ML20085E220
| ML20085E220 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/11/1991 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20085E217 | List: |
| References | |
| NUDOCS 9110180105 | |
| Download: ML20085E220 (18) | |
Text
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F AUACHMEMLB PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSE NPF-11 AND NPF REVISED _PAGES UjiLLQHE(NPF-11) UNIT TH0 (RPE.lal 3/4 6-3 3/4 6-3
-B 3/4.6-1 B'3/4 6-1 9110190105=911011 PDR ADOCK 05000'373 P PDR ZNLD/965/19
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CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued)-
restore:
- a. The overall integrated leakage rate (s) to less than or equal to 0.75 L,. and
- b. The combined leakage rate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam isolation valves and valves which are hydrostatically leak tested per Table 3.6.t-1, subject to Type 8 and C tests to less than or equal to 0.60 L,. and
- c. The leakage rate to le/s than or equal to 100 scf per hour for all four main steam lines through the isolation valves, and
- d. The combined leakage rate for all ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to 1 gpm times the total number of such valves, prior to increasing reactor coolant system temperature above 200 F.
SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be demonstrated at the followir.g test schedule and-shall be determined in conformance with the criteria specified in Appendix J of 10JFR Part 50 usingje methods andyrovisions of ANSI l N45.4-1972j y - -((f}eet+omyetemp pens gnwQR6.Mifn
- a. Three Type A Overafl~fnt M d Containment Leakage Rat [ tel
^
be conducted at 40110 month intervals during shutdown at P,,
39.6 psig, during each 10 year service period. The third test of each set shall be conducted during the shutdown for the 10 year plaat inservice inspection.
- b. If any periodic Type A test fails te seet 0.75 La, the test schedule l p~
eviewed and approved by the Unlan an ewadna for subsequent c ) Commissiong!'f Type AType two consecutive testsA shall be to meet 0.75 L,,
tests fail a l 4yged h Cenimissm, Type A test shall be performed at least every 18 months until two consecutive Type-A tests meet 0.75 L,, at which time the above test l schedule may be resumed,
- c. The accuracy of each Type A test shall be verified by a supplemental test which:
- 1. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L,.
- 2. Has duration suf ficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.
- 3. Requires the quantity of gas injected into the containment or bled from the containmer.t during the supplemental test to be equivalent to at least 25% of the total measured leakage ]
at Pa, 39.6 psig.
LA SALLE - UNIT 1 3/4 6-3 Amendment No.18
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CONTAINMENT SYSTEMS LIMIT!NG CONDITION FOR OPERATION (Continued)
ACTION: (Continued) restore:
- a. The overall integrated leakage rate (s) to less than or equal to 0.75 L,, and
- b. The combined leakage ' ate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam isolation valves and valve which are hydrostatically leak tested per Table 3.6.3-1, subject to Type B and C tests to less than or equal to 0.60 L,, and
- c. The leakage rate to less than or equal to 100 scf per hour for all four main steam lines through the isolation valves, and
- d. The combined leakage rate for all ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the primary containment to less than or equal to 1 gpm times the total number of such valves, prior to increasing reactor coolant system temperature above 200*F.
SURVEILLANCE-REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of-10 CFR Part 50 using the methods and provisions of ANSI N45,4-1972f _ , g }t Q Q p g g g g g',"c g ,p g
- a. Three Type A Overall Integrated Containment Leakage Rate tests shall be conducted at 40 1 10 month intervals during shutdown at P,,
39.6 psig, during each 10 year service period. The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection,
- b. If any periodic Type A test fails to meet 0.75 L,, the test schedule for subsequent. Type A tests shall be reviewed and approved by the
~,_wh untm a exeqb r Commission. -- 4,If two consecutive Type A tests fail to meet 0.75 La , a Type is 73&ed by Sc ( consecutive A test shall be performed at least every 18 months until two c ,, d d 5ie % y Type A tests meet 0.75 L,, at which time the above test l' u.% ,.? schedule may be resumed.
- c. The accuracy of each Type A test shall be verified by a supplemental test which:
- 1. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L,.
) 2. Has duration sufficient to establish accurately the change in l
leakage rate between the Type A test and the supplemental test.
l 3, Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25% of the total measured leakage at P,, 39.6 psig.
