ML20148L428

Enforcement Conference Insp Rept 50-423/88-03 on 880308. Major Areas Discussed:Root Cause Analysis & Corrective Actions for Overpressure Transient That Occurred at Low Temp W/O Required Overpressure Protection Sys Being in Svc

ML20148L428
Person / Time
Site:	Millstone
Issue date:	03/14/1988
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20148L418	List: ML20148L428
References
50-423-88-03-EC, 50-423-88-3-EC, NUDOCS 8804010272
Download: ML20148L428 (46)

See also: IR 05000423/1988003

Text

{{#Wiki_filter:f t' . U.S. NUCLEAR REGULATORY COMMISSION REGION I Report No.: 50-423/88-03 Docket No.: 50-423 License No.: NPF-49 Licensee: Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06101-0270 Facility: Millstone Neclear Power Station, Unit 3 Meeting Location: NRC Region I, King of Prassia, Pennsylvania Meeting Date: March 8, 1988 $0 N, 9/d/88 Approved by: _ _ _E. C. McCabe, Chief, Reactor Projects Section IB Date Meeting Summary: The Enforcement Conference was held to discuss the findings of Inspection Report 50-423/88-03. The topics discussed at the meeting dealt with the root cau;e analysis and corrective actions for an overpressure transient that occurred at low temperature without the required overpressure protection systems being in service. Additional topics were discussed, including generic implications of the event and predicted plant response without operator action. 1 8804010272 880323 PDR ADOCK 05000423 i ' O DCD

t i . DETAILS 1. Participants Northeast Nuclear Energy Company (NNECo) W. Romberg, Vice President, Nuclear Operations S. Scace, Station Superintendent C. Clement, Millstone 3 Superintendent G. Van Noordennen, Supervisor, Generation Facilities Licensing J. Ely, Supervisor, Component Engineering J. Harris, Acting Operations Supervisor R. Joshi, Licensing U.S. Nuclear Regulatory Commission W. Kane, Director, Division of Reactor Projects L. Bettenhausen, Chief, Projects Branch No. 1 E. McCabe, Chief, Reactor Projects Section IB W. Raymond, Senior Resident Inspector S. Barber,' Resident Inspector V. Pritchett, Reactor Engineer, Reactor Projects Section IB R. Ferguson, Millstone 3 Project Manager, NRR J. Strosnider, Chief, Materials and Processes Section J. Gutierrez, Regional Counsel D. Holody, Enforcement Coordinator . 2. Purpose The Enforcement Conference was initiated by NRC Region I to discuss the Janu- ary 19, 1988 overpressure event without the required overpressure protection systems operable. Northeast Nuclear Energy Company was requested to identify and describe the event's root cause; broader implications; generic concerns; lessons learned; and, both their short and long term corrective actions. To ensure that the licensee was aware of the specific NRC concerns, Attachment 1 to this report hsd previously been provided to the licensee during the February 18, 1988 exit meeting for Special Inspection 50-423/88-03, 3. Discussion NRC management made a few brief opening remarks. NNECo then made a presenta- tion based on the outline included as Attachment 2. NNECo agreed that the events and technical issues described in the NRC January 18-29, 1988 special inspection report were accurate.

