ML20150D382
| ML20150D382 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/27/1978 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7812050188 | |
| Download: ML20150D382 (11) | |
Text
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l Wisconsin Electnc eowca couesnr i
I 231 W. MICHIGAN, P.O. BOX 2046, MILWAUKEE, WI 53201 November 27, 1978 i
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C.
20555
Dear Mr. Denton:
DOCKET NO. 50-301 INSERVICE INSPECTION OF SAFETY CLASS COMPONENTS TECHNICAL SPECIFICATION CHANGE REQUEST NO. 58 POINT BEACH NUCLEAR PLANT UNIT 2 In accordance with Section 50.59 of 10 CFR 50, Wisconsin Electric Power Company (Licensee) hereby requests an amendment to Facility Operating License 3
DPR-27 to incorporate changes to the Technical Specifications for Point Beach Nuclear Plant Unit No. 2.
The attached changes to the Technical Specification are in compliance with Section 50.55a paragraph (g)(5) of 10 CFR 50 which requires amendment of the technical specifications to conform the specifications to the updated inservice inspection and testing programs for safety class components mandated by Section 50.55a.
On February 12, 1976, notice was published in the Federal Register of a revision to 10 CFR 50 Section 50.55a which changed the inservice inspection and testing requirements for nuclear power plant safety class components.
- Briefly, this revised regulation requires that all nuclear power facilities meet, to the extent practical, the inservice inspection requirements of Section XI of the ASME Boiler and Pressure Vessel Code, edition and addenda as specified in 50.55a paragraph (b).
Licensees are required to update their inservice inspection program prior to each 40 month inservice inspection interval and update the pump and valve inservice testing program every 20 months.
If conflicts exist between the ASME Section XI code and the plant's technical specification, Licensees are required to submit a proposed amendment to the technical specifications for NRC approval at least six months prior to the start of the next inspection interval.
Your letters of April 26, 1976 and November 22, 1976 provided further guidance on this subject and included sample technical specification language.
We are presently in the process of updating the inservice inspection program for Point Beach Nuclear Plant Unit 2 to the 1974 edition Summer 1975 Addenda of the ASME Section XI code. As a result of this program update, and 0 a
- 7812959, g
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' Mr. Harold R. Denton - Page Two considering your guidance regarding sample technical specification language, we propose the following technical specification changes:
(Revised pages are attached) 1 1.
Specification 15.4.2.B. - The existing section has been completely deleted and revised language which follows the NRC guidance has been substituted. Licensee has included specific language to confirm that all containment isolation valves shall be tested in accordance with the provisions of Appendix J to 10 CFR 50 " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors".
We have also provided lan5uage to clarify that, per section 50.55a I
paragraph (g)(5)(iii), if the Licensee has determined that conformance with a code requirement is impractical, Licensee shall notify the Commission and submit information to support his determination.
2.
Specification 15.4.5.11 items A and B - These items which specified certain pump and valve tests have been deleted as they are superseded by ASME Section XI valve and pump testing code requirements.
3.
Specification 15.4.7 - This specification on main steam stop valve periodic testing is superseded by ASME Section XI valve testing code requirements and hcs been deleted.
4.
Specification 15.4.8 - This specification on auxiliary feedwater pumps periodic testing is wperseded by ASME Section XI pump testing code requirements and has been deleted.
5.
Specification 15.6.9.3 - Item B of this specification required a one t'me review of the inservice inspection program at the end of the first five years of operation. This review has been done in connection with this mandatory update of the inservice inspection program. This requirement may thus '
ieleted.
The changes herein requested to th tical specifications for Unit 2 are identical to the changes requested for Ut in Change Request No. 42 in our letter to Mr. Rusche dated February 17, 1977.
is requested that you approve these change requests at the same time so that $
Technical Specifications for Unit 1 and Unit 2 may remain identical in this area.
Licensee has reviewed the requirements of 10 CFR Part 170.22 regarding the schedule of fees for facility license amendments.
It is our determination that since this license amendment request is the result of a written order of the Commission, namely 10 CFR Section 50.55a, which requires amendment of the technical specification to conform the specifications to the updated inservice inspection and testing programs for safety class components, this amendment is not subject to such fees.