1 i LA SALLE - UNIT 2 3/4 6 '
I -
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31 4 .'6 CONTAINMENT SYSTEMS BASES 1
i 3/4.6.1 PRIMARY CONTAINMENT
, 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restric-tion, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.
4 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the
- total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 39.6 psig, P , As an I added conservatism, the measured overall integrated leakage rat 8 is further
- limited to it.ss than o' equal to 0.75 L during performance of the periodic tests to account for possible degradati8n of the containment leakage barriers between leakage tests.
1- Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of 4
the valvesi therefore the special requirement for testing these valves.
The surveillance testing for measuring leakage rates is consistent
~
with the requirements of Appendix J to 10 CFR 50 with the exception of exemption (s) granted for main steam isolation ~' valve
" leak te g nd g gg the airlocks after each opening.4- -
3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS f*
c t 5 3ra d lay i e Na!7 cw* Oas The limitations on closure and leak rate or primary containment air descrd ed ,%
locks are required to me:t the restrictions on PRIMARY CONTAINMENT INTEGRITY TE 14rdh i and the primary containment leakage rate given in Specifications 3.6.1.1 and No%.1s.71 i 3.6.1.2. Tie specification makes allowances for the fact that there may be
- long periods of time when the air locks will be in a closed and secured position during reactor caeration. Only one closed door in each air lock is required to maintain the integrity of the containment.
3/4.6.i.4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR 100 guidelines provided the main steam line systs.r from the isolation valves up to and including the turbine condenser remains intact. Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIV's such-that the specified leakage requirements have not always been maintained continuously.
The requirement for the leakage control system will reduce the untreated leakage from the isolation valves when isolation of the primary system and containment is required.
LA SALLE - UNIT 1 B 3/4 6-1
3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment at.nosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restric-tion, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.
3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitadons on primary containment leakage rates ensure that the total containment leakage volume will iot exceed the value assumed in the accident analyses at the peak accider pressure of 39.6 psig, P As an added conservatism, the measured ove all integrated leakage rat $. is further limited to less than or equal to 0, 5 L during performance of the periodic tests to account for possible de9',dati8n of the containment leakage barriers between leakage tests.
Operating experience with the main steam line i. solation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves.
The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J to 10 CFR 50 with the exception of exemption (s) granted for train steam isolation valve leak testing and testing the airlocks after each opening. 4 ~
~N 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the primary containment leakage rate given in Specifications 3.6.1.1 and 3.6.1.2. The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation. Only one closed door in each air lock is required to maintain the integrity of the containment.
3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting fro.n the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a f small fraction of the 10 CFR 100 guidelines provided the main steam line system from the isolation valves up to and including the turbine condenser remains intact. Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIV's such that the j specified leakage requirements have not always been maintained continuously.
The requirement for the leakage control system will reduce the untreated [
leakage from the isolation valves when isolation, of the primary system and containment is required.
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LA SALLE - UNIT 2 B 3/4 6-1 w Ws . w u Wte &, M-N~ u -ndw.u,,grtJed by ike Ce*4
ALIACBMEHI_C
. CORRECTIVE ACTION PLAN FOR TfPE C TEST FAILURES CONTRIIRITING TO "AS-FOUND" ILRT FAILURES A. Problems:
- 1. During the 1987 first refuel outage, the 1988 second refuel outage and the 1990 third refuel outage, the LaSalle Unit Two LLRT leakage exceeded 0.6 La, maximum pathway leakage (highest leakage valve in each penetration).
- 2. During the first refuel outage and the third refuel outage, the LaSalle Unit Two CILRT failed the "as-found" condition due to the penalty additions from Type C LLRTs (minimum pathway leakage). If a CILRT had been required during the second refuel outage, the minimut pathway leakage ("as-found" leakage of the valve with the lowest leakge in each penetration minus the "as-inft" local leakage 'or the valve with the lowest leakage in the penetration) would " ave been acceptable.
B. Root Cause of the Problems:
- 1. Containment Isolation Valve leakage.
- 2. Inappropriate local leak rate test method for scco types of valves.
- 3. Heak trending and comparison of valve /penetratten perfoimance.
C. General local Leak Rate Test program improvements:
LaSalle Stetion has improved the quality of their Type B and C test program by implementing proper ptnetration saintenance, License Event Report (LER) investigatten, and engineering analysis to determine short and long term corrective actions.
