ML20096D611

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Proposed Tech Specs Change 100 to Eliminate Main Steam Line Radiation Monitor Scram & Isolation Functions
ML20096D611
Person / Time
Site: Cooper Entergy icon.png
Issue date: 05/04/1992
From:
NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20096D605 List:
References
NUDOCS 9205180155
Download: ML20096D611 (11)
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l NOTES FOR TABLE 3.1.1

1. There shall be two operable or tripped trip systems for each function. If the minimum number of or>erable instrument channels for a trip system cannot be met, the affected trip system shall be placed in the safe (tripped) conditi on , or the appropriate actions listed below shall be taken.

A. Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

a B. Raduce power to less than 30% of rated.

C. Reduce power level to IPJi range and place mode switch in the Startup position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and depressurite to less than 1000 psig.

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2. Parmissible to bypass, with control. ro'd block, for reactor protection system reset J

in refuel and shutdown positions of the reacter mode switch.

3. This note dele te<' .
4. Permin91bic to bypass when turbine first stage pressure is less than 30% of full load.
5. IIdi's are bypassed when APPJi's are onscale and the reactor mode switch is in the run position.
6. The design permits closure of any two lines without a full scram being initiated.

When the reactor is suberitical, fuel is in the vessel, and the reactor water temperature is less than 212*F, only the following trip functions need to be operable:

a. Mode switch in shutucun,
b. Manual scram.
c. IPJ1 high flux. 120/125 ir.dicated scale.
d. APPJi (15?.) high flux scram.
8. Not required to be operable when pnmary containment integrity is not required.
9. Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels act to exceed 5 MW(t).
10. Not required to be operabla when the reactor pressure vessel head is not bolted to the vessel.

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COOPER NbCLEAR STATION TABLE 4.1.1 (Page 2)

REACTOR PROTECTION SYSTEM (SCRAM INSTRUttEETATION) FdNCTIONAL TESTS

!!IN1dMM FUNCTIJNAL TFJT FREQUENCIES FOR SAFETY INSTP.. AND CONTROL CIRCUITS Minimum Freuuency (3)

Grot o (2) Functional Test

_ Instrument channel Trin Channel and Alarm Once/3 Months A

Itigh Water nevel lu Scram Llscharge Volume CRD-LS-231 A & B CRD-LS-23^ A'& B CFD-LT-T 1 C & D CRD-LT-234 C & D I

-A Trip Channel and Alarm Once/tfonth (1) flain Steam Line Isolation Valve Glosure MS-UU-86 A,B,C, & D nS-IRS-89 A,B,C, 6 D A Trip Channel and Alarm Once/ Month (1)

Turbine Control Valve rast Closure TG.-63/OPC -1,2,3,4

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Trip Channel and Alarm Once/3 Months Tat bi ne First Stage Pressere. A Permissive MS-PS-14 A,B,C, 6 D A Trip Channel and Alarm Once/Monch (1) h.rbine Stop Valve Closure SVOS-1 (1), SV05-1 (2)

SV03-2 (1), SVOS-2 (2)

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NOTES FOR TABLE 4.1.1

1. Initially once per month until exposure (M as defined on Figure 4.1.1) is 2.0 x 10 ;

thereafter, according to Figure 4.1.1 with an interval not les:, than one month nor more than three months after review and apptoval of the NRC. The compilation of I instrument failure rate data may include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of CNS.

2. A det:ription of the three groups is included in the Bases of this Specification.

-3. Functional tests are not required when the systems are not required to be operable or are tripped. If reactor,startups occur more frequently than once per week. the maximum functional test frequency need not exceed once per week.

.If tests are missed, they shall be performed prior to returnina, the systems to an i operable status.

4. Deleted. l

.5. Test R?S channel after maintenance.

6. The water level in the reactor vessel will be perturbed and the corresponding Ir, vel indic stor. changes will be monitored. This perturbation test will be performed every mouth af ter completion of the monthly functional test program.

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NOTES FOR TABLES 4.1.2

1. A description of three grwps is included in the bases of this Specification.
2. Ca?ibration tests are ncc required when the systems are not required to be operable or are tripped but are required prior to return to service.
3. Deleted.  !
4. Maximum frequency required is once per week.
5. Response time is not a part of the routine instrument channel test, but will be checked once per operating cycle. The response time measurement will be the time segment from the time tho' sensor contacts actuate to the time the scram solenoid valves deenergite.
6. Physical inspection and actuation of these position switches will be performed during the refueling outages.

7 On controlled shutdowns , the IRM reading 120/125 of. full scale will be set equal to or less than 45% of rated power. All rape,e scales above that scale on which the most recent IRM calibration was performed will be mechanicall'f blocked.

B. The Flow Bias Scram Calibration will con.41s t of calibrating the sensors, flow

, converters and signal offset networks during oporation. The instrumentation is an analog type with redundant flow signals that can be compared. The flow bias trip and upscale will be functionally tested according to table 4.1.1 to assure proper operation during the operating cycle. Refer to Bases of 4.1 for further explanation of calibration frequeneles.

9. LFRM detectors shall be calibrated every six weeks of reactor power operatioc. above 20% of rated power.

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LIMITING CONDITIONS FO LOPERATION SUoVEILUNcE_RFOUTpFMENTg 1

3.1 - BAl&E (Cont'd. ) 4.1 BASEE (cont'd.)

