ML20079A873
| ML20079A873 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 06/30/1975 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Giambusso A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9105220320 | |
| Download: ML20079A873 (49) | |
Text
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HSF NORTHERN STATES POWER COMPANY
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August 29, 1975 --
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Mr. A. Giambusso, Director t Division of Reactor Licensing ::: 0, U. S. Nuclear Regulatory Comission Washington, DC 20555
Dear Mr. Giambusso:
MONTICELID NUCLEAR CENERATING PIAN't
--- - Docke t No. 50-263 License No. DPR 22 Semi-Annual Operating Report No. 9 January 1 to June 30, 1975 In accordance with Section 6.7. A.2 of Appendix A Technical spect-fications for Provisional Operating License DPR-22, we are submic-ting 40 copies of Semi-Annual Operating Report No. 9 covering the period January 1 - June 30,1975.
Yours very truly, W
L. O. Mayer, PE =
Manager, Nuclear Support Services 4 1hM/INM/ deb ,,
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*I NORTi(ERN STATES P0hER COMPN #
M]NTICEllD NUCifAR GINERATING PLWT DOCKlff NO. 50-263 LICINSE N0. DPR-22 REFORT 10 UNITED STATES NUCLEAR REGJ!ATORY 00ht!ISSION DIVISION OF REACTOR LICENSING SDtI-ANNUAL OPERATING REPORT NO. 9 JANllARY 1,1975 'HIROUGI J1NE 30,1975
s
,o TABLE OF CONTENTS 1
I. OPERATIONS SRf4\RY A. Chronological llistory B. Changes in Plant Design C. Perfonnance Characteristics D. Procedure Changes E. Results of Surveillance Tests and Inspections F. Containment Leak Rate Tests G. Changes, Tests G Experiments Requiring NRC Authorization
- 11. Changes in Plant Operating Organization I. Occupational Personnel Radiation Exposure J. Relief Valve Operation Sumary II. IVhliR OPERATICt1 AND 91LTfD0hNS A. Power Generation Statistics
- b. Shutdowns III. MAINTINANCE SLM4\RY IV. CilANGES, TESTS AND EXPERIMENTS V. RADI0 ACTIVE EFFLUINT RELFASES A. Gaseous Effluents
- 1. Gross Radioactivity Releases
- 2. Iodine Releases
- 3. Particulate Reicases B. Liquid Effluents C. Solid Waste VI. RADIATION ENVIRON 4NTAL RNITORING i
e 11
> i ./ '
- 1. OPERATm4S SRfuRY 1
A. Chronological flistory 1/1/75 Operated at 75% of rated power. Power was administrative 1y to limited to minimize background radiation and equipnent 1/3/75 contamination levels within the plant. 1/4/75 Power was reduced to 66% due to icing problems 5 the inlet stmeture. 1/5/75 Reduced power to 64%. 1/6/75 Operatal at 64% of rated power, to 1/8/75 1/9/75 Began scheduled outage to refuel the reactor and perfom miscellanecus plant inspections, modifications and maintenance. In addition to refueling, itens accmplished during the outage (1/9/75 to 2/6/75) included the following:
- a. Local leak rate tests of selected primary containment isolation valves,
- b. In-service inspection activities,
- c. Surveillance tests and inspections,
- d. Installation of a main steam pressure averaging manifold.
- c. Relocation of the dyrwell-to-torus vacuum breaker test solenoid valves to inside the torus,
- f. Removal of the reactor feedpump runout protection logic.
- g. Repair of #11 PJIR heat exchanger bottom head gasket leak.
I
- h. Repair and installation of a main steam safety / relief valve.
- i. thintenance of the operator for RCIC isolation valve FO 2076. The main gear train and limit switch gear train were cleaned and repacked with grease, and the torque and limit switches were adjusted,
- j. CRD mainteriance, including replacement of small mesh inner filters which were installed in four drives.