It is LaSalle County Station policy to maintain the primary containment leakage as low as possible. Administiative limits are set for each penetration / component. If these ilmits are exceeded, the component is repaired. On a case-by-ccse basis, a technical evaluation is performed to allow this administrative limit to be exceeded only if the overall containment leakage is acceptahiy 1o4. This policy is established by LaSalle Technical Surveillance, LTS-300-5, " Local Leak Rate Test. 0.6 La Accountability Program." The total maximum pathway leakage rate is reported to the Startup On-Site Review committee for approval prior to the end of each refueling outage. Devietion reports (investigative reports) are initiated as required to document any Type B and/or C failures and describe corrective actio,s. Type B and C testing is considered an important and effective pcogram where actions are taken to address problem areas independent of the Type A test.
LaSalle County Statloa has established motor operated valve preventative maintenanco en containment isolation valves which includes:
- 1. Motor operator gear box grease sampling which, if the sample is ur. acceptable, requires that the valve operator will be refurbished.
In addition, all motor operated contalaeont isolation valve operators are scheduled to be refurbished by the end of the fifth refueling outage for each LaSalle County Station Unit.
INLD/965/11
AHACHENLC.Jtontinuedl
- 2. Diagnostic testing of motor operated valve thrust capabilities as specified in commitments to Generic Letter 89-10 provides information on the capability of the valve to operate against its design pressure. This testing is scheduled to be completed per commitments to Generic Letter 89-10 on applicable valves by the end of the sixth-refuel outage on each LaSalle Unit. Valves that do not meet the acceptance criteria will have torque switch adjustments and motor operator replacements as required to achieve the acceptable calculated thrust band.
D. Specific corrective actions completed or planned on specific valves identified as problematic:
- 1. 2RE024/2RE025 Drywell Equipment Drain Sump Penetration:
The drywell equipment drain sump penetration isolation valves have been recurring LLRT failures. The cause of the failures has been attributed to dirt / foreign objects getting into the drywell equipment drain sump during extended or refueling outages. The dirt or foreign material would subsequently be pumped through the valves causing this material to settle on their valve seats. The result of this caused improper / incomplete valve disc seating and/or irregularities to the valve seat / disc during subsequent valve operation.
An evaluation was performed to resolve the recurring failures. It was determined that the best solution was to install screens at the bottom of the sump during extended or refueling outages. This would prevent foreign material from entering into the piping and isolation valve seats thus mitigating valve leakage. In addition to the screen installation, the Equipment Drain Sump has been placed on a periodic cleaning schedule which will also minimize foreign material buildup and transfer.
Sump screen installation ar.d cleaning tc.ok place during the latest refuel outhge on LaSalle Unit i during the Spring of 1991. The screens are scheduled to be installed on Unit 2 during the fourth refueling outage (L2R04). The results of Type C tests indicated a great improvement in valve leakage performance.
Leakage performance for the drywell equipment drain surap penetration will continue to be monitored to verify that the problem is resolved and does not require any further action.
- 2. 2RF012/2RF013 Drywell Floor Drain Sump Penetration:
The drywell floor drain sump penetration isolation valves have been recurring LLRT failures. The cause of the failures has been determined to be similar to that associated with the 2RE024/2RE025 drywell equipment drain sump penetrations. The corrective actions and corresponding results are similar to those achieved for the 2RE024/2RE025 drywell eauipment drain sump penetrations.
Leakage performance for the drywell floor drain sump penetration will continue to be monitored to verify that the problem is resolved and does not require any further action.
(
AUAChMENIAlcontinuedl 3, 2G33-f001/2G33-F004 Reactor Hater Cleanup Suction Penetration:
The Reactor Hater Cleanup (RHCU) Suction Isolation Valves have been a recurring LLRT failure contributing to the CILRT "as-found" failures.
The cause of the RHCU Isolation valve failures varied, but were mainly attributed to dirty or scratched seating surfaces. Table 2 shows the "as-found" Local Leak Rate Test results for both Unit I and Unit 2 Reactor Hater Cleanup Suction Isolation Valves since the first refuel for each unit.
LaSslie first revised the testing methodology for the RHCU isolation valves during the third refueling outage (L2R03). The isolation valves were previously tested simultaneously by pressurizing the volume between the two valvas. That method made it difficult to troubleshoot and determine leakage through each valve and identify the minimum pathway leakage to be added to the "as-found" CILRT results.