_ _. / .' ATTACHMENT 1 17 PAGE 1 . ! During the RHR isolation event on January 19, a potentially significant event existed when the operating charging pump continued to run after the established RHR letdown path was secured. Operator actions at the time were prompt and appropriate. A relief valve on the letdown line downstream of the orifices with a setpoint of 600 psig was capable of relieving RCS pressure during the RHR isolation. The engineering evaluation discussed above demonstrates that adequate margins remained to the reactor vessel limits. Although the RHR isolation and RCS pressure increase event on January 19 is a significant event that highlighted weaknesses in operating procedures, operator and technician knowledge, drawings, and controls, the actual safety impact of the event was minimal. One item that warrants further licensee and NRC staff consideration because of the potential significance of the isolation event is the question of "what would the outcome of the transient have been if the operators had not acted to mitigate the pressure increase." This question was presented to the lic- ensee for consideration in his evaluation of the event. 8.0 Summary of Findings Listed below are some issues to be discussed by the NRC and the licensee at an enforcement conference. . ! Evaluation of what the January 19th pressure transient outcome would have -- been assuming no operator action. - t Adequacy of drawings and of training personnel in their use, especially -- in interrelationships between drawings from dif ferent suppliers (e.g. , architect-engineer and NSSS vendor drawings). Assuring that work affecting SSPS and other safety-related equipment is -- performed by personnel having appropriate training, qualifications, and , experience. < Adequacy of controls allowing work on turbine bypass control circuits -- ' 1 without tags for pulling fuses, and the duty senior control room operator . on January 19th approving work without reviewing the applicable AWO. ) Adequacy and utility of Technical Specification curves prescribing -- pressure-temperature limits. Adequacy of procedures and training on the COPS design and interface with -- the SSPS. Use of alarn response procedures in post-transient reviews. -- Appropriateness of provisions that allow up to 31 days to elapse between -- surveilling a system for operability and placing it in operation, and assuring adequacy of supporting equipment configuration in the interim and afterwards, . 1 i . - - - - -

's , ATTACHMENT 1 . 18. PAGE 2 . Adequacy of the surveillance used to determine COP 3 operability. -- Lack of positive indication of COPS arming. l -- Development of a specific low temperature overpressure protection proce- { -- dure, or providing additional guidance in existing procedures to assure l that operators are properly aware of low pressure overprotection fea- tures' status and that they follow-up quickly, appropriately, and fully to transients and to losses of overpressure protection. Assumption by the operators that COPS was the available overpressure -- system when, in fact, it was not operable. This resulted in the viola- tion of TS 3.4.9.3 which requires at least one overpressure system to be available at all times when less than 350 degress F (VIO 88-03-02). The failure to notify the NRC about the overpressure transient via the -- Emergency Notification System (ENS) within four hours (VIO 88-03-03). - - - - - - - - - - - - - - - - - - - - - - - - - -

_ _ _ . _ _ _ . _ __ f s' ATTACHMENT 2 r , Docket No. 50-423 ! NRC ENFORCEhENT CONFERENCE . Millstone Unit 3 Cold Overpressure Protection System Falls to Operate During a Pressure Transient (Incident Date: January 19,1988) Presented By: Northeast Nuclear Energy Company March 8,1988 4 1M i 9 -M k ,m M J l illi ._ - -. - - < > . . - - , _ . - , _ . . . , . - _ . . _ . - . - - _ . . - - . . . . _ . . - ,_. . , . . - - _ _ _ _ - _ . _ _ _ _ _ .

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NRC ENFORCEMENT CONFERENCE EVENT SUMMARY PLANT CONDITIONS, JANUARY 16,1988 ' ARMING COPS, JANUARY 16,1988 PRESSURE TRANSIENT, JANUARY 19,1988 POST TRANSIENT REVIEW - SAFETY SIGN'FICANCE ANALYSIS - RHR ISOLATION CORRECTIVE ACTION - RHR ANALYSIS - ARMING COPS CORRECTIVE ACTION - COPS REPORTABILITY GENERIC IMPLICATIONS CONCERNS SUMMARY 1 of 15 .- _

< ,. 1 l l EVENT SUMMARY . On January 19,1988 with the plant solid in cold shutdown, temperature 135 *F, pressure 350 psia, a pressure transient occurred which challenged the COLD Overpressure Protection System (COPS). COPS failed to operate when required. The pressure transient was mitigated by manual operator action. l i 1 __. 2 of 15 - . _ .,. --

- . . .. .. < , PLANT CONDITIONS, J ANU ARY 16,1988 COLD SHUT DOWN,135 *F,350 psia RCS solid - Loops B & C filled and vented Loops A & D drained and isolated for RCP locking cap repair - Reactor Coolant Pump C running - Residual Heat Removal (RHR) A & B operable Overpressure protection provided by RHR relief valves (440 psia) - 3 of 15 ._

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ARMING COPS, JANUARY 16,1988 Technical Specification surveillances reviewed - Protection set Analog Channel Operational Test (31 day) . Reactor Protection Channel Calibration (18 month) - PORV block valve open (72 Hour) Verified support systems operable by discussion with l&C tech - Armed COPS by placing the PORV switches to ARM position . Overpressure protection provided by COPS - RHR train A tagged for system outage - __ 4 of 15 i . - , . - . - - - -

