ML20128H391
| ML20128H391 | |
| Person / Time | |
|---|---|
| Issue date: | 11/08/1984 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20128H385 | List: |
| References | |
| FOIA-84-790 PROC-841108, NUDOCS 8505300506 | |
| Download: ML20128H391 (38) | |
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-h ANNUAL OPERATING REPORT
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.g (Mame of Plant, Name of Licensce)
O s FOR(CalcudarYetc) o j Docket No.
e License No.
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8505300506 PDR Fogg 841108 ~~
PEDR084-790 PDR y /;
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O TABLE OF CONTENTS (Should contain at least all the headings set forth in this recommended format, with page numbers. Every heading here should appear in the body of the report, even if there is nothing to report, as on page 10).
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IllTRODUCTI0ff l HIGiLIGHTS 2
SUMMARY
OF OPERATII:G EXPERIENCE 3 PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMEflTS, AND SAFETY-RELATED MAINTENAf!CE 8 Amendments to Facility License or Technical Spe,cifications 9 Facility' or Procedure Changes Requiring NRC Approval 10 Te's ts and Experiments Requiring flRC Approval 11 Other Changes , Tests , and Experiments 12
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Corrective thintenance of Safety-Related Equipment 15 LICENSEE EVEllT REPORTS 18
. -0THER EVENTS OF IllTEREST 21 DATA TABULATIONS 29 Net Electrical Power Generation
.29 Unit Shutdowns and Forced Power Reductions 30 Number of Personnel and Man / Rem Exposure by Work and Job
. Function 33 UNIQUE REPORTIflG REQUIREMENTS 34 GLOSSARY 36 u- ..
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INTRODUCTI0ft
"(Facility name) is a (type of reactor) of Maximum Dependable
. Capacity, in 10le fiet, owned by (name) and located . The nuclear steam supply system is a (name of manufacturer and type).
The architect / engineer was (name), and the constructor was (name). The condenser cooling method is , and the is the condenser cooling water source. The plant is subject to license (number), issued (date) pursuant to Docket fiumber . The date of initial reactor criticality was , and commercial generation of power began "This report was prepared by (name and telephonc number). Major perscnnel char.ges duritig tbbear N,clugid^\ M'h ihn- . Lemp;es repieced P.
b Alister Burt (retWed) r 1as General Superintendent, Nuclear Generation.
Mr. Lempges was Superintendent, James A. FitzPatrick Nuclear Power Plant.
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, HIGHLIGHTS (Executive Sumary)
(This one- or two-page section should discuss briefly the most
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significant events of the year -- power output, ,o,ter reductions due to NRC restrictions and consequent power loss, major outages.
- other mil'estones. If the event is described in more detail else-where in this report, cite the section or page nuder.) HIGHLIGiiTS
- page or pages may be printed on paper of a different color.
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.D SUNtARY OF OPERATING EXPERIENCE L(In chronological order, briefly describe forced load changes
, and reference load following. operation.where applicable. If a forcedfload. change occurs, give time of occurrence, system and major component involved, and type of shutdown.-tzfor to
- corresponding Licensee Event Report. Include in this chronology.
any other-pertinent items of interest. Requirement: Technical Specifications ~and Regulatory Guide 1.16, Revision 4, item C;1.b.'(1)).
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g- z 7 ' -Y-SU MA'RY OF OPERATING EXPERIENCE The following'is a chronological description of Plant Operations including other pertinent items of interest for the twelve month period ending December 31, 1975.
1-l' At the beginning of this report period the plant was operating at 100% power (595 MWe).
'l-4 A load reduction begun at 0400, reached 400 M'le at 0525, a satisfactory turbine stop and control valve test was conducted and load reterned to 95% power (565) MWe) at 0805.
e 1-5 Following the required 23.85% margin to limit Fz24 hour hold peri calculation was[ Mat 95% power and a increased reaching 100% (595) MWe) at 0903.
1-31 The plant was at 100% power (5 M1- )
2-1 Load was reduced beginning at , 00 and reaching 400 MWe at 1417.
A satisfactory turbine stop an ontrol valve test was conducted.
Following the valve test load w s maintained while cleaning of the-condenser tube sheets gathbeing done. At 1615 a reactor turbine trip occurred from a Gngcm generator feed-steam flow misa;atch/
low level signal caused'by a low suction pressure-feed pump trip.
Low suction pressuge to he feed pumps was caused by cavitation of the condensate pumr / from low hotwell level. The unit remained at' hot standby while iijcellaneous maintenance was perfonned.
2-2 At 0404 th reactor was brought critical and the generator was phased to t rid at 0818. Load was increased to 30% and held until steam nerator chemistry stabilized within specs.
2-3 At 0327 load was increased to 50% and held for steam generator chemistry analysis. Chemistry was within specs. and load was increased at 0545 reaching the required 802 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> hold. point at 0745.
2-4 At 0756 load was increased reaching 100% at 1010. At 1825 control l-rod bank E.was constituted from control rods 6,7,8 and 9 and the manual incore-excore axial offset program was placed into effect.
2-5' The #3 steam generator wide range level transmitter which incorporates a high level override closing of the feed reg. valve feature failed high causing the feed valve to close. Alert ,
-operator action by going to manual mode and re-establishing l feedwater flow to the steam generator averted a plant trip. The )
-transmitter was subsequently replaced and control returned to automatic.
2-13 First truck load of new fuel from B&W arrived on site.
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c 2-19 At 0635 the. plant; reached the 28. billion KW/hr. production mark.
4 2-28 Load was reduced.beginning at 0301 and reaching 400 M,We at 0500. A' satisfactory turbine stop and control valve test was conducted and at 0541 a load increase began, reaching full load 4
-(595MWe)at'.1023.
3-5. The new 84,000 gallon waste holdup tank was placed in service for
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receipt of radioactive liquid waste.53-11_"A" Boron waste storage tank was changed ovEr Tron hydrogenated liquid' waste.to dagasified liquid waste. The new high' pressure
- waste gas system was placed in operation and the old system
' isolated.
3-23 Beginning at 1635, because of a malfunction i 3he "A reheater drains tank nonnal-level control valve, load.ug reduced to 555 lNe. The problea was corrected and at 1703 faid was increased, reaching full load (5921Me) at 1845.
. 3-24 At 1318 lead was reduced to 580 MW take "A" reheater normal level control valve out of service $rrepairstoabottom
. flange l leak. Replacement of a gas was completed and load -
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increasedat1828reachingfu1[oa (593.MWe) at 1845.
3-26 Because of an excessive le: Lag' from a packing ~ gland on LD-MOV-200in-thecontaint.brt,tothevalvestemleakoffheader, a load reduction to come i.f th'e line begin at 1302. . At 1357 -
the generator was r . ved from the grid and a cooldown of the .
the primary syste comnen ed. '(Abnormal Occurrence flo. 75-1).
-4 27 Following _repackingpf LD-M0V '200 in a partial cooldown condition of the primary system, the plant was-heated up to operating condition, during the outage, including plugging _three leaking tubes in 4A.
.feedwater heater.
3-28 The reactor was brought critical at 01200,. turbine rolled at 0343 and the generator phased to the grid at 0500. Load was increased
' to 1800 MWe by 0715 and held for chemistry checks. At 1430 load was. increased reaching 555 (- 300 MWe) at 1655 and held for 80"
'.(475 MWe) at 2040 and held for chemit,try checks. At 2115 load was increased reaching 547 IGie at 2235. Load was increased gradually as dilution of the primary system progressed.
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= 3-29 Load increased slowly through the day reaching full load (593 P,le) at 1229.
'4-2 The circulating water intake de-icing system was removed from service at 1337.
p ~4-11 End-of-life for Core V occurred at 1928. At this time C B 0 ppm, all rods were fully withdrawn, output was 596 !!We (1820 MWt), Tavg was 555.8 F and main steam heaker pressure 5
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at 684.8 PSIG.' '
4-25 From 0900 to 1515 load was reduced to 580 MWe during the conduct of:incore-excore correlation check.
26 End of Full Power Life for Core V occurred at 0659 on the 26th, commencing plant coastdown.
4-27 Beginning at 0610 load was reduced, reaching 400 H.le 0 0756.
Following a successful turbine-stop and control-valve test
. load was increased commencing at 0953, with the plar.t bach at full load (587 MWe) at 1600.
4-30 At the end of the month the plant.was opera conditions with load 0 574 MWe,, Tavg 543 Jods all out,F}frthg "0" under boron, #1, #2 and #3 Turbine control valve eJT de open.
5-16. Load reduction from 538 MWe and FVlavg. commenced at 2130.
5-17 Generator separated from g i at 02 for scheduled Core V-VI refueling and maintenance c. Reactor shutdcun 01800 and cooldown of primary sy em$ started at 2105, t
5-18 Held primary system aN50 and 600 PSIG for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, maximizingsteam;negjorblowdownattheoptimumphosphate solubility poin .
5-19 Primary system cooJed down to 125 FO . Residual Heat Removal system in operation.
_5-20 Containment integrity broken and personnel ' hatch opened at 1210.
5-23 Primary system drained to refueling reference level, all four
. loops isolated, conoscals removed and bullet noses installed.
5-24 Started detensioning reactor vessel studs.
0 5-28 Removed reactor head to storage pad. Filled cavity using purification pump.
5-30 Removed drive shafts, upper internals and started fuel movement at 2240.
6-1 At the beginning of the month the plant was in a cold shutdown condition with the residual heat removal system in operation, the refueling cavity full' and fuel movement in progress The refueling water storage tank was emptied for scheduled maintenance.
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7-1 The generator was phased to the gid at 0538 ending the Core V-VI Refueling and Maintenance Shutdown. (45 d., 4 hrs.,
36 m. including 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 12 mins. on line heat soaking for overspeed trip setpoint check and adjustment to 1850 RPM) and load increased to 150 MWe by 0948 and held for steam generator chemistry stabilization on AVT.
I 7-3 At 0800 load was increased reaching 462 MWe at 1220 and subsequently to 100% power at 2010.
7-5 At 0528 a broken oil pressure line on the left hand turbine trip valve necessitated an emergency load redttetion with the generator separated from the grid at 0553. Re' pairs were completed and the generator phased to the grpt 1116.
7-6 Plant reached 100% power at 0850. /
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7-14 An oil leak developed in the turbine a[ tost oil system piping and a load reduction begunat q340! The generator was separated _from the grid at 0 ^/. 7/filo', ting repairs the unit was phased at 0913, load in ased to 25% and held for secondary chemistry stabilj io 0212 and 100% power (572 MWe) 7-15Loadincreasecommenc$edat' achieved at 0735. /
7-25' A tube leak dqElopw in q L "A" Condenser and was satisnctorily controlled by ~.. addition of " Wizard" compound to the intake.
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8-1 Plant operatin at 100% power (566 MWe).
8-10 At 0330 load was reduced to 400 MWe, a satisfactory turbine stop and control valve test conducted and ten leaking tubes plugged in "A" waterbox. Load increased started at 0910 and the plant was at full power (581 MWe) by 1230.
9-1 The plant was operating at full load (575 MWe).
9-7 Comencing at 0500 load was reduced to 400 Mde, a satisfactory turbine stop and control valve test performed and the plant returned to full load (580 MWe) at 0913.
9-14 Because of an oil-soaked insulation fire on gland steca lines under the governor end of the high pressure turbine a load reduction commenced at 0852 and was terminated at 0933 with a load of 481 MWe on the unit. The plant was restored to full load at 1424. No damage was caused from the fire.
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PLANT OR PROCUDURE CHANGES, TESTS. . EXPERIMENTS, AND SAFETY-RELATED MAINTENANCE Amendments to Facility Licens~e or Technical Specifications Facility or Procedure Changes Requiring NRC Approval
' Tests
- and Experiments ' Requiring NRC Approval
' Other Changes, Tests and Experiments ---
-(Requirement: _ Technical Specifications and Part 10 of the Code of Federal. Regulations 50.59 (b))
Corrective Maintenance of Safety-Related Equict.:ent (Do not include any associated with outages or power reductions that are discussed elsewhere in this report).
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' Amendments ' to Facility License or Technical Specifications On' January 28, 1975 the Nuclear Regulatory Comission issued Amendment
. #1 to-the Connecticut Yankee Facility. License No. DPR-61. Paragraph 2.c.(2) was amended to as follows:
"(2) TECHNICAL SPECIFICATIONS
, The Technical Specifications Contained in Appendices A and
= B, as revised, are-hereby incorporated in the license. The
, licensee shall operate the facility in accordance with the Technical Specifications, as revi'rerby issued changes-thereto through Change No. 1."
The amendment incorporated Change #1 to the. Technical Specifications into the license. Change #1 to the Technical Specifications consisted of changes to Section 3.10 Control Group Insertion Limig and Section 3.18 Power Distribution Monitorino anc Control. This cifanges the power distribution monitoring. requirements of the Technical Spec'ficai.'ipns to utilize manual axial offset monitoring.- '
On May 20, 1975 the Nuclear Regulatory C ission iss.ued Amendment #2 to the Connecticut Yankee Facility LAcen e No. DPR-61. Paragraph 2.c.(2) .
was amended to read as follows: "\
'(2) TECHNICAL SPE CATIONS The Te calf Specifications contained in Appendices A and B,
, asrevisef,arehereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications, as revised by issued changes thereto through change No. 2."
The amendment incorporated change #2 to the Technical Specifications into the license. Change #2 consis,ted of changing Section 4.4 Containment Testing, Sub section I Integrated Leakage Rate Test to comply with the provisions of 10 CFR, Part 50. Appendix J requirements with respect to TYPE A tests.
On June 20, 1975 the Nuclear Regulatory Commission issued Amendunt #3 to the Connecticut Yankce Facility License No. DPR-61. Paragraph 2.C.(2) was amendea to read as follows: ,
"(2) TECHNICAL SPECIFICATIONS u The Technical Specifications contained in Appendices A and B, L as revised, are hereby incorporated in the license. The
- license shall operate the facility in accordance with the Technical Specifications, as revised by issued changes thereto through Change No. 3."
The amendment incorporated Change #3 to the Technical Specifications.
Change #3 provided for Cycle VI new thennal hydraulic safety limit curves i: 9
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Facility or Procedure Changes Requiring NRC Approval There were no facility or procedure chahges during the report year
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that required. prior approval by the fluclear Regulatory Commission.
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_ Tests and' Experiments Requiring-NRC Approval
'During this reporting period, Florida Power and Light requested and received approval from the Nuclear Regulatory Commission to participate
-in a Power Ramp Demonstration Test for Westi This test Lwas conducted as outlined in" Westinghouse T&pgttouse PWR's.
gcalReportsWCAP-8529 (proprietary) .and WCAP-8531 (non-propribt3S). A summary report of the test conduct and results is enclos onkt'he fo] lowing pages.
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Other Changes, Tests and Experiments Plant Design Change #94
' Plant Design Change #94 entitled " Installation of Totalizing Meter in the Demineralized Water Storage Tank (DUST) Makeup Line" was completed during the month. This change concerns the installation of a direct reading totalizing flow meter in the makeup line leading to the DWST from the water treatment demineralizers.
This change provides a more accurate method for determining the secondary plant makeup requirements.
The safety evaluation for this change considered it not to be an unreviewed safety item because the possibility of an accident or malfunction of a different type than any evaluated previously has not been created, and the probability of an ace,ident to equipment previously evaluated will not be increased. '
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Plant Design Change #111 'j Plant Design Change #111 entitl "
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ge in Closing Scheme for Service Water Motor and Air Ope edalvesDuringaPartialLoss of Station Power Supply" has beQ completed. This change involved, (1) ;nstallation of two SG-29rehys across Buses 4, 5, 6 ard 7, (2) installation of one contacFitom each SG-2 relays on each DC bus to type HEA trip and lockoht relays, (3) installation of one contact from each HEA relay in%e ' closing circuits of'MOV 1, 2, 3 4 and A0V 8 and 9. ,
o In the original "w/* of AC Power / Core Cooling" scheb.e only one service water pump was plac in service due' to emergency power supply limitations.
Because of this limitation all usage of service water except that necessary to support core cooling requirements were automatically isolated on a loss of. AC power signal by closing M0V.#1 and #2 (turbine valves) and ~A0V #8 and #9 (service water return valves from the boron recovery system and
. spent fuel pit cooler). Following the installction of redundant full size emergency power generators and additional core cooling equiprent, the core cooling systems were reparated into two independant " trains" cach service water pump in each train starts autcoatically on a ' core cooling signal and there is sufficient paaer from each emerger.cy generator to add an additional service water pump if necessary, however, the service water isolation valves closing scheme was not changed and these valves close on a " partial loss of AC power". Closure of these valves on a partial loss of power was unnecessary and undesirable, however, the requircnent for closing on a total loss of AC was retained to provide the necessary service water supply should one pump be out of service for any reason and the redundant emergency generator-fail to start.
The safety evaluation for this change considered that on a partial loss of AC power the emergency diesel generator would automatically start up one service water pump for core cooling requirements. In addition the E.
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failed bus would be activated by the operator through coninon ties with the good bus, therfore, there is no reason to . isolate the service water on a partial loss of AC power since we'have all-pu.nps available. Changing the: closing scheme from a " partial loss" to a " total loss" of AC will
-increase. The operating reliability of the system and does not result in the degradation of the design or of its safety function. This does i not constitute an unreviewed safety item because the availability of two -!
service water pumps fullfills the requirements of the station'and the q probability of an accident previously evaluated will not be increased. '
Plant Design Change #116 Plant. Design . Change #116 entitled " Radioactive WaN Disposal System Modifications" has been completed. This chanj;e upgraded thc entire radio-active waste disposal system and associatedAupport systems. Associated design changes have been or will be reporteh3eparately. These changes were required to comply with the release lists established by the proposed Appendix I of 10 CFR 50 and 'nafl implementation of the "as low as practicable" philosophy wit rqadrds to radioactive releases.
The safety evaluation for this ch ge determined, through analysis following Safety Guide #24 guidel es, that the new design would result in a lower whole body dose f'\he tite boundary than that previously determined and therefore do(s n'ot constitute an unreviewed safety item.
b The Commission was advipI f these changes on December 5,1972 pursuant to Section 50.59 ('t of f10) CFR 50.
b Plant Design Change !26 Plant Design Change 26 entitled " Change to the Main Feed Pump Suction Trip Scheme" was completed during the month.
This change consisted of changing th'e feed pump low suction pressure trip circuitry logic from a 1/1 trip to a 2/3 trip. In the original scheme the loss of fuse in the trip circuit could cause a plant trip without valid cause. Changing to a 2/3 trip logic avoids unnecessary plant trips from instrumentation malfunctions while still providing the necessary protection for valid low suction pressure conditions.
The safety evaluation considered the proposed change did not ccnsititute an unreviewed safety qucstion and would improve plant reliability.
Plant Design Chance #136 Plant Design Change #136 entitled " Ion Exchange Building Gantry Crane" has been completed. This change involved the installation of a " gantry crane" with rails running North and South, on top of the Ion Exchange Building and Boron Waste Storage Tank dike wall. This was necessary to
. provide crane service for changing radioactive waste system filters located in an extension of the ion exchange building associated with the upgrading of radioactive waste facilities and associated support systems (PDC #116).
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. . . r The. safety evaluation for this change . concluded that the probability of occurrence and the consequences of an accident or equipment malfunction would not be increased by this design change. The possibility for a different type of accident other than that previously evaluated did not exist and that the margin of safety would not be reduced. .It also concluded that the change represented sound' engineering judgement.
Plant Design Change #144 Plant Design Change #144 entitled " Condensate SodiuTAnalyser" has been completed. This change involved the installation of an in-line sodium analyser capable of receiving a continuous sas(le stream from the condensate system, analysing it for sodium c6nyent, indicating results locally and in the control room. This changU p'ovided continuous monitoring capability for the detennination f condenser inleakage.
The safety evaluation concluded this char' I
did not constitute an unreviewed safety item.
Plant Design Change #145 Plant Design Change #145 entitled ".Cond&1 sate Inline pH Monitoring" has been completed. This change inv jcM the installation of an inline pH monitor capable of receiving a c@ntinuous o sarple stream from the condensate system, analysing it for pH, indicyting results locally and in the control room. This change provided the corftinuous monitoring capability for detennination of, and changeshist condensate system pH.
The safety evaluation conclud .. his change did not constitute an unreviewed safety item.
Plant Design Change #14 Plant Design Change #146 en tied "Feedwater System Inline pH, Oxygen and Conductivity lionitoring"has been completed. This change involved the installation of a pH electrode, conductivity cell and a dissolved oxygen analyser capable of receiving a continuous sample stre:m from the feedwater system (alternately from the condensate system), analysing it for pH, conductivity and dissolved oxygen content and indicating results locally and in the control room. This change provided the continuous monitoring capability for detennination of, and changes in, feedwater system chemistry parameters. .
The safety evaluation concluded this change would improve the feedwater system chemistry parameters and did not constitute an unreviewed safety item.
Plant Design Change #148 Plant Design Change #148 entitled " Containment Personnel Hatch Hydraulic Motor Control Center #2 to MCC #5.
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j Corrective Maintenance of Safety-Related Equipment l
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Mechanical and Electrical Maintenance .
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NARIRE OF LER OR OUTAGE MArruNCTION EQUIPMEt:T MAIt:TEnAt:cE NUMBER CAUSE RESULT CORRECTIVE ACTION Corrective Rust & dirt Valve leaked Cleaned up valve & lapped.
Radiation Monitoring MR-14576 buildup through seats Valva 3200B Control Rod Drive Fans Preventive NA NA Replaced bearings IW3A & B MR-15107 Corrective Poor design Soft seat "O". Remachinuiseat groove to Reactor Makeup Water MR-12808 ring would fall better retain "O" ring Valve 529 out of groove in seat Reactor Coolant Pump Corrective g Fau , /./
Ofl leak from M per reservoir Replaced upper reservoir gaskets IPlB MR-15103 g, 4 gas
- , y Safety Injection Valve Corrective J ' Dirt buildup Valve leaked Lapped seats 853C MR-13422 through NA liA Inspected internals for CVCS Letdown Valve MOV- Preventive '
RR-12523 damage 285 CVCS Charging Pump IP2A Corrective Worn packing Leakage around keplaced plunger packing MR-14517 plungers Main steam Step Valves Preventive NA !!A Inspected seats & repacked MR-13914 Electrical Dreakers Preventive NA NA ' Inspected & multiamp tested all 480V breakers NA Inspected & repaired as Contair.mont Ventilation Preventive NA' Panu IWlA, B, C , D '
necessary Preventive NA NA Inspected internals & meggar Pressuri::er tested heaters
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. Instrumentation and Control Equipment Maintenance' I
NATURE CP LER OR OUT' AGE MALFUNCTION EQUIPMENT MAINTENANCE NUMBER CAUSE e
RESULT CORRI4hVE ACTION Contair. ment Personnel Corrective Deteriorated Air leak around Airlock t1R-13332 ' Replaced epoxy & rebuilgs-electrical penetra tion penetration penetration. during test CVCS Charging Pump 2P2C Corrective Worn packing Leakage around' Replaced pump packing MR-15034 plungers-CVCS Charging Pump 2P2A Correctiva Worn packing Leakage around Replaced pump packing MR-14 316 plungers CVQS Charging Pump 2P2A s
, Corrective Improper Cror%U$ad Replaced crosshead shims f MR-15040 ljis i tr nent CVCS Charging Pump 2P2B Corrective MR-15048
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Nothal [: car Worn belts Replaced belts.
Component Cooling Corrcetive- Normal wear Leaky seal Replaced mechanical seal 2P11B MR-14388 j 4.
CVCS Charging Pump 2P2A Corrective Worn packing Leak around t
Replaced packing on. plungers MR-14563 plungers CVCS Doric Acid Blender Correct ive Failed Bypasa valve 2B5-356 MR-14579 Leak through Installed new diaphragm diaphragm valve CVCS Charging Pump 2P2A Corrective Worn packing MR-14340 Leak around Replaced packing on plungers plungers Centaittment Preventive NA NA Performed annual surface inspection Ratctor Trip & Bypass Preventive NA NA Inspected Dreakers mm - -
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y -(b LICENSEE EVENT-REPORTS Present LERS in chronological order with a detailed narrative-
-describing eventi its consequences (including offsite releases),
. associated power reductions and outages, and corrective actions.