The new test method individually tests the valves in the normal direction (trom inside the containment). This method allowed the test engineer to easily identify the exact leakage through each valve.
An engineering evaluation determined that the present valve design (single flex wedge gate) was adequate for the RUCU suction application. The evalcation concluded, however, that a double disc gate valve would greatly improve leatage performance for this particular application. With the new design, each valve has two discs and associated seating surfaces, thus doubling the barrier to valve leakage.
LaSalle contacted other utilities to conduct a strvey on the performance of double disc gate valves in the RHCU suction application. Clinto, Power Station uses the double disc gate valve design for the same application and has experienced no valve leakage problems. Hope Creek Nuclear Station has not had any recurring problems with their double disc gate valves since installation in 1985. Fitzpatrick has recently installed double disc gate valves for the same appilcation as LaSalle Station. The previous single flex wedge gate design at Fitzpatrick experienced Leakage Rate Test failures. After inittai installation, fitzpatrick experienced LLRT problems with the new design caused by improper initial installation.
The valves were retested with satisfactory results after maintenance.
Brunswick has also recently installed and subsequently experienced problems with their double disc gate valves. The valve problems are presently under evaluation; however, early investigative results show that the leakages in the new design are the result of manufacturing flaws in the valve.
The RHCU Suction Isolation Valves are scheduled to be replaced with the new valve design no later than the end of the fifth refueling outage for Unit 2, (L2R05) scheduled to begin in September 1993.
Although the RHCU valves have contributed to the "as-found" CILRT failures, only in one instance in eight LLRTs (including Unit I results) did the valves account for total CILRT failure (Unit 2, third refuel) due to excessive minimum pathway leakage. Past experience for Units I and 2 has shown that the RHCU valves can be successfully INLD/965/23 -
i ATTACHMENT C (ran11auedL n
repaired (if_ required) and tested to perform their function as- ~i containment isolation valves. LaSalle Station is confident that the-proposed resolution for the-RHCU Isolation Valves will greatly improve the overall containment: leakage performance.
4c _21NO31: Traversing Incore Probe (TIP) Air Purge Supply Penetration;-
1The TIP Purge Air-Supply Isolation t.lve, 21NO31, is a single containment isolation valve. This con.ponent failed its LLRT during- ~,
the.first LaSalle Unit 2 refueling outage and contributed to the "as-found" Type A CILRT failure. The cause of the failure was .
_. attributed to a dirty seating surface in conjunction with corrosion.
-The valve was-disassembled, cleaned and the internals were replaced.- ,
The 21NO31 valve was originally tested in the reverse direction. The test procedure, LTS-100-32 "Drywell-Pneumatic System Discharge Isolation Valves Local Leak Rate Test 1(2)IN017, 1(2)INO31,-and
-1(2)IN018," was-revised to test the component in the normal direction which is from inside the containment.-
]
Since the initial failure of 21N031 during the first refueling-outage-(L2R01),:the valve has performed satisfactorily with no leakage.
5, 2E12-F053A "A"; Residual Heat Removal Shutdown Cooling-(RHR SDC)-Return Penetration: '
The'"A" RHR-SDC' Return Isolation Valve, 2E12-F053A, failed its LLRT during the first Unit 2 refueling outage and contributed to the "as-found" Type A CILRT failure. The cause of;the leakage was identified to be a defective valve disc. A new-valve disc was installed and the valve' seat was refaced.
The repair was successful: based on subsequent testing performed during -
the next-two refueling outages for LaSalle Unit 2. Since the initial LLRT_ failure,-testing results showed minimal or no leakage.