Mr. Harold R. Denton - Page Three As specified in 10 CFR Part 50, we are enclosed three signed originals and thirty-seven c:.;;ies of this license amendment request and proposed technical specification change.
The next 40 month inspection interval for Point Beach Unit 2 will begin on June 1,1979. Our proposed inservice inspection and testing program will be submitted at least 90 days prior to that date.
Should you have any questions concerning these proposed technical specification changes, please contact us as soon as possible.
Very truly yours, J
f3
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Sol Burstein xec ive Vice President Attachments Subscribed and sworn to before me this Il 7 R 0 6 f d %v r 0Ai,itl')b'-
@ 404 d 9'/d?<o NwA w Notary Publigitate of Wisconsin My commi W.i expires O d le /'T/ O,
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15.4.2 IN-5ERVICC JNCPrC ION OF sal *CTY CI.ACS COMPONENTS A 4>11cably g 1
'oplics to in-service inspection of Safety Clacs Componento.
j_cctivec, To provide annurnnec of the continuing int egrity of the safety class nystems.
Speci fi ca ti_og A.
Steam Generator Tube Inspection Requirements 1.
Tube Inspection Entry from the hot-leg side wl.th examination from the point of entry completely around the U-bend to the top support of the cold-log is considered a tube inspection.
2.
Sample Selection and Tecting Selection and testing of steam generator tubes shall be made on the following basis (a)
One steam generator of each unit chall be inspected during inservice inspection in accordance with the following requirements:
1.
The incervice inspection r.my be limited to one steam generator on an alternating sequenca basis.
This examination chall include at least 6v of the tubes if the resulta of the first or a prior inspection indicate that both generatorr are performing in a comparabic mannar.
2.
When both ateara generators are required to be examined I
by Tabic 15.4.2-1 and i f the condition of the tubes in one gencrtitor is found to be nore severe than in the other steam generator of a unit, the stcan generator campling sequence at the cubsequent innervice it.opection chall be modified to excmine the steam generator with the more severo condi * '..;n.
(b)
The minimum samnic size, inspcction result clannification and the associatra required action shall be in confornance with the requircnonts specified in Table 15.4.2-1.
The resultu of each sampling examination of a steam generator j
shall be classifiud into the fc110 wing three cateqorien:
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f Defect in an imperfection of such severity that it exceeds A tube f
the ininimum acceptable tube wall thickncun of 50".
containing a defect in deicctivc.
giug_nJng_ JJmit in the imperfection depth beyond which the tube munt i,c removed f rom cervice, becaur4e the tube may become defective' prior to the next scheduled inspection.
The plugging limit in 401, of the nominal tube wall thickness.
6.
Corrective ricanuren All tubes that leak or have degradation exceeding the plugging limit cha31 be plugged prior to return to power from a refueling or incervice inspection condition.
7.
P_cyorto (a) Af ter each inservice examination, the number of tubes plugged in each steam generator shall be reported to the Commiculon as soon as practicabic.
(L) The complete results of the stean generator tube innervice inspection shall be includcd in the Operating Report for the period in which the inspection was completed.
In addition all results in Category C-3 of Tabic 15.4.2-1 shall be reported to the Coceniccion prior to resumption of plant operation.
(c) Reports shall include:
1.
11 umber and extent of tubes inspected
. O Location and percent of all thicknecs penetration for 2.
cach indication 3.
Identification of tubes plugged (d) Reports required by Table 15.4.2 Steam Generator Tube Innpcetion chall provide the infornation required by Specification 15.4.2.C.2 and a description of invectigations conducted to determine cause of the tube degradntion and corrective measures taken to prevent recurrence.
In-scrvice Inspection of Safety Clacs Components Other Than B.
Steam Generator Tubea Inservice incroction of ASMU Code C] ass 3, Clacs 2 and C1 css 3 1.
components shall be performed in accordance with Section XI of Doller and Preusure Vessel Code and applicabic 7ddenda as the AhpJ:
Section 50. 55a (q) modified by Section required by 10 CFR 50,
- 50. 5 Sa (b), except whero crecific written relief is granted by the Where difficulty tinC, purnuant to 10 CFR 50, Section 50.55a (g) U) (i).