I-1
l
- k. Installation of a cobalt free sample probe on the feedwater systun. i
- 1. thin turbine inspection and maintenan,-.
l 1/19/75 The actuating am of the refueling bridge limit switch, , which provides a rcd withdrawal block when the bridge '- is over the core came loose fran the shaft. The lifnit switch was proper,ly adjusted and the actuating am set screw was retightened using 1oc Tite on the threads. 1/20/75 1inear indications were detected during ultrasonic exam-ination of three welds in loop A and one weld in loop II l cf the reactor recirculation systern discharge valve 4-inch bypass lines. Segments of loth loops containing the indicatlons were replaced, f 1/23/75 Puel sipping was cunplotcd. 1/20/75 Core reloading was empleted. ' 2/5/75 In-service inspections of the jet imps, feedwater spargers, selected austenitic pipe welds and miscelle - cous reactor cmponets were completed, h'ith exceptic. of the recirculation systun 4 inch bypass lines, tho . l results cf the inspections were satisfactory. 2/5/75 Canpleted the primary syston leakage test. ! 2/6/75 Established inithl criticality for cycle 4 lleatup i continued and !!PCI, RCIC and relief valve operability tests were perfonncd. t 2/7/75 Generato1 on line. Power was a.lintained at approx-to imately 15% of rated to provide a soak period for the 2/9/75 new fuel. t On 2/9/75, reactot water cleanup c',4tboard isolation valve, FD 2399, became inoperable due to a shorted relay coil. , The relay and a control transfonner were replaccd. 2/10/75 lucreated power to 73% of rated, to 2/11/75 2/12/75 Power was reduccd to 66% of ratcd for control rod with-drawal. 2/13/75 to Power was increased to 94% of ratcd and then reduccd to 2/14/75 58% of rated for control rod witidrawal. I-2 i
,. .~
s . 4 On 2/14/75, it was discover (d that a stiffener had been tack welded to the hinge am of lilCI exhaust line chwk valve, !!PCI 9, without the proper engincering analysis aid safety review haveing been perfomcd. The valve was disassunbled and dye penetrant inspectcd. No indi-cations of cracks were observed. thintenance personnel were provided ndditional training on administrative requironents. 2/15/75 Power was increased to 100% of ratni. 2/10/75 Operated at 100% of rated power to 2/17/75 On 2/17/75, it was discoveral that a hydraulic shock suppressor locatcd on the shutdown cooling suction line of the R{R system was inoperabic due to loss of oil. The suppressor was replaced. 2/18/75 Power was reduced to 50% of rated following a trip of loth off gas rocaabiners initiaicd by a f alse hydrogen analyzer trip (probably due to moisture in the sample systm). Reconbiner systm was returned to service and pwer was increascd to 100% of rated. 2/19/75 Operated at 100% of rated power. to 2/20/75 2/21/75 Power was decreasal to 82% of rated to rcduce release rate frca reactor luilding vent following failure of recmbiner systm hydrogen analyzer sample ptap. Sample paap was replaced and pwer was returned to 100% of rated. A noticeabic spike in reactor water conductivity and 41 1 occurred following backwash and precoating of a condensate dmineralizer, apparently due to resin Icakage. 2/22/75 Oprated at 100% of rated power. One filter elment in to the D condensate dmineralizer was found unseated and was 2/26/75 rescated. 2/27/75 hhen D cordensate demineralizer was returncd to service a reactor water conductivity spike and decrease in Idi j was observed concurrent with It pwer decrease, indi-i cating resin Icakage. Pwer gradually increased to 100% of rated. I 2/28/75 Operated at 100% of rated power, to 3/1/75 3/2/75 Pcuer was reduced to 97% of ratcd to enable processing of a condensato demineralizer while maintainirig low differ-ential pressure on the dendneralizer after strainers. 3/3/75 Power war increased to 100% of rated then reducal to 9M of rated to piocess a condensate dmineralizer. Power was returncd to 100% of rated. I.- 3 A