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-initiate the core standby cooling 2. The factor M is the exposure equipmenc. A high drywell pressure hours and is equal to the ,

scram is provided- at .the same aumber of sensors in a group, setting as the core standby ecoling n, tires the elapsed time systems (CSCS) iniciation- to T (M - nT).

minimize the energy which must be 3. The accumulated number of accommocated during a loss- of unsafe failures is plotced as coolant accident: and to prevent an ordinate against M as an return- to criticality. This instrumentation is a backup to the h FW M 1.

-reactor vessel water level 4. After a trend is established, iutrumentation. - the appropriate monthly test ,

j_ interval to satisfy . the ,3oal A reactor mode switch is provided will be the test interval to which actuates or bypasses the the left of the plotted

-various scram functions appropriate points.

to the part.icular plant operating 5 A test interval of 1 month s atua, Ref, paragraph VII.2.3,7 will h M M M G m il a trend is established, which is based on system availability

The manual scram function is active analysis and good engineering in all modco,-thus providing for a manual neans of rapidly inserting 3"dD**"* P 1 "* E'#8'I"E experience, control- rods -during all modes of reactor operation. Group (B) devices utilize an analog sensor followed by an ' amplifier and-The APRM (High flux in P, tart Up or a bi stablo trip circuit. The Refuel) system provic'es protection sensor and amplifie r are active against excessive power levels and components and a failure is almost short reactor periods- in the always accompanied by an alarm and; ',,

start-up and intermediate power an indicatiot, of the source of ranges. trouble. In the event of failure, repair or substitution can scart 1 The IRM system provides protection. immediately. An "as-is" failure is one that " sticks" mid scale and is not capablo of going either up or 1

-down in response to an out of-limits input. This type of failure for

. analog devices is a rare occurrence and is detectable by an operator who

-observes tha t- one signal does not

-track the other chiee. For purpose of analysis, it is assumed that this rare fail."re will be decected within cwo hours.

The bi-stable trip circuit which is s . a part of the Group (B) devices. can sustain unsafe failures which are (6) Reliability of Engineered Safety Features as a Function of Testing Frequency, I.M. Jacobs, " Nuclear Safety", Vol 9, No. a, July-Aug.

1968, pp. 210.:12 Change-No.

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. COOPER NUCLEAR STATION' .

<u F T/0LE 3.2,A:(Page 1).

' PRIMARY CONTAINMENT nND REACT 04 VESSEL IS01ATION INSTRUMENTe ' ION -

.b .. _ b Minimum Nwnber Action Required' oiLOperable When Compenent; Ins tr+ unent .

Components Per . Operability is;.. .

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. : ' 3 m4s itcan Line !!!gh RMP-RM-251, A,B,C,&D . s 3 Times ~ Full Power 2 .E- ]

M M stiaa ag wser tea W ter Level NBI-LIS-101, A,B,C,6D'#l 2+4.5 in.' Indicated Level 2(4) A or B n avut izw lov tow Weter NBI-LIS-57 A & B #1 2-145.5'in-. Indicated level 2 A or B

%i NB1-LIS-58 A 6'B.81

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. 145, 146, 147, 148, 149

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'MS-dPIS-116 A,B,C,&D 6 150% of Rated Steam 2(3) B

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-NOTE.%-FOR TABLE 3.2;A

1. . Whenever Pr' nary Containment _ integrity is required there

. shall.be-two o,erable or L tripped trip systems for each function..

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2. If the minimum number of operable instrument channels per-trip system requirement l cannot be met by : a . trip system, that trip system shall be tripped. If the a

(- requirements cannot bo met by both trip systems, t).e appropriata a: tion listed below-1 shall be taken.

'A Initiate an wderly shutdown and have the reactor.in a cold shutdown condition in 24 houra.

-B. It.itiate an orderly load reduction and have the Main Steam Isolation Valves shut-.within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

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D. Isolate the Shutdown Cooling mode of the RHR System.

E. Isolate the Reactor Water Sample Valves.

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-3. t Two. required for each steam line.

4. These signals also start the' Standby Gas. Treatment System and initiate. Secondary Containment isolation.
5. Net required in the refuel, shutdown, and startup/ hot standby modes tinterlocked with the mode,switen).
6. ' Requires one channel 2 rem each physical location cor each trio system.

7 Low vacuum -isolation . is bypassed when che turbine stop is not full open, manual

bypass switches are in bypass and mode rwitch is not in RUN.
8. The instruments on this tab.e produce primary containment and system isolations. The following: listing-groups the system signals and the system isolated.

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! solation Signals:

1. P4 actor Low Low Low Water Level -(2-145.5 in.)

J2. .Maia St am Line Low Pressure ( 2825 ps ig in t he R'.*N mode )

3. Main Stean Line Lvak Detection ($200nF)
4.  : Condenser Lov-Vacuun (27' Ho vacuum)
  • - 5. Main Steam Linn High Flera is1501t of rated-flowi Isciations:

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HQIfts-FOR TABLE 3.2.D 1, [ Action required when component operability is not assured.