Cite all applicable LER,' outage, and. plant chang $ umbers, but-amplify. reports previously submitted, if possible. Requirement:
' Technical Specifications and Regulatory Guide 1.16, Revision 4,
-- itemC.l.b(2)'(a-e).
Item C.l.b. '(2) (b) requires reporting of LERS pertdining only to outages and pover reductions. A yearly tabulation of all LERs would be more informative but i:, not mand:.'. cry.
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. LICENSEE EVENT REPORTS a,
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- During the reporting ~ period of a' total of thirty events classified as
' Licensee: Event 1 Reports,took place with. reference to Unit.No. 2.
These are' summarized as follows:
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751) Valve" Stem Leakoff Causes Containment Radiation Monitor Alarm -
March 26, 1975 ,
_ . Valves in the reactor coolant system, two:(2) inches or greater in
- size, have a . dual' valve -stem. packing arrangement with a valve stem
-leakoff: system connected to the valve'benigen the- two sets of.
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- packing,;to' direct lcakage from the innjcmacking gland to a . drain
' ; collection tank via a cooler. At-0530cpflarch 26, 1975, a.high Jtemperature alarm was rec'eived at he outlet of the valve: stem leakoff cooler. . Tota 1' reactor coo nf system leakage at 0800 was Lealculated to:be 0.3 gpm. ~ /.f inve igation to find the cause of the alarm.was initiated. 'At 1
main _ control: room for. the vole l}00lan e control tank levelalarm was received in th
" Rate-of-Change High" followed shortly an alarm from the containment.
- radiation monitoring sys . . increase in the, containment- '
y : sump level was -observed at 1302an ' orderly reactor _ plant shut-the leak rate to the reactor contain-
~down mentt forwas. commenced. ,At} ted 4This; 15 minutes,indu:a gpm.exceeds the allowable ,
technical speci .dc?afon limits for reactor operation. A containment >
s entry was mad been pressuriz fand
{ revealed that the valve stem leak-off line ha '
, to;the point of causing the valve. stem leakoff systeo Jrelief valve to ift and relieve.the pressure to the containment sump. ~ The plant and reactor was shutdown, the syste'n cooled down _
_ and depressurized to 3850 F and 550 psig. : LD-M0V-200 w'as . repacked
, ' while in its _back seat position. The plant was heated up and returned!
to-power. ,
The: apparent cause of the'above occurrence was determined to be a failure of the inner set of packing on' the letdo'wn isolation-
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ivalve LD-MOV-200 allowing-a reactor' coolant system leakage'in' excess of_ that' allowed by the Station Technical Specifications for
_ reactor operation.
, m . All reactor ~ coolant leakage was' directed either to the normal collection 1 stank or to the reactor containment. Both of which are designed to contain reactor coolant system leakage and since no radioactivity was: released to the environment, we conclude that the public health
}~ .and safety was not compromised as a result of this occurrence.
Both~ the; inner and outer sets of packing on LD-MOV-200 were replaced with' n ,
new packing. A containment inspection was conducted to check the.other 4.
'" system valves for excessive leakoff and misadjustment. LD-M0V-200' ioperability was checked -satisfactory.
h 19 4
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E 5a ,
n.
75-2 Failure of Steam Generator Seismic RestraintC_ May 21, 1975 On may 21,1975 during preparatory work associa.ted'with the installation of a new steam-generator support and seismic restraint system, one of the original system long bolts
-(#55) on the #3 steam generator. was found broken. A
,long bolt alarm whould have activated on the bolt position monitoring. system, however, no alarm had been-received.
However, a "short bolt" alarm for #7 microswitch position
.was indicated but the associated short bolt was checked in the normal position. The monitoring system was designed to alert the operators to a-possible1 bolt problem and was intended to _ serve temporarily until a new stermTg'enerator
< holddown system could be designed and igstalled. Prior to the discovery of the broken bolt no " bolt alarm" had been received, nor was one present at time of discovery.
.When the possibility of a broken lpng t became known, progress towards removal of the~boit osition monitoring
. system was temporarily suspende a an investigation initiated to detemine why the , arm system had failed to identify the condition.
A check of the electrical ci aiasy installed disclosed t. hat the microswitch giving the "short bolt" al on the broken long bolt and vice versa. Furtit;r checking revealed'that'the identification n'u-bers for 17 (short bolt) and !S (long bolt) microswitches were rev vsed. The satisfaciery electrical continuity checks p fopAd during installation of the monitoring system therefore d " not find the improperly marked lccation numbers.
A complete investigation of the holdd6wn support systea was conducted
.and revealed two additional broken "long" bolts on #4 steam generator (bolts #51 and #59). !!o alarm had been received to indicate that these bolts were broken and an examination of the alam system indicated that the system was functioning normally. In fact all alarms were functionally tested and '
performed properly. The two affected microswitches were removed for shop testing to insure that the two switches were not damaged by the bolts " breaking". . Shop tests as weil as functional functional tests inplace indicated that no m11 function of the microswitches existed; consequently it appears that upon breaking, the two bolts did not " rise" sufficiently to actuate the alams.
h the original installation it was assumed that upon breaking the bolts would " pop up" a significant amount, based on observation of the original twenty-four broken bolts. The alam system
~
microswitches were set to actuate with only sever thousandths of Additionally; an inch movement but were not designed for great accuracy.
.it is likely that the settings drifted during the seventeen months service.- At any rate, tests indicated that slightly more movement would have actuated the alams. Plant records and normal reports will document any further findings of the investigation, however, since 'this system is being removed from service it does not appear to warrant further reporting at this time.
20 f
, , (t 3 n
OTHER EVENTS OF INTEREST
~
'(In chronological order, present a narrative description of nuclear and.non-nuclear events not discussed elsewhere in this report that had a major impact on the plant. Include for exbmple, the-discovery and ' implications of failed fuel; the results of-eddy.
- current testing, ultrasonic tests, visual exa.ninations; loss of condenser capability;. failed turbine assemblies; etc. Requirement:
. Technical Specifications and Regulatory Guide 1.16, Revision 4,' item C.l.b.(4)).
t Also, present narrative. descriptions of each single relco.se of L
radioactivity or-single radiation exposure specifically associated with an outage that accounts for.rnre than 10 percent of the ,
allowable values. Requirement: Technical Specifications and
- Regulatory Guide.l.16, Revision 4, Item C.1.b (2)(f)).
1 Reference to previously subnitted reports are acceptable and are considered to fulfill annual report requirement's. Inclusion of a summary report and results would be desirable as presented on Page 11.
L 21 k
, -1 Fuel Perfomance-Off-gas. activity through the January 1 throygh June 30, 1975 report ficant fuel failiares.
Off-gas limits. release' rates were49cless than 1% of((/ perio Periodic comparisons of actual cont 1 d -i.nventoa.in-core to control rod inventories predicted by nor.'ra7.(zed computer progrars indicate reactivity anomolics less he6?'ll .3.
' Reactor power level wa>1NI.ted to 50% of rated power from April 26 through June 30,1975%/juse of concern that LPFJ4 vibration at high core flow rates had dam' ed or wnuld damage fuel channels.
No refueling or fuel excmination was perfomed.during this report period.
b 22
_. t 3
f
.+ ..^ , ,
. . I;;7 >
c Fuel Perfomance Report, End-of-Cycle 3, Dresden Unit Three The Unit Three reactor became critical to begin its third fuel cycle at 0630 hours0.00729 days <br />0.175 hours <br />0.00104 weeks <br />2.39715e-4 months <br /> on June 5,1974, following an 87-day refueling outage.
The generator was loaded at 1243 on June 6,1974. A month-by-month summary of hours critical, number of scrams from critical, number of shutdowns, and average operating offgas release for Cycle 3 is given. in Figure 1.
,- ~Majo~r events during Cycle 3 include the #filliing:
t'
- l. 24-74 Excessive feedwater pipigq/ vibrations (8 days out).
- 2. 9-20-74 Leak in feedwate pi@ g (7 $ys out).
- 3. 10-31-7.4 Fuel damage 11r ing control rod movements (50%
.derating f emac.nder of cycle).
4 .~ 4-14-75 Beginn.ing E0C-3 void reactivity tests by. G.E.
Ik
. Cycle 3 was ended atJ030 0 April 16,1975, with a manual scram of the unit for the EOC-3 shr;n reactivity test by G.E., thus beginning the third refueling outage.) The bundle with the highest exposure at E0C-3 tias DO 055, located at/39-40, with an exposure of 13750 t'/dD/T.
Flel Performance Data The following outline addresses topics in which the Nuclear Regulatory Commissicn has expressed specific interest.
I. General.
A. Fuel vendor: General Electric Company.
B. Fuel'1oading data:
- 1. Fuel assembly type numbers:
- a. Initial core, DD.
- b. First reload, Generic B, GEB.
- c. Second reload, DDB.
, ore loading map, Cycle 3: See Figure 3.
P 23 o -
I
J.f4 'p J' . . - , . . . . . . . - . . . ,
% N 3
Unit 1 St' eam Generator Tube Inspection -
A program of eddy current examination and tubesheet cleaning was
- conducted on both Unit 1 steam generators during the period of 7 i April 27 to May 2,1975. The eddy current testing was conducted j in accordance with the Prairie Island Technical Specifications. *
! The tubesheet cleaning was performed as a preventive measure to -
assure continuous steam generator tube reliability and integrity.
The eddy current inspection program was planned to provide examination coverage complimentary to the baselinmamination I
- of those regions where evidence of tube wall deterio'raticn had 3 been detected in other plants and to determine if there had s j been any significant change in tubes previouslpexamined in the ;
baseline.
j 7 Examinations for defects were performed examinationsofsteamgeneratorNo.114lat40 HZ with supplemental )
t P,,5 KHZ to obtain information _
on the height of sludge deposits surfburidihg the tubes. The same probe and equipment was used for the 400 KHZ Idd 25 KHZ examinations with the :
l oscillograph recorder speed at m, hec and 25 mm/sec, respectively. 3 All data was recorded on mag 2Jh tajie and. oscillograph strip charts =
and a data sheet was filled o$t uring each shift. All tubes examined at g 400 KHZ were examined a nd tbc U-bend to the top support on the -
opposite sides, while thh 5 KHZ examination was only perforced up i
. to the first tube 3 Titorf. ]'
The total number of
\
e legs inspected at 400 KHZ per steam generator -
was 246. They are cdegorized as follows:
i
- No. 11 Steam Gen No. 12 Steam Gen Inlet Outlet Inlet Outlet !i Required Sample (3%) 136 68 136 68 Examined for 1st time 138 71 136 75 Baseline Tubes w/o 16 0 30 0 g EC indications Baseline Tubes with 12 9 0 5' _
EC indict.tions . ]
The extra number'of tubes examined over the minimum required was d done in order to provide a more symetrical pattern for tube inspection. -
i The evaluation of the eddy current results disclosed no tube degradation ;
or significant change in indications during the period of operation since September, 1974. Of the new tubes inspected, there were 13 tubes with :
permeability variations and four tubes with dents. One tube, R30-C21, in =
steam generator No.12 outlet contained a small 0D indication ;
d 2
i 24 I
7 ; _. .
. . . co V*' :'* .
Eddy' Current Test Results e
. Site: 'Prairic Island Unit 1 Steam Generator: No. 11 Inict Test Frequency: 400 Kitz Date: 4-30-75 COLT 41N TYPE OF INDICATION LOCATION
-ROW l
- 5 30 Dent 12" above tubesheet 51 Dent 30" above No. 5 support 5
5 78 Dent T above No. 2 support
-*6 41 Dent 10" and 20" above tubcsheet f-~
'*7 .
39 Permeability variation 30" above tubesheet
'51 Permeability variation Betvecn No. 3 and No. 4 support-8 Pcmeability Variation Tuber.heet to No. 5 support
- 10 33 38 Permeability. Variation Betucen No. I and No. 3 support
.
- 10 .
- 12 .42 Dent 6" above tubesheet
- 13 .33 Dent 6" above tubcsheet .
-63 Dent 48" above No. 5 support 13 Permeability Variation Between No. 2 and No. 3 suppcet
~
17 78-
- 19_ 46 -Permeability va5" tion 30". above tubesacet
- 19 '47 Permeability V.itjation 36" above tubcsheet
- 20 37 Dent .[, 6" above tubcsheet q '
- 20 43 Dent / 6" above tubesheet 23 75 permenbht#
hy/a*tdition
/ 30" above No. 4 support
- 29 44 Dent g:'" 4" above No. 7 support bI fy eR J p
- Reexamination of baselinc examined tubes with E.C. indications.
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^ Summary 'of Containment Penetration Tests
- . ' -The:following containment renetrations were-tested during the
' report period.
(#/24 Hrs.-)f NUMBER. TITLE LEAKRATE.(id 4,0 PSIG)'
! P-4 Pressurizer Relief Tank Vent 0.01 I:
!' P-11 Liquid Sampl.e Lines- '1.45
- P-15, 16, I 17, ' 18 Steam Generator- Blowdown Lines 10.99 l P-22 . Space Heating Condensate Return Line 0.00
- ,I 'P-80 Auxiliary Containment Spray from' Fire 0.52 System
.P-81 Auxiliary Steam Generator Feedwater Sup;,1y 1.06 P-B Electrical Penetrations 17.13 P-71 Primary Vent Header 0.00 P-23-C Dead Weight Tester 0.06 P-23-D- Air I:onitor P.g Je . 0.01 P-A Perso w/
gliHa* i 14.16 P-13 i.up Discharge 0.40 P-33 Ppf cling Cavity Purification _1.45 .
P-62 Service Air to Containment 0.03 P-66 Drain Cooler - Component Cooling Supply 6.54 P Drain Cooler - Component Cooling Return 14.31 P-70 Instrument Air Supply 0.01 P-12A Valve Stem Leakoff 0.32 P-12B Neutron Shield Tank Sampler O.03
. P_- 14/ P Vapor Seal Head Tank / Pressurizer Relief 0.05 Tank Drain
.P-23A Containment Open Bulb System 0.54 P-23B Containment Closed Bulb System 0.03 P-28: CCW to RCP Oil Coolers 10.32 P-29 CCW From RCP Oil Coolers 0.65 P-30 Space Heating Steam Supply 0.01
.)
. ij a e'
( (#/24 !!rs.)
NUMBER TITLE LEAKRATE (0 40 PSIG)
P-34 CCW From RCP Thermal Barrier 0.42 j P-38 CCW to RCP Thermal Barrier 0.42 Containment Purge Air Exhaust P-39 3.66
-l.
P-40 ContainmentPurgekirSupply 0.00
- i. .P-61 CCW From Neutron Shield Tank -0.00 l
0.0
~
.PD Dome Vent F1ange .
-l
- - P10 Reactor Coolant Letdown 0.55 P41 Loop Drain l{cader
/
eg 2.02-3.
P64 Air Monitor Sa .. ' > @ rom Containment 10.57 P60 CCW t p pIik Tank Cooler 1.49 PC~ Equigp.%.'-ctch
. 0.0 P50 Fuel Transfer Tube 0.0
. .P3- Safety Injection 0.0
.P69 Loop Fill 0.0 P68 Primary Water to Containment 2.00 PB Electrical Penetrations 29.09 P24 Safety injection Recirc. Lines 0.19 PA Personnel Hatch 0.20 Running Total through December 1975 = 164.94 # Air / day 0 40 PSIG 28 s
- y. .-
w
. ii.T . c 4.: ,
Izf DATA TABULATIONS Net Electrical Power Cencration Unit Shutdowns and Forced. Power Reductions
'(Tabulate'information submitted to Gray Book in form specified
- by. Appendix D of Regulatory Guide 1.16, Revision 4. Indicate--
plant status.during the outage, proximate cause, and system or major- component involved if equipment nalfunction was responsible;, report each single release of radioactivity associated'with the outage that. accounts for more than 10 percent of the allewabic instantaneous values'; refer to applicable Licensee Event Report;. report previous and reduced peecr-levels;
' describe maintennnce perforncd during outage and ccrrective actions taken to prevent .uurence.)
Number of Personnel and Man /Ren Exposure.by Work and Job Function (Requirement: Technical Specification and Regulatory Guide 1.16, Revision 4, item C.1.b (4)J )
Plant Organization and Personnel
~(Optional Chart) 29
'$ 6
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3/26/75 F 24.0 A 1&3 l 75-01 CH Valve X Repacked gland of letdown system stop valve. During manual shutdown ,
# spurio'us high start L ;
up rate signal tripped reactor NI 15.0 F Miscellaneous Maintenance Items:-
Plug 3 leak.ing tubds in feedwater heater.
5/17/75 S 1076.4 C 1 1 -
RC Fucil XX Core V-VI Refueli ,,
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and Maintenance
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C-Refueling 3-Automatic Scram.
. 0-Resulatory Restriction 4-Other (Explain)
E- Operator Training and License Examination 3. Columns to be completed by NRC.
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UtlIQUE REPORTING REQUIREMENTS (In:this chapter, report information required by the Technical Specifications that has not been covered elsewhere in this report --
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GLOSSARY
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Technical terms used in the Annual Operating Report-should be consonant with the definitions provided in Regulatory Guide.l.16 and the monthly Operating Units Status Report (the Gray Book).
C Terms, words or phrases not previously defined by the flRC that are used_in the Annual Operating Report should be explained in this-Glossary. In particular, llRC revieviers require definitions or -
explanations of abbreviations and nomenclature not commonly used throughout the industry.
Suggestions for additions to the glossaries in Regulatory Guide 1.16 and the Gray Book or proposals for improved definitions may be addressed to the appropriate units of the fluelear Regulatory Ccamission.
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.. ~ , - STATUS REPORT ON
[
ANTICIPATED TRANSIENTS WITHOUT SCRAM i
FOR GENERAL ELECTRIC REACTORS 1 I
i M I DECEMBER 9, 1975 4
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TABLE OF CONTENTS i
Page ;
1.0 Introduction.................................................... 1-1 ;.
1.1 Genera 1.................................................... 1-1 1.2 Scope of Review............................................. 1-1 1.3 Outstanding Issues.......................................... 1-1 [..
2.0 Anticipated Transients.......................................... 2-1 I
< -. c4 3.0 Analysis Assumptions............................................ 3-1 - -
3.1 Pi tn t Co n d i t io ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -1 3.2 Opern ting Equipment & Sys t ems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 4.0 Analy t ical Me thod s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1 Organization of Codes...................................... 4-1 4.2 Analytical Methods of Plant Transients..................... 4-1 l r
1 5.0 Staff Independent Calculations.................................. 5-1
_.p,
( ;') 5.1 Staff Mode 1................................................ 5-1 5.2 Dif ferences Between Staf f and Vendor Mode 1. . . . . . . . . . . . . . . . . 5-1 5.3 Computational Results...................................... 5-2 5.4 Summary and Conclusions.................................... 5-3 !
6.0 Conformance to WASH-1270........................................ 6-1 6.1 Limiting Transients........................................ 6-1 l *l 6.2 Reactor Coolant Boundary Pressure.......................... 6-4 j
=== 6.3 Fuel Condition............................................. 6-8 f 6.4 Con tainment Cond ition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 '
6.5 Radiological Consequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-13
, 6.6 Post-ATWS Shutdown......................................... 6-14 6.7 Reactors Protection System Susceptibility to Common Mode Failures.............................................. 6-17 7.0 References...................................................... 7-1 [h),
Appendix A Chronology of Review................................. A-1 ~
Appendix B Staff Independent Calculation........................ B-1 ,'.
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1.0 INTRODUCTION
1.1 General In September 1973, the Atomic Energy Commission published the " Technical Report on Anticipated Transients Without Scram for Water-Cooled Power L Reactors" (WASH-1270)1 establishing acceptance criteria for anticipated r V - ..
transients without scram (ATWS). These criteria were developed because * ,f
- of the staff belief that a fully satisfactory methodology for analyzing the reliability of protection systems from the standpoint of common i mode failures was not available at that time, that these types of failures had occurred in protection systems and that the potential t
consequences of some postulated anticipated transients without scram 1
might be hazardous to the public. Subsequent to the publication of c .i -
WASH-1270, the staff met with the General Electric Company on a regular j.
i basis and reviewed their evaluation model, the results of analyses of [
i anticipated transients without scram, the diversity of the systems [.
relied upon to mitigate the consequences of ATWS, and the susceptibility
. of the Reactor Protection System to common mode failure. A chronology f-of the review is presented in Appendix A. ( ,
f l The General Electric Company in conformance with the requirements of ~
l
-,.s..
1 I Section II-B of Appendix A to WASH-1270 submitted analyses of anticipated PfC .
1 4 ^'
transients without scram, NEDO-20626 and NEDO-10349 , and an evaluation -
, of the susceptibility of the Reactor Protection System to a common 3
mode failure, NED0-10189 . l
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1.2 Scope of Review 'kV
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The review of the GE ATWS analyses included the anticipated tran- .
sients expected to occur, the initial conditions and system parameters .
~ assumed in the analyses, the reliability of systems, the analytical techniques, the results of analyses of ATWS events and the design of the Reactor Protection System. Using the requirements of WASH-1270 as a guideline, the staff reviewed each relevant aspect of the GE .
model and analysis. The discussion of anticipated transients is presented in Section 2, and the initial conditions, system parameters and operating systems assumed in the analyses of these transients are discussed in Section'3. .The analytical techniques and computer programs are reviewed in Section 4. The independent calculation conducted by the staff is described in Section 5 and Appendix B. The
- 7 comparison of the GE analyses to the requirements of WASH-1270 is .\s ; _
presented in Section 6.
1.3 Outstanding Issues We have identified certain outstanding issues in our review which require that GE or the applicant provide additional information or t
modify their present design. These items are summarized below and are discussed further in the sections of the report as indicated.
- 1. GE must provide analysis of additional anticipated transients
- (Section 2).
1
- 2. Analysis assuming both the llPCI and RCIC fail to operate must be f
provided (Section 3.2.2.).
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- 3. Analysis assuming two safety / relief valves fail to reclose must be provided (Section 3.2. 2) .
f-GE must demonstrate that the feedvater pump trip system, the 4.
recirculation pump trip system, and the standby liquid control system have acceptably low unreliabilities (Section 3.2.2). 4
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- 5. Diversity must be provided in the circuitry used to trip recir-culation and feed pumps, initiate standby liquid control system, t
and start HPCI and RCIC (Section 3.2.3).
i
- 6. Diverse logic design must be provided for automatic MSIV closure, i
( ' r actuation of relief / safety valves at the relief set point and .
containment isolation (Section 3.2.3). I l
- 7. Additional information required on the AWS evaluation model ,
- (Section 4.2).
- t o
- 8. GE must demonstrate that piping must not be stressed beyond the - #
I emergency stress intensity limit (Section 6.2). '
1
..).5ll-l 1
1 .H RHR valves must be demonstrated (Section 6.2).
. l 9
/ 1-3 _' -
G ^
l
- &~yy g.manman . ,
, m- ., m: ? .;;~e.z ~ g;.: 3;~.
. ;.e u - :;;;;g y; cey;; 91.e ,, . a ., ,- , . , m .w- r- . ; s. ;9y~. < n 7 ,y
>~
. g? ,
9f.r3 8
3'
,, v{<,,
i :td 10. The structural integrity of the pool must be demonstrated (Section 6.2 and 6.4).
.. c7 T40?
l,$7&
....y 4E 11. An estimate of the number of failed fuel rods must be provided
- l. *
/' .
(Section 6.3).
l
- 12. Assumed boron mixing characteristics must be shown to be con-
", - servative (Section 6.6). .-.
+
- 13. Analysis showing long-term effects on the vessel, containment and the doses must be provided (Section 6.4, 6.5, and 6.6).
~
l 14. Diverse means of interrupting power to the scram solenoids must j be provided (Section 6.7).
.i 4
I 1
i
)
J 4 1-4
,, .g . _ _ . . . _
- _ _ . _ _ . . ..
2'.0 ANTICIPATED TRANSIENTS During the operation of a water-cooled reactor power plant deviations '
from normal operating conditions are expected. These " anticipated k transients" might ec' cur one or more times during the service life of ,,
{
a plant. They are thus distinguished from " accidents," which hava a l much lower likelihood of occurrence. There are a number of anticipated I a..
}:
transients, some of quite trivial nature and others that are more -
[
I significant in terms of the demands imposed on plant equipment. j Anticipated transients include such events as a loss of electrical load that leads to closing the turbine stop valves, a load increase such as the opening of a condenser bypass valve, or a loss of feedwater ;
i.