- 6. 2E12-F0538 *B" RHR Shutdown Cooling (RHR SDC) Return Penetration:
The "B" -
RHR SDC Return Isolation Valve, 2E12-F0538, failed its LLRT-dur_ing the Unit 2 third refueling outage (L2R03) and contributed to the "as-found" CILRT failure. The cause of the excessive leakage;was determined.to:be the failure of the motor operator to properly drive ,
.the disc far enough into its seat. The motor operator was refurbished and the valve was retested with satisfactory results. During_the refurbishment, the only-item replaced that may have contributed to the failure was the-torque switch. The torque switch was replaced for administrative reasons with no specific problem with.the torque switch noted in the. work request documentation; therefore, the root cause of the valve 1 failure to fully close was undetermined. LaSalle-Administrative Procedure, LAP-300-31, " Motor Operated Valve Program,"-
is being revised to require root cause determination for any motor operated valve failure to prevent recurrence. This procedure' revision will-be completed before the next refueling outage. This failure is an isolated occurrence based upon a review of MOV Hork Requests. In addition, as mentioned in Item No. C above, LaSalle Station has an ongoing preventative maintenance program for MOVs which includes
=
ZNLD/965/23 -
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'AUACIMNLCIRontinuedl
'[ test!ng'andrefurbishmente lhese measures help assure that the 2E12-f053B penetration has successful LLRTs in the future.
Prior-to th'e'LLRT failureJ the 2E12-F0530 valve was tested wtth -
satisfactory: test results during the first and second refuel outages .
-for LaSalle Unit 2.
-7. 2HG005A/2HG006A Hydrogen Recombiner 2H501A Unit 2 Suppression Pool
- Discharge Penetration:
The Hydrogen Recombiner 2H601A Unit 2 Suppression P wl Discharge Isolation Valves. 2HG005A and 2HG006A, ht.ve-had recurring LLRT fatlures. During the first LaSalle Unit 2 refueling outage, the excessive leakage contributed to the'"as-found" CILRT-failure. Valve seat trregularities were determined to be the cause of the LLRT failure.-. Valve seating surfaces were lapped and the valves were retested sattsfactorily.
During the second and third refueling outages, only one of the two isolation valves continued to have leakage problems (2HG006A). In each: case, the valve. was determined to have a dirty and irregular -
seating surface. .The seating surfaces were cleaned and lapped and retests were performed with satisfactory results. Since the first ,
refueling outage, the Type A CILRT "as-found" results would have been unaffected by the single valve LLRT'fallure. -
Because.of.the 2HG005A/2HG006A Isolation valve LLRT failures, all Hydrogen Recombiner Primary Containment Isolation Valves are presently
. under evaluation to determine the long term corrective action-required to. prevent recurring LLRT-fallure. There are six other isolation valves of the'same size and function as 2HG005A and 2HG006A associated with the Unit 1 and Unit 2 Combustible Gas Control Systems (note-the ?
comparison in Table 1 to the 2HG005B/2HG0068 penetrations). The only other tsc.lation valve with a similar failure is the Hydrogen
-Recombiner 2HG01A Unit 1 Suppression Pool Discharge Valve, 1HG005B, which failed LaSa11e's LLR1 administrative limit. There have been a 3 total'of four LLRT failures in 16 LLRTs for these four penetrations and three of these are associated with the 2HG005A/2HG006A penetration. The fourth failure was the IHG005B/1HG0060 penetration .
- which was corrected by: cleaning and lapping the valve seat of the 1HG005B.- This Unit 1. penetration has-since been Local. Leak Rate tested-three times satisfactorily.
There are four other penetrations associated with the supply lines to the Hydrogen Recombiners. The Hydrogen Recombiner 1HG01A Unit 1
-Drywell. Suction Valve, 1HG001A, and the Hydrogen Recombiner 2HG01A Unit'1 Drywell Suction Valve, 1HG001B, each failed their. respective Lt.P.1s during the Unit-1 third refuel and were repaired by lapping the -
valve seats. Both of these penetrations were satisfactorily Leak Rate
. tested during-the fourth Unit I refuel. The Unit 1 Local Leak Rate
- Test failures were shown by test to be due to omy one of the two ,
. valves-in each penetration; therefcre, the "as-found'? CILRT was- not affected.
The 2HG005A and 2HG006A are on the initial list of containment isolation l valves to be tested in any non-refuel outage of greater than 30 days.
ZNLD/965/25 a 2, . . - , . - --,,_a.._.._. - _ . , . . _ . _ , , . _ . _ . - _ . . . . . _ . _ _ _ - _ . _ ,
-l
. AIIACHSBLClcoatinutd1 See Table ITfor a tabulation of the historical-Local Leak-Rate Test 2
performance of the above Unit-Two valves.