A. (1) If radiation level exceeds 1.0 ci/sec (prior to 30 rain. delay line) for a period greater than 15 consecutive minutes, the off-gar isolation valve shall close and reactor shutdown shall be initiated immediately and the ,

reactor placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

A.  :(2) Refer to Specification 3.21.A.2.

B. A minimum of one instrument channel per trip system shall be operable when -handling i.rradiated fuel inside sacondary containment , _ and when '

moving loads inside secondary containment which have the potential to damage '.rradiated fuel. If this requirement cannot be met b-fa trip .

system, then that trip ' system shall be tripped. If this requirement cannot be met by both trip systems, then the following actions shall be

!- taken:

G Cease handling of irradiated fuel inside secondary containment and remove the load from over the irradiated fuel via the most direct path, or

~(2) LIsolate secondary containment and start SBGT, O. During-release of radioactive wastes, the effluent control monitor shall

  • be set tc alarm and automatically close the waste discharge valve prior to exceeding the. limits of Spee?fication 3.21.B;1.

D. Refer to Section entitled " Additional Safety Related Plant Capabilities".

E.  : Refer to . Section 3.2.D.S and the requirements for Primary Containment Isolation on= high noin steam line radiation, Table 3.2. A.

2. Trip settings to Ocorrespond to Specification 3.21.B.l.

- 3. Trip settings to correspond to-Specification 3.21.C 6 a.

- 4. . Minimum number of channels shall be _ cne during mechanical vacuum ptmp operation.

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COOPER 14UCLEAR STATIOri ,

TABLE 4.2.A (Pago 1)

PR IMARY COffrAI!!MEttr At!D REACW. VESSEL ISOLATIO!3 SYSTEM TEST A!!D CALIBRATIO!1 FREQUEriCIES i

Instnment Function Test Freq. Calibration Freu. Owck

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J3 tiBI-LIS-101, A,B,C,&D Once/ Month (1) Once/3 Months 0nce/ Day  !

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unct Lev,1 i.3I-LIS-57, A% B #2 Once/ Month (1) Once/3 11onths once/ Day  !

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!!B I - L IS- 5 8 A & B #2 i4B I - L15 - 57, A& B #1 Once/ Month (1) Once/3 Months Once/ Day

, e 1. - ! s w< Oter Lv .s l llBI-LIS-58, A& 9 #1 u li t ;h .iadiat lon RM P-F11- 2 51, A,B,C,&D Once/ Month (1) (13) Once/3 Honths (14) Once/ Day l aa $

- - 2  ;, t o. , ten MS-TS-121, A,B,C,LD Once/ Month (1) Once/ Operating trone j

' 122, 123, 124, 143, 144, Cycle _;

145, 146, 147, 148, 149, 150 i 1

(i a u t t, , F i r/w 11S- d PIS - 116, A.E,C,LD Once/ Month (1) Once/3 Months tione I

4 22 117 Once/Mc. nth (1) Once/3 M<_nths flone 118 Once/ttonth (1) Once/3 Months Ilone I 119 Oncc atont h (1) Once/3 Months tione l l

2.a . _w h. e s s . 113-PS-134, A,B,C,LD Once/ Month (1) Once/3 Months 11cne f

.. n i itR'PS-128, A&B Once/ Month (1) Once/3 Months tLane

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, ~ ,. wa MS-PS-103, A,B,C,LD Once/ Month (1) ence/3 Months tione RWCU-dPIS-170, A &B Once/ Month (1) Once/3 Months lione

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tlich Flow RWCU-TS-150 A-D, 151, 152, once /Montin (1) Once/ Operating IIone a: # High Opace 153, 154, 155, 156, 157, Cycle 158, 159, RWCU-TS-81, A,B,E,F RWCU-TS-81 C,D,G,H J

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' ' UNITED STATES 3

NUCLEAR REGULATORY COMMISSION o, I f sc ADVISORY COMMITTEE ON NUCLE AR WASTE WASHINGTON D.C. 20555 v

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f January 21, 1992 MEr40RANDUM FOR: James M. Taylor Executive Director for Operations FROM: F.' My Executive Director, A 4W Z

SUBJECT:

38TH ACNW MEETING FOLLOW-UP ITEMS $

Based on discussions regarding methods for improving implementation and- follow-up of ACNW recommendations, a summary of " Actions, Agreements, Assignments, and Requests" made during each ACNW meeting is sent to 'four office following erh meeting.

Attached is a summary of the " Actions, Agreements, Assignments, and Requests" made at the 3Pth ACNW meeting, December. 18-19, 1991, that deal with requests made of the NRC staff or tatters that are pertinent to NRC staff activities.

Attachment:

As stated cc: H. L. Thonpson, EDO

-J. L. Blaha, EDO _

S. J. Chilk, SECY, E. J. Jordan, AEOD R. M. Bernero,.NMSS T. E. Murley, NRR E. S. Beckjord, RES A. L. Eiss, NMSS C. Abbate, NRR W.' Brown,. OCM/IS S. Bilhorn, OCM/KR J. Kotra, OCM/JC R. R. Boyle, OCM/TR ,

W. D. Travers, NRR D. M. Crutchfield, Mr.P P. Gulnn, OCMfGV e i

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l SUMMAkY OF ACTIONS, AGREEMENTS, ASSIGNMENTS, APID REQUESTS 38TH ACNW MEPTING -

DECEMBER 18-19, 1991 During its 38th meeting, December 18-19, 1991, the Advisory Committee on Nuclear Wasi e discussed several matt.ers, completed and authorized the reports noted below.