-g, - flow, i t- l qy The following is a list of events analyzed by CE in the NEDO-10349 2
and NED0-20626 reports.
- 1. Primary coolant flow decrease. These transients include failure of recirculation pumps and a recirculation flow controller malfunction causing decrease in core flow and could result from ;
t-
- faults in the power supply. I
- 2. Reactor Water Temperature Decrease. These transients include i
malfunction of the feedwater control in a direction to increase feedwater flow, loss of a feedwater heater, shutdown cooling o malfunction and inadvertent activation of auxiliary cold water systems.
t .
A.
2-1
>;g@g N t xpg ;
kl. I $
m-- - - m - ,, - m,~
<. . ,, ii my ge.mpex ., , ,.e w.gsu u , g:gg w .. . .. ,: g , . . . , q.y ; , .d n
. )
{.(6)
., [ ' 3. Reactor Coolant Flow Increase. These transients include a I malfunction of the recirculation flow controller in a manner to
';;fj' cause increasing primary coolant flow and the startup of a
~~A:
/'*/ recirculation pump.
ry
- 4. Reactor Water Inventory Decrease. These transients include loss of feedwater, pressure regulator failure, and opening of .
~
condenser bypass valves. _
- 5. Primary Pressure Increase. These transients include loss of load (generator trip, turbine trip and loss of condenser vacuum),
closure of main steam line isolation valves and a malfunction of the pressure regulator.
The staff believes these events have a frequency of occurrence of si, _
one or more times in the lifetime of a plant and are appropriate for the ATWS analyses. However, we believe there are other transients I }
j that may occur one or more times in the lifetime of a plant and we l require that GE analyze them as ATWS events unless GE can demonstrate l their probability of occurrence is low enough that they need not be
]
! included in the ATWS evaluation. The transients for which ATWS i
l analyses are required are:
5
- 1. Rod withdrawal from zero power and full power. CE must provide analyses of in- and out-of-sequence rod withdrawal at zero power and at full power.
- 2. Loss of normal onsite and offsite power i
)
k l 2-2 i
,, g . . . _ . . . . . - . _ _ _ _ _ _ _ _ , . _ . . . _ . _ _ - . . .
j c . ,
1 i
f'~ ,
- 3. Stuck open safety / relief valve. 1 1
1 l
- 4. 100?F step loss in feedwater heating. .
1 e
I-1 l'
i % . ,; ;
. - e 4
r" "
. ^t N. '
I l4 i
WW
(- .
f p
4
. ),
- r 4
e
,.',s 73 , - 2-3 -
u-
<x? h .
(%V,$" A p$.4e 0
-s
. v;,
F# }.4
< s,lP. . g . 3S
.r-
... uw . < + ~a r .
- .. . m vc . er puw.:, m l3v.p2i
' ~ '
- : . . c.,; .. .
. .. ..e...,..
e e s .- - - - _ __ . _ . _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _
3.0 PLANT CONDITIONS AND OPERATING SYSTEMS 3.1 Plant Conditions The NRC staff aim was to obtain an overall conservative assessment of I
the consequences of anticipated transients in which scram did not ,
s c
occur and yet provide a realistic evaluation of the cours; of the e i
i event. Therefore, the staff asked that GE use the normally anticipated or expected values of the initial conditions and system parameters ' ,;,
such as power, level, flow rate, and pressure. The GE analyses ,
i followed the staff guidelines on the assumed plant parameters, except in some cases, such as initial power level, GE assumed conservative values. l D~ The initial conditions and the equipment characteristics used in the 2 '
analysis are provided in Tables 3-1 and 3-2 of NEDO-20626 . These initial conditions and equipment characteristics are representative of t
i BWR-4, BWR-5 and BWR-6 reactors. Therefore, each applicant referencing '
2 NEDC-20626 must provide a comparison of their plant parameters with Tables 3-1 and 3-2 and demonstrate the applicability of the NEDO- t. ,. O 20626 report to their plant. i ,
~
N.
4 3.2 Operating Equipment and Systems The purpose of ATWS evaluations is to determine the expected course of action, given that a common mode failure prevents a scram action when
- 4, 9 -
demanded. All other components and systems may be assumed to operate normally provided all of the following criteria are met.
~
ej . 4 b
3-1 i
,n,, 9-a n k
n i
- , .3 3 4.su;;.;n g. p .~m, . . .;; , ygm,;, gu,y p . g.
.. , ; : :,w w ~ n.y. y
.= .
b h j w w; L.
~ ' j}' 1) Failure of the component or system did not initiate the transient -
g>. c being analyzed.
~
- 2) Operation of equipment shall be assumed if it makes the ATWS
- ;f event more severe and would be actuated as a consequence of the l+ $4
- ,ys
- :
"'^$ course of the event.
- 3) The function of the component or system is not disabled .
~' ' ! -
as a consequence of the transient being analyzed.
l
.j
- 4) The probability of an anticipated transient resulting in unacceptable consequences due to failure of the reactivity shutdown system and the equipment relied upon to mitigate
< consequences of ATWS would be of the order of 10-7
. . . Per
- =
reactor year. (s}
-w-
- 5) The initiating signal and equipment assumed to operate following an ATWS is diverse from the reactor reactivity 3
shutdown system.
I The GE ATWS analyses assume operation of the following systems.
I a) Recirculation Pump Trip I
r j b) Feedwater Pump Trip i
+
c) Standby Liquid Control System d) Relief / Safety Valves e) High Pressure Coolant Injection System (BWR/4) or High Pressure Core Spray Systems (BWR/5 and BWR/6) f) Reactor Core Isolation Cooling System j
i 3-2 t
.n .- ,~ - . .
-~
g) Main Steamline Isolation Valves h) Containment Isolation
- 1) Residual Heat Removal System j) Service Water System k) Component Cooling System !
t 1
The GE assumptions are consistent with the staff criteria 1 and I 2 above, however the remaining criteria (3,4 and 5) are not satisfied.
~
I 3.2.1 Functions Disabled Under ATWS Conditions i GE does not consider that an ATWS event would affect the functioning and operability of any component or system. The staff believes ;
that under the high pressure resulting from an ATWS, the functioning
- 1
/~ '
of the RHR system isolation va.'6ves could ,be imIaired. t The staff !
u therefore requires that the operability of these valves after ,
?
being subjected to the peak ATWS pressures and temperatures be !
demonstrated by analysis or test.
3.2.2 Reliability of Systems In order to meet the safety objective of WASH-1270, the combined t
unreliability of the systems required to mitigate the consequences _
of ATWS events must be sufficiently low so that their failure ', l ,
need not be considered. In addition to cor.aon mode failures ,
which are discussed in the following section, these systems must 25 have a combined unreliability due to random lailures of approximately '
~3
. 10 per demand.
3-3 7,y 3,s ..
- Jl:
kd, N fig
. - y , . . . . . , y . m ,- y;y - m t - - -w.--cy ++
. . . - - .,- Scoww :3v..cc .v,;;r ;m,; m.yy.4p::q. , ;- 7 .:, .~ , , . mg.:.y y. . , m-
<4 4
2;sX~, ~
n
.. The long teon cooling systems (items i, j, and k above) which are (;' .
.- y.
- required to function during every shutdown are assumed to have high .
I enough reliability for normal and post-LOCA operation and therefore -
3 yg?:,t are presumed to have sufficient reliability for ATWS events.
.;- ~.-
x ;.
'~
The staff has reviewed the main steamline isolation valves (item g) and the containment isolation system (item h) and concluded that these system are sufficiently reliable and their failure need not be con-sidered in the analysis of ATWS events. However, based on the staff review the remaining systems may not have sufficient reliability.
Three of these systems, the Reactor Core Isolation Systems, RCIC
, (item f), the High Pressure Coolant Injection System, HPCI, or the High Pressure Core Spray System, HPCS, in later plants (item e) f w.
~' and the pilot operated safety / relief valves (item d) do not have y4.. -
a sufficiently low unreliability. Therefore GE is required to demonstrate that the WASH-1270 limits are not exceeded even if a) one pilot operated safety / relief valve fails to open at the 1
relief set point or b) the RCIC and HPCI both fail to operate or i
! c) two pilot operated safety / relief valves fail to reclose. In inter plants the failure of RCIC or HPCS instead of the failure of
{
the RCIC and the HPCI must be considered. The three remaining systems, f
I the Standby Liquid Control System, SLCS, (item c), the feedwater pump trip (item b) and the recirculation pump trip (item a) are 3-4 J'
' ~
additions or modifications to current designs and have not been described in sufficient detail to enable the staff to judge their reliability. However, GE is required to provide detailed design descriptions of these systems which demonstrate that their unrelia-1 bility is sufficiently low. .
I 3.2.3 Diversity of Systems Required to Function During An ATWS i, '. ,
i GE takes credit for the automatic trip of the recirculation pumps I 'N and the motor driven feedwater pumps, and for the automatic
(.
initiation of the liquid control system, the RCIC System, and the
- I r
HPCI (or HPCS) System. GE, in their response to staff questions
[.
(NEDO-20626, Section 7.2), have described a conceptual design of l f
1 the proposed system for automatic initiation of these functions. '
, The proposed system would provide equipment diversity in the
.; .~,
signal processing and in the actuating devices which would preclude failure mode common with the reactor protection system. However, [ ,
the same signal sensing equipment as in the reactor protection system would be used, since it is claimed that functional diversity n ensures that no common mode failure of any one type of sensing y..
equipment can cause failure of both scram and bacicup shutdown , $'
4 systems actuation. The staff does not have sufficient information '
to completely agree with this contention. We require that a I
- w detailed preliminary design and/or additional analyses be submitted . , .
.dy i which demonstrates that the actuatifon of these systems is diverse
,[l,
- y. .
'l from the reactor protection system. '
- .'j'[
'y j,
3-5 "O' I l
. d'.-
t 6 ,w '
u
, m,,,
n~ .
,u , - a,c.,wua;g.>r.- - , :.,, - 2 3 . ., m y : ; ;: ,g ,; g , .
, , :v w +.> : r . n + - -
m - .
?; :&, f
. , f" ^.;
"ik The proposed conceptual design would initiate the liquid control \.
system on high reactor vessel pressure or low water level, but only after a shorc time delay (on the order of 5 seconds), and -
, 'd1J only if a high neutron flux signal then existed. The purpose of
'Vy 06' the time delay is to cause the flux reading to be taken after a l normal scram would have reduced the flux to essentially zero l 1
, l (thus eliminating the need for operation of the liquid control
., j system).
1 There are a number of anticipated transients (Table 2 of NEDO-i s 10349) for which a scram is initiated only by a high flux signal.
- Therefore, with the proposed design, a common mode failure of flux sensing equipment could result in both failure to scram and 6
failure to initiate the liquid control system. Thus, we will t require that GE either provide analyses to demonstrate that an i
i effective scram will be initiated for all postulated transients 1
j by other than the high flux sensors, or submit a detailed design t which incorporates the required sensor diversity from the reactor 4
4 protection system.
1 i
',, -j The proposed initiation system would also trip the recirculation pumps on either high reacto~ vessel pressure or low water level.
I i This logic design utilizes the functional diversity of the existing reactor protection sensing equipment. A common mode failure of either the pressure or level sensors would not prevent a reactor scram for any anticipated transient. Therefore, the proposed
\
l sl 3-6 b
--- ~ ,..a- . , -. -, -.m.., .,,
, -2 _ _ _ _ _ - - - . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ ___ . . _ _ . _ _
design will meet the diversity requirements of WASH-1270 if the actuation logic is ivplemented (as proposed by GE) with equipment that is diverse irm that of the reactor protection system. We will require that CE submit a detailed design of this initiation system which demonstrates that the required diversity is provided. i i
1 The proposed conceptual design of the automatic initiation system ' ., . .
for tripping the motor driven feedwater pumps and for starting the RCIC and HPCI/HPCS system is not presented in sufficient '---
detail to support a conclusion that the required diversity is f t
provided. We will require that GE submit a detailed design that :
i demonstrates the diversity of this system (from the reactor
_ orotection system) from sensor through actuating devices, or that O
gf analyses be provided that do not take credit for automstic .
initiation of these functions. ,
6 In all A'IWS events, the closure of the MSIV's is required in I l.
order to preclude radiation releases in excess of 10 CFR 100
['-
~ guidelines in the event of fuel failure (NEDO-10349, Page 12-13). [ , ;.y --
. :(- o This valve closure is initiated by high steamline radiation -
signals. We have concluded that this provides the required i'
l diversity in the signal sensing and processing portions of the : ,
-Qf '
MSIV actuation circuitry. However, it is not clear from the ,[
- %f information submitted that the associated logic is of diverse W.27' design. We will require rhat GE demonstrate the diversity of this logic or modify their design to provide the required diversity. RN,'
e l0,T'.
c
- .,y N
- ,
3-7 ' 't s
pg- : y a__,.......-
~. . .. . - - . :>.. w ,.g n ory g. c 8;. .
- g .;y, pgp . . _ . , , , , . .c. , g
,W f
CN -d$;
x <twA ; GE takes credit for automatic actuation of the relief / safety ,._
WL jl,%
(; c . :. valves at the relief set point. The initiating signal for relief valve actuation is provided by pressure sensors. In some designs
); , the pressure sensor / actuator is integral to the relief valve and nf
'[@' . 2 c ,
provides the required actuation system diversity. In other
- Q: : ,
designs a separate pressure sensor (s) and associated logic is i
used to actuate the relief valves. For this type of design we
.Jt '
?1 have concluded that the necessary sensor diversity is provided.
However, it is not clear from the information submitted that the ,,,
'l associated logic is of diverse design. We will require that GE substantiate the diversity of this logic or modify their design to provide the required diversity.
GE does not claim credit for containment isolation. However, c.
{' o .
2 credit is claimed for operation of the relief / safety valves in order to mitigate the consequences of ATWS. It is not clear from i
l ; the information submitted that this will not result in release of i
j radioactive fission products into the containment. We will j require that GE either provide analyses (see section 6.5) demonstrating that containment isolation is not required to mitigate the consequences of ATWS, or provide automatic actuation of containment isolation.
'A{
j The existing containment isolation actuation system design utilizes initiating signals (high radiation level in the containment purge exhaust, high steamline radiation, and many others) that are 3-8 ;
J l
- -w.-
_ -__m- --_-____.- - - _ _ _ _ - _ __ _ _.__ _ - . _
i
,, e . _. . _ . . . . . .. --. ... . . . _ _ _ _ . _._._ __ .. .
/*'
l functionally diverse from those of the reactor protection system. I 1
However, the diversity of the associated logic and actuating devices has not been substantiated. If it is determined that automatic containment isolation is required for A W S, we will .
require GE to submit detailed design information demonstrating b )
1 that the required diversity is provided.
Eg2 * >j
' ~ _ ~
i f
a t
[ ..
- e f
f t ..
l:, .
I
.s 6 y'
- 9 1 b.' '4,
, se e.
6 . ' = h s.
.i.,
',. e; m.?
s%-
A%.-:
u e " , .w.
"... .?y
,-o ;y;&. .
q.?
GllF.y , . ,
,y eE.
r m
(~fb g4 ' e.
' tdk i.-- 4
'~'
$1 n
f 3-9 1:
.h h *G " 'w ' , " - " -
g
- t # -
.. 7. .. 1 .. . . . . .
.* hp _i% 3 A59.b C .~ , A.
, ; .*w*. . . .;A43 s o s *6 *,.* , ^ - -*. wa '
-- - 4/ e
4.0 ATWS EVALUATION MODEL
/
4.1 Organization of Codes The primary computer code used by CE to perform ATWS analyses is
- RED [. The REDY code is an analytical model of a direct-cycle ,
BWR, is used to perform the basic system transient and calculates l the. core and system average transient conditions (power, flow,
(.y.g temperature and pressure) in the reactor. The response of individual l 6
fuel assemblies is calculated using the SCAT and CHASTE6 codes. t SCAT performs the fuel assembly hydraulic analysis using the b i
system parameters calculated by REDY. The CHASTE code calculates I
the heat-up of a fuel assembly using the heat transfer parameters l
calculated by SCAT and the system parameters produced by REDY.
- [p-The response of the containment is calculated using the GE pressure i 7 f-suppression containment analytical model . This model calculates i I
the pressure and temperature in the containment using the relief
[
[~ .
and safety valve discharge flows calculated by REDY. {
The interaction between these codes is shown in Figure 4.1.1. .
The REDY code is discussed in the following section. The containment j ,...
model, SCAT and CHASTE have been previously reviewed for LOCA '
analysis and, therefore, are not discussed here.
. . . .-i.6, .,
4.2 Analytical Methods of Plant Transients
, $5 i 5 ,
The NEDO-10802 report documents the analytical techniques and ;
. . . . .i C., methods used for the transient analysis of the direct cycle BWR. '
?_ l
- i'; : ',2 : ' '
M 4-1 [ *N gn y?,
4o
-.m : = w1 u_ sk y - ,. m ~M. =-
~. ~ --: --- ,
e v J 4a . , 7 ;QW, fi
..g. q .:. v. t /yg 4
4
. g. ,;
4 , , -
.--;f .;>.,~ ,
~ . s
%'- ':myr.p' ,':,
, s <
{.V "
C . R
~
- w. a -:: a. -
?
l -I.
e
- i.
Figure 4.1-1 Interaction of Codes i
I t
Heat transfer Core heat up, g SCAT - - CHASTE )j goefficient Cladding oxidation ;
.e 4 ,
9 Fuel internal .-
n heat generation :'
.e.,
- Pressure, CONTAINMENT Stm. Flow ,i[
MODEL REDY '
temperature through relief valve a,?
A vessel pressure Q
d u +1 n
Yo ts D
'.C
'?, '
s i
- (e .
4-2 .
'l t . ., , ,..,, ,
.'( ' , .
, ~.
$ i1
B 4
,, g . _ _ _ . . . . _ . . . _ _ _ _ . . _ . - . _ _ . .
1 The equations used in the nuclear, mass and energy, momentum, and 1 control analysis of the plant are presented. This review is
' limited to the structure and form of the equations. The specific form and completeness of the equations were evaluated as well as ,
their range of applicability. The values of parameters used in i
the equations are not addressed, as they are generally plant .
3:
specific.
4.2.1 Nuclear Analysis The neutronics model that is used in the GE plant transient code b is based on point kinetics. The basic assumptions made in this .
model are (1) the entire neotron population responds as a single 3~ energy group, (2) the neutron flux is separable into its spatial
^l' and temporal components with system response due only to the , _
fundamental flux mode, and (3) the reactivity feedbacks can be !.
calculated for an average channel to represent the total core ll response.- Violation, for any plant transient, of any one of .
-. these assumptions may make the point kinetics model invalid. , ,
L Four types of reactivity inpuc are considered in the point kinetics model. These are void, Doppler, scram, and control rod movement reactivity. No explanation is provided in NEDO-10802 of how the
- m. .
reactivity difference between two reactor states, occurring -
during a transient, is calculated for the four reactivities #^0 considered. No information is provided on the method of solution ,
e V.,
e ,f -
J.- 4-3 :::.w" Ta Nb1?Ct.
MpM 4.~. -
.(
m.
9 *
. a - .~m.r . '. un' :
. ; n . ,. m . , . . .. - .
.w: - - . . .
r .n. m :r eme. ,- 'e. u a: x : p p ...;;;;;:23 p ;
_- .w n
.. l
. [hb .n ,. !
P N A[ of the point kinetics equation. General Electric should provide ijr.,
n,
- fy$- i
^^; descriptions of how reactivity differences for two reactor states ,
is determined and of the method of solution of the point kinetics ,
. . ~ .
equations.
( 3,;
l'
{. A description of the calculation of the average fuel temperature is given on page 2-9 of NEDO-10802. This average temperature is
.. ,; a volume weighted quantity. However, the uranium-238 absorption
, emphasizes the surface region of the fuel rods. General Electric should _
provide a justification for the use of this average fuel temperature
.,1 in computing reactivity changes caused by changes in fuel temperature.
3
-t i 8 9 j Information has been received in Amendments 1 and 2 to NEDO-1 j 10802 on the effect of using a constant power shape during the 1
/;
I course of a plant transient. Comparisons of the model have been (_
1 made with a turbine trip experiment performed at Nine Mile Point i.
-. I 1 in April 1973, with data obtained from the Monticello plant, t
i 1
,'-] and a one-dimensional, three-group time dependent axial neutronics
( j calculation replacing the point kinetics model of the plant t
transient code. For the Nine Mile Point 1 turbine trip good s
agreement was obtained for the peak pressure out to 2 seconds (10
}
psi higher pressure predicted at peak value) and for the neutron i'
l flux out to 0.8 seconds. Good agreement was also obtained for
.i j
the comparison of the model with Monticello data and the one-dimensional, three group model. The results indicated that the two most important parameters, peak vessel pressure and MCPR, I
4-4
+b** w- - e ev..a._ sue e+e .w ep-wa o ,-eisaw +mwww.
, g --ale-<- e* ' ' ' ' = "-
e -
'^
show little sensitivity to axial variation of the neutron flux.
This insensitivity to the axial flux shape is caused by the long time constant of the. fuel and integral nature of the GEXL cor-relation. Therefore, we conclude that our concerns on the sensitivity i
to the axial power shape during plant transients of the vessel pressure and MCPR have been satisfactorily resolved and that the use of a constant axial shape is an appropriate approximation.
Based on our review, we conclude that the information provided in f the nuclear analysis area is satisfactory. However, General !
I Electric should provide additional information for the items p '
{
listed in Section 4.2.6. ;
l- .-
4.2.2 Mass and Energy Analysis
- The mass and energy equations describe the fluid from the feedwater pump, through the reactor and recirculation loop, through the main steamlines up to and including the turbine control valves and bypass valves. The described portion of the plant is broken into nodes, wherein mass and energy balances are formulated to describe the plant. -
Spatial variation in thermodynamic properties are neglected in ')
the core plenum region as well as the rest of the plant. This is -
acceptable as the relationship of thermodynamic properties to -
n, ,
pressure is a weak one for the small pressure drops at the operating <1 4-5 . .. . ~ . ;
/*
Q;C
'4pl
.. o q
.$5W!i y B ..
I
~~ ,
3 hy@'"'1"'M N* ;"','"
...w., ..; a ,
- . r, , + ;cw :. .a , a :yqa- .,,,c.p.2,3,,-( x y mz.3 ygQz :,,_. y , .
pressures of the core. Also, the kinetic energy terms of the
.Nfk,
'yys;;g fluid have been ignored in the energy balances. This is acceptable ~
,;g.,., >. *
, e> as these terms, and the changes in the terms, are small relative s
m ., , -
$$,4 to the heat transport and work terms of the energy equations.
- . :n% '
'i&,.
'.' In general, the energy equation is structured to solve for the
- . * .X
- pressure in the node for which it is written. Generally, thermodynamic
. equilibrium between the phases was assumed for the two-phase -
.9
. ag,3 '. ! ,
? .
nodes. An exception to thermodynamic equilibrium is the treatment of the vessel pressure rate. In the vessel pressure node, the .-
>;;y.
J~ , mass transfer at the steam water interfaces, other than carryunder in the bulkwater, we'e r assumed to be negligible. This assumption
~
1 j is conservative for pressurizing accidents as it maximizes the
- -;. mass of the steam.
i.',
~"
The treatment of carryunder from the steam separators has not
, been satisfactorily resolved. The entrained mass of steam bubbles
^
f j are assumed to be in thermal equilibrium with the bulkwater.
5 This is nonconservative for pressurizing transients as the steam
- mass and volume in the bulkwater are minimized. Furthermore, in 9
NEDO 10802-2 , it is stated that the carryunder can vary as a
.j ' 1 function of separator inlet flow, quality, and water level.
T.
However, for analysis, a constant carryunder fraction of 0.3% is chosen as it is conservative relative to a determined global i
! average of 0.2%. The stated limitations are that at low qualities, t
4-6 m-e h w e-h9 e e e. -'t6emp we,' ese = se gww y== g,,,,,.py,=ge_
'~
X < 0.09, combined with low water levels, the carryunder may reach 0.4 to 0.5%. This is substantiated by Figure 6 of NEDO-10802-2, which shows separator test results as a function of carryunder, separator inlet quality, and water level. In summary, the non-conservatism of the steam mass and volume in the bulkwater i
remains to be addressed by General Electric. Furthermore, the staff will require estimates of carryunder as a function of time whenever X < 0.09 for all future ATWS transients in order to r
evaluate the conservatism of the constant value used in the analysis. ,
F tt -
The mass balance equations are well formulated and are acceptable. l Compressibility in the subcooled region of the core is neglected, J which is acceptable as the fluid is in a liquid state. Compressibility
\n 1 O of the vapor, wherein the pressure and mass vary, is accounted .
for in the model.