LE. Corrective Actions-for.the Local Leak Rate Test program: ;
1 '. ' Alternative-Leakage Test-Program:-
In addition to the LLRTs performed during refueling outeges for ,
--"as-found" maximum pathway leakage. LaSalle County Station will perform LLRTs on penetrations which are determined by Item No. 2 -
and/or Item No. 3 below to be:susceptibic to excessive leakage. These
.LLRTs will be performed during any LaSalie Unit Two non-refuel outage; of; greater than 30. days.
- 2. LaSalle' County Station will develop and implement an improved trending
. program for tracking penetration and valve leakage rate performance.
The program will_ include several of the parameters listed in Item No.
E.3 below-and is planned to be computer based. .
3.---Identification of valve type, service, and manufacturer for leakage.
rate; performance comparison for all Primary Containment Isolation Valves that are-required to be' Local Leak Rate tested to help determine anyl patterns or groups that are either-exceptionally good pe formers with minimal.or no leakage or-are poor. performer _s with:
/eral cases of high leakage.
,, Application of_ engineering evaluations to develop recommendations for the. improvement of'tast methods and their implementation orLthe repair, modification or replacement of problem Primary Containment
-Isolation Valves which have been. identified as described above.
The test method improvement-will include a technical review of LLRF
~
procedures for poor LLRT test performers to verify the proper 1 type of test for=the penetration involved (for example, test direction from the Primary-Containment outward versus between a pair of isolailon-valves in thecsame penetration or from the outside in), test boundary adequacy..and other aspects-of the test method for clarity and user friendliness.
ZNLD/965/26
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n F.
AIIACliMENLC_(sontinuedl TABLE 1 Loc at Le aLRa t e_Ies t_Ee rf ormanc e_S i nc e_1987 VALVE 1987 1988/1989 1990- LEAKAGi CONTRIBui10N
____ .L2ROL__._L2R02. *
- _L2R03 __.30 "AS-E0VHD" CILRI_._
< L2E0J _L2ROL.
2RE024 29,1 5.9 5.2 0.0 *: ,'/. '
2RE025 2RF012 0.51 9.19 225.5 N/A* 110.32
~ 2RF013 2G33-f001 17.7 200.1 GROSS 8.67 GROSS 2G33-f004 21NO31 250.9 0.0 0.0 250.88 N/A*
2E12-f053A 88.2 0.0 2.65 87.83 N/A*
2E12-f053B 0.46 0.0 65.7 N/A* 63.28 2HG005A 121.6 31.9 111.0 60.57 0.5 2HG006A 2HG0058 0.37 0.32 0.37 N/A* N/A*
2HG0060
- No work performed on valves.
- Non-CILRT outage.
I
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m AUACHMEHLC_Lcontinuedl TABLE 2 Loc tLL e ALRa t ele tfor man c elor_Un tt_ Land _2 '
Re ac tor _ Hat e r_ Cleanup _S u ctioLI s olation_Vah ei VALVE 1st FORCED /SURV 2nd 3rd 4th REEUEL QUIAGE-*. ** REEVEL' _REEUEL - REEUEL*
(values shown are "as-found"/"as-left")
Unit 1:
I 1G33-F001 1.71/1.72 1,84/1.84 2.7/2.7 GROSS /37.4 23.382 /0.0 1G33-F004 28.063 /0.0 Unit 2:-
2G33-F001 17.74 /0.37 200,1 5/3.31 GROSS6 /3,33 7
2G33-F004 GROSS /4.19 NOTES:
- Non-CILRT outage.
- Unit-1 forced maintenance and surveillance outage.
- 1. Incomplete seating. surfaces on 1G33-F004. The valve seating surfaces were cleaned and lapped.- -The retest was accepted based on low total maximum combined leakage.
- 2. Incomplete seating surfaces on IG33-F001. The valve seating surfaces were cleaned and lapped and the valve was repacked.
- 3. Incomplete seating surfaces on 1G33-F004. The valve seating surfaces were cleaned-and lapped.
- 4. Incomplete ~ seating surfaces on 2G33-F001. The valve seating surfaces'were cleaned and lapped. -The seating surface was slightly damaged and a crack was found in the disc of 2G33-F004. Lapped the valve seat, machined and lapped the new disc to match the seat and repacked 2G33-F004,
- 5. Incomplete seating surfaces on 2G33-F001. The valve seating surfaces were cleaned and lapped. Replaced the disc on 2G33-F004 due to crack found the previous outage. Lapped the valve seat, machined and lapped the new disc to match the seat and repacked 2G33-F004,
- 6. Incomplete seating surfaces on 2G33-F001. The valve seating surfaces were
-cleaned and lapped.
