ML19341C522
| ML19341C522 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/25/1981 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | Clark R, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| TAC-07711, TAC-08404, TAC-7711, TAC-8404, NUDOCS 8103030689 | |
| Download: ML19341C522 (6) | |
Text
I 1
1 Wisconsin Electnc ecara coupasr 231 W. MICHIGAN P.O. BOX 2046. MILWAUKEE, WI 53201 February 25, 1981 T
D' Mr. H.
R.
Denton, Director gdL
// D N/{C#'I Office of Nuclear Reactor Regulation i
8 9
U.
S. NUCLEAR REGULATORY COMMISSION Washington, D.
C.
20555 h
un o '.,,6 s.
m
=
A coSI@'cer [7' Atten'. ion :
Mr.
R. A. Clark, Chief 9-u,,, %
Operating Reactors Branch #3
/
Gentlemen:
DOCKET NOS. 50-266 AND 50-301 ADDITIONAL INFORMATION FOR T.
S.
CHANGE REQUEST NO. 28 APPENDIX J CONTAINMENT TESTING REQUIREMENTS POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 In accordance with your letter dated January 27, 1901, we are herewith transmitting the additional information you have requested so you may finalize your review of the exemptions to 10 CFR 50 Apper. dix J we requested.
We would like to bring to your attention the fact that we have always provided timely responses to your requests for information; however, in view of your extended schedule for review of a request initially submitted in 1975, we consider your request for a response within 30 days to be inappropriate.
Our responses to your information request are contained in the attachment to this letter.
They are numbered to correspond to the identification of your January 27 letter.
We trust these responses will satisfy your questions and concerns regarding the subject change request and that your review can be expedited.
Very truly yours, C.
W.
Fay, Director Nuclear Power Department Attachment
\\
%0,\\ t Copy to NRC Resident Inspector 81030306F7
ADDITIONAL INFORMATION TECHNICAL SPECIFICATION CHANGE REQUEST NO. 28 2.1.1 The requested deviations and associated justifications from the requirements of 10 CFR 50 Appendix J, Section III. A. l(d), with regard to the venting and draining of systems in preparation for a Type "A" test are provided below.
l.
The containment service air supply line, penetration 33C, is used during the Type "A"
test to pressurize and depressurize the containment.
The isolation 4
requirements for the test and the temporary piping installed for the test prevent its being tested in accordance with Appendix J.
We have previously requested an exemption from the requirements to vent this line to containment atmosphere during the Type "A"
test.
Instead, we perform a Tupe "C" test on the isolation valves in this line according to Appendix J,Section III.C, and add the leakage measured in this test to the overall leakage measured in the Type "A" I
test.
2.
Appendix J,Section III.A.l(d), requires the performance of a Type "C" test on the containment isolation valves of the residual heat removal system.
This is due to the fact that the residual heat ~ removal system is required to maintain the plant in a safe condition during performance of the Type "A" test and, therefore, will not be drainee. and vented to containment atmosphere.
We previously requested (Reference 1), and are again requesting, that an exemption from the Type "C" testing of the residual heat removal system containment isolation valves be granted to allow for the testing presently required by the approved Technical Specifi-cation Section 15.4.4.IV.
The system is presently treated and tested as if it was an extension of the containment boundary as the i
system is "in use" fcllowing a design basis accident.
The present testing requirements include a 350 psig hydrostatic test of the entire system, except for the containment sump piping.
A 60 psig hydrostatic test is parformed in that section.
A total system leakage test is performed per the requirements of NUREG-0578, Item 2.1.6.a. -
y p
wi
, It can be seen that any leakage past the containment isolation valves of the residual heat removal system would only contribute to the total system fluid inventory, which in itself is leak-tested to assure that the overall system leakage is below the limits set to provide for the safe operation of the plant and the public health and safety during post-accident conditions.
You are fully aware that the residual heat removal system is always flooded and available for service.
Removal of this system from active service availability for a Type "C" test of the system's containment isolation valves would present a significant potential 4
risk to the reactor core in the event of a major accident and as such is mimical to the public health and safety.
Removal of the entire core for the purpose of conducting such a test is totally unwarranted.
Considering the "in use" nature of the system, that the additional information gained by Type "C" testing of these containment isolation valves is of question-able value, the additional radiation exposure to plant personnel in performing the tests, unnecessary additional risk in having to perform a full core unload, and the totally 'dequate system testing programs presently provided, we believe that Type "C"
testing of these valves is not justified.
2.1.2 There were no valves isolated from the Type "A" containment integrated leak rate test performed on Unit 1 in October 1977 because of excessive leakage.
However, there were six penetr-tions isolated from the Type "A" test for purposes of maintaining the plant in a safe condition, for providing a flow path for establishing containment test pressure, and for providing an instrumentation channel for pressure test data.
These penetrations were Type "C" tested following recovery from the Type "A" test.
The resulting leakage data were added to the results of the Type "A" test.