C 3/4/75 Operated at 100% of ratcd power. 3/11/75 3/12/75 Power decreased to 99% of rated following processing of a condensate donincralizer. Pcue. was reducal to 74% of {' ratcd to allow inspection of the "J' condensate domin-eralizer. L 3/13/75 Power was ircreascd to 94% of ratal, with fcur of the five condcnsate dmineralizers in service. , 3/14/75 Operated at 94% of rated power. 3/15/75 Increased p;wer to 96% of rated. 3/16/75 Power was increased to 100% of ratcd following repairs to the D" conJensate demineralizer. The relairs includal ! replacanent of several f!1ter clunent gas sets and seating cups. In addition, a small tear in the' resin trap was rcpaircd. ' 3/17/75 Operatcd at 100% of rated power. to ! 3/22/75 3/23/75 Power was decreascd to 43% of ratal following trips of both off-gas s econbiners (initiatcd by a false hydrogen analyzer trip) and subsequent controller problms associated with #12 recirculation ptrup. The #12 recirculation }unp was manually tripped and the recirc crosstle valves were opened. The reccrnbiner systan was
- returned to service. The speed controller for #12 recirc-ulation ptrnp was replaced and the ptrnp was returned to ,
serYlCe. 3/24/75 Power was increased to 100% of rated. 3/25/75 Operatcd at 98 100% of rated power. to 4/16/75 4/17/75 Began power decrease for control rod sequence intenhange. 4/18/75 Power was decreascd to 35% for the control rcd sequence to int erchange. Power was increased following the interchange 4/21/75 achieving 100% of rated by 4/21/75 4/22/75 Operated at 99 - 100% of rated power except for a short + to duration 21 power decrease on 4/29/75 due to a recirc-5/6/76 ulation Junp speed controller problun. l 1-4 I h i l _ - - - . - - _ _ _ . . . _ . . . _ . . _ _ - . _ _ _ . _ - . . _ , . _ . . - _ _ _ . - _ _ _ _ _ _ _ - . _ _ _ . _ . ~ . - . . .
On 5/5/75, the control circuit for valve, 50 2078, steam supply to irlC turbine, was found de energizal. Investigation revealal that the motor control unit undervoltage relay coil had opened. The relay coil was replaced. 5/7/75 Power was reduced to 94% of rated to minimizi: pinnt radiation background and contamination levels. 5/8/75 Power was gradually increased to 9M of rated. 5/9/75 Operatcd at 9M of rated power. to 5/10/75 5/11/75 Power was ralucal to 9M of ratal for control rod drive withdrawal stall flow testing. 5/12/75 Operatcd at 9M of rated power to 5/14/75 5/15/75 The reactor was shutdown for a planned maintenance outage to to inspect two safety / relief valves that Fad high discharge 5/16/75 line temperatures. Other items accouplished during the outage includal the following:
- a. The reeirculation pimp discharge valve 4 inch bypass lines were visually inspected for signs of leakage.
There was no evidence of leakage,
- b. Several isolation valves located inside the drywell were repacked,
- c. The hydraulic shock suppressors located inside the drywell were inspectcd and none were found to be in-operable.
- d. The 10Cl exhaust line check valve (10El 9) hinge am was dye penetrant inspected. No indications of crack-j ing were observed.
5/17/75 Control rods were withdrawn and reactor heatup cccinenced, j to Power was increased to 78% of rated by 5/18/75
; 5/18/75 5/19/75 Power was reducal to 60% of rated for control rod with-to drawal. Power was increased to 91% of rated by 5/20/75.
5/20/75 1-5 l i l
l 5/21/75 Operatcd at 89 91% of rated pwer. to 5/24/75 5/25/75 Pwor was raluced to 75% of ratal for control rcd exctrise testing, t 5/26/75 Power was increased to 82% of ratal. h'hile perfonning a surveillance test , "A" Core Spray injection Valve, FD 1751 fallnl to reopen by means of the motorized operator. The valve was returncd to it's nonnally open position by means of the local handwheel. The valve operated properly during subsequent testing. Tests and inspectlons to detennine the cause of the malfunction were unsuccessful. 5/27/75 Pcuer was ralucal twice (to 50% and 461 of ratal) following trips of both off gas recunbiner trains. The recombiner system was retun.ed to nonnal operation and power was increasal to 80% of ratal. The recanbiner system trips were caused by spurious signals receivcd when substation switching was performed. 5/28/75 Pcuer was increasn! to 89% of ratal at which time a reactor serrn cecurral due to a malfunction of the reactor pressure control system. A pulsation snubber in the mechanical pressure regulator sensing line was found to be plugged. S/29/75 The pulsatico snubber was replaced and reactor startup ccrneneni. Reccnbiner train !! was shutdown due to after-condenser level control prob 1cus. Investigation revealcd catalyst in the af ter-condenser and in the flash tank. The reccub-iner was inspectal and it was found tMt the upper catalyst retention screen had failed. The systun was flushcd anJ the retention screen was modified. The train was then returned to service. Operating procalures were revised to minimi:e the differential pressure experier cd across the reconbiners. 5/30/75 Pwer was increascd to 91% of rated. 5/31/75 Operatal at 88 - 91% of rated pwer. 6/1/75 Pwer was reduccd to 70% of ratcd for control rcd exercise testing. Upon cunpletion of the testing power was increasal to 90% of ratcd. I-6
