TXX-4386, Forwards Response to NRC Question Q022.22 & Revised FSAR Text Re Containment Integrated Leak Rate Test.Changes Will Be Part of Amend 54 to FSAR Scheduled for Transmittal in Jan 1985
| ML20101F494 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/21/1984 |
| From: | Beck J TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| TXX-4386, NUDOCS 8412270192 | |
| Download: ML20101F494 (5) | |
Text
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a TEXAS UTILITIES GENERATING COMPANY Log # TXX-4386 mayw4r rowen. wo nonru ouve nruser, o...I
- DAM.AN,rEKAN T3208 FiIe # 914.2 December 21, 1984
' f?"5?af,."
- Director of Nuclear Reactor Regulatio'n Attention: Mr. B. J. Youngblood, Chief
-Licensing Branch No. 1 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION
' RESPONSE TO NRC QUESTION Q022.22
Dear Sir:
Attached please find the response to NRC Question QO22.22 and revised FSAR text.
These changes will be part of Amendment 54 to the CPsES FSAR which is scheduled for transmittal in January of 1985.
Should you have any questions in this matter, please contant this office.
Respectfully, W.
= = ' -
J. W. Beck BSD/grr Attachment c - J. J. Stefano 8412270192 841221 8p PDR ADOCK 05000445 A
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e CPSES/FSAR QO22.22:
'It is stated in the FSAR that the methodology of' ANSI N45.4-1972 will-be used to conduct the ILRT.
The staff -
Technical Review Team (TRT) has found that the methodology of ANSI /ANS 56.8-1981,. instead of ASNI N45.4-1972,_.was used. in performing the. test;- ANSI /ANS 56.8-1981, however, has not been endorsed by-the staff.
L In reviewing the ILRT sumary report, dated May 6, 1983_, we note that the " mass-plot method" of ANSI /ANS '
56.8-1981, was used to calculate the containment -
leakage rate. Although we find this _ acceptable, the applicant is requested to identify and justify any other differences in applying' ANSI /ANS 56.8-1981 in lieu of ANSI N45.4-1972.
R022.22
-The CILRT Procedure (1CP-PT-75-02) was compared with FSAR comitments to conduct the CILRT in accordance with 10CFR50 Appendi.~. J and ANSI N45.4-1972.
The results of that review indicate no deviations, other than use of tne " mass-plot method" for calculation of containment leakage rate and isolation of. the three 7
electrical penetrations during conduct of the test as a result of applying ANSI /ANS 56.8-1981 in lieu of ANSI i
N45.4-1972.
These deviations have been accepted by the staff in the NRC request for additional information letter dated August 23, 1984.
FSAR Section -14.2, Table 14.2-2, Sheet 59 of 60 and Section 3.8, Paragraph 3.8.1.7.2 have been revised to include this informa tion.
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022-44
.w CPSES/FSAR since the CPSES Containment is not a prototype containment).
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"3.8.1.7.2
. Initial Leakage Rate Tests Containment leakage ' testing is in accordance with all the requirements of 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors (2/5/73).
A reduced pressure-test program is used, as describe'd in 10 CFR Part 1
~ 50, Appendix J.
The reduced pressure test is at-50 percent of the
~ calculated peak Containment internal pressure. For calculated peak Containment internal pressure, see Section 6.2.1.
A preoperational Type A integrated leakage-rate test is performed at the calculated peak Containment internal pressure and also at the reduced pressure.
Type B tests of components and Type C tests of Containment isolation valves are performed in accordance with 10 CFR Part 50, Appendix J..
The maximum allowable leakage-rate (La as defined in 10 CFR Part 50, Appendix J), related to the maximum Containment leakage under design basis pressurization accident conditions, is 0.10 percent of the weight of contained air, at the calculated peak Containment internal pressure in a 24-hr period.
L I
For a discussion of the test objectives and the acceptance criteria, see the Technical Specifications for Containment Tests.
Test methods are in accordance with ANSI N45.4-1972, Leakage Rate Testing of Containment Structures for Nuclear P.eactors (03/26/73), with the exceptions of isolated penetrations and the use of the mass-plot 54 method per ANSI /ANS 56.8-1981 as stated in Table 14.2-2 (Sheet 59 of Q022.22 60).
3.8-61
e 4
s CPSES/FSAR Table -14.2-2 (Sheet 59 of 60)
CONTAINENT INTEGRATED LEAK RATE TEST TEST SUPetARY OBJECTIVE To verify the primary reactor containment overall integrated leakage rate is within acceptable limits.
PREREQUISITES 1.
Fluid system conditions are established'as applicable to simulate post accident conditions which extend the boundary of the Cont,ainment Building.
' 2.,
Containment component and isolation valve leak tests have been satisfactorily performed.
3.
All containment isolation valves-have been closed by normal actuation methods.
TEST METHODS 1.
Perform the containment integrated leak rate test per Appendix J of 10 CFR Part 50.
2.
Perform the leakage rate calculation by using the mass-plot methodology as described by ANSI /ANI 56.8-1981.
54 Q022.22 3.
If during the performance of a type A test, excessive leakage occurs through locally testable penetrations or isolation valves, these leakage paths may be' isolated and the Type A test continued untti completion.
The sum of the post repaired local leakage rate values will be added to the UCL per ANSI 56.8-1981.
hinne simm
. mm a i i s an.i. ism m..
.,a CPSES/FSAR Table 14.2-2 (Sheet 59A of 60)
. ACCEPTANCE CRITERIA The Containment Integrated Leak Rate Test meets the requirements of Appendix J of 10 CFR Part 50.
Note:
' The containment structural integrity test described in FSAR Section 3.8
. may. be performed concurrently with the Integrated Leak Rate Test.
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