< ,- PRESSURE TRANSIENT. JANUARY 19,1988 . I&C calibrating the Main Steam Dump Valves P-12 Interlock must be disabeled to stroke the steam dumps - l&C tech A researched the fuse to be pulled to disable P-12 - l&C Tech B reviewed and concured with the fuse pull i - l&C tech A discussed the fuse pull with the SCO - Fuse 61 pulled in the reactor protection system - RHR suction valve starts to close - RHR low flow alarm - Operator action to mitigate the pressure transient - RHR pump stopped RCS pressure increasing - Stopped the charging pump - Increased letdown flow - Reestablished RHR -

__ 5 of 15 __

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< . l POST TRANSIENT REVIEW The fuse pulled not only supplied power to the steam... . Deenergization of the relays closed the RHR suction valves... - Investigation showed COPS setpoints were attained but the... . The failure of COPS to operate was do to the lack of input from... . 6 of 15 . -- .y ,_. ,

_ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ' < . l c 1 , SAFETY SIGNIElCANCE COPS failure challenged overpressure limits per 10 CFR 50... - Actual Peak pressure was 526 psia. - Actual low temperature was 134 F. - COPS setpoint curves are conservative. They include... Appendix G limits not exceeded when conservatisms removed. - , 7 of 15 __

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~l . ANALYSIS - RHR ISOLATION Electrical Schematic for steam durep does not show branch loads. , - SSPS vendor drawing showing all fuse load is complex. - I&C Technician A not qualified in SSPS. - l&C Technician A did request assistance, but not formally. - I&C Technician B did not fully review the consequences of the... - l&C Technician A did not request assistance from foreman. - The procedure for setting steam dump positioners was a generic... - ROOT CAUSE Incorrect use of drawings by a technician not qualified in... - 8 of 15 .

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CORRECTIVE ACTION - RHR Memo issued to discuss proper use of drawings. - Tail board discussion of incident with all departments. - Training to incorporate lessons learned - April 15, 1988. - Policy set to ensure only qualified personnel work on SSPS. - Procedure for setting steam dump positioners - May 15, 1988. - Conduct of Maintenance procedure to inicude correct line of... - Train additional I&C personnel for SSPS qualification. - i 9 of 15

_- ' , . ANALYSIS - ARMING COPS GOP to arm COPS was not adequate for plant condition. - Survellances to arm COPS did not ensure pre requisites. - Relationship between SSPS and COPS not fully understood. - No direct indication that COPS is armed. - Support system reviews for arming COPS were inadequate. - I 10 of 15

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. CORRECTIVE ACTION - COPS Procedure for arming COPS - complete. - Annunciator response for SSPS General Warning - complete. - Review feasability of providing direct COPS armed indication ... - Operator Training will be revised to re emphasize the... . Tag SSPS test switch to SS when COPS is armed. - 11 of 15

I _ , . . REPORTABILITY Initially classified as not reportable. - Reclassified as reportable per 10 CFR 50.73(a)(2)(vii). - Should have been reportable via ENS per 10 CFR 50.72. Resident was initially informed on January 19, 1988. . Resident was kept informed during investigation. . 12 of 15 , IIII *

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. GENERIC IMPLICATIONS CY has mechanical relief valves for cold overpressure. - MP2 - COPS is a single setpoint for PORV's. - Use of drawings for pulling fuses at MP1, MP2 and CY, - Qualified personnel only to remove fuses in I&C systems with... . Review other potential procedures for fuse pulls. - . 13 of 15 .