4 6/2/75 Operatal at 90% of rated pcuer. to 6/5/75 6/6/75 Power was decrensal to 80% when one turbine stop valve drifted closed during an exercise test. A slight readjustment of the stop valve bypass hardwheel corrected the probicn. Power was increascd to 90% of rated. 6/7/75 Operatal at 90% of ratnl pcwer. 6/8/75 Power was reduced to 70% of rated for control rod drive exercise testing. Upon ccupletion of the test , pcuer was increascd to 90% of ratcd. 6/9/75 Operated at 90% of ratal power, to 6/12/75 6/13/75 Power was reduccd to 46% of ratcd following trips of both off-gas roccrabiners. The reconbiner systein was returned to operation and power was increased to 88% of ratcd. 6/14/75 Operated at 88% of rated pwcr. 6/15/75 Power was reduced to 70% of rated for contrc: rcd exercise testing. Power was returned to 88% ot .mito.! upon ccupletion of the test. 6/16/75 Operated at 86 - 89% of ratcd power, to 6/22/75 6/23/75 Power was reduced to 70% of rated for control rcd exercise testing. Pcuer was returned to 85% of rated upon ccepletion of the test. 6/24/75 Operatcd at 84 - 85% of rated power to 6/30/75 17
6 B, gpnjesinPlantDesign, Cinnges to the facility as describcd in the FFN1 are includtd in Section IV of this report in accordance with the requironents of 10Cril50:59(b). These changes incinic the following
- 1. A pressure averaging manifold was insta11cd on the main steam lines to minimite pressure disturbances during stop valve testing.
- 2. The such inner that thefilter innerassunblics on toOW's filter is attachcd 628 the top and of the 635 stop were miified piston to assure tiot a plugged filter does not affect the scram time.
- 3. The drywell-to-torus vacutta breaker test solenoid valves were leakage fran the containment via the air line frm the solenoid valves to the vacutra breakers.
- 4. The reactor feWpenp runmt protection feature was removW fran the feedwater control systun to improve plant reliability. (
5. A cobalt free sample probe was installed on the feedwater systen to be used for detemining cobalt levels in the feWwater to the reactor. C. Perfomance Claracteristics itst plant perfonnance claracteristics during this reporting period continued to be as predictcd. Administrative limits m reactor power to minimite in plant background and contamination icvels were continued fran the previous reporting pericd until January 9th, at which tSne the plant was shutdown for refueling. Prior to shutting down, power was limited to approximately 66% of rated. Puel sipping operations during the outage identificd 42 leaking fuel assanblies. These assunblies, which were all part of the original core, were replaccd. In addition,12 assemblics which were classified as possible leakers ard 26 of the highest exposure assemblics were replacW. Off-gas activity innediately following the outage was significant tlat administrative lifnits on reactor Ix)wer were again initiatcd. On 6/24/75 and 6/25/75, ten initial core fuel assemblics which had been identified as leakers and runaved fran the core at the end of Cycle 2 and 3, were inspectcd using underwater television. Visibic cracks were The remining observed in several corner fuel pins of the assanblics. initial core 7x7 fuel is scheduled to be runovW frun the reactor during a refueling outage schcduled to begin in Septenber,1975. Other significant equipnent malfunctions or performance deviations are briefly s unarized in Section I.A. 1-8 y
./Q
- m _
4 D, prmodure Cinnges The folicuing cinnges in procedures were necessitatcd by itans B 1 and C, alove, or were required to improve the safety of plant oper- ' ations:
- 1. Surveillance and operating procedures were issued or revised as necessary to satisfy new Technical Specification requirements aM !
to accuranodate equipnent modifications, l l
- 2. A preventive maintenance procedure ard associatcd operatiens l controlling dcarment was issued for the scram pilot valves.