REPORTS e Er23rjup Plan for the__Adv i sory Corgitf_pe on Nuclear Wasto (Report to Chairman Selin, dated December 23, 1991) e Geoloqic Datina of OuaternArv Volcanic Featurec a.n_p Materi_a_ls  !

(Report to Chairman Selin, dated December 24, 1991) e HIGHLIGJITS OF CIET2]XJfATTERS CONSIDERID_BY THE COMMITTEE e Syr,tems Aqulvsis Approach to Reviewina the Overall Hich-Level Waste Procram The Conmittee wac briefed by Mr. Alex Radin on the report of the Monitored Retrievable Storage Review Commission. The Committee will continue to investigate the feasibility of ,

using a systems analysis approach to review the overall high-level waste program, including the short and mid-range technical rilestones for handling high-level waste, with the goal of developing its recommendations as to the scope of the review and the advisability of undertaking it.

  • Neetina with t.be NRC Commigrsioners The Committee met with the Commissioners to discuss items of mutual interest. The principle topics of discussion were: s The reports to Commissioner Rogers on the NRC staff's performance assessment and computer modeling capabilities for HLW and LLW disposal facilities The recent Working Group r.cetira 1 geologic dating A status icport on the feasibility of 3 systemn analysis approach te reviewing the Overall High-Level Wat.te Progran.
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'38th ACFW Meeting 2 December 18-19, 1991 e Election _pf ACNW Officers The Committee reelected Dr. Dade W. Moeller and Dr. Mattir. J.

Steindler to the positions of Chairman and Vice Chairman, respectively, for calendar year 1992. ~

  • M GW Future Activities
  • Tho. Committee agreed to defer indefinitely the Working Group meeting (scheduled for January 15, 1992) to discuss

-the need for, and status of, proposed changes to 10 CFR '

Part 61.

e The Committee agreed to extend the 39th ACNW meeting to provide adequate time to discuss the Committee's long range plans. The 39th ACNW meeting vill be held January 15-17, 1992. .

e e The mac berc discussed a proposed agenda for the 44th ACNW Meeting to be tentatively held on June 24-26, 1992, in Richland, Washington. The members recommended that a ,

public meeting be held either at Pacitic Northwest Laboratories or the DOE Richland Regional Operations Office as appropriate.

Facility tours will be scheduled before and after the 44th ACNW meeting with representatives of the U.S.

Department of Energy Panford Facilitiec and the U.S.

Ecology low-level waste disposal facility. Items of -

possible interest include:

Grouting Program for LLW N-Reactor Decommissioning Performance Acrossment and Decontamination Easte Tank Stabilization and Hydrogen-Control Hydrology Modoling Capabil!. ties e Dr. Pomeroy requested a_ meeting with the neabers of MLW NRC staff to discuss the use of " expert judgnent" in perfornance assesn=nnt.

  • The conmittee agreed to def er o statun oriet inq cn the Licensing Support Systen. The ACHW staff .c t l i provide inf ormation or, the c.tu w, cf tN !4 var t t u the no nbe ' .

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38th ACNW Meeting 3 December 18419, 1991

  • The Committee asked to be kept informed on the NRC and the Environmental Protection Mency's ef forts to develop joint guidance on mixed waste :.esting and storage, e The Committee agreed to invite Mr. Michael Mattia, i

Lirector of Risk Management, Institute of Scrap Recycling Industries, to brief the Committee on practice and procedures of the recycling industry in dealing with radioactive materials found in the recycling proceas.

  • The Committee agreed to indefinitely defer further work on the impacts of the Clean Air Act on uraniu,n uill tailings and the proposed revision of 40 CFR Part 61, Subparts I, T, and W.

Appendix A summarizes the ite:as proposed f or future Ineetin'gs of the Committee and related Working Groups. This list includes items proposed by the Commissioners and NRC staff as well as ACNW nembers, n

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APPENDIX A. FUTURE SCHEDULE j 39th ACNW Committee Meeting January 15-17, 1992 Svutoms Analysis Approach to Reviewino the Overall Hiah-bgvel Waste Procram - The Committee will continue deliberations to investigate the feasibility of - a systems analysis approach to review the overall high-level waste programs, including the short and mid-range technical milestones for handling high-level waste, with the goal of developing its recommendations as to the scope of the re. view and the advisability of undertaking it.

Re. vision to NUREG-120_0 Q - The Committee w'll review and comment on a proposed revision to NUREG-1200, St?ndard Review Plan for a Low-Level. Waste Facility.

Staff Technical Position on the IdentificatioD of Fault Displagfment and Seicmic Hazards at a Geolocig Repository - The Committee will complete its review and comment on the draft Staff Technical Position on the " Identification of Fault Displacement and Seismic Hazards at a Geologic Repository."