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CONCERNS Pressure transient outcome assuming no operator action. - Adequacy of drawings and of training personnel in their use. - Work on safety related equipment performed by trained... - Adequacy of controls of work. - Adequacy of curves prescribing pressure temperature limits. - Procedures / training on the COPS design / interface with SSPS. - Surveillance interval / equipment configuration control. - Are procedures inadequate to insure COPS operability?... - Development of a specific low temperature overpressure... - Use of alarm response procedures in post transient reviews. - Lack of positive indication of COPS arming. - The failure to notify the NRC within four hours. - 14 of 15

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. SUMMARY- COPS not fully operable for 63 hours January 16 - 19. - RCS was protected against overpressure until RHR suction valve... - Max RCS pressure 526 psia. - Corrective Action extensive and prompt. - 15 of 15

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COPS armed, Mxle 4 entered (Note 1) . Nov. 2_ Nov. 13 Dec. 4 Dec. 8 Jan. 14 Jan. 16 _ _ _7 Trains of MIR Operable (Note 3) . Dec. 4 ~ Nov. 13 ' Ibc Vessel Head Benoved ~ Dec. 8 Jan. 14 _- - Code Safety Valve Removed . (ICS Vent) Notes: 1) On Nov.1,1987, mode 4 was entered at ' 1730; COPS had been armed at 1655. But, there is an SS log entry that states "SSPS restored" at 1910. '1herefore, COPS may not have been available for 3 hours ' and 40 minutes. , 2)' Time period of the events umaxining loss of Mm, SSPS, and COPS. ' 3) 3 loops isolated (after both trains of

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.- .. . < , SECTION 4 POSTlRRADIATION TESTING , 41. CAPSULE REMOVAL The first capsule (Capsule U) should be removed at the end of the first core cycle (1st refueling) as shown in Table 41. Subsequent capsules should be removed at 5,9, and 15 EFPY (Effective Full Power Years) as indicated. Each specimen capsule, removed after exposure, will be transferred to a postirradiation test facility for disassembly and testing of all the specimens. TABLE 41 SURVEILLANCE CAPSULE REMOVAL SCHEDULE _ Orientation . Capsule of Lead Removat Expected Capsule I identification Capsules *I FactorIbl Time Fluence (n/cm ) 2 U 58.5* 4.00 ist Refueling 3.6 x 10 Y 241 * 3.69 5 EFPY 1.3 x 10I'l V 61 * 3.69 9 EFPY 2.4 x 10Idl X 238.5' 4.00 15 EFPY 4.3 x 10 W 121.5* 4.00 Stand By -- Z 301.5' 4.00 Stand By -- a. Reference Irradiation Capsule Assembly Drawing, Figure 2-4. b. The factor by which the capsule fluence leads the vessels maximum inner wall fluence. c. Approximate Fluence at % wall thickness at End of Life. ' d. Approximate Fluence at vesselInner wall at End-of Life. l ' . i ' l 41 t I l .. ~ ,J _ _ _ . . , _ , _ .......m, .. .

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.* . . . , , , , , TABLE A 4 TNDT.RTNDT AND UPPER SHELF ENERGY FOR THE MILLSTONE UNIT NO 3 REACTOR PRESSURE VESSEL CORE REGION SHELL PLATES AND WELD METAL ! Upper Shelf 'll*l NDT[e][bj RTNDT Material Energy ('C) (*F) ('C) ('F) (J) (f t Ib) Intermediate Shell Plates: 89805 1 -40 -40 16 60 126 93 B9805 2 - 51 - 60 -12 10 122 90 89805 3 -40 -40 -18 0 145 107 Lower Shell Plates: B98201 -46 - 50 -12 10 104 77 89820 2 - 34 - 30 4 40 103 76 B9820 3 - 34 - 30 -7 20 108.5 80 ' a. Data obtained from Combustion Engineering, Inc. Reactor Vessel Material Certification Reports. b. Drop weight data obtained from the transverse material properties (normal to the major working direction). c. From impact data obtained from the transverse material properties (normal to the major working direction). II THDT RTNDT Material Energy . ('C) ('F) (*C) (*F) (J) (f t Ib) Intermediate and Lower ' Shell Longitudinal Weld Seams and Closing Girth -46 -50 -46 - 50 271 200 Wold Seam (Weld Wire Heat ' No. 4P6052, Linde 0091 Flux, Lot No. 0145) d. Data obtained from Combustion Engineering, Inc. Wire / Flux Weld Deposit Material Certification Test No.1332. A5 .