- 7. Incomplete seatings surfaces and severe packing leak on 2G33-F004. The valve seating surfaces were cleaned and lapped and the valve was repacked.
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. ALTACHMENLD SUPNARY AND PROPOSED TYPE A TEST SCHEDULE TYPE A TEST E IE DISCRIPl10N CONDUCIED P. ASS /IAIL_
JANUARY 1987- L2R01 YES FAIL OCTOBER 1988 L2R02 NO2 ____
HARCH 1990 L2R03 YES FAIL JANUARY - MARCH 1992 L2R04 NO I ----
SEPTEMBER - NOVEMBER 1993 L2R05 YESI ----
MARCH - MAY 1995 L2R06 NO I ----
NOV. 1996 - JAN, 1997 L2R0/ YESI ----
NOTE 1: Proposed Type A test Schedule NOTE 2: Minimum pathway leakage for determining leakage penalties for an "as-found" CILRT would have been acceptable, ZNLD/965/28
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. ALIACliMENl_E !
SIGNIFICANT llAZARDS CONSIDERATION i Commonwealth Edison has evaluated the proposed Technical Specification Amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92, operation of LaSalle County Station Unit I and Unit 2 in accordance with the proposed Amendment will not:
- 1) Involve a significant increase in the probability or consequences of an accident prevlonsly evaluated because:
The proposed amendment allows the station to waive the requirements of Surveillance Requirement 4.6.1.2. If an exemption is granted by the NRC.
10 CFR 50.12(a) states:
"The Commission may, upon application by any interested ps1sr or upon its own initative, grant exemptions from the requirements of the regulations of this part, . "
Allowing the use of the exemption process towards the Cll'l test requires LaSalle Station to show that the granting of the exemption is authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. Also, special circumstances are required to be present fer the granting of an exemption.
One of the special circumstances that < bob apply in this instance is 10 CFR 50.12(a)(2)(11) which states:
" Appl, cation of the regulatio) *n the particolar circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule..."
The granting of such an exemption by the NRC staff requires LaSalle Station to show that unacceptable containment leakage has been identified and corrected. Alternatives to the testing requirements of Appendix J must assure that Primary containment leakage will continue to be within limits.
Exceeding the allowable leakage rate during the performance of the Containment Integrated Leak Rate Test (CILRI) is indicative of either a passive or a structura? component that is leaking or that there is an inadequacy ir, the Local Leak Rate Test (LLRT) program. When the failure of a CILRT is due to a passive or structural component, the only test for adequate repair would be the CILRT. For a LLRT program inadequacy, the CILRT would serve as a means of verification of the results of the test program. The more frequent performance of the CILRT as required by LaSalle County Station Technical Specifications due to the significant contribution of Local Leak Rate Test failures is redundant to the performance of LLRTs; under such circumstances there is little or no benefit to be gained by performing a Type A test on an accelerated schedule.
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! AUAttttiENT _ E (rAttt.inuedl I .
Analyzed accidents at LaSalle Station that involve a potential off-site radioactive release include as an assumption the minimum pathway Primary Containment Isolation Valve leakage. The performance of Local Leak Rate Tests (Type B and C) identifies leaking valves and penetrations. l'he verification of "as-found" and "as-left" local leakage assures that Primary Containment leakage will be within the analyzed limit assumed for accident analyses. The possibility exists that slightly increased local leakage rates may lead to slightly increased dose consequences in the event of an accident. However, the potential dose consequences are maintained to acceptable minimum levels if Primary Containment Isolation Valve leakage is maintained below analyzed limits. Any exemption request to the requirements of Appendix 3 must satisfy 10 CFR 50.12(a)(2)(iv) to verify there is no adverse impact on the public health and safety. LaSalle's proposed amendment requires Staff review and approval of any exemption from the survelliance requirements of 10 CFR 50, Appendix J. Therefore, there is no significant increase in the dose consequences of any accident previously cvaluated.
Containment leakage is not considered an initiator of any previously evaluated accident; therefore, there is no increase in the probability of an accident previously analyzed.