Presertt.ption at the Low-Level Waste Forum Winter Meetina - TLe Committee will discuss a paper being prepared by the ACNW for presentation at the . Low-Leval Maste Forum Winter Meeting. The papcr will be based - on reports recently - issued by the ACNW on various low-level radioactive waste topics Working Group.Mewtings 1

Systens Analypis Approach 19 Reviewina the_Oypr.AlLtilgh-hevel Wasto frqqr_qm, February 19, 1992, 7920 Norfolk Aver,u e , Bethesda, MD

{ (Larson) -

The Working Group aill continue to discuss the feasibility af a systems analysis approach to reviewing the overall high-level waste program, including the .short-~ and cid-ra..ge technica) nilestones for handling high-level waste.

s- @ m e , - og- a wm A +2 -eg N e. sp ye 4 pa - . s Dm11n..AnL Ean26.Apr: 1 21-22, 1992, n2e bortolk Avenue, fwuanda, MD (Gnugnoli) -

h Wct Lity Cruep will cii u wu th Matcrual evidence aM the ;w.: ant 141 fer <* 1 : 3,;, t e rmangen i t. no , M;+,7< c i n

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  • 38th ACNN Meeting 5 December 18-19, 19'91 Etsidual Contamination Clean-up_CriteriO (Date to be determined) ,

7920 Norfolk Avenue, Bethesda, MD (Gnugnoli) -

The Working Group will review the guidelines for radionuclide centamin:stion limits for unrestricted use of sites and facilities that 3.re or have been under NRC license, or were at one time under AEC license.

Methods for Assessina the PrpEsngs of Natural Resources at the Proposed HoW Repositqry Site, (Date to be determined) , 7920 Norfolk Avenue, Bethesda, MD (Larson) -

The Working Group will discuss methodo.logies for the assessment of the potential for natural resources at the proposed high-level waste repository site at Yucca _

Mountain. The relationship between natural resources and the potential for human intrusion will be emphasized.

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%,. ,p February 14, 1992 MEMORANDUM FOR: Ja'mes E Taylor Executive Director for Operations

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FROM: 'Raygond F. raley v/ L Executive Director, ACNW e

SUBJECT:

39TH ACNW MEETING FOLLOW-UP ITEMS s.

Bused on discussions regarding methoun for improving implementatien and follow-up of LCNW recommendations, a summary of " Actions,

, Agreements, Assignuents, and Requests" made during each ACNW meeting is sent to your office following ecch meeting.

-Attached is a supr:ny of the " Actions, Agreements, Assignments, and L Requests" made at the 39th ACNW meeting, January 15-17, 3992, that F deal with requests made of the NRC staff or matters that are l pertinent to NRC staff activities. M

Attachment:

As stated

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cc: 11 . L. Thompson, EDO [

J. L. Blaha, EDO T S. J. Ch31k, SEOY  ?

L E. J. Jordan, AEOD 2 R. E :Bernero. NMSS g T. E. Murley, NRR  ;

E. S. Beckjord, RES -

A. L. Eiss, NMSS _

p C. Abbate, NRR s -W. Brown, GCM/J$ _

. S. Bilhorn CCM/KR  ;

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, .4 SUM %kY OF ACTIONS, AGREEMENTS, ASSIGNMENTS, AND RE: QUESTS 39TH N NW MEETING -

JANUARY 15-17, 1992 During its 39th meeting, January 15-17, 1992, the Advisory Committee on Nuclear Waste discussed several matters, and completed or authorized the report and memoranda noted below.

RE.EORI

  • NRC. . S ta f1_'Le.phrd.g.al_fo_n.i t ion o n "Tbf Identification...of_ Fau)t ,

01splacqLent_.pnd Sti_gmin_ Hazuga. at a_.fdtologin .Repositorv a.

(Report to Chairman Selin, dated 'anuary 24, 1992)

MEMORANDJ1 9 Et.gndard._Egy.ipf Plan Jot v ae Beview of q_Lic00.ClL hDpli.G#t.192

)

_for a___ Low-Lovel R.qdioactive Wants Fac ility_._(NUREG-12 OO)

(Memorandum to Richard L. Bangart, Director, Division of Low-

[ Level Waste Ma'mgement and Decommissioning, NMSS, dated 4 January 2.3, 1992)

= .Euemaries of t)m S40_tSmber t 1991_J:ERI WorlWllgp on EPA's_RLii

, Et.andgras alid of the_ December 111)_ Egg,ipt.l igr_ Risk Analysig IRRA) Annual.]Leetina (Memorandum to Commission 3r Rogers from Raymond Fraley, dated Janitary 29, 1992) e Epoumer.-t on Interj1gtif;Lrtal Perspfdgt;ives on _ low-L.e2LQ1

_ a Raditactlye Waste .DJsnosal (Memorandu.a to Conmissioner Remick from Raymorid Traley, dated January 28, 1992) c JilGHLIGETE py__CJEIAIN METTEP.S CONSIDERER.FX THE COElllTTEl' t 1. .ita_DdnTJ a Revigv Plan &gy thq_Feview of_.3.l: wngg ApylicatlRD LQL.D_. Low-IBVe l R4d_igaqtil.v3_Waotg _Q1&nos.111_ Facility MUEEQ-12.99.1 The Co.?xittee r-aviWed and commented on a proposed reviolon to NUREG-1200, Standard Revi.ew i>lan f cr a Low-1.evel Radioactive a haste Disposal Facility.