_. . . . -. .. .- . . i l TABLE 3 5 SUMMARY OF MILLSTONE UNIT NO. 3 REACTOR PRESSURE VESSEL IMPACT TEST RESULTS FOR ._ INTERMEDIATE SHELL PLATE B98051 AND CORE REGION WELD AND HEAT AFFECTED4 . MATERIAL .._ Upper Shelf 41 J 68 J 0.89 mm Energy (30 ft Ib) (50 ft Ib) (35 mils) Material (USE) Index Temp Index Temp Index Temp (J) (f t Ib) (*C) (* F) ('C) (' F) ('C) (* F) Plate B98051 (Longitudinal 180 133 -15 5 2 35 -1 30 Orientation) . Plate B98051 (Transverse 151 111 - 15 5 -16 60 2 35 Orientation) Weld 194 143 - 37 - 35 - 26 -15 - 29 - 20 _ Heat ' Affected 193 142 - 90 - 130 - 71 - 95 - 76 - 105 Zone > - ! 37 - - t b - .

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. ' ' " # PAGE d 0F d3 ' !40 ATHEAST UTILlilEl $ERYlCE COMPANY . ' HUCLEAR ENGINEERING AND OPERATIONS GROUP ' GENERATION MECHANICAL ENGINEERING A5ME SECTION til CLA55 2 AND 3 AND ANSI B31.1.0 PlPING ANALY115 PROJECT ASSIGNMENT: 84-00 % CALCULATION NUMBER: bi*C03 A73 6P PLANT: fnMc7 /f of fAvfrF TITL E: //f 9& TT'"N'#FW lW / 7 (vevet Rc ?? (FPr QA CATEGORY 1 REYlSON 0 gy ppk 5/n/e 3- n ff r ''{ 'y./ ' "^hg " ,

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.CONECTL CT.MtqE JhIT CURVE FOR HYDf0 STATIC AND EAX STSTING .

. - . . _ APPLICABIE FOR 22.0 IVE FULL KMER YEARS. (T e 10 ,P - . . ERBOR = 50 PSIG)~ .

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, . _ . _ . . . . . . . _ _ . . .. .. .._...1 .. . . .. t -.. . . . . . _ . - .. 3, _,. .._ . . . . .. _ . . , ,, l . _ . . -. -3600 .-

- - - - - . - - - - - - - - - - - - - - . . - . - - - - . - - - _ - -- - - - - - - -... - . -. . . . . .- . .. -4 k.2400.. - .. -..,. . ._ _. . ... .-,... . - . . - - - . . . . - - . _ . . ._. .. .__._ . . . _ . . ---2200 - -- -- - - - - -- . - - . -- ... -- -. - - - - - - - - - -- . FJ i !2 SYSTD1 HYDROSTATIC EST - - - - _.2000 PRESSURE (P=2160) e y e . . . _ .. _ .. ..... . .._.- b ~.~~ - SYSTD1 IZAKAGE 'IIST . . - _ _ -

PRESSURE (P=2000) ' ' _.1800-- --- - - - - - --- -- - --- --- - - . - - - - - - --- ---- -- - - - - - - - MDD1 MINDUI SYSTDI _ _ _ . . . _ .. . _ _ . - . _ . . . . . _ . _. D'PY HYDFO TDP IIAR TEST TDP , E ~1600~ - - - - - - - - - - c. . . - b '- ~ --~~ ~ ~ ~ ~ ~ ' ' - ' ~~ 22 245 F - - - - -235 F --- -- ._1400,, -- . . . _ _ . . _ ._. . . _ . . _ . _ _ . _ . _ _ _ . ! i .-. . I , ., ~ S?JT C0FD.TIQ1 [. 3 -1200- - - - - - . - - - - - - . - - - --- n

i. . . _ . _ _ . -- .. ._ . . _ . i . . _. . _ . . _ _ . . _ . , ' . 000 ,

. - - - - - - - . - - -

---

.I > . . . - - . - - - i - - - --.-- .- - . - . . ..-. ..- - . .e ! p - ... 800.. - -. - - _ . . _ . _ _ _ _ _ - . _ - _____ -- ~ .___ .I i . . .. -- . . . _ - _ - - - . ... i . . _ . . . . . . . _ . . _ . . . . s- F 600.. - -fDR RESIDUAL HEAT RDD/AL--- ----- - - - - , .N. ' . _ . . . . , _. .. SYSTD1 OPEPATIQi . .... .__. . . _ . . . _ . ._ p m m, . - . . - - . .. _ . _ . _ . . . . . _ . . _ . _ . _ _ _ _,,; . / ' -