The core exit quality is defined in terms of exit vapor flow and I..
core inlet flow. This definition is valid during steady state,
~
but is in error during a transient as core inlet flow is not .I
~
equal to core exit ficw. However, to assume core exit flow equal to core inlet flow is conservative for pressurizing transients.
During a pressure increase, voids in the core collapse, allowing ,,,
more liquid mass to be stored in the core. Thus, with exit flow '
equal to inlet flow, the steamflow is overestimated, which is conservative. I em
- .V *y m.
a?Qi; y >
-c >
iEg 4_7 <
}%
e.o k-Ip Ml
. . . ,e,~.c.. ns. ; 7+ ; , ; . sa . . _ e . u. ~ -. .
, ,, , ,. . .m . y w m v..y
. @;c>w?;.w 7- " u.vi.gy3y".;*v;teu.F Pc a
?< . y *.
?
Y Q'.d M.E The modeling of ffow through the separator, safety valves, and c
- .9;x'a a p .% (cv .+
.ggfy relief valves, was also reviewed. The modeling for the main ,
steam line isolation valves includes flow area as a variable.
. gig:" '
{pr;, ' Flow through the relief valves and safety valves is pressure cs N ]'{
dependent. The modeling of these flows is acceptable to the f .' '
staff.
t Bubble formation in the core is an important mechanism to
- p T;3.c reactivity control of the nuclear power plant. For purposes of conservatism, it is valid to assume an instantaneous collapse L. '
}
of bubbles and voids with pressure increase as positive reactivity is maximized. However, upon increased power, the j reformation of bubbles occurs over a finite time, such that
~
- reality and conservatism are not served when the reformation -
\;
is assumed instantaneously. The concerns have been expressed I
l by the staff and no response has been obtained to date.
q Amendment number one to General Electric Company licensing Topical Report NEDO-10802, " Analytical Methods of Plant Transient
=
9 Evaluations for the General Electric Boiling Water Reactor"
- g has been reviewed. In response to a question on the void sweep
,] model, transfer functions were provided comparing the analytical s
model with experimental data. The experimental data was for two different velocities, which were stated as 4.0 f.p.s. and 5.5 f.p.s. These velocities are undefined and require clarification.
Furthermore, it is noted that at the lower velocity of 4.0 f.p.s., a natural frequency of the system exists at approximately 4-8
4 . - - - _ . . . _ _ _ _ _ _ _ _ _ _ _ _ . - - . _ . . - _ . _ _ _
1.3 H:. The gain is approximately -6.5 db and the phase is undefined at the natural frequency. The analytical model is conservative in gain relative to the experimental data up to this natural frequency, but does not simulate the natural frequency. Due to the low amplitude of the gain at this frequency, this is probably adequate. The analytical model is nonconservative relative to phase, but this is not a critical -
parameter if the gain at the natural frequence is small.
5 In Section 2.6.2 of NEDO 10802 , three different models for f void sweep simulation are presented. In NEDO 10802-1 , it is l i
further stated that steam transient time and bulk boiling i i
boundary shift are very significant parameters in reproducing -
- - the dynamic response of boiling channels. Also as expected, j p
[(g]v axial power shape influenced void sweep. The factors of power '-'
shape weighting and boiling boundary motion have been incorporated i e-into the method of calculating void sweep as reported in NEDO 10802, in form 3. This method is characterized as a second order, variable time constant transfer function. Based on the i.'
l favorable frequency response comparisons between the analytical , ,
. and experimental data presented in Figures 1 and 2 of NEDO 10802-1 , form 3 of the void sweep model is acceptable. Form 1 and 2 of the void sweep model can not be reviewed because of d f$c lack of comparative data. In summary, the second order variable di
~
. time constant transfer function for void sweep, presented as e
Li
- 3. 4-9 -
- s. ~
- 4.; m ,t
.p .
'A i
e di 9
77 ,-~. - . 7. y ,, ,
..4 . - , . . ; .- .
.. .- : :, a s- m.eg,,.,g. yg.y .q.. .. m p; gg.y ,, _ ,
4
%;4 a swi
[) ,$$, form 3 in NEDO 10802, is an acceptable model for ATWS transients. .[7 s.
ee4. The other models, form 1 and form 2, presented for void sweep .
3 g4a. are unacceptable. .
MSi
//G$
- +7t+- J.
A concern arises as to the range for which the void sweep s
- ~.y,
' ;$!' model may be used. Do the transfer functions presented for wn 4 t-n 8
,.3s, the experimental evaluations (Figure 1 and 2, NEDO 10802-1 )
j t?R p .* .
II,I?) scope the operating range of the plants simulated? If not,
., : x _
, .2.fy .
provide the transfer functions with experimental data an1 the ,,,
analytical model comparison that cover the operating limits of
.,e.r l the reactor. Also, provide an explanation of the parameters that describe the experimental operational state of the plant.
9 w
- h;h Because of the importance of axial power shape upon critical ,
E^
t heat flux, the staff has been concerned with the time variation > -
of axial power shape. In the analysis, General Electric uses a constant axial power shape, independent of time. In response 8
a to the concern (page 8, NEDO 10802-1 ) detailed calculations
.. .f were made to study the effect of changing power shape during a
/
'1; :,A
~ [ 5! Ccmparison of these results were made with results obtained
- I from a constant power shape model. The average surface heat
- flux for the constant power shape model is conservative relative to the variable power shape model. The reason is that the
{
i flux skewing results in a decreased effective power input to 4-10
,, . g . . _ . . . _ . . _ . . . _ _ . _ _ . _ . . ._. . - _ .
~
to the fuel. This is augmented by the filtering effect of the fuel time cons, tant, which diminishes the importance of exact detailing of flux shape changes.
Based on these results, the use of a time independent axial power shape is acceptable as conservative in the evaluation of j surface heat flux.
4.2.3 Momentum Analysis The momentum equations are generally of a mechanical energy ,
type (Bernoulli equation) for a stream tube, written for one j i
dimensional flow, and implemented for the branches connecting i the various pressure nodes of the model. The pressure effects
resulting from rotationality of the fluid, friction, area l-i
( .
changes and bends are taken into account by introducing a form D.;
loss factor. For the core, the form loss factor is also void fraction dependent, which is a desirable feature. The remaining terms in the momentum equation account for inertia, gravity, and pressure.
For the jet pumps, aninertia free momentum balance for the momentum exchange region is written. Although idealized, this ,
i is acceptable because of the smallness of the region relative t ,
?:.p;c ,
to the total flowpath and the treatment of losses in the "'
.)
- a j L
momentum equations of the adjacent flowpath. j The treatment of the non jet-pump recirculation system is 5
presented in Section 2.10.1.2 of NEDO 10802 . As stated, g[
g,:9 m;30
~s..
4-11 hhbf' ;
1
=
%+.
,,__\ _ {. ( .g, , . . _
. . < - . .m : o . . :npe:;w.w: m,.a. n. : r.. .- . . ' x w r'v ~ *,t.Y %~;;-
F. '
fu
?$;?!
q: :, ;; equation 135 of this section is incomplete and erroneous. The . .-
p y ;, .
h corrected version of this equation is stated in NEDO 10802-2 .
i g; The form of the momentum equation reviewed has been previously
' L; 9N)s'l f, evaluated by the staff l0
. Based on the previous evaluation,
.Ms.~ vM
.'.?.,.f.
4 ,, . ,' g:
the momentum analysis presented is acceptable for ATWS studies
-3]
.g~
only for the following conditions:
.c ,
, ; f'1 ' 1. The flow in all paths remains subsonic, M < 0.7. (The
'n > ,
,Qj p only exception is choked steam flow at the turbine inlet .
,' nozzles, relief valves and for the isolation valve closure transient.)
'yI 2. The form factors utilized in the analysis are applicable g' .
to the geometry of the flowpath, are applicable to the
. m. g
N operational state of the fluid, agd have experimental )
basis.
t
- l, 4.2.4 Controls 1
The systems presented in the report consisted of the vessel
- level, recirculation pump drive, pressure regulator and turbine
'i
. , . , controls, and the reactor safety system. For the most part, L.y
.'G i
,1 che controls are presented in the form of block diagrams
- 3
- l containing transfer functions including gains, leads, i lags, and time constants. Furthermore, non-linearities such i
as deadbands and limiting values of signals are presented.
The type of control modeling presented is acceptable in a general sense. However, General Electric must provide a model 4-12
and demonstrate the adequacy of the model for each plant analyzed by the code. Quantitatively, this is to include the time constants, gains, transfer functions and nonlinearities required to model the control system as well as the overall response characteristics of the control system.
I The type of modeling presented is acceptable in a general
{I sense. However, the modeling used in each specific plant will have to be reviewed for plant unique characteristics, and for the parameters used within the transfer functions.
- 4.2.5 General Items !
i Although not described in the report, the numerical integration
'~
technique utilized for solution of the equations was questioned.
General Electric responded by stating that the integration technique (GEDAC computer program) is the same as described in the Numerical Integration Documentation section of the GE I
compliance (NEDO-20566) with Section II of Appendix K to 10 CFR 50. The technique is an explicit technique, and uses an Euler extrapolation procedure. These techniques were reviewed under the LOCA effort. The LOCA analysis however did not l-discuss the solutions to the point kinetics equation, and
- i additional information has been requested from General Electric ,
[' ~ ~
in this area.
'( { .
4.2.6 Outstanding Items and New Concerns e 1. The condensation rate of the carryunder from the steam *
/7 separator is evaluated by assuming thermodynamic equilibrium
.y (f: .. y, e..f i' ~
4-13 [
n j M m g cve :m w 1 m n " ~mm.rmt "Y ~"
< ,. .. .~ ; ...a., .p mg.ygg.,g.= g.y q9 ;7,:pem:ysm;g;p .
9 .
- ,w -s,.-
bM* between the steam and water. In Section 2.6.6 of NEDO
.. r g . ' im T' 2 ::p (u.; g e -
5 Gl.r~ 10802, Vessel Pressure Rate, it is stated that the rates ;g; '
(- of mass transfer at the steam water interfaces are negligible
- i./-+s-- other than the carry under in the bulkwater. For pressurizing -
.D.,e w
type transients such as the turbine trip, the assumption k)$E
..s.
i[ of thermodynamic equilibrium is nonconservative, as a jb ~ # . finite time is required to achieve equilibrium. General L .'
q,[f, Electric must evaluate the pressurizing effects due to
$'?] the mass and volume of the steam in nonequilibrium.
-e
,' 2. A specific item in the validation o,f the model is the
~'
void reactivity feedback model. During an ATWS event A
.l resulting in pressurization, it is conservative to assume
"' l immediate bubble collapse with pressurization, thus j ,
resulting in immediate positive reactivity feedback. The (' -
reformation of bubbles and voids at higher surface heat l
i flux levels appear to be modeled as an instantaneous event, and is thus nonconservative in the time feedback of negative reactivity. Bubble formation and bubble t .- .i growth are time consuming events and thus it would appear
'.t. -
- ' P logical to model the reactivity feedback in a manner reflecting the physical phenomenon. General Electric r .
must provide justification for modeling negative reactivity l feedback as an instantaneous result of bubble formation 1
and an estimate of the bubble growth time constant. If this time constant is 0.03 seconds or greater, General 4-14 _s/
b, . . ._ _. __ .- . . .. . ~ . . . _ . . . . _ _ _ _ . _ _ _ _ _ . _ . _ ,
Electric must also provide studies which show the sensitivity of pressure to this time constant. .
l l
- 3. The information provided by General Electric in NEDO-10802-01 pertaining to void sweep frequency response is not comprehensible. The intent of the original concern was to obtain the experimental justification for the void I.. _
sweep model, a definition and explanation of the parameters within the model, and the operating range of the model.
General Electric is required to address these concerns t
and to provide clarification to the terms and parameters f currently defined in the void sweep frequency response !
curves. +
_r' , -
l p,
- 4. The'information provided by General Electric on page 16
[.
t..
' -. + ~
^
of NEDO 10802-01 in response tc a thermodynamic concern l:
is extremely confusing. The response refers to equation 42 in NEDO 10802. In the response, it appears that core s
exit quality has been confused with core leakage. As n-presented in NEDO 10802-01, it is difficult to see how
, flow into the saturated node is accounted for by the L
stated equation. General Electric is to amend the response I and clarify the relationship between exit quality and 9 .
leakage, if any. .-
S. General Electric should provide additional information in the nuclear analysis area for the following items:
c.
ilh
., c.
I? ~b.
\
v.; n~
4-15 $
- 9 m{
. . j?t a q tF'
,. -+ n , 'b , 's e .u -...o M ,8 % .% ' y t%L s * *' s X[,+ > i +. t o as ? ~
t s %
. .. -. . . L . 2:L---- . _ . 2.. L 1. u:.2:Q th? M iM E=1hn M E W LE M S Ynhl
~
1 q . .
L jh f -
- y
j; g* (a) the method for determining the reactivity difference ;.,
, %: , <* l
..L A between any two reactor states for all components of l reactivity
-l
.Q,. (b) the manner in which the radial and axial spatial
.m: . -
r[h
.y .
importance is included in determining the reactivity of a given reactor state m .
e (c) details of the method of solution of the point
]
-- } kinetics equations ==
.s
,{, (d) justification for using a volume weighted average J
fuel temperature in determining the Doppler reactivity 4
j contribution.
, . 3
, 4.2.7 Safety Analysis .
y /- )
In sumary, the uses of the analysis methods described in NEDO l
l 10802 for ATWS transients is acceptable, subject to the following limitations:
. 1. Satisfactory answers are obtained to all concerns and k'i
, a; <
questions described in Section 4.2.6.
r 2.
. i (l Controls: The type of modeling presented is acceptable
, in a general sense. However, the modeling used in each specific plant will have to be reviewed for plant unique j characteristics and for the parameters used with the
, i i transfer functions, l
t M
4-16
.,,w _r wy . . . . . . - gem. ee e e... e --=- -
- 3. Momentum Analysis The use of the mechanical form of the momentum equation is acceptable provided flow in paths remain subsonic, M <
0.7. Also, the form factors used in the analysis must have experimental basis.
t
- 4. Mass and Energy Analysis [
p.. ,
(Upon response to stated concerns in Section 4.2.2.) ['
i i
- 5. Nuclear Analysis (Upon response to stated concerns in Section 4.2.1.)
i i
- .=
% 0 h
~
e
+ " .
/ ~ h. m E
'($ ?.1_
., J_ '
% 4 9
6
- 4 4 a
- . . - ..,g,_
,e , _ ,
4-17 yfgi;gv ES
- .- *ek f. N m_
% 1. ) ' '
- Q ' , ' , , . Y -_W '
.?~ : , n; . . r.; w :: . . . . . . - . :.. u . . . _ , , ~
. a
. n. . . . _ . . . -
5.0 STAFF INDEPENDENT CALCULATION 5.1 Staff Model For the analyses of this review, the staff has used the RELAP3-B 1
computer program . Update modifications were made to the code for this study which 1) corrected inaccuracies in the steam l ,
tables, 2) installed a corrected Hench-Levy CHF computation, f :;
i
- 3) adjusted the void reactivity feedback calculation,12 and 4)
- L i
implemented a restart logic sequence to simulate relief valve .
I behavior.13 I
i Figure B-1 (Appendix B) gives the plant nodalization diagram ,
used in this analysis. The core is represented by two channels, one a hot channel consisting of 13 nodes, and the other an
-(- t D average channel comp'rised of 5 nodes. The jet pump recirculation loop is also modeled. Heat removal and coolant mass inventory ,
are simulated with a steam leg and input flow rates for exiting ..'
steam and entering feedwater. A more detailed description of the staff calculational nolel is given in Appendix B.
. t*
~
5.2 Difference Between Vendor and Staff Models ,
l l
The basic BWR-4 simulations used by GE and the staff are j discussed in Sections 4.2 and 5.1, respectively. A comparison
< O.
of various model features is presented in Table 1. Though i-
- p, -
both are similar in their description of reactor components, Y- ' ,
discrepancies in calculational results are attributed to two ,
l a i principal differing items. The first is void description. GE i^
f&: ;
ll _': :., .
y:en.
@}
g 5-1
~. - .
N
^
. < Q,W . _
m ; N ,
.N
. . . n_s . .
ws~ w w .mes - - -
... . :y, .. 4 . - ~ _
s ,
. .- . u %;u;w=v .x.5;;;;;;;s~zt:2. ,y x g1
+
'#I:
v*3 3 48-jf computes a core void map for steady-state conditions and , !
f'. . ~ , '
y ;l. '
?Ei supplements this with a dynamic void sweep model. The staff .<
c~
- 4 assumes a homogeneous void model for all conditions. The amount of calculational disagreement introduced by this
+-
. ~.,g G difference has not been quantitatively assessed. The second
/.
[*t;)/ item is the Main Steam Isolation Valve (MSIV) flow. In the GE ci .3
- model, MSIV flow is functionally dependent upon valve position
- and pressure gradient across the valve. Steam exit to the
.],
turbine / generator from a devnetream node, is an input specification.
j The staff describes MSIV flow through a tabular input dependent
. 1
~
on time. This difference influences the effective closure rapidity, a factor which staff parameter studies indicate is
.t important to the severity of the transient consequence (See 1;
1] Section B-IVb).
l \ --
j l 5.3 computational Results b
The base case transient considered by the staff, was the Main Steam Isolation Valve closure. In addition, cases varying void reactivity feedback, MSIV closure, recirculation pump trip parameters, relief valve capacity, and fuel rod gap con-
/ hf."; ductivity were analyzed by the staff and GE for the BWR-4 4 product line. The BWR-4 design is the most limiting BWR for i
A ATWS transients.
'! Good agreement between the staff and GE base case calculations l { was obtained; peak steamline pressures were 1377 and 1370 psia, respectively. The results of the staff sensitivity
,/
5-2
_ .,,.g - _ . _ __ _ _ _ _ ___ _ . _ _._
t
'~
studies are presented in Table B-1, and a maximum pressure of 1548 was calculated (90% Relief / Safety valve flow). The calculation assumed that the feedwater flow terminated at 9.5 seconds. By assuming the GE value of 6.0 seconds, the peak 4
pressure was reduced to 1483 psia. This compares with the GE value of 1450 psia for this sensitivity study. ,
Staff calculations showed peak cladding temperatures below 1590*Ffor all cases. In all cases hot channel rods experience critical heat flux. Computational results are summarized in -
the tables and figures of Appendix B.
5.4 Summary and conclusions The following results and conclusions are based upon the ,
calculations and sensitivity performed by the staff. This '
3- .,;
evaluation shows that the following parameters have an influence on the peak calculated steamline pressure:
~
- 1. MSIV Operation - The peak pressure ranged from 1352 to '
,,, 1420 psia (See Section B-IVb). l.
l.
- 2. Feedwater Flow - The peak pressure increased from 1377 to ,
I 1421 psia for an increase in feedwater termination time from 3.0 to 9.5 seconds (See Section B-IIf, B-IVb, and B- g VI). i?;' Ti 50:. ..
~
- 3. Phasing of MSIV Closure and Feedwater Flow - The interplay of MSIV operatica and feedwater flow was shown to be
?b .
5-3 h
1
- 1. _b 'I '
., . . . :y . ,, . p. a w a , ,a g v. >
ww m.:;: . , wg.
D?
%;.pv important in the determination of the source and severity [n < >
?l : 1 .
of an AWS transient (See Sections B-IVb and B-VI). ,
4
- 4. Void Reactivity Feedback - The peak steamline pressure
- increased from 1365 to 1468 psia for variations in void
.d
~ D7 reactivity from 4.58 to 14.3c/% (See Section B-ivc).
- 5. calculational Results - For the base case submitted ,
variables, there was excellent agreement between GE and l~ I staff analyses (7 psi difference in peak steamline pressure). "
In submitted sensitivity results, the agreement was also
-i good. Subject 'to aptroyal of the GE model, and further
- required hot channel and fuel damage analyses produce i
satisfactory results, the GE calculated results are i
acceptable. ) _
l 5
)
[ 'I '
- < . , a ..
, 'lV Y
l 4
1
(
I 5-4 wy- , e-u w-er',ee--s.e g-,A g N98" * * ^t-
f 6.0 CONFORMANCE TO WASH-1270 6.1 ,
Limiting Transients GE analyzed the ATWS events and reported their results in NEDO- . 10349 in March 1971. Subsequent to the publication in WASH-1270 of the staff guidelines on the analysis and limits, GE reanalyzed the more limiting ATWS events and reported their results in NEDO-20626 in October 1974. The results of the GE ATWS analysis are I summarized in Table 6.1-1. In their later analyses GE calculated lower dome pressure because GE assumed higher relief valve capacities -
than those used in the NED0-10349 analyses. Further, GE also l calculated lower containment pressures as reported in the NEDO-20626 2 (7.6 psig) report than that in the NED0-10349 (46 psig)
. [{
report due to early initiation of the liquid control system which reduces the power and consequently less steam is released to the '
containment. I Table 6.M shows that the MSIV closure results in the worst vessel pressures and the highest peak fuel enthalpy. This event also results in the most severe fuel duty and consequently a larger I i, .u
- amount of steam is dumped into the suppression pool than any ?
f other ATWS event analyzed. ,
The MSIV closure event could result from a spurious signal. The .fbr m.
sequence of events for this transient in the event of a failure -u_
i~2 of the reactor scram system may be summarized as follows. A trip ;
of the MSIV valves would initiate a reactor scram signal. This .
Jh.
- kb 6-1 .W ~
1.f' "i3.1
,{ h
$5 y * , $ .,
a .: ,n. .
v , v. , ; ,;y .
y' ,,543:;
.: j., fg. ,.
l* ,
i! .g.;;;. -
- r. , . .. ,, -
--7
- + ^ %me - aFpa
_ . . _ _ _ - , .. _._.. ... . . . _. _ . . - . _ ' _ _ - - -. ; L. -
Table 6.1-1 Results of ATWS Analysis +
t Peak Dome Peak Containment Peak Pool Temp. Peak Fuel Enthalpy Pressure, Psig Pressure Psig 'F cal /gm ,
1500 Design See Section 6.4 See Section 6.3 -
CRITERIA NEDO NEDO NEDO NEDO NEDO NEDO NEDO NEDO TRANSIENT 20626 10349 20626 10349 20626 10349 20626 10349 Loss of Condenser vacuum 1324 1441
- 46 * * <142 142 -
MSIV Closure 1345 1535 7.6 46 163
- 150 150
?
i Loss of 100*F feed water /
heating
- 1094 * <46 * * * <150 ,
Feedwater control Max.
- 1138 * <46 * * * <150 T
m Demand I Pressure Regulator Fail open 1245 1453
- 46 *
- 148 145 Loss of FW Flow 1240 1205 * <46 * * * <150 Inadvertent Opening of a i;
} Safety valve Depressurization
- 38 * * *
- Jr C
/ ~', t Core Coolant flow Increase
- 1099 * <46 * * *
,, <150 (
.A Loss of Auxiliary Power 1163 * * * * * * * ',
{9
- I c.
l.4
. y
- Results not provided by GE
- j
+ Most limiting results are obtained for a BWR/4 design. @
- L~, p V$
I s . ,
, . e, . _ . . _ _ _ _ _ _ __.._ _.
~ s i
8 scram signal is ignored as part of the assumptions in these analyses, as are the subsequent reactor scram signals generated (for example a common mode failure in the logic protection system could prevent scram from any of these signals) as the transient proceeds. The MSIV clostire would stop the steam flow out of the 1
reactor causing a pressure rise in the core and a collapse of 1
l l
steam voids. Because of void reactivity feedback this gives a positive reactivity contribution which causes the neutron flux i
and reactor power to rise, however the Doppler reactivity coefficient j reduces the magnitude of the peak neutron flux. At three seconds i j
into the transient the recirculation pumps trip initiated by !
vessel pressure exceeding 1150 psig. The recirculation pump trip !
e
_ results in a decrease of the coolant flow through the core and s.
more steam voids are produced. The neutron flux then decreases
.eQW . , ,..
due to the negative void reactivity feedback. During this period the feed pumps are also tripped to reduce inlet subcooling into 3
the vessel and to increase the voids. The HPCI and RCIC flow i m
initiated by the high containment signal reach the vessel at I
-. r I' l about thirty seconds to prevent the core uncovery. The borated i ^;[
, water initiated by high vessel pressure reaches the core at [._
g'.