2. Fo?2DmW. MslyfiiA..ApprpstLip_RnyinviD1 %Lt& mali 111:3h:Lnci Emiv_Instn Tne h ai++ae cont;nad di w unlehn e# tbo (c a i td i t ty cr oppO inq a g e nn oip .a rt. to the 4ralynto of the mo r a l l 4 f,
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39th ACNW Meeting 2 January 1" 17, 1992

} 3 Sigff Techniqal P,p.ai,t ion on, the Identification of Fau]1 Dingla_qPJient add _ Se IsmiS. Ha eards at.A_Geoloaic Egoository The Committee completed its review and comments on the draft Staff Technical Position on the "Idantification of Fault Displacement and Seismic liar.hrds at a Geologic Repository."

4. Report o,n.Jeetina with the Director of the Division m of Low-LOXel_ Waste mad 34eM9Jit and Dogommidsionino Dr. Moeller repot' ed en a meeting he had wf.th Mr. Richard Bangart, Director, LLWM, and Mr. Paul Lohaus on-December 20,.

1991. The ateeting participants discussed the proposed revisions to 10 CFR Part 61 Licensing Requirements for Land Disposal of Radioactive Waste, low-level waste performance L assessment, residual contamination limits, and several other items of mutual interest.

5. B_e p p_ tt ' o n MeetIDg with the Chlaf. Geosciences and Systems RST19tmance Brangh. MWikt

> Dr. .Pomeroy reported on his meeting with Ms. . Margaret Federline, Chief, Geosciences and Systems Performance Branch, on January 14, 1992. Topics discussed included the use of expert judgment in high-level waste performance analysis, und

' the IIRC's capabilities in performance ansessment and computer modeling.

6. DOE Study Pla rl The Committee discussed the status of the DOE Study Plans currently under review by the NRC staff. The Committee discussed the possibility of conducting an indepth Committee analysis of the DOE Study Plan for Charneterization of Yucca Mountain Regional Furface-Water Kunoff and Stream 1lov. The Conmittee concluded that it would select a different study plar in the futare for an inderth Ccamittee analysis.
7. MFw rutyttAqilyitha

. The Comittee agreed to inrmrinitely derer thu ,sorking croup necting on the residual rndloactive clean-up criteriu l t% 2 b f or imre s tricted 054 cf conta n t nu ed siten thu a rer ' tar ha n t4mn um3er ERC 11ccum . ine staff a4de r n ic t N;d et O% ne working en the desveiep e t of u n r;r M 4 1 1 o . the Q u ia?ec wx11 $vs;t t h e. reautig of

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39th ACNW Meeting 3 January 15-17, 1992 Washington. The ACNW staff Will finalize the meeting dates and agenda with representatives of Pacific Northwest Laboratories and the Richland Regional DOE operations office.

Appendix A summarizes the items proposed for future meetings of the Committee and related Working Groups. This lism includes items proposed by the Commissioners and NRC staff as well as ACNW naambers .

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39th ACNW Meeting 4 January 15-17, 1992 b

APPENDIX A. FUTURS SCHEDULE 40th ACNW Committee Meeting February 20 21, 1992 (Tentative Schedule)

Systems Analysis Anoroach to Reviewina L.tp_QygA]l.lli.gh-LQyc),_.Wasig Proaram (Open) -

The Cor.mittee will continue to consider the feasibility . of using a systems analysis approach to review the short and mid-range technical milestones for handling apent nuclear power plant fuel, with the goal of developing reconmendations as to the scope of the review and the advisability of undertaking it.

The Committee will discuss the results of a February 15, 1992 working group meeting oi, this topic.

U.S. Env_ironmental_ Proi;_ection_Acency'sjligh . level waste standar.Ap.

(4 0 _ CFR .Pe_r.t_2 331 (7 pen) .

The Committec will be briefed by representatives of EPA on norking draft /4 of 40 CFR Part 191.

Reoort on the EP_RI Follow-or __ Meeting (Open) - 'The Committee vill hear-a report on the Electric Power Research Institute meeting, helf February 4-6, 1992, on the U.S. Environmental Protection Agency's high-level waste standards (40 CFR Parc 191).

ILe.l. ort Qn, the Low-Level _W3Lsle Forum Wint_er Meetina (Open) -

The Committee w;11 hear & report on the Low-Level Wante Forum Winter Me.eting held in San Diego, California, on January 29-31, 1992.

RG2G 1_gn the Meg.1dna with Dr. _ David McLrrison (Open) .

Dr.

Moeller will report on his necting with Dr. David Morrison, Chairman, Nuclear Safety Research Review Committen.