- 2i_.200- ' _MINDut PPESSURE (P = 300 PSIG). TOR ' ' ' ' ! RE/CIOR CDOIA'TT PmP OPEPATIQi' --

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. - - . .. . - - ._ OCNNDCTICOT. YANKEE REACICR COOLvc SYS'IDi HEA20P _ - I

' LIMITATICES FOR 22.0 tuRIVE FitT.L POWER YEARS.

-

_ . . . i (T = -10 ' , p RIOR- = 150 psIg) - ERIOR E - { . :.s .l. . .-- . - - . .. i _ _ . . . . . .

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J h ' 1800 - k - . ,' ' OPFP77DY4 ~~~X - ~- . - - - _ - . - - - - - - - - -. - i ie h .1600. - - - - - - - _ . - - - . . - -.l . . - - - - - - . . - . - - - - - - - e ~. . ... . ...-. 4 -. C- l b _1400_ __- _ . - . . ... ._ .-. . .l PATE Og J50D/HR UP 'IO_. I 20 A';D 100 F/HR UP 'IO - , 550 ~ , . , 1200. . ~. ---...- --.. . .. . j - . -- .- - . - . _ . - _... .

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6 400 . --.---.../......--...---- .-- . . -

i ' >< I p . @ ' ' ' ./.. b 4 / . . . . . . . . . . ' !! I- 4 00- ' - - - - - - - - - - - - - - - - - - - - - - - - - =- - - - - - - .. r ai - - - - -. - --. .2--.-. ..! CRITICALITY ... . . .__: i - 400 - --- ---& T=2 ) p - . - - - - - - . - - . - ,' .- - _ . - . llc f a / .a _ . . . . sJ00 -. _ . RINDU4 PRESSUPE, A i / i MAX 1121 PESSUT .4P = 450 PSIG)JOR - (P = 300 PSIG) IVR RESIDUAL HEAT RDWAL SYSTD1 CPEPATICt1 . +: .

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. ' _ i LIMITATICNS .*tR._.22.0 ptLC.T_IV.E . FULL . POWER.._YE_A.RS. . _. - J4,:EWV + _. u .9 .,.. P ..= . ..

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. e : . 4, , i : :1 2. _ Ei > , j -: . _, _ _ _ _ . . . . . . . . . ... .. ,, ,._, ._. . . . . . . . Q_2,60C . . ' . . . . . . .. _ . _ . . _ _ . . .. .,. .. _ . . . - . _ . . . . _ _ . . . . .. . ,__240L '. . . _ _ . . . _ . . . . - . . . . . _ _ . . . . . . . - _ . . . , ___. . . _ . . . . . . .. _ 2200 ... . R _ . . _ . . . . . . . . . . _ . _ . . _ _ - _.. _ ...____. g b .. .....).... . . . . - . . .c . . . . . ._ : _ . $ . 2000. ... ... .- . . . . . . _ . _ _ . . . - . . . - . - - Lwp ~.. . _- . 0 " __1800. _ . . _ . . C001DCQ1 PATE OP 100 F/HR DN1 TO 200 F ' . . _ _ . _ . . . .. AND 307/HR TO 707. _ 1600 . . . . . _ . . _ . _ _ . .__ _ . _. . .. . . . . . . _ . . _ _ . . - . . _ _ . . _ _ _ . . . _ OPFP77D1Ci 2CUF - - - __ _ _ _ . ._. - - - - a h__l400. ._ _ _ . . . . _ . . . . . . _ _ . _ _ . . . . .. .. . .. . . . . . 1200. . . _ _ _ . . . . . . __ .. . . _ _ _ _ _ . . _ i < g _~ _ . _ . . - . . - . . . _ . . . . . . ._. . .. . _ . ..g .100a. _ .. .. .. .... _ .. - __ ._. .__ _ _ . _ _ _ ,, , b) - - . _ . . . . . _ . . . . . . ._ . _. . .+.. .. .. .. . ._ , -e t=- _ BOR _ ._ .____ . - _.._.... . _ _ . . _ . . . . _ m 4 Gl .. - . i . .