)
ic forty seconds and reduces the power. F
[u ,
Before the staff can conclude that the MSIV closure ATWS event is [
h[h n;:
most limiting the staff requires that GE provide analyses of the }#.1(jy? ,'
4 $4 ' .
in-sequence and out-of-sequence rod withdrawal ATWS events, loss l of normal onsite and offsite power ATWS event, stuck-open relief / safety ~~
eh [
[" valve and a 100*F step loss in feedwater heating ATWS event hM-b 6-3 mvg 9
- M ,$
4 -
% ' (-' ..f,.J "4 E g %
~
g , .c ++ u g L ~ ,,- ; c. ' . --*' '
- m Vs : ;. :lE 4 5 .' N;_ f_3. ;Q< t 'l Q'Qb v' N '
, ~_p_
,. y . , m y m x.::; w ; w n s p p p y 2 4~. ;
- s. . -
W, '
., ,f m l,;
yy;:'Q
! g. C unless GE can demonstrate that the frequency of occurrence of these transients does not contribute significantly to the frequency of
, occurrence of all transients. The staff also requires that GE provide
~
, long term vessel steam flow rate and the effects of stuck open safety /
)Q ,
relief valves on the poo1> temperature and the radiological consequences.
,1:;d, y;, ~,
' 6.2 Reactor Coolant Boundary Pressure l
6.2.1- Stress Limit and Basis .
.]
'i f The licensing position contained in WASH-1270, "Tec % ical Report
~j on Anticipated Transients Without Scram for Water-Cooled Power I
]
Reactors," requires, for those plants defined as Class B, that
-(
i "The calculated reactor coolant system transient pressure should be limited such that the maximum primary stress anywhere in the 1
! system boundary is less than that of the ' Emergency Condition' as g j.--
! defined in the ASME Nuclear Power Plant Components Code, Section 1--
1 t
III.14" l
a 6.2.2 Pressure Limit
_i l ~! We have reviewed the component by component emergency condition stress intensity and corresponding maximum allowable reactor
'*I ,'j.[
,:,(..
coolant system pressures reported by GE in Enclosure 1 of their 16
[y u--,; letter dated 10/7/74 from I. Stuart to V. Stello. This letter C. ;3-discusses the allowable pressures for the major reactor oolant i system components such as reactor vessel and those in the recircula-tion system. We have also reviewed Section 4.2.1 of NEDO-20626 I
" Reactor Coolant Boundary Pressure Limit."
/
. 6-4 >
= - _
t.
l!
While we do not agree with some of the reported allowable pressures of some specific components, we do agree with the GE conclusion as stated in NEDO.-20626, that if the reactor vessel pressure is ,
kept to 1500 psigor less the components of the reactor coolant system should be below the Emergency Condition Stress Intensity Limit as defined in ASME Section III. ,
We have independently calculated the maximum allowable pressures ,
using a simple but conservative method. The ratio of the Emergency Condition Stress Intensity Limit to the Design Condition Stress Intensity Limit was determined for the types of materials reported l h
by GE in the 10/7/74 letter. This ratio, always larger than one, }
when multiplied by the component design pressure provides a
- .- simple but conservative value for a limiting emergency condition Sq
' pressure. S-y The pressures thus determined are minimum values that a component with minimum ASME Code acceptable wall thicknesses could be y expected to withstand without being exposed to stresses higher ;-
~
than permitted in ASME Section III for emergency condition events. Generally component wall thicknesses could be expected ,
to be greater than the required minimum and thus a given i component could be exposed to somewhat higher pressures without j-ny exceeding the emergency condition limit. The actual maximum x%.
J 2 S .);
allowable pressure would have to be evaluated on a component by 4.;-j
, ;.p. f, component basis. .
E . -
t WKS- <
. .q .
1 7 Cpi-l 6-5 " V $4 g
- w . 9 , fi 't^ k m s. w su N ;;, , Jg. en s, ~ = , --
- lf.p.m :; ' k i * % :.p,- ~ .
L,- . . , _qw y ; m p,ygsp_. .
, ~ , . . _ u, . .- . -
,.~.. , ,3 ' %p, . .m .
. e
+
4 nni We note that G.E. designs their reactor coolant system components j f c-s
' E ',. for a temperature of 575'F. In our evaluation, for convenience, "a'-
x we have used this same temperature for determining code allowable
]
ji stress values.
';fai:
gyJ;;' The information in NEDO-20626 indicates that the peak fluid
%g x- n temperature, coincident with peak system pressure, may be about 20*F or so higher. However, for this small temperature difference .
-I the code allowable stress values would remain essentially the 11 same. Also, it is unlikely that the temperature of the components ,,,
-i i
could reach equilibrium at the fluid temperature during the 4
relatively short dura ~ tion of these transients i.e. less than 25 s
.* seconds.
6.2.3 Limiting Transient p I?
We have reviewed the results of the analyses and note that the f)
-s- '-
j MSIV closure transient results in the highest reactor coolant
-l j system pressure. For this transient G.E. has calculated the peak system pressure assuming that all safety and relief valves funct.on. Even assuming that one relief or relief / safety valve
_;g fails +.o open at the relief set point, the staff believes the
.se t j,f,% : emergency stress limit will not be exceeded. The staff conclusion ll ,
is based on the results provided in Table 6.3 of NEDO-20626 where g .'
~;
GE has calculated that peak calculated pressure is below 1500 psig for all classes of plants (BWR/4, BWR/5 and BWR/6) when only 90 percent relieving capacity is assurzd.
i I s
\
l ! 6-6 l ; :
6
5.2.4 Conclusions We conclude that GE plants, subject to the reservations noted below, can be exposed to pressures up to 1500 psig without exceeding the Emergency Condition Stress limit criterion established in WASH-1270.
l There are plant to plant variations and since material evalua.ed to l
date deals only with the major Reactor Coolant system components, the l following should be verified by G.E. or by individual Class B plant licensees. ,
t
- 1. NEDO-20626 and the 10/7/74 G.E. letter referenced above do not specifically discuss the maximum allowable pressures for piping, isolation valves and any other components of auxiliary subsystems -
_ _ which interface with the reactor vessel, recirculation feedwater, f) and steam piping inside the drywell and are exposed to reactor .
operating pressures. Examples of typical subsystems are High Pressure Coolant Injection, High Pressure Core Spray, Low Pres-sure Core Spray, Low Pressure Coolant Injection, Residual Heat Re_ oval, and Standby Liquid Control System. Verification must be provided that piping and isolation valves for all such subsystems .l that could be exposed to ATWS pressures would not be thereby stressed beyoad the Emergency Stress Intens1ty limit. ;
l 4
- 2. Isolation valves for those auxiliary subsystems that must func- #: 'O f:,;[
tion to mitigate the consequences of ATWS 1.e. HPCI, HPCS, RCIC, RHR, and Standby Liquid Control System, must be shown to have sufficiently low stress levels so that no permanent deformation 1 rn.
n,y .1, 6-7 k
(4;
.T
- s. .
f WQ u
> ;e - .
.. . - ,, , :c ,: ,,~>,c w
- T ' "!W - '
' W.~ , a j$ /: -of any part of the valve would result which might inhibit valve f
..o operability.
Yfl 't
[
- 3. Verification must be provided that the hydrodynamic loads asso- -
(([d
- 4 L - cisted with the large relief valve mass flow rates and extended V blowdown times resulting from ATWS would not result in loss of x 3:
structural integrity or operability of any safety related com-ponents which may be installed in the suppression pool.
- 5 1
- 4. Each applicant shall demonstrate to the satisfaction of the NRC l 5
==
- i. staff that the information in NEDO-20626 and the 10/7/74 letter
,j from 1. Stuart to V. Stello is conservative for his particular
'I plant.
I
- ~li 6.3 Fuel Condition
[*
! 6.3.1 Fuel Limit and Basis
" ~
The licensing position contained in WASH-1270, " Technical Report on
} Anticipated Transients Without Scram for Water-Cooled Power Reactors,"
-m ,
requires for those plants defined as Class B, that "the calculated A
reactor coolant system transient pressure should not exceed a value for which tests and analyses demonstrate that there is no significant y?
f,1,cp g' safety problem with the fuel," and for the fuel thermal and hydraulic
$_ t t j performance WASH-1270 requires that:
i
) ?
"(1) the calculated average enthalpy of the hottest fuel pellet
,l should not result in significant cladding degradation or r j j significant fuel melting.
t
)
o-6-8 i
$ ]
, 4 l
l l
- e. , - - -emes ,mm. -av - -.,e
P ' . .m.. ~ . . " . . . ~ -
. . j I
(ii) a calculated critical hcat flux event should not occur unless the calculated peak cladding temperature can be shown not to result in significant cladding degradation."
t 6.3.2 Evaluation, Results and Conclusions G.E. has proposed specific limits for the fuel rod during an AWS which are described as General Electric's interpretation of the WASH-1270 criteria. These are:
a) Peak fuel enthalpy: 280 cal /gm i b) Maximum cladding oxidation: 17% (by volume)
We have reviewed these proposed limits in the light of the possible consequences of a BWR AWS on the fuel rods in the core and have concluded that while these limits are necessary they are not suffi-f:4 cient to be reasonably certain that a fuel rod which meets these
- zy limits will not fail by some other mechanism when all possible BWR A WS events are considered.
In addition to these limits proposed by G.E. we require that the
_ following limits be used in calculating failure for an ATWS event.
For the purposes of radiological dose calculations, any fuel rod .
7 bundle for which maximum critical power ratio (MCPR) is less than the !
minimum MCPR for a specific plant or 1.05 for generic calculations will be considered to have failed. In addition, if calculations for ,
7 o - ,
, an AWS event show that an increase in fuel rod power occurs on a rod bundle for which the MCPR is greater than the minimum at which failure l
}
sn. .
6-9 u: .
' HAM Nd.h.[AN* ?Mhd$ hN M - 1.-l '.-$'
~ . , , , ., .i ' B: ' ' % F -
~
.w" a ' .x . ;.s an. ~: 'bw :.e ; , ;. ;. - n. . . . .
7--- ---------,,
, ,, , g .q.
. .. ,wfp% .g .
y;, -
- 'A *
.a
. ., : f '; , .
. . . . ./ _
l .g' l l 3
O
/ 9 y _> is assumed, G.E. should perform an analysis to determine the number of ' "' '
4 e- :
$z o
fuel rods which fail as n consequence of pellet cladding mechanical .
- j -
4 interaction. A combination of these analyses will provide an accept- .
ably conservative estimate of the number of failed fuel rods for use
,,.7 -
E in radiological dose calculations.
9,Y.
In response to staff questions in NEDO-20626-2 G.E. has stated that
" detailed analysis of BWR fuel designs for resistance to cladding a
collapse during ATWS events are scheduled for completion in late 1975.
t j To demonstrate the margin between 10 CFR 100 limits and the estimated
~
consequences of an ATWS, the failure of 1% of the core fuel rod plena ,
't i identified as an upper limit in NEDO-20626 is assumed for analysis."
i
,- , Assuming only plena collapse ignores the coilspse of the cladding into m s I l 4 axial gaps in the fuel column with subsequent axial wrinkling of the - - M 4
,- l cladding. This mode of failure should be considered by G.E. in the calculations planned for completion in late 1975.
i The staff expects G.E., based on the guidance given in the above e
paragraphs, to provide an estimate of the number of failed fuel rods.
3 5
,- The number of rods which fail by each mechanisms should be listed i 5
1,;' separately; that is, the number of failed fuel rods due to: 4 l a) Collapse of the cladding into axial gaps in the fuel column or 5 with collapsed plena. =
- a '
,6
=!
{ b) Cladding for which the MCPR is less than the minimum plant MCPR $
i l
- or 1.05 for generic calculations. a
-1 -
2 2 } c) Pellet-cladding mechanical interaction. 3
]
l
- 6-10 #
1
=
a E - - -s -- . --
,. . -. 4. . ...
1 I
The staff also requires that GE analyze the Loss of Normal Onsite and Offsite AC power, Loss of 100*F Feedwater Heating and In and Out of Sequence Rod Withdrawal at zero and full power ATWS events and demon-strate that the MSIV closure is still the most limiting ATWS event. ,
1
. l 6.4 Containment - !
l 6.4.1 Containment Pressure Limit and Basis The licensing position contained in WASH-1270, " Technical Report on
. i Anticipated Transients Without Scram for Water-Cooled Power Reactors", l j i
requires, for those plants defined as class B, that " Calculated maxi- i
! I mum containment pressure should not exceed the design pressure of the j ;
i containment structure. Equipment located within the containment that :
is relied upon to mitigate the consequences of ATWS should be qualified by testing in the combined pressure, temperature, and humidity environ-ment conservatively predicted to occur during the course of the event". A further requirement has been identified for BWR pressure suppression containments. Reactor operating experience has indicated that potential instabilities in quenching of relief valve discharge l I
= flow could occur at certain steam mass flux conditions and suppression pool temperatures. Therefore this region of mass flux / temperature where instabilities might occur must be avoided during the ATWS event. !
6.4.2 Limiting Transients General Electric has provided ATWS evaluations for typical BWR/4, 5 _ , , -j 2
- and 6, Category B 2 actors in NED0-20626 . For containment, the main
, steam line isolation valve (MSIV) closure transient is the'most severe !
., j
-t"
-a 6-11
. A.
g h
,C%.::6 a $ , ,- fA -^ ',
- m. s s. . , . - .w .. . ; ,.
+- -
.~./,#..-4 L W 4 a s .c - 3 s m:- -
. ,,a: - .
a e-' .a,=~ .-
W
- NE -
1 Mai ATWS event analyzed since it results in the highest total mass and O l TRii?; 'y~
Mk energy release to the suppression pool.
9:4_.) '
j GE has performed analyses of the suppression pool heatup and con- *
, ,; n -
ji tainment pressure response for a time period of 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> using the y-k analytical models described in NEDO-10320 and NEDO-20533 18 . The calculated containment pressure rises for Mark I, II, and III type containments were from 3 psi to 7 psi and pool temperature increases '
q of 40*F to 68'F The containment pressures correspond approximately
.g =
f to the increase in vapor pressure due to the temperature rise of the suppression pool. We have asked GE to provide the pressure response
- of the reactor vessel during this 2.5-hour time period, but have not l
l ,
yet received this information.
i -
i, -
- m. '
6.4.3 Staff Evaluatitu and Conclusions
.; The analyses in NEDO-20626 indicate a maximum average pool temperature increase of 67.6*F. Based on the GE recommended maximum operating suppression pool temperature of 95*F, the average pool temperature I -
exceeds 162*F following the MSIV closure ATWS event. Further, the analysis of a stuck open safety / relief valve ATWS asked for in section 6.1 may result in higher pool temperatures than the MSIV closure ATWS.
l' GE has recommended 170*F as a local pool temperature limit. Since the t
~{ local pool temperature is expected to be about 10*F above the average
,] pool temperature, the maximum local pool temperature following the ,
~I j ATWS event is about 173*F which exceeds the GE limit.
l t
6-12 ,
l l Y
Furthermore, until CE can provide information to substantiate the selection of 170*F pool temperature limit the staff requires that a local pool temperature of 160*F not be exceeded.
6.5 Radiological Consequences 6.5.1 Dose Limit and Basis
{
f The licensing position contained in WASH-1270, " Technical Report on Anticipated Transients Without Scram for Water-Cooled Power Reactors," l requires, for those plants defined as Class B, that " Calculated t
i radiological consequences should be within the guidelines values set '
forth in 10 CFR Part 100."
l 6.5.2 Evaluation Results and Conclusions f
The two most important considerations in the determination of the !
e (l l radiological consequences of an ATWS are the possibility of fuel pin failure, and a possible release of activity from the containment prior )
i to its isolation. We have identified these concerns in Section 6.3 l 1
and our letter to GE respectively. The answers to these questions are still outstanding. We cannot arrive at a final evaluation of the
=
radiological consequences until these concerns are addressed by GE. l However, to demonstrate the severity of the possible radiological ,
i consequences we have calculated offsite doses for the MSLIV closure transient, based on conservative assumptions concerning these outstand-ing items. These assumptions are:
6-13 e
h ' ' ^;
M k r.w i
l 71 f NI5NENE '
s ,,
.a .. ,, . -
- w. . .
, , + .~ J L ' v, .1
- s-
- s N s P y a) The transient results in the failure of 1% of the fuel pins. !
s u n. y k
W-p C 's is based on a similar illustrative assumption used by GE in ,
answer to one of our concerns)9
/
M. w.(;;; b) The containment is manually isolated (i.e. an assumed operator
, .g action time of 10 minutes) and remains intact throughout the 1
transient. This calculation is summarized in Table 1. The doses a
I calculated with these assumption are several orders of magnitude above the guidelines of 10 CFR 100, and therefore are unacceptabl~. e l Based on these calculations automatic, rapid isolation of the containment and the demonstrated, long term integrity of the containment is required. However, this may be insufficient to t
meet the dose limit and a final evaluation of the radiological
? consequences of an ATWS must await the satisfactory resolution to /
\
Section 6.3 or demonstration by GE that the dose limits will not be exceeded if the purge lines are not isolated, unless accept-
) able signal provided, in those containments with continuous purging.
. ' {!
.' 6.6 Post-A'1VS Shutdown hl 6.6.1 Safe Shutdown Condition v .- :>
2 7, D, , After a plant experiences an ATWS event, it is imperative that the i
capability exist to bring the plant to a safe shutdown condition. A
]
.; safe shutdown condition would be achieved if the reactor is brought to I
j suberiticality, the long term heat removal systems (e.g., Residual I
i Heat Removal) are in operation, the long term cooling systems have 4
f i &
6-14
_ , _ _ .. _ _ . =
g . _ ~ . ~ .._ m.. -..~.w..
/
TABLE 1 Radiological Consequences of an MSTV Closure Transient Without Se am Power Level 4321 MWt Failed Fuel Pins 1%
Fraction of fission product inventory in gap 10%
Decontamination factors for iodine:
DF liquid-for-steam 10 l DF suppression pool 2
_ _ Release to environs 3.22 x 10 4 des '
g.- .
Breathing rate 3.47 x 10-4 ,3/see Q_.f ,
-3 3 6 X/Q 1.0 x 10 sec/m j i
2-hour thyroid dose 16,200 rem (10 CFR 100 guideline 300 rem) -
t I
I i
e t
?
r a l l
3 i i
j g... \
( ; } 'c I l 6-15
'ti iU
$; ' W;d U $
.. .. . p. . . ..s .;. . m p4 '.; w p.-E m h g 4;2FeW W.1w-
^ .-F.7 u ~ , e'm >, ~
~
q w u x :- J ;; '< b '
hei = : ;+ :- -
a
, 2,- w ; ~ - , . - .
W l ,,, > -
.~7-
. .. : 4 v; +- %,'
&. ,Q ec y . sufficient capacity to bring the plant to a cold shutdown condition, Til?'
v.
{. . . and no limits of WASH-1270 will be exceeded in bringing and maintain- ,
. t, ing the plant to a cold shutdown condition.
- - j
- g. m gy
- y,p 6.6.2 Limiting Transient and Results pg
- ,
.e 2 .
GE has stated that the MSIV cicsure is the most limiting ATWS event
- } since it imposes the most severe conditions on both the fuel, the 1
vessel and the containment. Before the staff can concur with GE that m j
-1 MSIV is the most limiting ATWS event, GE must provide analysis of events identified in Section 6.1. The GE long term analyses assume
- . i
{ the use of boron poison insertion in the vessel. The ATWS results
- t l would depend to a certain extent on the efficiency of mixing of the l I j boron poison with the water in the vessel. We believe that the model Jl W}l4
- . y used by GE assumes homogeneous mixing in the lower plenum and the ^
-.". =
g f'" j staff requires that GE provide bases and details of their model and - -
l-1 include sensitivity studies to account for nonhomogeneous effects.
(.'> d The GE long term analyses performed so far are also incomplete in that p;g, 7 , GE has not demonstrated the ability to maintain the vessel level, the j;; ability to remove heat in the long term, the ability to maintain
. :1, , 1 Pg j'; containment temperature and pressure within limits and the ability to y ff!,
i.tyffe c, isolate containment and not exceed the radiological limits. The staff 33.-
,"',' , , addressed these concerns in their letters 14'15 to GE. However GE has
+c ; not yet provided responses to these concerns.
i Presuming the reliability and the diversity guidelines of Section 3 l are satisfied an acceptable analysis to demonstrate that the plant can be brought to a safe shutdown without exceeding the WASH-1270 limits 3 Y
l 6-16
( ; ,. .
should be performed. The analysis must address MSIV closure ATWS event and the other events identified in section 6.1 and include all parameters affecting the vessel inventory, containment pressure and temperature limits and the radiological calculations, approximate times of all operator actions, the sources of water, their capacities and their amounts required must also be provided. The analysis must also estimate the impact of the loss of normal onsite and offsite AC power ATWS event on the ability to bring the plant to a safe shutdown condition.
6.7 Reactor Protection System Vulnerability to Common Mode Failure WASH-1270 specified that the areas of the reactor protection system particularly vulnerable to common mode failures be identified and corrected. With this objective in mind, we have reviewed the GE analyses (NEDO-10139 17 3
~
4 i
, -10189 , and -10349 ) and the design of the GE l reactor protection system for the BWR-4 and 5 product lines. For f these designs we have concluded that, because of the diversity and l
i redundancy in the signal generation and logic portions of the system, l
for each transient the failure probability of the system depends on l i
that of the output devices, i.e., the scram contactors which de- j energize to in turn de-energize the a-c scram pilot solenoids which s remove the air supply to respective diaphragm type scram valves (2 per control rod) and thereby effect a scram. We have further concluded I that no reliability credit can be given for the backup scram feature
. f because it utilizes normally closed contacts from the same scram P **,
ep 6-17
,e * .' !' )
r>/t v h>, s
,[ '"
t 7 .
.. ., ~J %' kh " J 'i' 5'/ - ! *
~
.,w . . . i .+ ei c n=5 m J f L- - w=M
.. , . c ,, .
. u .w . . o -: -
$. yfi_.,s
]g-contactors (to energize two backup scram pilot solenoids from a 125 v (,
d-c source which in turn depressurize the instrument air header to the scram valves to effect a scram), and because of the long response time
.[ , of the backup scram feature, i.e., one to two minutes compared to the
%jn K,.R 0.5 to 6 seconds for normal scram.
.e(-
- These contactors are identical in manufacture, design, operating s 1
-) .
principle and operational requirements. Therefore, we conclude that *
.l these scram contactors are particularly vulnerable to conunon mode
- failures and that either a diverse means must be provided for inter-i l rupting power to the scram solenoids for normal reactor scram, or the
'{
j.
design of the backup scram must be modified to provide diverse con-tactors to energize the backup scram solenoids and to provide a scram g:,
- $ ~ , response time equivalent to normal scram. ,.
, t a _
The review of the BWR-6 design for the reactor protection system is
, j presently underway. This review will be completed when final pre-s liminary design information is made available. As required by WASH-1270, areas of the design that are particularly vulnerable to common
, mode failure will be identified and corrected in the course of this
.py
.'N.
9 review.
i I
f
, 1 i
6-18 ,
e J
_. ~ _ . .. ._ _ . - , , , . .
w, e, -
f I
l
7.0 REFERENCES
- 1. Technical Report on Anticipated Transients Without Scram for Water- ,
Cooled Power Reactors, WASH-1270, U.S.A.E.C. , September, 1973.
- 2. Topical Report: " Studies of Design for Mitigation of Anticipated '
Transients Without Scram," NEDO-20626, The General Electric Co., October, 1974.
- 3. Topical Report: "An Analysis of Functional Common Mode Failures in '
GE BWR Protection of Control Instrumentation," j NEDO-10189, The General Eeletric Co., July, 1970. l
- 4. Topical Report: " Analysis of Anticipated Transients Without Scram," .
NEDO-10349, The General Electric Co., March, 1971.
i
- 5. Topical Report: " Analytical Methods of Plant Transient Evaluations i for The General Electric Boiling Water Reactors,", {
NEDO-10802, the General Electric Co., February, 1973. ;
- 6. Topical Report: " General Electric Analytical Model for Loss of Coolant Analysis in Accordance with 10 CFR 50 Appendix K," -
NED0-20566 (draft), August 1974. !