- 3. Cable spreading rocun penetr tions were addcd to the critical systans list to assure that a second level review is perfonned prior to working an these penetrations.
4 prxcdures were issued to assure canpliance with actions requested in htC - lE Bulletin 7410B concerning primary syston leakage inside primary containment.
- 5. Security Guard procalutes were issucd.
- 6. Several procedures were issued for functionally testing radiation protection instntmentation.
7 A startup checklist was issued to insure that pertinent startup itans are canpleted as required. 8 A special procedure was issued for detemining the weight on each ; of the toms support columns, l
- 9. I A procedure was issued for inspecting the RlR heat exchangers.
- 10. The refueling platfom preventive maintenance procedure was revised l to include a step for inspection of the bridge position interlock limit switches.
- 11. The diesel generator starting system preventive maintenance precedure was revised to incitde a step for inspceting the control air filter. ,
., 12. The jumper bypass form was revised to require an additional verifi-cation signature.
- 13. The recanbiner syston operating procedures were reviscd to minimize the differential pressure experiencui across the operating recombiner when returning the alternate recambiner to service.
E. Results of Surveillance Tests ard Inspc<tions The results of all surveillarr:e tests and inspections required by the Technical Specifications were satisfactory except as noted below: 19
, Test Date Reason for Test Results of Test i
- 7. January 10, 1975 Sectniary Containment Dmonstratcd capa-Capability Test perfonned bility to maintain in accordance with Section at least 1/4" water 4.7 C.1.a of the Technical vacutrn under calm S)ecifications to ensure wind conditions with t at the seccndary contain- a filter train flow ment system meets leakage rate less than acceptance criteria prior 4000 CIN.
to refueling.
- 8. January 22, 1975 Local leak rate test of The measured leakage the valve set A0-2377, was 5.08 SCHl.
A0 2378 and A0 2381 following maintenance.
- 9. Febrtary 1,1975 Local leak rate test of The measurai leakages the torus and CRD manways were 0.0 SCHl.
prior to startup.
- 10. February 5, 1975 Local leak rate test of The measur(d leakage the drywell equipnent was 3.09 SCHl.
hatch prior to startup.
- 11. February 5,1975 local leak rate test of The measured leakage the drywell head prior was 0.05 SCHl.
to startup.
- 12. February 7, 1975 Air lock leak rate test The measural leakage during startup, was 0.0 SCHl.
- 13. hby 18, 1975 Air lock leak rate test The measured leakage during startup, was 0.92 SCHi.
G. Changes, Tests aM Experi.'ients Requiring NRC Authorization The only change during this reporting period which required NRC authorization was the C>rje 4 core wrtich was loaded during the January,1975, o, This change is described in a report tit icd , 'Wnt icel incrating plant Third Reload Submittal", which was sulnitted to the NIC Deccaber 11, 1974. This report supplcnentcd the topical report, "GE/IMR Generic Reload Licens-ing Application for 8x8 Fuel", Novcaber,1974. I
- 11. _Ch3 nSes in plant _ Operating Organization
'. Mr. R. L. Scheinost was appointed to the position of Quality Engineer, effective June 1,1975. The functions of this position were previously perfonned by resident personnel obtained frun a consulting finn.
I - 11
' et l .
- 3. Ibring a refueling outage weekly surveillance test, a Ral Witldrawal block was not received when the refueling bridge was over the core and the reactor mode switch was placed in the STAIGUP position as reported in Mr. L 0 bbyer's letter to Mr A Giambusso, dated January 29, 1975. The refueling bridge travel limit switch was properly adjusted and the actuating arm screw was retightencd using loc-Tite on the threads.
- 2. naring routine monthly surveillance, the Core Spray injection Valve, FD 1751, failed to open upon signal from the control room as reportol in Mr. L 0 bhyer's letter to Mr. A Giambusso, datcd June 6,1975. naring subsequent testing the valve operatol normally. Investigation revealed no apparent cause of the initial failure.
F. Contairrnent Leak Rate Tests Test Date Reason for Test Results of Test
- 1. January 9,1975 Local Leak rate test of The measure leakage the CRD manway perfomed was 0.0 SCRI.
prior to opening for CRD maintenance.