2.1.3 Please initially note that ANSI N45.4-1972 does not require a 24-hour test period as you have indicated in your correspondence.
Section 7.6 of ANSI N45.4-1972 allows for a test period shorter than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as long as "
it can be demonstrated to the satisfaction of those responsible for the acceptance of the containment structure that the leakage rate can be accurately determined during a shorter test period.
p.
-+
. l Justification for the use of a test duration less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided in the procedural excerpts below.
The Technical Specifications, Section 15.4.4.1.A.2, state that "the test duration shall not be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless test experiences of at least two prior tests provide evidence of the adequacy of shorter test duration".
The procedure used for the integrated leak rate test on the Unit 1 containment in October 1977 stated that "the test duration shall be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; however, based on the adequacy of two prior tests, the test duration may be shortened to no less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided there is evidence of a 95 percent certainty that the leak rate test is within allowable limits".
Please note that the containment integrated leakage rate test procedures used during the Unit 1 Type "A" test of October 1977 were prepared by Nuclear Services Corporation, approved for use by the Wisconsin Electric Power Company Point Beach Nuclear Plant Manager's Supervisory Staff, and reviewed by the NRC prior to implementation.
2.2 As we have previously stated, we strongly disagree with the Appendix J,Section III.D.2, requirement to test the operated containment airlock (whether it be a complete airlock test or a reduced pressure test between the "0"
ring door seals) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of every containment entry.
Apparently, the reasoning behind this additional test requirement is to provide for a means of assuring that the door seals have not been damaged or seated improperly during airlock use (as conveyed to us in Attachment "A"
of Reference 3).
We are again requesting that an exemption be granted from the 72-hour containment airlock testing requirement of Appendix J,Section III.D.2, to allow for our present Technical Specification testing requirements, plus an additional procedural change requiring a visual examina-tion of the "0" ring seals following periodic containment inspection entries.
The following considerations justify our above request:
1 In the numerous Type "A" containment tests and Type "B"
airlock tests, there has never oeen a test failure caused by leakage through tae door "0"
ring seals (see Reference 2).
This large amount of inservice test data surely verifies the ability of the air!ock door seals to seat properly.
\\
2.
A visual inspection of the "0"
ring seals will be required following periodic containment inspections upon exit to assure that the "0" rings were not damaged due co or during containment airlock use.
3.
The design of the airlocks is such that a full design pressure test between the "0" ring seals is not possible.
In addition, there appears to be no viable method of extrapolating the results of a reduced pressure test between the "0" ring seals to an equivalent design pressure leak rate.
l 4.
The implementation of the 72-hour airlock testing requirement of Appendix J,Section III.D.2, will have a potential negative impact on containment and reactor systems surveillance at the Point Beach Nuclear Plant.
This negative impact relates to the likelihood that fewer containment entries will be made for purposes of surveillance testing, inspections, and radiation monitoring (see Reference 2 for containment entry i
practices).
5.
Existing airlock testing procedures require a contain-ment entry to install a strongback on the containment.
4 side of the inner door.
This containment entry, in turn, would require an additional 72-hour airlock test.
The requirements of Appendix J,Section III.D.2, would dictate that this "round robin" testing scenario be implemented on the present design of the Point Beach Nuclear Plant.
It becomes obvious that this leads to the absurd result of a continuous airlock testing activity.
2.3 We cannot guarantee that there will be sufficient fluid inventory within the piping system, downstream of the containment spray system's containment isolation check valves, to assure that the valves' seating surface t.
will be water-sealed for 30 days following a design basis accident.
We are not requesting a total exemption from having to l
Type "C" leak test these valves.
However, we are requesting an exemption from the requirements of Appendix J,Section III.C.2.a, in the method of testing these valves.
,=
,r--,----g
--,,s
't~w w
--v
~-'
, Appendix J,Section III.C.2.a, requires that these valves be pressurized with air or nitrogen.
Since the valves are installed in a vertical piping run, we cannot drain that piping section sufficiently to expose the valves' seating surfaces to the test medium required.
We are requesting that an exemption be granted to allow us to test these valves with the undrainable water in the system present.
Considering the following, we feel that the request is justified:
1.
A total system leakage rate test is performed per the requirements of NUREG-0578, Item 2.1.6.a.
2.
In the event of backleakage through these valves, there are numerous other valve boundaries that would also have to be breached to allow for a release to the environment.
3.
If unacceptable backleakage was identified, the line could be manually isolated via the closure of the manual isolation valve located at the containment penetration.
References 3
(1)
Wisconsin Electric to Nuclear Regulatory Commission letter dated December 12, 1975.
i (2)
Wisconsin Electric to Nuclear Regulatory Commission letter dated July 18, 1977.
( 3)
Nuclear Regulatory Commission to Wisconsin Electric letter dated May 31, 1977.
(4)
Nuclear Regulatory Commission to Wisconsin Electric letter dated January 27, 1981.
1 i
_ - - _ _ _ _, _. - - - -