. , - l
- 7. Topical Report: "The General Electric _Presst!re Suppression Containment
[6/ ,
Analytical Model," NEDO-10320, April, 1971.
- 8. NEDO-10802-01, " Analytical Methods of Plant Transient Evaluations '
for the GE BWR Amendment Number 1," April 1975, R. B. Linford, General Electric.
- 9. NEDO-10802-02, " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor" June 1975, S. E. Baham,
. R. B. Linford, General Electric.
I
- 10. " Evaluation of LOCA Hydrodynamics" November 1974, Regulatory staff, i Technical Review, U.S. Atomic Energy Commission.
- 11. RELAP 3-B, manual, A Reactor System Transient Code, Brookhaven National Laboratory, December, 1974. i
- 12. Monthly Report, Melvin Levine (BNL) to Francis Schroeder (NRC),
January' 31, 1975.
- 13. Letter, Melvin Levine (BNL) to Leo Beltracchi (NRC), October 24, 1974.
- 14. Letter, V. Stello to I. Stuart, January 28, 1975.
- 15. Letter, W. Butler to I. Stuart, April 9, 1975.
J 7-1 DW3 n m V nsi gQ vnm m vyyy~gw-y~~ y
, llj ;] 16. Letter, I. Stuart to V. Stello, October 7, 1974.
.., V i;j 17. Topical Report: " Compliance of Protection Systems to Industry Criteria, General Electric BWR Nuclear Steam 4
- Supply Systems," NEDO-10139, June 1970.
. I,,. 18. Topical Report: "The General Electric Mark III Suppression d,yW$13 Containment System Analyrical Model," NEDO-20533,
'T!?'
N ~;j June 1974.
- fs. 4,;x.
. t;.. q i
.s*' ,
'j a
T #
F -
+,
-l
/ ,
v .*'*S f&-
x y-
,, 3, . -- t.
.y L; -'
., * ' .; b
%w a
.(
- )
e
+
t
_- c., .s -
. L.a. :l e . :
QS -n,j
~ . *. '
.:bygu
?
ph'Gr??$ ityAj]
b,.ty'-
i'
. , . y V .
4 k
)
i i
I
\
l 7-2
-w=* - = v~* ~ * * *
- l
~ ~ _ . . . . .
'w ' . e t . is er , , _ _ .
- ,. .y ~ . . -
t l
t Appendix A Chronology of Review September 1973 WASH-1270, " Technical Report on Anticipated Transients Without Scram for Water-Cooled ,
Power Reactors," issued, j November 1973 Meeting between staff and GE - discussed schedule of ATWS analysis, ATWS problem i areas and staff implementation philsophy. j February 1974 Meeting between staff and GE - discussed GE progress on ATWS evaluation, especially their 5 calculational methods. i-June 1974 Letter sent from staff to GE requesting f additional information on NEDO-10349, "Analy- g sis of Anticipated Transients Without Scram." i July 1974 Meeting between staff and GE - discussed i possible ATWS " fixes." ~
August 1974 Meeting between staff and GE - discussed
- - f.
h M August 1974 Meeting between staff and GE - discussed ATWS events being analyzed by CE.
October 1974 Responses to staff questions on NEDO-10349
' received.
October 1974 GE report NEDO-20626, " Studies of BWR Designs of Mitigation of Anticipated Transients
. Without Scram," released.
~
December 1974 Letter sent from staff to GE requesting additional information on NEDO-10802,"Ana- g lytical Methods of Plant Transient Evalu-l ations for the General Electric Boiling Water ,
Reactor."
January 1975 Letter sent from staff to GE requesting additional information on NEDO-20626.
March 1975 Letter received from GE with schedule of responses to staff's January 1975 questions. ,
April 1975 Letter sent from staff to GE requesting
,,, s additional information on NEDO-20626.
A, W?.
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-. ..w,, a .- . ap c.
??:nb.N : -z : . :
^ fk 2 -b
?T ji3 April 1975 Letter received from GE with schedule of responses to staff's April 1975 questions. F
?' "./
- 4 April 1975 Partial response to staff questions on NEDO-10802 received. .
h.2..j e,fi) June 1975 Partial response to staff's January 1975 i.y51 questions received.
RQ2_ j July 1975 Partial response to staff's April 1975 questions 7-l received.
July 1975 Remainder of responses to staff questions on -
a,s f -! NEDO-10802 received.
W N ~ .' e I
e A
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, y e4 '
5 s
o s' ?
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A-2
- O -
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i e-APPENDIX B STAFF IDENPENDENT CALCULATION B-I. Staff Independent Calculations For the ATWS analysis, it was postulated that a GE BWR-4 reactor experienced a main steam isolation valve closure initiating a pressure excursion with the actuation of high pressure relief I
valves and the concurrent trip of two recirculation pumps. This l
~
hypothesized accident produces the most severe overpressure [
transient. Parameter studies were performed to gauge sensiti-l vities to variance in MSIV closure characteristics, void reac-
{
tivity feedback magnitude, pump trip setpoint and delay time, gap
- - conductivity, and relief capacity. These studies were performed e
, using Brookhaven National Laboratory modified RELAP 3-B computer i 1
program.
B-II. Model Description RELAP3-B obtains a time-dependent thermal and hydraulic descrip-tion of a reactor by intergrating a set of differential equations subject to certain algebraic relationships.
t A mass and energy balance is obtained from junction conditions, l heat output of reactor regions, heat removed by heat exchanger J' regions and heat added by pumping power. By using the equation-of-state of water (in the form of tab hs) a pressure and a
- . quality are obtained for each volume. Assumption of a linear !
B1 As
??
. CINE.
TEtikh M *" " " v "'" " 7 v T T m W v " m y yi y R 7 7 3 G' y v.3 7""L T y '
a, . -
. > ~
. , . +.. . . . .
~.:
'[ t
- .,;d$' density model for steam separation within each volume allows y, dide 4-
..p . : .] fluid conditions to be determined for each junction. A one- ,
q '
, dimensional momentum balance (flow calculation) is then obtained -
s ,,
djl.Y. for each junction. The fluid conditions within a volume, the Nr4mN N.N S$R,1 flow through a volume, and the reactor power level are then used
< to establish the reactor heat flux and the fuel element tempera-tures. Update modifications to RELAP3-B for this review were ,
,U
'9 developed at Brookhaven National Laboratory. These include:
, r .+ ,..
~)
y
/ :J' 1) Correction of the RELAP3-B steam tables (Prandtl number and =
L*i ~< !
conductivity tables).
"NQ , 2) Implementation of Hench-Levy CHF correlation.
Q' '
- 3) Substitution of void reactivity feedback for density e.
l:~ feedback.
ftid f gt,.m', 4) Implementation of a logic sequence to simulate relief valve ;\ _
4 <
behavior.
- .j
.x- } Figure B-1 shows the nodalization pattern used by the staff to 7
a , ;'
,7L simulate a generic GE Model 218 (BWR-4) reactor with RELAP3-B.
J'h The fueled section of the core is represented by a five node e a,p .
5E L%;c , (volumes 2 through 6) average channel assumed to reflect the 9e .,
N l average core behavior and a 13 node (volumes 15 through 27) hot
- +f,, j
, 6 >" channel. The two jet pump recirculation loops of the reactor are
% represented by one node. The heat removal system effects are 1 s' [ simulated by tabular input MSIV closure and feedwater flow rates.
Hysteresis and reaction delay times for relief valves are ;
simulated by a trip-and-restart logic.
)
B-2
. .g _ ..
B-III.a Specified Characteristics Some reactor characteristics are input specified according to vendor design. The following are included among those behaviors.
l B-2. The valve is assumed 100% open from the time of !
transient initiation until 0.8 second transient elapsed ;
time, and then for the next 2.0 seconds (until 2.8 seconds !
elapsed time) the valve ramps linearly to a position of 1.0%
open. It -then closes on a linear ramp in another 0.2 seconds (total 3.0 seconds elasped time to full closure). ,
MSIV flow rates for staff calculations are directly propor- *
- 5 fjl tional to MSIV position.
- 2) Relief /Safet'; Valve Operation Relief / Safety valves, closed at the start of the transient, open upon the attainment of the pressure of 1105 psia, and -
==
close again when the pressure drops below 1060 psia.
Associated with this valve action is a 0.5 second delay time
. and a 0.1 second closure / opening time. For Model 218 BWR-I-
4's, a relief valve capacity equal to 89% of design rated I core steam flow (Nucisar Boiler Rated, NBR) is specified.
This capacity was determined sufficient to satisfy pre-ATWS safety requirements and has been included here as a design
, specification (NEDO-20626-1, References 3, 4).
SS>
- 9 w= gh B-3 ' ;lg /
%., s y,-c
.u li 4Q,;;
- v ,me; e.e.ma.,s.. ... .a v . . . .m. .%,w,w-wetawas > "'MM
's ' L ' .
e
~:
~" -
n.e . m u M.L-W,
...,o
,A-
-A ' 3) Recirculation Pump Trip ,
. The recirculation pumps are set to trip at a main steamline *
.,;, {
pressure of 1165 psia with a 0.53 second delay time.
-l"%- #j
- 4) Cap Conductivity l]m?;[,[
s
'E An average rod gap conductivity of 1000 Btu /hr-ft *F was assumed, and the hot channel gap conductivity was 650 ,
2 2 I ' a, ,
Btu /hr-ft *F.
.c- ,;
', , :l 5) Void Reactivity ""
i
.y, For the base case a void reactivity curve corresponding to a
^ " dynamic void coefficient" of lic/% was used.
x, t
The basic void reactivity feedback curve used is shown in Figure B-3. The graph pictured is labeled 6.28c/% in ad-
~ ~
herence with the convention used in GE documentation and
' , J4 defined and specified by the vendor.2 The void coefficient may be derived from the curve by multiplying the design
, j e < operating void fraction (.4025) by the slope of the curve in its operating void range (.4025 x ($(6.4 - 4.85)/(50 - 40%]
'M .1 ;e
= $.0628/%).
%if T*
CE has calculated its most limiting void feedback (for AWS)
,.y ,
< to be that with the greatest negative value. This depen-
$. dence upon void feedback was confirmed in staff studies.
'_ (See Section B-IVC). For the AWS calculations, CE has selected a conservative value of .11c/%.
-I s
B-4 s' i
1
. y. . z .~ _. .
- 6) Feedwater Flow The feedwater flow used for most staff calculations was a function diminishing to zero flow in 9.5 seconds. GE cal- I culations used a feedwater flow which dropped to zero flow ,
I at 6 seconds (Figure B-7). This difference has a conserv- '
ative influence on the staff calculations; the use of the f
i 9.5 second feedwater termination rather than the 6.0 second -
I termination results in a 44 psi feed steamline pressure t increase.
B-III.b Sensitivity Studies In addition to the base case, the following parametric studies r.
./ jf g were performed:
Q, ,
- 1) Four different MSIV closure functions (See Figure B-4) l
- 2) 90% relief capacity
- 3) Pump trip setpoint of 1215 psia
- 4) Pump trip delay of 1.53 seconds Average rod gap conductivities of 600 Btu /hr-ft 2
- 5) -
- F and 2000 Btu /hr-ft2 , .7
- 6) Void reactivity curves corresponding to dynamic void co- 1 l
efficients of 4.58c/%, 6.28c/%, 8.8c/%, 11c/%, and 14.3c/%. 'l 9
Ni l D B-5 l
.n4'esA.,5
< l
y y
.<9j B-IV Calculational Results and Discussion .
- a. Main Steam Isolation Valve Closure
).
.- . i.d (described in Section B-1) g ',;d
'# y f Figures B-Sa through B-So present graphical results for selected 4
variables. For analysis of the transient, the following param-eter values were assigned:
(. -
,; . 4 1. MSIV closure - the GE supplied behavior was used (Figure B-gg j 2).
, d
? 2. Void reactivity feedback - an 11c/% curve was used (ref-
- 9. . ). '
erence 2).
- 3. An average rod gap coefficient of 1000 Btu /hr-ft *F was
'. used. /
- 4. Recirculation pump was tripped at 1165 psia with a 0.53 delay time.
.c 5. "89% capacity" was used. This is specified in NEDO-20626
,1 (Reference 3) as the normal full flow. In a sensitivity run this capacity was further reduced by another 10%.
P
.g ac Figures B-5a through B-So give predicted results for the base
- As * ' ' '
case. Power (Figure B-Sa) is seen to stay near its steady state
[ power level until about one second of elapsed time, when the
- pressure increase (Figure B-Sb) attributable to MSIV closure (Figure B-2) causes void collapse, inserting a large surge of positive void reactivity (Figure B-5g). This results in a power B-6 .
. . _ , . ~ . , , ,...n.. ,
_ - . = , - - - - - -- --- - . , - - - - -.
(.:_,g- _ . _ . . - _ . . .._ _ _ __ . - .
n rise. As the MSIV continues to close it presents a resistance to exiting fluid (Figure B-Si) and core flow. At about two seconds this increased resistance retards core flow (Figure B-5h) to such t a degree that core voiding increases which in turn terminates ,
I the power surge. The power continues to decrease until at three seconds the isolation valve completes its closure preventing ,
further steam escape. Again the power increases due to void collapse. This power rise is faster than the previous one - but <
i is terminated by the opening of pressure relief valves at 3.05 l seconds. The power falls quickly when the escaping steam causes
(
increased voiding (3.15 seconds). However, the same valve action f which causes the voiding also results in lower core flow re-sistance and 0.7 seconds later (3.85 seconds) core flow is seen to increase sweeping voids from the fueled region and cause another power spike. When the pressure exceeds 1165 psia, the recirculation pumps trip and, as the inlet flow decreases, core '
voiding again increases reversing the power trend. During this last power decline, with a reduced coolant mass unable to remove a sufficient amount of the energy inventory, core hot channel nodes 18 and 19 experience MCHFR's of less than 1.0 (5.05 seconds). A sharp decrease in heat flux at these nodes and a t-rise in peak cladding temperature is observed achieving its peak ,
of 1386*F at 8.4 seconds. Steamline pressure continues to rise e until it reaches a maximum of 1377 psia at 11.45 seconds.
M : a,-
B-7 M4 k,s mvMMW' MM 'r N'".?'T** """",""?? ? "." .1
+.
- $C ' { Y..
yR: Q: ~ >-
$ c.; B.IV.b. MSIV Closure Studies Six different MSIV behavior cases were analyzed. Five of these 3
_ g 7. ,
are shown in Figure B-4, and the sixth is the "GE" model (Figure t7.M:y B-2). The GE model was uesd for all other parameter studies.
- . " All MSIV sensitivity runs were made usiyg an 8.1c/% void reac-tivity feedback curve and a 7.15 second feedwater shutdown
' .M function. Calculational results are given in Table B-1. The
,' f) MSIV closures reflected in Figures B-2 and B-4 were analyzed to establish steamline pressure magnitude and phase response to the various behaviors. The following conclusions were drawn from the MSIV closure study.
The phasing of MSIV closure and feedwater shutdown is a strong , g..
4 1
factor in determining the severity of the pressure surge. Pressur Y- -'
is amplified by rapid valve closures and slow feedwater shut- ,.
f s .U downs. Also the greater the lag time between MSIV closure and fee'd water shutdown, the higher the peak pressure is reached.
! . This is readily ceen in the results for Models 3 and 4. In both ae' cases, feedwater shutdown began at 2.85 seconds and finished at
- 5
7.15 seconds. In Model 3 MSIV closure began at 2 seconds, 0.85 i seconds prior to feedwater shutdown; whereas in Model 4 MSIV action started at 0 time. The 68 psia difference in peak pressure t (Model 4: 1420 psia; Model 3: 1352 psia) is attributed to the 2 j second discrepancy in feedwater shutdown lag relative to MSIV closure.
)
J
! B-8 v- - - , , - - - - o
,.. ii . . - . __ _ =
+ % *64, - . - -
~
s i I
Comparing results from runs for Models 1 and 4, it is seen that I Model 4 calculations reflect a peak steamline pressure of 1420 psia and Model 1 yields 1357 psia. Both MSIV models initiate valve action at 0 time and have the same feedwater lag. It is noted however that in Model 4 the valves close more rapidly for the first second and in the span of the first 3 seconds provide 0.875 full relief flow seconds and in Model 1 the valves give 1.5 full relief flow seconds. The 63 psia difference in peak pres-f sure is attributable to the difference in rate of closure. The !
I "GE" valve closure model is seen to produce results conservative j with respect to the group average. It is also evident, however, i i
- y{ that this sensitivity study has demonstrated a range of 68 psi W'ia
~
- resulting from variance in MSIV behavior during the first 3 -
seconds of transient time, thus showing the importance of def-inition of this input operation. g B.IV.c Void Reactivity Six cases were run postulating the same conditions as those of the base case, but varying the value of void reactivity feedback.
Results for these studies are presented in Table B-1. It is !
noted that increasing the magnitude of the negative feedback amplified the severity of the transient. This is observed from the trend of higher steamline pressures (1365-1468 psia), higher (
peak clad temperatures (1200-1534*F), and decreased time to s attain MCHFR (6.14-3.54 sec.) for increasing magnitudes for 3"9ln
- "7 negative void feedback (4.58-14.3c/%).
43Gg B-9 ; ;
gg
,7 g-. , - , . m - ,3 y . ,..q g -
. : : c m. ,'
)
Le . ...
W, E! Q;
, /t .,
- r b}S
,a f '.
(s
.hE[7; B.IV.d Gap Coefficient The base case was varied in two sensitivity runs to demonstrate ,
3.:i, .
. 2 , ~ the influence of rod gap conductivity. The base case assumed au I N? I average rod gap conductivity of 1000 Btu /hr-ft *F. Analyses
','- were also assuming values of 600 and 2000 Btu /hr-ft 2 *F. A check
.. , e ~ . e
- rgr ;. g of Table 1 reveals that a decrease in gap conductivity hinders .
46 "
the rod's ability to transfer heat to the coolant thus reducing
^E:/
?
- "?;; void generation and shutdown reactivity. The 600 Btu /hr-ft2 """
- F
~
i l case therefore achieves the highest power, develops the highest steamline pressure, and reaches the highest peak clad tempera-ture. Accordingly the 2000 Btu /hr-f t *F run produces the least i
j n conservative excursion.
4 .
The peak calculated steamline pressure was decreased by 20 psi by.
- increasing the gap coefficient fram 600 to 2000 Btu /hr-ft *F.
- , B.IV.e Relief / Safety Valve Capacity The base case specified a relief valve flow capacity of 88.9% of
^t the design core steam flow. Decreasing this by an additional 10%
. ;, g reflected a marked influence. The peak normalized power increased
'.~7 e from 3.90 to 5 14, the peak steamline pressure increased from 1392 psia to 1348 psis, CHF decreased from 5.33 seconds to 4.8 .
seconds, and the peak clad temperature rose from 1295'F to 1510*F. An additional 90% relief run was made, revising the feedwater shutdown curve from the "9.5 second" function to the s
B-10
, . . .. .. ;, .. 4_ ,
-.3 . .g 4
GE - specified "6 recond" function. For this case, the peak steamline pressura wen 1483 psia and the peak cladding temper-ature was 1392*F B-IV.f Recirculation Pump Trip Sensitivity Runs were made assuming a 1215 psia recirculation pump trip setpoint (instead of the standard 1165 psia setting). In cases ,
using the "6 second" feedwater shutdown function this resulted in .
t calculation of a higher peak steamline pressure (1484 psia vs i 1377 psia) and a higher peak cladding temperature of 1554*F (vs i 1364 *F) . ;
i i
In another analysis, the recirculation pump trip delay time was increased frw 0.53 seconds to 1.53 seconds. This yieAded a peak steamlinc pra.ssure of 1470 psia (vs 1377 psia) and a peak clad-ding tempersture of 1554*F (vs 1364'F). 3, B.V CE Analytical Results ard Comparisou Po Staff Predictions The GE model (Figure B-6) represents all plant equipment simu-lated by the staff model (Figure B-1). Specified correlations used, however, differ particularly when describing natural phenomena such as void generation and disposition. Many char-acteristics, such an MSIV operation are the same in bot.h GE and staff models by design definition. Results of the GE analyses are presented in Figure B-7 and Table B-2.
e B-11 9
m%f
~. .
i *; . , ,
1% b
't?:l 7:. ; . ; $
- , w For the base case, GE has calculated a peak steamline pressure of
' > , 1370 psia; this is 7 psia lower than the staff calculated 1377 , C psia. This discrepancy is considered negligible and may be
- ' ~;,.
attributed to differences in void recctivity feedback and MSIV
.w. ;
flow during closure. The GE void model includes a void sweep
' ' .}i[y correlation, whereas the staff uses a homogenous flow model. ,
.j n;;.C )
r AiV
?- Though the magnitude of influence this exerts on computed results T' .~. 'KE.Q(
r J. l f3'"l is as yet unascertained, calculational differences are produced. """
.. 4 The GE MSIV treatment, because of the intact steam leg, can assign a functional flow versus valve position dependence. In j the staff analyses, the dependence was assumed to be one to one i
(i.e., 60% opening - 60% flow, etc.). It is thought that this du
- produced MSIV flow disagreements. TheMSIVstudy(SectionB-IVb). _
i demonstrates the steamline pressure sensitivity to MSIV behavior, and its chronological relation to feedwater shutdown. Because i i the actual difference in MSIV flow between staff and GE calcu-1 i lations is not known, the resultant calculational discrepancy has
]
- not been ascertained. In NEDO-20626, GE reports boiling tran-
- f. sition at about 4 seconds. Staff calculations show this event at
- i. around 5 seconds.
l i 1 I l B-12 i he v. --.- , . , -*-ma -
. = ,
. g i i
~
.. a-
- t g- - ... ,
g N TABLE 1 j i j COMPARISON OF KEY MODEL BASES i k f
;; GE Staff '
Void Reactivity Void Dependent Void Fraction-Dependent Polynomial IIeat Removal Turbine Performance Simulation Indirect . Lumped into MSIV and Feedwater Properties MSIV Flow Functionally Dependent on Valve Proportionally Dependent on Valve
; Position; Valve Position Time Position; Valve Position Time , , Dependent Dependent 'E MSIV Closure Time 3 Seconds (Incl.) (Time Ramp, 0.8 3 Seconds (Incl.) (Time Ramp, , Second Delay) 0.8 Sec. Delay)
Core Average Channel 1 5 jj Axial Nodes i Core Hot Channel N.A. (Separate Code) 13 2
-; Axial Nodes t
[ ; Void Model Steady State-Transient Core Void Map-Void Sweep Model Homogeneous , f Gap Conductance Fixed Variable Variable - Temperature Dependent Turbine Trip Immediate Immediate Relief Valve Setting (psig) 1080 to 1100 + 1% 1090 { Recire Pump 1150 psig 1150 psig 7 Trip Setting n (k I-c- f"
? ?L p ,
- y
- g /- 1: .5 .
i
^
, . f:
l i N
.q h '.. 4 7, ) !
- . ';; si I,' ,
TABLE 1 (Cont'd)
.{
v. GE Staff Pressure Control Static Pressure Regulatory Input Steady-State - with System Overridden in Transients Transient Reaction k Feedwater Control System Functionally Dependent Tabular Input Power Generation Model Point Kinetics Point Kinetics i , f 7 A B r. g e N-*
* -e - ,W p+=- p= -- 9 % = men ggangp> ws-m -MeMe ew- r g e- f '=%+4 , e
- M=E
,' . ); , ' ',' [ 3 r* . :s '.,
. g.
t-t- ~' ~ [:y + ..
#;;- ; ' _,' T,. ..
YI er i ..#,s
- q. p, t r c. ,
~,
- d*5 -
e .g -
. . - 4, 's TABLE B-1 STAFF RESULTS FOR MAIN STEAM ISOLATION VALVE CLOSURE ACCIDENT Peak Peak Peak Steamline Claddirg Normalized Pressure, Temperature, Power PSIA Time of "F Case (time) (time) CHF, Sec. (time)
Base 5.11 (3.15) 1377 (11.45) 5.05 1364 (8.9) Valve The following Cases Used "7.15 Second" Feedwater Function Model 1 2.685 (2.69) 1357 (11.25) 5.92 1286 (8.6) . Model 2 3.10 (3.37) 1379 (12.25) 5.52 1290 (9.25) Model 3 5.91 (3.15) 1352 (12.0) 5.80 1216 (8.9) - Model 4 3.68 (1.25) 1420 (12.25) 4.27 1306 (8.33) ; Model 5 -
. 1384 (11.25) - - -
GE 3.09 (3.18) 1392 (12.25) 5.33 1295 (9.0) Void h Reactivity The Following Cases Used "9.5 Second" Feedwater Function
-.~.