O 15.4.2-3c
in encount ered in par rotning an inspection in accordance with the code, t he modified pl.m or alternative nethod will be fully doeurnanted for cubnmpmnt revic w by the tur. Ubould the Inc not agree with the document ed ch ino.";,
the item in quention wil] be re-examined in a onitually aarceabic manner.
2.
Irmervice tenting of AS!U Code Claan 3, Clasa 2 and Clans 3 puvim and valver, shall be par for t o. d in accordance wi th f:cet ion XI of t he ADlu: Boiler nnd Prmir,ure Vesrel Code and applicabic Addenda an required by 10 CI'R 50, Section 50. 55a (g) podifi,d by Section 50. 55a (b), c.s: cept where Fpecific written relief is gr anted by the 1:I;C purau: int to 10 CI P. 50, Sect ion 50. 55a (g) (. -) (i).
Where difficulty in encount red in perforning a tent in accordance vith the coda, the nodified plan or aJ tur nat ive raethod will be f ully documanted icr uuhocquent review by the NHC.
Should t.he !!RC not agree vith the document.ed changen, the pump or valve in quection will be retested in a mutually agreeable manner.
3.
Containment isolation valves will be tested in accordance with
)
Technical St acificatien 15.4.4 inctead of Section IFV-3420, Valve Leak Rate Test.
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That is, tho appropri.ito pump notor breakern shall have opened and cloned, and all valvec nhall have completed their travel.
D.
Containment Spray Syntem 1.
System tects chall be performed during reactor chutdouns for major fuel reloading.
The tent shall be performed with the isolation valves in the spray cupply lines at the containment blocked closed. Oper ation of the cyctem is initiated by tripping the normal actuation instrumentation.
Tbc motor breahors for the pumps shall be placed in the " test" ponition for this test.
2.
The test will be consincred satisfactory if visual observations indicato all components have operated satisfactorily.
3.
The spray nozcles shall be checked to verify that they are not obstructed at intervalc not execeding five years.
C.
Containnont ran Coolers Each fan cooler unit chall be tested at each refueling to verify proper operation of the backdrnft acmpers and the cervice water i
bypass valvec.
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15.4.5-2 i O l
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Danis The Safety Injection Systera and the Contair. ment Spray Syntem are principal plant Safety Systcmc that are normally inoperative during reactor operation.
Complete systens tests cannot be perforned when the reactor is operating because a safety injection signal causes containrcht isolation and a Containment Spray System test requires the syr tem to be tc;tporarily disabled.
The inethod of ascuring operability of thece r,ystems ir therefore to combine syctems testa to be perforned during refualing shutdownc, with store frequcnt conponent tectc, schich can be perforned during reactor operatien.
The syntena tents demonstrate preper automa t:ic operat ion of the Safety Injection and Containuent Spray Systems.
With the puivpa blocked from ctarting a test nignal ir applicd to initiate autmatic action and veri ficaticn mde that the componento receive the safety injectjen in the proper sequence.
The test demonstrates the opcIntion of the valves, pump ci rcuit breakers, arnl automrtlic circuit ry.
O 15.4.5-3
- e During reactor operation, i he inctrumentration which in depended on to initiate safety injection and containment opray ic generally checked weekly and the initiating circuits arc Leuted month.ly (in accordance wit h F1 ecification 15.4.1).
Other cyctems that are also important to the emergen-y cooling function are the accumulatora, the Component Cooling System, the Service Unter Systera and the containment fan coolerc. The accumulators are a passivo safeguard.
In accordanec_with Specification 15.4.1 the water volume and pressure in the accumulators are checked periodically.
The other systems mentioned operate when the reactor is in operation and by these means ero continuously monitored for satisfactory performance.
Referencen (1)
PSAR Section 6.2 O
15.A. 5-4
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15.4.7 MAIN ST! AM STOP VAINES The main steam stop valves are tented in accordance with npecification 15.4.2.n.
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- ' 15,4. 0 AUXILIA!!Y l'l'1!DUATr:n SYST1:f t e
The auxiliary feedwater 1>unips are t.ected in accordance with cliccification 1 5. 4. 2. 11 O
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