- 2) Create the possibility of a new or different kind of accident from any accident previously evaluated because:
Both Local Leak Rat? Testing and Containment Integrated Leak Rate Testing as specified in the LaSalle County Station Safety Analysis Report were evaluated in Secticn 6.2.6 of LaSalle's Safety Evaluation Report, NUREG-0519, and fotnd to be acceptable. Local Leak Rate Testing identifies and verifies corre. tion of penetration leakage. Local Leak Rate Testing penalty additions :aused the f ailure of "as-found" CILRTs for LaSalle Station (two "as-f>und" CILRTs for LaSalle Unit Two failed as a direct result of incorpor, tion of the minimum pathway leakage determined by local Leak Rate Tests). Local Leak Rate Testing will provide adequate assurance of the continued ictegrity of the Primary Containment without increasing the frequency of Containment Integrated Leak Rate Tests. Following Staff review and approval of an exemption request, primary Containment integrity will continue to be maintained as designed and previously evaluated.
Because there are no changes to the facility or operation of the facility as described in the UFSAR, this amendment request does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3) Involve a significant reduction in the margin of safety because:
At the design basis accident pressure specified in Technical Specification 3.6.1.2.a. a 40 percent margin to the maximum allowable containment leakage rate (La) is maintained for total maximum pathway leakage determinations.
This limit is determined from Local Leak Rate Tests. Administrative guidelines have been set for each penetration or valve to ensure that abnormal leakages w111 be corrected. Repairs or corrections to penetrations demonstrating excessive leakaae are performed unless the total Type B and Type C leakage is maintained at less than 0.6 La. Although there exists the possibility of slightly increased leakage rates for a small sample of penetrations, all repairs are required to restore leakage rates to less than the administrative limit at the next refueling outage.
ZNLD/965/31
. A11ACliMEMI_E_1 continued 1 Local Leak Rate Test. in conjunctial with a comprehensive corrective and preventive maintenance program for p'enetrations determined to be poor performers will assure that the Primary Containment Integrity will be maintained without additional Containment Integrated Leak Rate Tests.
The proposed amendment request allows pursuit of an ekemption from the requirements of 10 CFR 50, Appendix J as recommended in IEN 85-71. To meet the guidance provided in IEN 85-71, LaSalle is required to propose a Corrective Action Plan that demonstrates CECO's committment to ensuring Primary Containment leakage rates are maintained to acceptable levels.
Local Leak Rate Test minimum pathway leakagc rate penalties were the direct
-cause of "as-found" Containment Integrated Leak Rate Test failures for LaSalle Station; therefore, verification of an adequate margin of safety is assured by the conduct of Local Leak Rate Tests. Section 6.2.6 of the Standard Review Plan (SRP) assures that CILRT tests are performed in accordance with the requirements of 10 CFR 50, Appendix J. Lasalle's proposed exemption process requiring Staff pre-approval ensures Primary Containment integrity will be maintained and would therefore result in no sign!ficant reduction in the margin offsafety.
Guidance t.s been provided in " Final Procedures and Standards on No Significant Hazards Considerations," final Rule, 51 FR 7744, for the application of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which are-and are not considered likely to involve significant hazards considerations. This proposed amendment request most closely fits the example of a change which may either result in some increase tio the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are :
clearly-within all acceptable criteria with respect to the system or component specified in-the Standard Review Plan.
This proposed amendment reque3t does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions for operations. Therefore, based on the guidance provided in the federal Register and the criteria established in 10 CFR 50.92(c), the proposed amendment request does not constittte a significant hatards consideration.
ZNLD/965/32
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AlTACliMENT E ;
i ENVIRONMENTAE ASSESSMENT STATEMENT APPLICADIL11Y REVIEW Commonwealth Edison has evaluated the proposed exemption request and Technical Specification amendment request against the criteria for the identification of Itcensing and regulatory actions requiring environmental assessment in accordance with 10 CfR 51.20. It has been determined that the proposed changes meet the criteria for a categorical exclusion as provided under 10 CFR 51.22(c)(9). This conclusion has been determined because the changas requested do not pose significant hazards consideration or do not involve a significant increase in the amounts, and no significant changes in the types, of any effluents that may be released off-site. Additionally, this request does not involve a significant increase in individual or cumulatise occupational radiation exposure.
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