~, .
3;;2 4.58 2.95 (2.9) 1365 (12.45) 6.14 1200 (10.25)
"2Y 6.28 l 2.86 (2.54) 1390 (12.35) 5.61 1213 (9.0) g 8.01 3.90 (3.18) 1392 (12.25) 5.33 1295 (9.0) ;
11.0 5.13 (3.15) 1421 (12.5) 5.05 1387 (8.9) g 14.3 8.30 (3.00) 1468 (13.05) 3.54 1534 (9.55) Gap Coefficient , f~ _ 600 5.30 (3.21) 1434 (12.65) 3.56 1440 (7.625) 2000 4.28 (4.45) 1413 (12.25) 5.05 1309.5 (9.25) Pump Trip
- 1.53 Delay 8.65 (4.65) 1530 (14.35) 4.80 1579 (8.90) 1215 Set 8.66 (4.45) - -
5.05 1309.5 (9.25) ,. 1 Relief 6-90% 5.14 (3.5) 1548 (16.1) 4.80 1510 (10.25)
$5B
~ . . < - , , .-..>.. u. v,w e d . d. t
.e , .
,_5
. . . ~,
TABLE B-la 0 ' '<
; STAFF RESULTS FOR .
i CASES RERUN i USING "GE" FEEDWATER FUNCTION ("6 Second") .
'L.
Peak Peak Peak Steamline Cladding W Normalized Pressure, Temperature,
/ 'S Power PSIA Time of 'F Case (time) (time) CHF, Sec. (time)
- l N '
Base 5.11 (3.15) 1377 (11.45) 5.05 1364 (8.9) -
-, ; k!'-
f/Ald N' Pump Trip 1.53 delay 7.85 (4.70) 1470 (13.05) 4.80 M^c ' , ' 1554 (7.7) . 1215 SET 7.86 (3.15) 1484 (13.30) 4.80 1557 (8.3) 90% Relief 5.10 (3.15) 1483 (14.35) 4.80 1392 (8.9) 6 _s
+
J-i I i l f l I
-4 I
i j 5
/
= - - - - - -
_u . - c; . . . - ., -. - . . -. TABLE B-2
\
GE ATWS RESULTS - MAIN STEAM ISOLATION VALVE CLOSURE Peak Steamline Peak Pressure Case Power Psia MCHFR PCT Base 9.31 1370 - - Cap Coefficient - - - - Pump Trip l P 1.53 Delay - 1420 - - i 1215 Set - 1385 - - 4 Relief 90% - 1450 - - I
- ,b',;:)
q, 8 y. 3 1 t . . I I-e e . ! OS m+ bh
. r,c , . u. . - ~ , . . , , .- _- ,_._ . . .
.. c s. e .....,_
, m . w %
.....+ .-.
t Vol.10. a h A. " *l.30 Ja steueF YALVE VoI. 9.J.ll.a
% ,, ,,aa ,_,
20NE
" "~ * " ' ' * " *
%I. F.'*
n .- 1 u. v a 1 '- si.2r ,he
%I. G." 2i ~
wi.25. 4: - Vol.I;.
'"~
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N$$ el2 w-" $3: - wi. 22. 2:= " wl.2e - .
**""4 %I.4." %I.7 vo .2d. >!= e,ir. I~
ca.z o
- v. . is. 4:= a.,/ ,.
8ve== wi.is.2:a c x==
%l. 3. "- , wi. i7. .
\
- - i<
J * ,,
%I.16 , Pump
%I. 2. ';*- wus % I 14 fl.
o 7 a n. . J.f. 17. f J. 20. g
'~
[(ws.w-av) *W Het- f Vol. 12. W.1. . LOWER PLENUM d l F lIOC h
.yg rA15- 4fbmP h War ;
i FIGURE B-1. BWR-4 ANALYTICAL MODEL DIAGRAM e n M, awn
""" ~ T m m " w?w m p:~ :,= y;_g
m1 3-J a a
~ l
[, 1 1 10 0-n s i n 3 t#
$', g
, .1 i E S. -
Z
,- , g ,. l
.~.. <
l O i 1
..v6e l ,,?; r;q ww.,
g , Q,' f e If
$ * ^ .% l
~,-
0 l b' : ? . oc i - T
.]
g , l
- a. .:n. ;.;~ ,-; )i
.. w n
~ .~ ' ' ] l O l C -
s i i . , .2. i 3 I . (,, w, e i I I s l . 1 t L1J J i
-l 1 I
I 0 b=======~~===.=====~====================
~
- I . a i 0.0 * * '
- l.0 2.0 2!t s
= =
IME (8ECONDS) FIGURE B-2. MSIV BEHAVIOR (G.E. MODEL-BASE CASE)
\
J f w... . . . _
~,-..----...--.n.-.. --
. . . . a. . _ _. .
.' .,. j . .;4 0.F 5h ki,_'_ bc - - -
..). pL.' .. ?.!
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-l-l
- l,yi -l I-l .
.p
. . . '.q.; ;;.l. .. ..i- ;- - l.{: - -
. i .
l.
. . . . . . .. 3. .i. - .r 7 .- rr- , ;.. . -
T ;. 4. i .. . . .H.H. th. ..i . N 1+- . +H -.v.,+4 - - l . p,.. l.. , , . -- t !,* .. -
~. U . 8 F.- -} p- j - ,Jlp ,l a,--
[J
. . .i.- 1.[ [.. .l.! .i.i
_. . I' 4 . t. l .{. . .l. L . j--
.}.
, - {- b l --m; 1--{ r-{ , .- {. F---
.l'_]. jl.
_+.. .]'. L l'.;. ,_ . ..
.. 7.
._,. .L i. q . . .I. .i 'rT-,rr C* _Lt. .,.i. .} .. 8. . 18-
- 1. .j . .. ;i. 4 - b r.g '-
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REFERENCES M% N t = i
- 1. REI1P-3B, Manual, A Reactor Systems Transient Code, Brookhaven National Laboracory, December 1974 .
..-s,
.~1.{f ' ', 2. Telephone Conversation, R. Wilson (GE) to A. Thadani (NRC),
June 9 and June 11, 1975. .
~,:
- 3. Topical Report: Studies of Design for Mitigation of Anticipated Transients Without Scram," NEDO-20626, the General Electric Co.,
' ' October 1974.
- . J
$IN 4. NEDO-20626, Amendment 1, June 1975 m
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)} Return Line Nozzle Cracking a
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~ BIBLIOGRAPHIC DATA SHEET -
NUREG-0312 4.TiTLa Amo suet Tta sean van == ==.ams 2. A a mens . Interin Technical Report on BWR Feedwater add Control ' :f
' ~' "T- '
;- T Rod Drive Return Line Nozzle Crackin9- .
1 REcmenn Accessics NO. . [
,7. AUTHOR 51 E
- 6. DATE REPORT COMPLETED Moes7:e l YEAR
.luly 1977
- 9. l'ERFORMING ORGANIZATION NAME AND MAILING ADORESS pac 44 EW Caser)
DATE MPORT ISSUfD asoseTM Division of Operating Reactors l YEAR .
- Office of Nuclaar Reactor Regulation July 1977 U
- s. a an=*i - :-
U.S. Nuclear Regulatory Comission Washington, D. C. Fjd e. Aaa aa=*J i2. sponsoMina oRoANizaTiou u.a AND MAluNG ADDMSS pne&W E4s Coel W as No. 9 ii. cowrRAcT wo. . -E 11 TYPE OF MPORT PtReco covEaEO mesheds meerl
- Technical Report is, men.sanENTARY NOTES 14 A mo esme) ^l is. AssTnAcT see === = w .
~:
- Recent reactor operating experience has revealed cracking of the inner surface of BWR fee &seter and control rod drive return line noules. The '
cracks are thermal fatigue initiated and in some instances have propagated through the cladding into the base metal. This report provides guidance 2 as to interia inspection criteria and procedural measures for operating reactors. a
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- I
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1 ' b' i i Manum.ript Comshted: July 1977 Date Pubirnbed. July 19?? I 1 t l I l-0
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. .?. ART I -: FEEDWATER N0ZZLE CRACKING . . . . . . . . . . . . . . . . .
.s 11 ,
9 3j, 3C ~
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- 1. 0 . IN T RO DUCT ION . . . . . . . . . . . . . . . . . . . . - . . . . . 1
, A I,
2.0 . STATEMENT OF PROBLEM . . . ....... .:.. . ....... 1 I
/ -3.0 CAUSE OF PROBLEM .
l.. . . . . . . . . . . . . . . . . . . . . . . 2- 2
'k:
~
- 4.0 ULTRASCNIC INV T!0N OF FEEDWATER N0ZZLE INNER RADIUS' . . . . . 3-5.0 INTERIM PROCE' .it MEASURES 'FOR OPERATING PEACTORS ....... 4
.'l-6;0 INTERIM CRITERIA FOR BWR FEEDWATEd N0ZZLE INSPECTION . .. . . . 5 h' ,
k i
! 7.0 RECOMEN DED ACT IONS . . . . . . . . . . . . . . . . . . . . . . . 11 ,
8.0 LONG TERM RESOLUTION OF PROBLEM . . . . . . . . . . - . . . . . . 13
. A TT ACHMr.NT.1 .
. . . . . . . . . . . . . . . . . . . . . . . . 17 t-A T T ACh ME N T 2 ~ . . . . . . . . . . . . . . . . . . . . . . . . . 18 ,
f ATTACHMLNT 3 . ... . ...... . . . . . . . . . . . . 19 -
'P*RT II - C:1000L w03 DahE :ET:!O ' hE N0llLE rRACKING . . . . . . 21 1.0 l'i?c0 M *:04 .
............. ... ... . . . 22 2.0 S'ATENN! Cr mi p:08 TEM ....... . . . . . . . . . . 22 3A C?.uSE Cr !"E N St.iM .......... . . . . . . . . . . 23 3.0 :E T OV . . ........... . . . . . . . . . . . . . . 24
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' ,! FEE 04ATER N0ZZLE CRACKING 1
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- g , s. ~ , .Of tM ~ experience has revealed significant degradation [
l^- 9 (cracking)'of.the inner surface of BWR fe.edwater nozzles. While the !p problems of feedwater nozzle cracking have not been completely re-t s- solved, the NRC staff believes that sufficient information is available
}[
w ;-
.~ from various sources to warrant a discussion of the elements of the Q problem and of potential remedial measures, both long and short tern. '
l1 , which are under consideration. The purpose of this report is to p susunarize this problem area and to present an interim staff position C regarding this generic issue. J ' j It should, of course, be recognized that information on this subject {i 1 is continually being generated, hence the staff position and cosaments l ; are subject to further modification. In thi, regard, we have requested 1 s licensees and the General Electric Company to inform us promptly of y j any information regartling this subject that results from on-going e . y; . , programs and related experience. The following discetssion is based primarily on information suppifed f** to the NRC by the General Electric Company, and on the staff's case-F by-case reviews of feedwater nozzle inspection results from a number f, of operating EWR facilities. The staff has prepared, as part of this { report, an interim inservice inspection position to ensure an appro- ] , priate conservattw treatment of this potential proelen at operating E facilities untti a long tern solution is developed. t . 2
.6 2.0 STATEMENT OF THE PROBLDI The feedmeter nozzles of essentially all operating BWt'sV which have f
been f r.spected to date havs been found to have blend redivs cracks. l some of which propagated through the cladding into the base metal. ; In several reactors, similar cracks wre found in the mzzle bore. If Cracks have been observed at all operating BMt's inspected to date with the exception of Browns Ferry 2. unich has had less
'han one year of operation.
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. - f The deepest. cracks found to date were in the nozzle bore and s P:i were of a total depth of about one and one-half inches. Examples h -QA i
of such cracking are given in Attachment 1. Analyses by the NRC .-J 1 g + staff, which are in agreement with those done by GE and field [a, data,from operating BWRs, indicate that the initial crack growth j j f_' rate is high up to crack depths of about 1/4 to 1/2 in. Further growth is slow but would accelerate with increasing depth. Even-J A tually, the cracks present a repair problem if, in removing them l g[ by grinding, the ASME Code limits on nozzle reinforcement were _1 exceeded. The crack depth equated with the reinforcement Ifmit will depend on the details of nozzle dimensions (see N8-3330 in f
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Sect. III, ASME Code). [ t Feedwater nczzle cracks are of concern to the NRC staff because: (1) reactor pressure vessel integrity is considered extremely y important to safety. (2) there are uncertainties about the rete i at which the cracks are growing, (3) current nozzle repair pre- 3 cedves require that cracks be ground out thus removing metal p- 3 3 from a relatively high stressed region of the reactor vessel, - g and (4) considerable radiation exposure is received by personnel - g " performing inspections of the nozzle region and repairing crocks _ in the nozzles. Although such cracking of the pressure vessel y ~ [ nozzles is incertant to safety, the I.RC staff believes l - i i that cracking that has penetrated the vessel cladding will grow at a slow enough rete such that the cracking does not pose a j f : criti.:a1 safety concern today that terrants ismediate action. { Rather, the staff believes that sufficient time is avellable, due k to the conservative design of the reactor pressure vessel, to i
- perwit continued operation of the affected facilities eAtle studies -
m t **ta aunt s ennt t eve on an capedited sc$edu'e. - r . . 3.0 CA'E E Of fME_fRORE3 - { The WRC s ta
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,i nozzle region during periods of low feedwater temperature when the
[ flow may also be unsteady and perhaps intermittent. Attacnment 2 E
! is a trace of such temperature fluctuations as they were measured in
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a mock-up of a typical feedwater nozzle. Once initiated, the cracks I i! are believed to be driven deeper by the larger, relatively low )! frequency, startup/ shutdown pressure and thermal cycles. The latter
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1 i ; result from significant changes in feedwater temperature during l .. . flood-up of the reactor vessel and when feedwater heaters are put ['
., into, or taken out of, service. During nonnel power operation,
[ j the plant feedwater heaters maintain the feedweter temperature at l l about 180*F below the reactor water tamperature. At low power. l { when the feedwater heaters are not in service, the temperature I 5 I differential can be 400*F or more. We believe that the basic cause of the thermal fatigue cracking problem is this relatively large ( tempe% ore differential between cold incoming feedwater and the [' ' hot reactor vessel water during low power and flood-te operations.
- y. 4.0 ULTRA 50ll!C titSPECTI0ft$ 0F FEEINATER an?RE IleIER RADIUS f- '
A number of ultrasonic (UT) emanination techniques presently are used h to inspect the feedmeter nozzle inner radius from outside the vessel, h l#ille these inspection methods are useful, their current reltability is limited det to the unleve character and location of m the thermal fatigue cracks. E 1 Sased on our review of the avellette field esaminetton results, we conclude that the UT sothods. when applied to nozzle geometry, have P not demonstrsted a level of evitaotlity that new1d allow UT to be $ used as a sole bests for a decision to pewit continued reector vessel 3 operation. To taspesse conf tesace in tMs authod. == ewourage the k E E E E F E E F l-gg __ _ _ _ _ _ . _ _ _ _ . . _ _ _ . . - - - -- - - - -
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j cJ nozzle inner radius examinations. Should future developments and -
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w examination results demonstrate the UT techniques reliably and i consistently detect thermal fatigue cracks in the nozzle region. ;g
, these techniques could then be used as a basis for modifying the Interim Criteria discussed balow.
g 5.0 INTERIM PROCEDURAL MEASURES FOR OPERATING REACTORS Because of the current incoglete status of studies and design ' " efforts to resolve the nozzle cracking issue and because hardware y changes and other long term remedial measures will require consider- 5 able time to implement at operating facilities, certain interim _1 revisions in operational practice are desirable. 3 c In general, the NRC staff has concluded that IMt facility operators ? l should moni*or feedwater temperature and flow during low power a operation. In addition, operating procedures should be revised to - minimize rapid changes in feeheter flow and/or temperature, to i minimize the duration of cold feeduster injection, to avoid condi- .' tions that may lead to inadvertent or unnecessary high pressure coolant injection (HPCI) system actuation, and to avoid the :ntro-duction of cold water from the reactor cleanup system. Reactor p operators should attempt to Itait the temperature differential = between water entering the feedwater nozzles and the reactor vessel water to no greater than the norinal differential at full power. - They should also avoid feedwater tesperature transients to the exte.7t practicable. It has been demonstrated that by carefully
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f bringing feedwater heaters into service, the maptitude of feedmeter a temperature transients can be significantly reduced. While these sttps are not espected to eliminate the nozzle cracking problem. we , believe that they should help to minimize the extent of cracking until pemenent changes are mode. ' The F sta " has
- eld numerous discussions with GE and with 19 censees on a c a se - t -r a te t.a s i s ' o on s i de r this issue. In geceral licensees y i
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have been implementing the BWR Feedwater Nozzle Inspection interim W9 ' criteria (Section 6. below) for more than a year and have adopted j many of the staff's reconnendations for minimizing the cold water a;
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ya flow through feedwater nozzles. There are extensive developmental ; i 4 efforts to perfect ultrasonic nondestructive examination procedures v
~ #
underway at GE and several utility groups. Cladding removal is underway in three reactor vessels and other utilities are planning i
; similar action. In sunnary, the staff believes that the feedwater nozzle cracking problem is being mitigated satisfactorily on an interim basis while long term remedies are being developed. Attachment 3 is a table sunnarizing feedwater nozzle cracking experience to date.
6.0 INTERIM CRITERIA FOR BWR FEEDWATER N0ZZLE INSPECTION 6.1 Introduction h' Analyses performed by the NRC staff and by GE indicate that thermal fatigue cracks in feedwater nozzles can, initiate and grow rather rapidly to dr.lchs of 1/4 to 1/2 inches primarily _ due to water temperature fluctuations in the vicinity of the [ nozzle during operations with unheated fee &eter flow. This ( growth can be experienced within the first fuel cycle of ~ operation. Further g.mswth of these cracks is at a slower _ calculated rate. The results of feedmater nozzle inspections reported to date appear to confirm the analytical predictions. $ There are uncertainties associated with the analyses- opera.- # tions and inspections however, as indicated by some discrep. 2
=
.l ancies between the inspection results from one facility .
to another or even between nozzles of the same reactor. The ,
; objective of this inspection program is to ensure that no crscks grow to a depth where t5ey beccue safety significant or where (l
the repair procedures to elisinate them would pose a problem. Although a detailed review of the results of fee & ster nozzle . in'.pection and repair at the seny facilities (see Attachment 3 J H d
,.,,.i..._..---
for a summary), domestic and foreign, which have taken such a actions would be beyond the scope e f this docurrent, the i activities at two plants should be mentioned.
.~ W The Niagra Mohawk Power Conpany, Nine Mile Point (NMP), facility 2 l and t'te Jersey Central Power & Light Company, Oyster Creek '
fac' lity, were among the first BWR plants to go into operation - (both in 1969). As such, they have accumulated a relatively '
=
l large number of startup/ shutdown cyclesU (both ateut 100) and ' have " sister" reactor oressure vessels. Using a machine designed . for the specific task, both utilities will remove the feedwater ' n nozzle stainless steel cladding duri..g the 1977 refueling outages. j As of old-June, the four nozzles at Nine Mile Point (NMP) were - finished and the job at Oyster Creek vas underway. After mechin- -- ing at W. penetrant testing revealed 5 crack-like indications - S on one nozzle. one indication on an adjacent nozzle and none on the otner two. The depth of local grindout required to re-move the single indication and the deepest of the other 5 was ' about 1-1/2 inches; the length and width of the oval (region of . contour grinding) were about 9-1/2 in, and 4 tr.., respectively. The four nozzles at the Oyster Creek facility were penetrant 7 tested before machining and although the degree of cracking . - varies f a not;.le to nozzle, the longest were of the same magni-j tude as those at le@. We therefore expect the final grindout to ' be approximately the same as for W. The removal of such a ] large amount of meterial from the pressure vessel makes an 4 analysis of the reconfiguration, relative to the t.SME Soller _y and Pressure Vessel Code, mandatory. i _ For the W reactor, the appitcable Code for nozzle reinforce-rent is Section III, article NB-3330. Detailed calculations have shown that excess reinforcemect rvmained at the deepest j local grindout, using a coctervative method of calculation. - 7 [f the change in reac tor thermi power f rom noreina l l [ 3 zero to i 5 oper a t 1.v'a i i c ve l a nd re t u rn t o tem - nee Sec t . 6 '
. t,e'aw.
1 1 d E 7
m 7 Section II! of the Code also requires a fatigue evaluation - hich was perfonned in a conservative manner for the deepest grindout showing that there was adequate margin between the resulting calculated fatigue life (to crack initiation) and the vessel design life of 40 calender years. Based on past ! ~ crack propagation analyses, an undetected crack remaining in ! the nozzle after rework will not grow significantly during the following fuel cycle. The adequacy of any other such clad removal machining operations will be evaluated similarly to that for NMp. The NRC staff has considered a number of alternative approaches j' for monitoring and limiting the growth of feedwater nozzle cracks in operating BWR's during the interim period while a long tern = solution is being developed. On November 19, 1976, the General , - Electric Company issued a Feedwater Nozzle Interim Examination Pecomendation (FMIER) as Service Information Letter (SIL) - No. 207. This doctment, in effect, measures service time in terms of the number of startup-shutdown cycles. The staff also - examined service time as measured in terms of the number of hours that the reactor water is hot while the incoming feeduster . is cold, that is, the duration of unheated feedwater flow. We 4 believe that this is the mode of operation during which cracks c
~
are initiated and grow to the order of 1/4 in. in depth, hence, the number of cold feedwater hours i? an appropriate paraineter . J-for at least the first fuel cycle. Subsequent crack groir.h ! appears to be more closely related to the thermal and pret sure 3 stress cycles associated with startup and shutdoom. Althot.@ ~ there is not necessarily a direct relationship between cold 4 - feedwater hours and startup/ shutdown cycles. these approaches art not entirely inconsistent and, in view of the statistical nature of the inspection results to date, we find the approach in SIL No. 207 to be generally acceptable. . I I d
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Until other procedures have been qualified .. toLthe~
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of the NRC, we have concluded that IMt' feedwate: nozzles should' be inspected in accordance with the program sat forth in Section 6.3 below. The NRC staff approach generally agrees with the inspection frequencies and actions recomended by GE; how- o ever, in some cases the staff's conclusions are based on the . i number of hours of cold feedwater ficw as well as on the number of startup/ shutdown cycles. Therefore, licensees should Feep t adequate records of these parameters. In addition, licensees I should provide temporary instrumentation to monitor detailed ~ feedwater temperature and flow during at least several startup , and shutdowns. The NRC recossends that ultrasonic (UT) pro- N cedures be used in conjunction with dye penetrant (PT) testing . to the extent practical to expedite their develepsent. 6.2 Background Infr.rmation
- The recomended plan, described below, is appitcable to all WRs with feeduster nozzles that do not have the thermal sleeve welded to the nozzle safe end and for plants that went critical -
after 1968. Other IMts will be evaluated on a case-by-case , basis by tne NRC staff. In the context of this document a startup/ shutdown cycle is .
^
I defined as a reactor thermal poor increase free nominally zero , and subsequent return to zero which produces both pressure and I temperature changes and involves the addition of any amount of . 2 3 cold feedwater through the feedwater nozzles. UT refers to ; ultrasonic inspection performed from outside the reactor vessel. PT refers to liquid penetrant inspection perforsed from inside '. - the reactor vessel. '- 6.3 Inspection Program . g The tollowing is an outline of the piocedures for performiag tne required inspection program. - y n 2 e E g ha h a hm
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. :1. ' At each scheduled refueling outage, perform an external ~~ i .. a J.W -
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UT examination of all feedwater nozzle safe' ends, bores, ' ' T~' ';_ - and inside blend raiti. If indicationsare found in the ' 5 safe end, evaluate per ASME Section AI. If reportable ! indications are found in the nozzle bore or the nozzle corner, proceed with the sparger removal PT inspection and d w I repair called for in item 3 below. -j
- 2. Determine from plant records the number of startup/ shutdown cycles for the reactor.
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- 3. . The first feedwater nozzle inspection should be performed after about 50 startup/ shutdown cycles but prict to 70 -
h cycles. The followirg should be performed: -h 4 (a) externally examine by UT all feedwater nozzle blend .
