judgments which are believed to be conservative and which will be confirmed or denied by periodic testing. If indicated to be necessary, the deterioration allowances will be altered based on experience. The durations for the integrated leakage rate tests are established to provide a minimum level of accuracy and to allow for daily cyclic variation in temperature and thermal radiation.
The frequency of the periodic integrated leakage rate test is keyed to the refueling outage schedule for the reactor, because these tests can best be performed during refueling shutdowns. The a
(_
initial core loading is designed for approximately 12 months of power operation; thus, the first refueling outage will occur approximately 18 months after initial criticality. Subsequent refueling outages are scheduled at approximately 12-month inter-vals, although larger periods may be utilized.
The specified frequency of periodic integrated leakage rate tests is based on three major considerations.
First is the low proba-bility 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-m_
E operational testing, which is consistent with 0.1% Icakage at design basis accident conditions and absence of any significent
= '
stresses in the liner during reactor operation.
Second is the more frequent testing, at the full accident pressure, of those
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portions of the containment envelope that are most likely to I
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.
F T
Third is the tendon stress surveillance ptogram, which provides assurance that an important part of the structural integrity of the containment is maintained.
Semi-annual leakage integrity tests of the purge isolation valves
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is established to identify excessive degradatien of the resilient seats of these valves.
Simultaneous testing of redundant purge valves from a leak test connection accessible from outside con-y
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tainment provides adequate testing during power operation. The semi-annual testing method is identical to the Type C purge iso-g 1ation valve test performed in accordance with 10 CFR Part 50, Appendix J.
For leakages found to be greater than 18,000 SCCM, iA re; airs shall be initiated to ensure these valves meet the ac-j[
ceptance criteria, b
tp Amendment No. 24 3-50 EIe
3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (Continued)
The limiting leakage rates from the shutdown cooling system are judgment values based primarily on 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 either by normal system operation or by hydrostatic testing gives an adequate margin over the highest pressure within the system after a design basis accident.
Similarly, the hydro-static test pressure for the return lines from the containment to the shutdown cooling system (100 psig) gives an adequate margin over the highest pressure within the lines after a design basis accident.
A shutdown cooling system leakage of one gpm will limit off-site exposures due to leakage to insignificant levels relative to those calculated for direct leakage from tLa containment in the design basis accident. The safety injection system pump rooms are equipped with individual charcoal filters which are placed into operetion by means of switches in the control room. The radiation detectors in the auxiliary building exhaust duct are used to detect high radiation level. The one gpm leak rate is sufficiently high to permit prompt detection and to allow for reasonable leakage through the pump seals and valve packings, and yet small enough to be readily handled by the pumps and radioactiva vaste system. Leak-age to the safety injection system pump room sumps will be re-turned to the spent regenerant tanks. Additional 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 acceptance criteria for leakage from the shutdown cooling system or the penetrations, it may be possible to effect repairs within a short time.
If so, it is considered unnecessary i,nd unjustified to shutdown the reactor.
The times allowed for repairs are consistent with the times developed for other engineered safeguards components.
A reduction in prestressing force and changes in physical condi-tions are expected for the prestressing system. Allowances have been made in the reactor building design for the reduction and changes. The inspection results for each tendon shall be recorded on the forms provided for that purpose 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 tendons.
If the force-time trend line, as extrapolated, falls below the predicted force-time curve for one or more surveillance tendons, then before the next scheduled surveillance inspection, an in-vestigation shall be made to determine whether the rate of force reduction is indeed occurring for other tendons.
If the rate of reduction is confirmed, tha 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 inepection schedule and the criteria and initial planning for corrective action.
If the force-time trend lines of the nurveillance 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.
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DISCUSSION As a result of the nuaerous reports on unsatisfactory performance of the resilient seats for the isolation valves in containment purge and vent lines (addressed in OISE Circular 77-11, dated September 6, 1977),
Generic Issue B-20, " Containment Leakage Due to Seal Deterioration", was established to evaluate this concern and establish an appropriate testing frequency for the isolation valves. The Commission's letter dated July 28, 1981 provided additional clarification on the test program and recommended a test frequency of at least once every six months for passive purge systems. The Omaha Public Power District herewith submits a Technical Specification change that conforms to the Commission's recommended testing program and six month testing frequency of the purge isolation valves.
The proposed leakage testing program, to be conducted at intervals not to exceed six months + 25%, is identical to the Type C purge iso-lation valve leakage test that has previously been conducted every refueling outage, or intervals not exceeding two years. A leak test connection that is accessible from outside containment is used to pres-surize the line between the redundant purge valves, inside and outside containment, to 60 psig. The leakage measurements are then made using appropriate instrumentation / equipment. Should the measured leakage rate exceed 18,000 standard cubic centimeters per minute (SCCM), corrective repairs will be initiated immediately to bring the leakage below 18,000 SCCM. The 18,000 SCCM leakage figure has been used as the maximum allowabic leakage in the performance of Type C leakage tests and in-service inspections for these valves. This specific limit of 18,000 SCCM leakage provides a means to measure and identify excessive seal degradation and ensures that corrective measures are carried out as i
needed.
To ensure Technical Specification adherence, the measured purge valve leakage rate is recomputed with leakage rates for all other Type B and C penetrations, and the integrated leakage rate must be less than or equal to 0.6 La.
If it is determin'd that the integrated leakage rate is greater than 0.6 La, repairs shall be initiated immediatel". The action requirements to be followed if the Icakage criteria is not met are covered by Technical Specifications 2.6(1)a and 2.0.1(1).
Accord-ingly, the existing action statement, paragraph 2 in Technical Speci-fication 3.5(3)b, allowing 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for repairs before action is re-quired has been deleted, since it was significantly less conservative than 2.0.1(1).
ATTACHMENT B
JUSTIFICATION FOR FEE CLASSIFICATION The proposed amendment is deemed to be Class III, within the meaning of 10 CFR 170.22, in that it involves a single safety con-cern.
I ATTACHMENT C