. , . . .

e, h . - h L 600.. - ... __ __ -. ... .. . - ..___ .. y .. _ . __ .. . M'XDU1 PRDSSURE (P = 350 PSIG) ... . . ._ . _ . _ . .. , n g . 9 _ ADO . ..SYSmt CPEFATICN . / ._._ ' ~*. / f . .; ,. ._ BOD.- - MDJUD1 PRFSSUE dP = 30D PSIG).JOR i

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l 10 0 11 0 12 0 13 0 I40 15 0 16 0 17 0 18 0 19 0 200 TEMPERATURE (DEG. F) ,

..

-_ _ - __ .- . l > .. . 1' EVALUATION OF JANUARY 19 TRANSIENT ASSUMING K0_DEERATOR ACTION o USE SECTION XI, APPENDIX E, EVALUATION OF UNANTICIPATED OPERATING EVENTS. 0 SECTION III, APPENDIX G CONTAINS ASSUMED BOUNDARY CONDITIONS: 1/4 T FLAW, ASPECT RATIO OF 1/6. -- "BOUNDED" HEAT-UP OR COOLDOWN RAMP RATES. -- , KIR CURVE. -- J FLUENCE THROUGH THE END OF AN OPERATING , -- ' ! PERIOD. ! O TRANSIENT CONDITIONS, EVEN DESIGN BASIS CCIDENTS ' FROM CHAPTER 15, ARE HQI CONSIDERED OR CONSISTENT i WITH THE P-T LIMIT CURVES. ,

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. ARTICLE E-1000 ' INTRODUCTION L1100 SCOPE TABLE E1 MAXIMUM ALLOWABLE PRES $URE AS A FUNCTION This Append.ts prondes acceptance criteria and 0F T, - AT.,, FOR !$0 THERMAL PRES $URE . . guidance for performing an engineering evaluation of TRANSIENTS (AT/At < 10*F/ht) the effects of an out oflimit condition on the structur. For design pressures greater than 2400 psis al integrity of the reactor sessel beltline region. Showing compliance with the criteria in either E.1200 T,- #Ts,, WadngnAApaW or E 1300 assures that the beltline region has adequate m Prniet (80 structura] integrity for the unit to return to aerdce. + 25 and prester 1.1 x Destem + 15 fa00 +10 2250 0 2000 -10 17S0 -25 1500 -S0 1200 E.1200 ACCEPTANCE CRITERIAu , , , , 3,,, ' Adequate structural integrity of the reactor vessel $ beltline region is assured if the fs!]oming applicable 200 7, 50 criterion is satisfied throughout the event: (a) For isothermal pressure transients (i e., AT/At ' I h E "* ' N O T E "' **' ** t* " ***"'d'

4 10*F/hr). the maxirnum pressure does not exceed the allowable salues of Table L1 at any value of T, - R Tur. ' (b) For pressurized thermal transients (i.e., AT/At > l'fFAr), the masimum pressure does not enesed the design pressure and T, - RTuris not less than 'ALUATION BY ANALYSIS $5'F. (a) Adequate structural lategrity of the reactor , if compliance with the above applicable criterion is sessel beltline region is assured ifit can be shoma by ' ( not shown, adequate structural integrity can be ana1 sis using the input of Table E.2 that the 3 - assured by satisfying the guidelines and criteria fo!!owing eriterion is met throughout the event: specifsed in E lM)0. ~ , 1.4 (K. + K.) + KfK. , , where

e K .,= stress intensity factor due to membrane stress i , ' K,= stress intensity factor due to thermal stress i t

K,= stress intensity factor due to residus) stress i K,= fracture toughness per A2ticle A 9000 i 'r, b & wh n w emtsat te=rentun, and Ar,/Ar b the (b) If complianc4 with the atose criterion cannot be sanmum sansme et tementun r,in mar o** bout senat eA Tor h * Wahat ad;*stad nfere== ta=Fntun (foe weld or shown, additional analyws or other actions shall be b.w matenar) et m inia, ..rf.= or the emw . : taken to assure that acceptable margins of safety *d. l < drw===: t> Reruhtory on i to Rev. 2. be maintained during subsequent operation.