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radii, nozzle bores, and safe ends; (b) remove a sparger , from one nozzle, flapper wheel grind and PT examine both , 2 the nozzle for the removed sparger and accessible portions a of the other nozzles. If any cracks are detected, remove all spargers and completely examine all nozzles. Remove
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all nozzle cracks. ,
- 4. For those plants where the feeduster nozzles have been PT examined but the flapper wheel cleaning or removal of all i detected cracks was not performed, the nozzles should be 3 reinspected as per item 3 above at the next scheduled re- ;
fueling outage. g
- 5. For those plants share the feeduster nozzles were PT examined
-d per GE Fleid Disposition Instruction (FDI) recousandations, j and all detected crocks were removed, subsegeset PT esamin-ations of the nozzles should be performed at the earlier of: j (a) every other scheduled refueling outage, or (b) at the scheduled refueling outage after 20 but prior to 40 startup/ -
shutdom cycles after the last PT examinction. 5 2 3 ll i
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m$.0 r6EIf Mhe feedwater spargers have forged tees,' the PT i i a examination may consist of flapper wheel cleaning and ~ ~ ! PT enmination of the accessible portions of the nozzle. If the feedwater spargers do not have forged tees, a j sparger should be removed fror one nozzle. Then that
~ nozzle and the accessible portions of- the other feedwater i nozzles should be flapper wheel cleaned and PT examined.
If any nozzle cracks are found, remove all spargers, clean. - ; q examine and repair the nozzles, i In addition to the chove procedure, the depth of cracks that i
'senetrate into the base metal or that are in excess cf 1/4
'nch deep should be measured and recorded and a record should ,
! be made of the circumferential and axisi position of each crack.
t !- e The sum of the total depths of all cracks that penetrated into I 1 base metal or exceeded 1/4 inch should be determined as well j i as the clad depth of several locations. If any crack exceeds l 3/4 inches total depth or if any crack penetrates deeper than 1/2 inches into base metal, a safety analysis report which f includes a discussion of the proposed repair procedure should he submitted to the NRC fbe review and approval prior to further f i action. I Inspection results should be communicated (orally or in writing) to the NRC as soon as practicable after results are obtained. Such an approach will better insure that the NRC staff can respond to ifcensees' requests in a timely sanger. Spargers and therunt sleeves should not be re-inserted in the vessel untti the results of the inspection have been discussed with the NRC. In addition to prompt transmittal of inspection results. Iicensees are requested to coordinate their planning for feedmeter nozzle r
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yh , as;+ [ , inspections with the NRC as early as practicable' to' minimize - y , ~ downtime and potential questions as to NRC requirements. ] ? In determining the inspection frequency of a specific facility, _ the NRC staff will also consider any remedial measures a licensee may previously have taken to mitigate the feedwater nozzie g cracking problem such as: '
. Measures to reduce significantly the duration of periods of ! .
1 low feedwater temperatures; =
. Significant reduction of maximum AT betwecn reactor water and -
incoming feedwater by system modifiestions and/or operating
- procedure changes; '
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. Appropriate and approved design modifications to the nozzle- '
l thermal sleeve-sparger region that shield the nozzle bore i h and blend radius surfaces from significant coolant temperature l d r transisats; l N. . Repairs that result in a more highly fatigue resistant nozzle -- N surface condition. as demonstrated by analysis and testing; and/or ' d
. Other remedial measure that mintsize thermal fatigue cracking. l [- 2 i
i 7.0 RECol W NDED ACTIONS 2
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! 7.1 Feeduster Nozzle inspection I -3 l }
In view of the status of the overall BWR feehater nozzle crack-Ing experiences to date, we have concluded that it is appropriate ' 4 and necessary to thoroughly inspect such nozzles unen affected nuclear facilities aen shut done for refueling. Accordingly.
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- f licensees should submit their proposed inspection plans to the NRC. Including the mder of nozzles to be inspected, inspection -
technique (s) and acceptance criteria to be utilized, methods of 7 1 nozzle surface cleaning, planned actions if UT or PT indications h
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h: unheated feedwater flow, and operating transients / conditions experienced which are gemene to the nozzle cracking phenomenon l such as inadvertent HPCI initiation or unstable feeduster flow (', control. These should be submitted at least 90 days prior to the projected start of the reactor refueling outage. < 7.2 Occupational Radiation Exoosure I 10 CFR Part 20.1(c) states that licensees should make every reasonable effort to keep radiation exposures "as low as is reasonably achievable" (ALARA). The inspection and repair - mector vessel nozzle crack. has a potential for signific
, , occupational radiation exposures because of the high red ion
,rg levels in the work amas and m1atively long stay time r uimd
[ to perform the necessary work. - Consequently licensees are requested to provide a description of the plans and procedures that will be implemented to keep . radiation exposures ALARA during proposed nozzle-related work. ; l The description should address the following areas: 1 (1) Training programs, including use of mock-ups, which will , I be used before beginning the actual mpair to ensure minimum stay times for completion of the job. (2) Special tools whica will minimize personnel stay times. (3) Shiveing used to siduce radiation levels. (4) Use of decontamination (such as hydrolasirq) to reduce radiation levels. i 4
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. - ; 1 93 g Upon completior of the nozzle related work. licensees are re . g N quested to provide a description of the experience so that other p$,
fd licensees can benefit from it in planning their ALARA programs. I~ 1 The description should include the following: (1) Dose rate information in critical areas before and after decontamination; shielding installations and their efficacy, f (2) Numbers of workers involved in the entire operation. d
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(3) Total man-rem exposure for the operation and man-ren break-A down by specific phases and by occupation, if available. j
. J 8.0 LONG TERM RESOLUTION OF THE PRO 8LEM %
The above staff interim criteria are primarily meant to establish ; a basis for continued plant operation during the time required to ji
~
develop and implement long ters solutions to this problem. As further f information becomes available the staff will continue its review in M this area and will issue final criteria at the apprcpriate time. M x 4 The ultimate remedy which will preclude feedwater nozzle cracking may ; y require a codination of individual measures to eliminate the sevem thermal transients or make tu nozzle less vulnerable to them. Such measures could include reduction of the fee & ster to reactor water ! $g touperature differential during low power operation, an improved
- h therumi sleeve-sparger design to reduce bypass flow which exposes "
the nozzle surface to fluctuating water temperatures, and removal of 3 clad from the nozzle surface which is believed to provide a surface l g more resistant to fatigue cracking. ;," Ra%ction of the feedwater-to-reactor water temperature differential say require bcth system redesign and operational changes to eliminate 1 unheated feediater in the sparger-cazzle region during low power and f other operations e the main feeduster heaters are not ir service. I I e i U {
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Since water passing through the:feedwater nazzle!can originate from E sources other than the feedwater system, these sources-(such as the! ' , - (lg{M Q b ~ eactor water cleanup system) may also need to be examined.- One ~j w particular source of cold water worthy of mention is the previously [pm; . a mentioned high pressure coolant injection system. An , aptable method of eliminating this source of cold water has not been estab-Nj gj m. - lished. ~ '
}- i Because of the importance of this issue, the nuclear industry will !- 3 4
need to devote considerable attention to investigate and implement I < system and operational changes to reduce feeduater to reactor water r x [3 j j-temperature diiferentials during all modes of operation. For example, if .]; the feedwater is always maintained withN 150*F to 250*F of reactor
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water temperature, the thermal stresses due to temperature fluctuation d within the vessel and feedwater temperature transients will be reduced.
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the number of fatigue cycles to crack initiation will be significantly increased, and the thermal component of startup-shutdown : tresses that t.,
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g could cause cracks to grow will be significantly reduced. Such an Q increased effort will also need to investigate potential system changes to eliminate the need for frequent sparger and thennal sleeve removals l }j t ,r w for inspection and repair of feedwater nozzles.
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One proposal that has been considered as a long term solution is the a _
.-Q welding of the thermal sleeve to the feeduster nozzle to preclude
- bypass leakage of co?d water between the therms) sleeve and tM nozzle, 3 y
_- At this time the NRC staff has reservations regarding the efficacy of ; this solution. Weld cracking could resalt from therms) and/or vibre-tory stresses in spite of analytical and design efforts to minimize Nf them. The main concern with the welded design, however, is that d3 9 ' neither the sleeve-to-nozzle weld nor the nozzle bore are accessible l for liquid penetrant examinattor. It is recosuanded that alternative designs with better Hspectability be considered. For example, two 5
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" j utility groups are installing baffles to restrict water circulation '
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Bypass flow should be limited by a'tighthfit;of the thermalssleevef # QUA 3 - j ng:y!Q,,,
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3, in the nozzle. ~ 0ther thermal sleeve designs incorporating piston: sf rings or bolted flanges- to eliminate bypass-leakage should ~be.- f t Jc considered. }[b t
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Another approach towards the minimization of this problem concerns
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the removal of all cladding in the feedwater nozzle blend radius and bore regions. Analyses submitted to the NRC ir.dicate that clad j h[~d removal can significantly increa:e nozzle fatigue life. In this j (f [ approach, the end result would be a nozzle with a clean smooth i; , surface without flaws or damage from the clad removal process. I v il g Several utilities with operating reactors have already decided to h1 implement clad removal. $ U . f The question of introducing a corrosion problem as a result of A
-h ,, . y removing con siderable stainless steel claddk1 was censidered by dM fj G
both GE and the staff. It was concluded that them will be no M sp
- c. . : problem; carbon steel has been used in contact with reactor water p j with no adverse effects. New 8WR reactor vessel nozzles will not
_l' be clad. Initially, cladding was applied to minimize rust accumu- T{
! Q lation in the reactor water thereby maintaining visibility during IN refueling and minimizing radioactive corrosion product carry-over 16
' I 9 into the clean-up system. The relatively small area of exposed carbon steel will not 1spect the above objectives siptficantly.
, h In summary, because a large number of operating OWRs either have
. li! been or are likely to be found with feedseter nozzle cracks that 4
require repair in the near future, we recommend increased effort # to to develop taproved repair procedures that leave a more fatigue- , M, resistant nozzle surface ccadition. The present short tenu pes-cedure is to gHnd out cracks as they are found. This leaves the nozzle with an irregular surface, sometimes with local areas of abusively ground clad, and base metal esposed to reactor water. A e,.se. -- l ___n_.--. ~ ~ ~ " " - ~ ~ ~ ~ ~ ~ ~ ^~
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, ATTACHMENT 3
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?g U SUP9MRY;0F BWR FEE 0 WATER N0ZZLE CRACKING PROBLEMS
- FIRST ' STARTI FEE'DWATER
- . PLKiT ~ GREATEST TOTAL OPN. UPS~ N0ZZLE INSPECT. ACTION TAKEN w.M CRACK DEPTH Dresden 1 10/59 no nozzles - plant design radically different.
, Big Rock Point 10/62 ' not relevant - reactor vessel has different design
;i? Humboldt Bay 4/63 110 1 ,
1976 (TV). '77 Install new 3/4 in.
- sDarcer Remachine i ~
Lacrosse nozzle 7/67 no nozzles - feedwater enters recirc. line pump intake Nine Mile Pt. I 11/69 109 1976 (UT). '77 Renachine
, 1-1/2in.
nozzles (4) Install 4 new spargers Oyster Creek 9/69 97 1976 (UT). '77 Same as Nine 1/2 in. Mile
', ^
Dresden 2 4/70 125 1975. '76 g Grind out cracks 1/2 in. Repicce spargers f.' Millstone 1 11/70 134 1974, '75 '76 Grind out cracks 0.55 in. M* 3 ' and replace spargers Drusden 3 7/71 93 1975 (same) 3/8 in. Monticello 3/11 91 1975 (same) 1/2 in. Quad Cities 1 4/72 112 1976 (same) 0.4 in. ~ Browns Ferry 1 10/73 68 1975 Grind out cracks. 5/32 in, repair spargers Browns Ferry 2 8/75 36 1975 Repair spargers; 1/32 in. no nozzle cracks Quad Cities 2 5/72 102 1975 Grind out cracks 3/8 in. and replace spargers Vermont Yankee 9/72 61 1975 (sase) 0.35 in. Peach Botton 2 2/74 65 1976 '77 (saae) 3/8 in. Peach Botton 3 9/74 46 1977 Grind out cracks 0.04 in. bot inc1weing nine foreign E.44 plants, at tent two of watch rrported cracking p+. espgpS$ h , 64
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-Fitzpatrick IL ^ 2/75' 50 1977(UT),'78~ Plan to renachine nczzles install new spargers -
Cooper 5/74 55 1976 Grind out cracks 0.175 in. 3 and replace spargers Pilgria ( 7/72 69 1976 (same) 3/4 in.- ; h Browns Ferry 3 9/76 21 Insp. planned at first outage (late 1978) , l Hatch 1 11/74 85 1977 Grind out cracks 0.04 in.
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Brunswick 2 4/75 62 No f/w nozzles inspecti.m to date ; , i f Duane Arnold 5/74 57 1977(UT) Hone 0 a
. ; -z Brunswick 1 10/76 No f/w nozzle inspection to date -
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IJ. t . . r, g . PART II
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l , .%, s..o r CONTROL R00 ORIVE RETURN "2, i i LINE N0ZZLE CRACXING f>. 9 - t .-4
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.1.0LINTRODUCTION _
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. Thire is one control rod drive (CRD) return line nozzie in BWR reaci.or l vessels, generally locatrd from 68 inches to 100 inches above the top b . 3 of the active fuel. The return line is typically 4 inches in diameter.
As early as 1974 a reneral Electric task force on austenitic stainless n , 1 steel piping noted the large measured thermal gradient in CRD return R line (CR0 RL) nozzles. Based on the unexpectedly high top to bottom thermal gradients in the nozzle, particularly at low flows, crack initia- ' { tion susceptability was cited and rerouting the return line was con- h- -i, r sidered. In addition, recent experience with BWR feedwater nozzles has !. demonstrated the occurrence of crack initiation in nozzles from thermal L cycling and further suggested the need to examine CRD return line ' i nozzles. GE issued Service Information Letter (SIL) No. 200 in October a.
\j 1976 recommending inspection of the nozzle and rerouting of the return -
line. This SIL was amended in March 1977 to provide for valving out . 1 the return line as an interim fix. J The staff has maintained an active involvement in this area through meetings with the General Electric Cospany and in case-by-case reviews of CRD RL nozzle inspection results from a number of operating 8WRs.
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The follcwing discussion and interim criteria hare been developed from 1 - currently available information and are subject to future modification. j Such interim criteria are needed and have been used to Justify continued f operation of boiling water reactors. 2 a 2.0 STATEMENT OF THE PROBLEM - Dye penetrant (PT) ins:>ections of the CRD return line nozzles to date 7. y 7
.f M " ali at dc9estic SWR plants have revealed cracks in three of the four plants ,
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inspected. Similar results were found at two overseas reactors. In !' M addition, cracks were found in the reactor vessel wall at Peach Bottom 2
.. [
in an area slightly below the CRD return line nozzle but still affected by the return line flow. 3 3 1 l The CRD RL nozzle examination results to date am summarized in Table 1.
,5-Cracking has been observed in both the blend radius and bore regions of l
the CRD RL nozzle. While most plants have a thermal sleeve in the CRD RL ' ( g nozzle, which would be expected to reduce the amount or extent of ' f cracking, cracks have been found at plants with and without sleeves. - 2 The cracking observed in the Peach Bottom 2 reactor vessel wall con- . f sisted of two horizontal cracks five and seven inches in length and - j 5 or 6 smaller cracks, located in an area six to twelve inches below ~ g the CRD RL nozzle. , g a 3.0 CAUSE OF THE PR08LEM
=
The underlying cause of crack initiation appears to be thennel fatigue, 7 j similar to that experienced with BWt feedwater nozzles. The thennel . cycling results from the low temperature (50*F to 100*F) condensate ' I water which enters the reactor vessel through the CR0 return line nozzle 5 during normal operation. Although crack initiation mechanisms for the < - t . feedwater and CR0 RL nozzles appear to be the same, there is a substan- - tial difference in the steady state stresses which ultimately affect crack growth rates. 3 a a y-n
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gn - CRD RL' nozzle crack l growth appears to' be' enhanced 'by the existence of a .. h[ continuous large thermal gradient from the top to the bottom of the nozzle, (550*F at the top. 50*F at the bottom), yielding high thermal w stresses. It'has also been stated, on the basis of stress calculations performed by GE, that control rod scram (wt.ich increases CR0 return flow j j from 15 GPM to 60 GPM for about three seconds) and scram testing does not
- significantly contribute to the ocule stress distribution. "'
- 4.0 P.EMEDY Effective long tern solutions to this problem require that the thermal cycling in the CRD return line nozzle be eliminated. Accordingly, the 7.
General Electric Comparty has made reconnendations, both interim and b_ f+ j , final, involving system modifications to accomplish this goal. 1 j The interim fix involved (a) valving off the CIW return Ifne to the 1 N~. reactor vessel, (b) reducing CR0 RL system flow. (c) raising the exhaust l water pressure to a level sufficient to permit the return water to enter the reactor ve,ssel via leakage past the sealing rings in the - I control rod drives rather than via the return line; (d) adding exhaust
- water filters and (e) testing of the modified system to verify that the control rod drives would operate properly.
1 The final system modification proposed rerouting the CR0 return Ifne in conjunction with the repair and capping of the nozzle. For 8W/2 l plants, GE reconsended that the return line be rerouted to the feeduster l I I
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primary containment and downstream of all motor i i -6 j 1 operated isolation valves. For BWR/3, 4 and 5 plants the return line i~" i . ce"1d be directed to the reactor water cleanup system downstream of the last motor operated isolation valve. , The reduction of cyc1fc thermal stresses in the nozzle could also be achieved, it appears, with an effective thennal sleeve. The available evidence from the Nine Mile Point 1 CRD RL nozzle inspection is encouraging in that thorough inspection (after cutting and removing the welded thermal sleeve) revealed no crack-like indications. Since the plant has operated for a significant period of time, being the third domestic BWR to go into operation, the favorable results might indicate that a well-designed thermal sleeve could be an alternative to system modifica-tions, although there are contravening considerations such as the' need to periodically inspect the nozzle. At this time, however, it is pre-nature to conclude whether or not such an approach is appropriate whom considering our overall safety objectives. 5.0 INTERIM STAFF CRITERIA 5ased on the information which is currently available, the NRC staff , i has determined that the following actions on the part of BWR licensees ' are appropriate in order to provide a sound basis for continu'ed plant , ! operation: l (it The CRD return line nozzle and the reactor vessel well below the nozzle should be inspected at the next scheduled refueling outage by dye penetrant examination, and in general, any crack indications should be repaired, generally, by grinding. The thermal sleeve, __ _ . _ _ _ - _ - - - -~ = = ---~ -- - ~~ -~ '
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this procedure. ; ( l (2) Rerouting of the CRD return line to either tha RWCU system or m feedwater ifne should be considered by all affected licensees ( - 3 and in those plants where cracking is observed, accomplished at , the earliest practical time. The staff recognizes that obtaining the necessary hardware may require long lead tires and therefore v implementation of tne reroute may not be possible at the most ; innediate upcoming refueling outage. Coincident with rerouting , of the return line, the CR0 RL nozzle extension should be cut 7, , off at the safe end and the nozzle capped. Thermal sleeves should 3 (.; be removed and all cracks removed by grinding. Complete clad , removal from the nozzle blend radius and adjacent bore region should be considered, and any weld-build-up areas used for attach-ment of thennel sleeves should be blended smoothly with the nozzle contour. (3) Implementation of the interim modification proposed by E i.e., valving out the return line, should be evaluated by each licensee j e - on a plant specific basis. The incidence of cracking in the CR0 AL j ; , nozzle. the time necessary to implement rerouting, and the avail-ability of the CR0 Rt, system as a source of makeup water tc the reactor vessel should be appropriately considered by each licensee. _._.__,_..__-m* -- - - - - ' " -
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T* }&mjlQ x . ? (4) The staff should.be keptiinformed .on:a timely basis of all .
~ V,^w;y:y
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- pertinent activities associated with nozzle inspection and repair 7 ' '" fi -
v and system modifications. Documentation of these activities 1
, should be provided promptly for staff review. - ? ; - ?
s v 6.0 FINAL STAFF POSITIONS f e :c The above interim staff positions are primarily meant to establish j
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information becomes available the staff will continue its review in e this area and will issue final positions at the appropriate time. J s t s 3 ^% 1 l l r [ 1 . .1 i i 4 4 __m
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$p- EXAMINATION RESULTS-
.G. .
p.. 4 y ;. - MAXIMUM w'ij .d.i; - CRACK DEPTH t -e YEARS
- START b A.- t
$$e OPN, UPS (CLAD 4 BASE) EXTENT THERMAL SLEEVE f!s%5 y; y Peach Bottom 3 2 45 7/8" General None 3 gf@qt x Peach Bottom 2 3 65
- General; rw..4: None
.- . also on v.c
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below CAD RL - e4 Nozzle W w CI Overseas 6 49 i .}. 7/8"
, General None L ,.. Reactor EO @W ,E Another Overseas 4 42 9
9/16" Ger.orel i.. om Reactor None .@i!Io Natch 1 2
%j f 85 5/8" Single Expanded Without tottom of Flange fMJA 1 hozzle )W ., V Rine Mile Point 1 7 109 --
None Welded, projects _ b into vessel sA several inches
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- Final results not available; 0.9 inch grind-out to date. '?
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ACT REQUEST
!6.'[A -!</-s77
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'd 7-/4-fy MR. J. M. FE_ ON, DIRECTO:
DIVI 5 ION 0 :U_"I OND RECC;D5 C ' ICE OF ACMIN!!TRATION U.S. NUCLERR :EGJLATOCi COMM. t..A 5 H I N G T O N 3 0.C. 2055* FREEDOM OF INFCMATION ACT REQUEST DEAR MR. FELTON RUR5UiNT TO THE FREEDOM QF I t, F 0 0' MATION ACT, = V50 EE2, A5 AMEtC ED , AND TnE R::0VI5IONI 0: 10 CFR RART 9, TnE UNDER5IGN:: MEREE6 REQUEITI THE OLLO:J IN .i
- 1. MEMORANDUM F:OM J. BUTLE:
TO L. RUSEN37EIN O f- ' MEET IN C-WITH DR. J. LE RE. n2 COM5U:-
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T I Of4' , DATED APRIL 1551
- 2. ' EVALUATION OF THE GLC'o I ,, ',' ;.
,1 IGNITER CONCEC- "C U5E IN NE 5EOUOYAM NU: LEA: :_ ANT' EY : .;. d-57RE7 LOU. C:ECA:ED =0; J, MI_-
MOAN.- NRC. DATED JANUAR'. i 1951
- 3. MA/ 24, 1952 MEMO:At4DUM F 09 f1 AR 5 M ALL EE Mate RE MCC I. IC A-TION OF NUOLEA: PEACTOR C O *.T;It MENT ATM05:nE:55 TO REDUCE ;II-F:OM MYDROSEN 03MEU57 ION.
4 ANS REPC *! C0;;E!PONDEN;E NOTE!. MEMD5 0: MI'..'E5 Cot 4CE: nit 4G TnE 040 03ED :ULE :- MYDROGEN CONTRC. *0: MAPr !I!
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#. , _ . , . . . . - . - j NING THE COMMITTEE TO REVIEW .l 6
GENERIC REQUIREMENT.. ,
- 5. ANY REDORT5, CODRESPONDENCE, ' l NOTE 5, MINUTES, OR MEMOS TC'OR FROM THE STAFF OR THE COMMI5- ,
f SION CONCERNING THE PROPO!ED d RULE (46 FC 62251) ON HYDROGEN CONTROL GENE:ATED'5INCE OCT05ER f 1983. ! IN ACCORDANCE WITH YOUR DOLICY, I HERESY REQUEST TnAT 75% OF THE FEE 5 FOR THE SEARCH AND RRODUC-TION OF THESE !TEMS EE WA*VED l AS THEY RELATE TO I55UE u 5 ! CONCERNING H/DK0 GEN CONTRC ADMITTED IN THE PERcY CL FRO-CEEDING, DOCr$7 N0!. 50-440/4A1. I AGREE TO ACCEPT THE REMAINING (25%) CHARGES. 5INCERELY, SU$4N L. nIATT i OCRE 8275 MUN50N FC. , 1 MENTOR, OM 44050 (215) 255-3155 l l l 1 I I \ i L-}}