Commit.119_kitivit192 (0 pen /Cleased) -

The Committee will discuss anticipated and proposed Commit ten activitics, futuro sceting agenda, and organizational uatters, as appropriate. The members ,

will . also discus.s- natters and specific iceues that vere not completed during provicus neatings, working Group Mirtings x

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36th ACNW Meeting 5 January 15-17, 1992 Th2_ Impact qf._kqng-Tera _ClimatfLaa.ngs_in the Area _of_the_.S.2RthcIn Dalin _a_ns Banac , May 26-27, 1992, 7920 Norf olk Avenue, Dechesda, MD (Gnugnoli) -

The Working Group will discuss the historical evidence and the potential for climate changes in the 30uthern Basin and Range and their associated inpacts on performance for the proposed high-level radioactive waste repository at 'lucca Heuntain.

tipMLq<1s for Assessina thy Preseng; of Nal; ural __J1gg.gurges a t_,,t.11q Pronosed HLW Reqqp_it.o v Site, July 29, 1992, 7920 Norfolk Avenue, Bethesdr, MD (Larson) -

The Working Group will discuss n.ethodologies for the ascessment of the pc tenti al for natural resources at the propased high-level waste repository site at Yucca Mountain. The relationship between natural tesources and the potential for human intrusion will be emphasized.

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- .. Commonwealth Edlaon 1400 Opun Place 1 Downers Grove, Ulinoir v615 May 7, 1992 Or. Thomas E. Murley, Director Of fice of Nuclear Reactor Regulatio' O.S. Nuclear Regulatory Commission Hashington, D.C. 20555 Attn: Document Control nesk __

Subject:

LaSalle County Station Units 1 and 2 In-Service Inspection Program Submittal of Relief Request RI-24 NRC_Dnckel_NmJi0-313MILM-374

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References:

(a) M. Richter (CECO) letter to T. Murley (NRC), dated Octouer 3,1991; Structur:11 Margin Evaluation for Reactor .

Pressure Vetsel Head Studs.

(b) M. Richter (CECO) letter to T. Murley (NRC), dated December 26, 1991; Relief Request RI-21 for Unit 2 Reactor /

Vessel Head Closure Studs.

.9 Dr. Murley:

Commonwealth Edison (CECO) is pursuing an enhanced inspection program for the reactor vessel head closure ctuds at its Bolling Hater Reactors.

This inspection program will allow CECO to make informed decisions on long-term inspection and potential replacement strategies for the studs.

To support implementation of tl.e enhanced inspection prog: am, code relief is reque'ted with respect to ASME Section XI sampi, expansion recutrements based on ihn results of the maancti particle inspectiont per formed on +ne removed s tudi . The attached relief reauest. RI-24, presents CECW 1 premcsed alternate sanp;p erpansion and examinatten methodology The Ptt#thed 7011tf teut:eit is arplic ele tc t~eth Unit' I and 2 atej t' it <eaMstsd *hrt N rel!+f ei+*nd iniz:n;p +N re W n;cr of the first if ynv !rt s tion inter u $ eith will tw r w ; etic M ?er v4:e m t's stitb ttfer3ig -ydop

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Dr. Thomas E. Murley, Director May 7, 1992 r Page 2 i m Please contact this office should further information be requirei 4

Respectfully, f -

JoAn[ti. Shields "

Nuclear Licensing Administrator

Attachment:

Relief Regtiest RI-24 for LaSalie County Station cc: A. Bert Davis, Regional Administrator-RIII B.L. Siegel, NRR Project Manager-LaSal!a O. Hills, Senior Resident Inspector-LaSalle R.A. Hermann, NRR Technical Staff

'J.A. Davis, NRR Tcchnical Staff K.D. Hard, Region III

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' RELIEF. REQUEST NO. RI-24 FOR LABALLE COUNTY STATION UNITS 1 & 2

ro_HPorcer . IprNTIFIClTION Code Class: 1

References:

Table IWB-2500-1 Paragraph IWB-2430 4

Examination Category: B-G-1 L

Item Number: B6.20 (In Place)

B6.30 .(When Removed)

Description:

Reactor Vessel Closure Stud Examination P.equirements CODE REOUIREMENT l

LaSalle County Station is committed to the 1980 Edition, Winter 1980 Addenda of ASME Sertion XI.

Table IWB-2500-1 requires a volumetric examination of keactor Vessel closure Studs if left in plece, or a surface and volumetric examination of Reactor Vessel Closure Studs wh o removed from the flange. Removal is not a requirement at any time.

IWB-2430 requires that additione.1 examinations be -

performed during the current outage if examinations performed in accordance with Table IWB-2500-1 ravcal indications-exceeding the acceptance -standards of Table

-IWB-3410-1. If indications exceeding the acceptance -

standards of Table IWD-3410-1 are found as a-result of the additional examinations, IWB-2430 requiren >

examinations'to be further extended in the current outa:se to include "the remaining number of similar

- compantats within the. name examination category. . . . "

3ASJ8 POR REIJZ Commonvaalth Edison Ccapany (CECO) discovered strosa cerrosion cracting (scc) in two r4 actor vennel closure studs at Druden Unit in lata 1968. Crco is currently a ulyzing the titud e terini nicrogttNeture and wetanicai perspert ica. CICO La aie.n pare dng a pec4etivs-pararau o f enh4xad stu4 ingwet tea v?d ra 45r4*d th@ % F f4 14 ct4 + f?T & C7 $*CT$4M Y87; AMd

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Pressure Vessel Head Stud Cracking," March 26, 1992.

The_ CECO progran is also intended to include some of the additional recommendations of Regulatory Guide 1.65.