358.3 .

. 2 o* .

TaMeE2 SECTION XI-DIVISION 1 1986 Edeles TABLE E2 . EVALUATION INPUT FOR PLANT AND i EVENT SPECIFIC UNEAR ELASTIC FRACTURE MECHANICS ANALY$t$ ver w e vehe Prosawee Eveat penaure time history <

Torgerstwo Event Wwerstwee tarve hietary eteet transke Eveat/p' ant specmc Aee/mishe se4tions O r t type Semi-elfiptical wfect Ase einmwm Wtletion 0.0 < a 5 1.0 h (Nete (1)) track see 4 Crock e4atsuen Long%dinal i K./K, location Swiface and run'ewm depth . Clad eMeets Cted to be considered in tlw I (Nrmal, stress, and fractwo mech.sMcs 4Wyses (hete (2)) Tra%'tu teve% ness K. per Article A 9000 Upper theT towthness (In cowrse of preparetken) Nace Flwence at the time of the trans%t $Nfttwrw Rep /atory Gwiet 1.99 Rev 2 pesidd strns Appropriate W:strtDvtion for the fabr4aUen process, se bnear

d.stnbet% mith +10tsJ et the ,

Ins 4e Serface and .30441 et the , eJth 6t w fact ' h0T[$; ' m a . the muim m cract destMn the tase meta! u

-' ' ! (2) The stresses dse to the dSereve t.eteten the kse meta' and Ela$$Ag thermat espa*Lsion Leeff.cients need ret De ten 64ered Wi the 66cthermal presve trans'ee.t ovatwation (Le , . A 7,/ A t < 107 / hr). l l l ! . l 6 e k ' i \\ i i 354 4 I I . ,

l J ! - - - - - - - - - - -

-_ --- 4 T ~ ' .. ' ', 220 i m ~ - t , ! 1s0 - 160 - 140 - K;g . & . & - x. ~ i 120 2 Y t; iOO - > , $ >-

a0 - EE 80 - lR8<

- .# ao - MTNOT JL 0 e a t t -100 -50 0 +50 +100 +150 +200 (T~ RTgof.*F FIG. A-4200-1 LOWER BOUND K,, AND Ke TEST DATA FOR SA-533 GRADE B CLASS 1, SA-508 CLASS 2, AND SA-508 CLASS 3 STEELS n ' . - - - - - - - - - - - c'-

. . _ _ . __ i w o n ( . ANALYSIS--APPENDIX E 0 o CURRENT ADJUSTED RTNDT ~ 90 F: f 0 ORIGINAL 600 AND 30 F SHIFT MEASURED IN -- SURVEILLANCE CAPSULE. I ! I CONSERVATIVE DUE To CAPSULE LEAD FACTOR ~ 4.0. -- 0 o Tc--RTNDT = 134-90 = 44 F.

1.1 x DESIGN @ o TABLE E-1, ALLOWABLE PRESSURE = Tc--RTNDT = 2. . I o TRANSIENT PRESSURE WoULD BE LIMITED To 2485 PSI BY CODE SAFETIES. . - OKAY. o PALLoW > PTRANriENT '

I ! -_. . - - - - - - . _ _ . - __ _ _ _ _ _ _ Q

, _- - . ... .. . - - - . I CONCLUSIONS O PRESSURE TRANSIENT OF JANUARY 19 HAS NO IMPACT ON MILLSTONE 3 STRUCTURAL INTEGRITY. O CHEMISTRY AND FLUENCE LIMIT RANGE OF ISOTHERMAL 0 TRANSIENTS OF CONCERN ~ 150 F OR LESS. 0 OPERATIONAL LIMITS HAVE SIGNIFICANT SAFETY MARGINS TO PRECLUDE COMPONENT DAMAGE. l ' .

l 1 . . . . . . - _ . . . _ ___ . . _ . , , - _ _ _ _ _ _, . . . . . . . . . . . _ - . }}