GE RICSIL 055 recommends that enhanced end shot UT be performed on "at least five RPV head studs either during the next refueling outage or at the next available opportunity." However, fer the remainder of the first 10 year ISI-Inspection Interval which encompasses the next two schedu)sd refueling outages (fifth and sixth) for LaSalle County Station Units 1 &

2, CECO plTns to perform enhanced end shot UT of all RPV closure head studs (68 in Unit 1, 76 in Unit 2).

The enhanced er.d shot UT technique developed by CECO utilizes a 3/4" to 1" diameter transducer with a frequency of 3.5 MHz or 5 MHz; the sensitivity of the examination is maximized by setting the background noise level at about 5% full screen height. This technique reliably detects a 0.3" deep saw cut notch from the top end of a reactor vessel stud. Any indications found with the enhanced end shot UT technique will be sized with bore probe UT. The bore probe UT technique developed by CECO reliably detects a 0.1" deep saw cut notch.

At each refueling outage CECO also plans to remove, if practicable, approximately 1/6 of the total number of studs (12 in Unit 1, 13 in Unit 2) from the flange of the LaSalle Reactor Pressure vessel for a wet -

fluore.iant MT. Studs which are normally removed each outage to allow for the installation of the_ Cattle Chute will be excluded from the sample because they are not exposed to the aqueous environment likely to cause 1 pitting. Cracking is believed to occur when pitted I studs are tensioned while still exposed to water at the end cf a refueling outage.

There are several reasons fo removing a na=ple of studs during the remaining two refueling outages in the first 10 year ISI Inspection Interval for surface exanination:

To provice data on ancipient st ud c; racking.

To 411cio f or addit ional uta llurg ica !

evaluation of cractint w: As n aa r. nrd

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cracked studs are found.

This information is necessary to make informed decisions on long-term inspection / replacement strategies.

Code structural margins will be assured through the enhanced end snot UT of all studs, and bore prabe UT sizing of all cracked studs. Enhanced end shot and bore probe UT results will be evaluated in accordance with " Fracture Mechanics Based Structural Margin Evaluation for Commonwealth Edison BWR Reactor Vessel Head Studs," GE Nuclear Energy Report GE-NE-523 0991, DRF 137-0010, September 1991 (submitted with a M.H. Richter (CECO) letter to T.E. Murley (NRC) dated October 3, 1991). The GE structural margin evaluation is based on conservative f..acture mechanics methodology and actual fracture toughness testing of material from one of the low-toughncss Dresden Unit II studs. If the end shot is found to be nonconservative, then an expanded sample with the more sensitive bore probe will be performed. This approach will assure that Code g structural margina are maintTined with out expanding the MT sample, d-Results of the enhanced end shot DT, bore probe UT, and MT will be compared in order to benchmark the minimum detection limit of-the enhanced end shot UT technique.

s-U The minimum detection limit of the enhanced end shot UT technique will.be judged against a conservative, -

bounding maximum allowable flaw size (established by the GE structural margin evaluation) which would be acceptable in all studs at the same time. If the mir.imum flaw detection limit of the enhanced end shot UT is'found to be greater than the maximum allowable flaw size, additional bore probe UT examinations will  : '

be performed.in lieu of the Section XI required MT sample expansion.

Expanding the MT sample if unacceptable surface indications are found would greatly-increase the critical path time and manrem burden during the outage.

And, as other utilities have found, it may be impossible to rcacve the desired sample of studo, withota dnxa ge , witMn the time constrains of a refMJ11TQ oMEajet it 18 Sutinted thAt CC*plete rsweval or all atude. A nusing me m str. wtuu , wo od t4tc & s4di t i cu t c ri t s ra l p lA ca p ed c+pM *

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, UT to be performed in place, and RICSIL 055 only recommends enhanced end shot UT of at least five studs.

l In accordance with Section XI, structural margin would still be assured by the enhanced end shot and bore probe UT. Yet much essential information could be gained by surface examination of a limited sample of studs. For these reasons, CECO requests relief from the MT sample expansion requirements of Section XI IWB-2430 for-the remainder of the first 10 year ISI

-Inspection Interval for both LaSalle County Station Units 1 & 2.

PROPOSED _ ALTERNATE EXAMINATION In lieu of the Code Requirement, at each refueling outage conducted in the applicable time period for L

LaSalle County Station Units 1 & 2 each LaSalle County Station stud;will be examined in place using enhanced end shot UT. Any flaws detected with the enhanced end '

shot UT will be sized using bore probe UT.

If an MT examination of a sample of studs reveals indications which are found by bore probe UT to exceed the maximum allowable flaw size, and were not detected by'the enhanced end shot UT, then sample expansion will proceed using bore probe UT in lieu of the Section XI required MT sample expansion. <

APPLICABLE TIME PERIOD -

This relief is requested for each refueling outage for LaSalle County Station Units 1 & 2, beginning with the fifth refueling outage for Unit I which is scheduled to begin September 26, 1992. It is also requented that the relief extend through the remainder of the first 10 year Inspection Interval for each Unit (1 & 2) which will be conpleted after that Unit's sixth refunling outage. The cixth refueling outage for LaSalle County Station Unit 1 is scheduled to ond in May of 1994. The sixth refueling outage for 14Salle County Station Unit 2 is scheduled to end in May of 199$. .

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