- 2. January 9,1975 Local le.sk rate test of The measured leakages the two torus manways were 0.0 SCRI.
prior to opening for torus access.
- 3. January 9,1975 Lctal leak rate test of The measured leakage valves A0-2377, A0-2378, was 3.29 SCRI.
and A0-2381 prior to main tenance.
- 4. January 10, 1975 Local leak rate test of 'lhe measural leakage the drywell head prior to was 0.03 SCRI.
renoving the head for reactor refueling.
- 5. January 10, 1975 local Icak rate test of The measurcd leakage the drywell equignent was 1.07 SCRt.
hatch prior to opening the hatch.
- 6. January- Type C local Icak rate Six valves requircd February, 1975 tests of selected primary repair to meet the contalirnent isolation Icakage rate accept-valves. ance criteria.
Licensee Event Reports submitted by L 0 bhyer to A Giambusso, on February 10,1975, include infomation I - 10 concerning corrective actions talen.
l ' l . 1 1. Occupational Personnel Radiation thposure
- 1. Radiation Ihposure by incrments in Reporting Perictl:
Ntsnber of Personnel thposure Pennanent Temporary increments .(Site NSP Txployees) (Off Site NSP Eryloyees) JWIM) _
< 100 26 55 100 - 250 10 12 250 - 500 12 5 500 750 9 10 750 1000 10 8 1000 - 2000 30 13 2000 - 3000 13 2 3000 - 4000 3 0
> 4000 0 0 2.
Ihposures by Duty Function in Excess of 500 nrm in Reporting Pericd: Number of Isposures > 300 mrca Duty Peminent Temporary Function ,(Site NSP Employees) .(Off Sit _e NSP Employees) Routine Plant Surveillance and Inspection 43 Routine Refueling Operations 29 Routine Plant Thintenance 22 Special Plant thintenance 16 33 Radwaste System hhintenance 23 6 Radwaste Preparation and Shipnent 3 Off Gas Systan thintenance 8 Some individuals appear in more than one duty function. Special Plant Fhintenance is maintenance performed during plant shutdown. I - 12
J. Relief Valve Operation Stramry ( i Following is a stunary of relief valve operation during this report period: ' Nttnber of Relief Valve Ntsnber Date , Ibw Initiated Jnitiations ' RV 2 71C 2/7/75 Minua11y during 1 operability testing RV 2 7JD 5/17/75 Rinually during 1 operability testing RV 2-71E 2/7/75 Minua11y during 1 ; operability testing ' RV 2 71G 2/7/75 Minually during 1 operability testing t 5/17/75 Mmually during 1 operability testing ; k I ( I-I s l i I - 13 l
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E. Shuttlwns.. A "Rcutine Operating Report Slutdom Re:ord" fonn, containing all of the infonnation requirtd by the Technical Specifications, has been established to facilitate reporting of outages. This so: tion con-tains copics of the record forms for this report pericd. I L t I i , - I i ,i : i i r II - 3
9 0002 9 IOU 1 i f) n!'[hA f _t f[; gin M !,ugligp[,la o u 9
- 1. Cause of the c,utar,o Rhtdultd refuelW mtw.
- 2. % t ho d o r c hu t t i ne down t he r oa< t ot t e.c. , T r i p au to.it i t rur.d wo, or rnanual ! y con t r o l l e:j .J.,I 1,o r a 4 e ch...t J aw i:
Muiually cont rolled deliberate shutdown.
- 3. Darat ion o f ou t re: .
1 t ier 3... rcr,c73tc, g r r ; , r,6 3 1/10/75 2333 Generator en l i r<e: 777775 ) gy, 4 Pla.t Stitut dararm t he oui at e ( e.r. , c .l J chu t .i .wn o r hot ute.oJ: ,-): Cold Shutdown S. Corrective Action inlen to prevent rc:;.eti+it n, if a;s;s t e p r i a t e : Not applicable, 11 -4
[ 4 6002 ; IOUTIfE griRA1ifn Mfubl CuV111WJ hrrq!J i i
- 1. Cause of the outar,e !
l Schedulcs) outage for inspcetion and unintenance of two relief valves. f
- 2. f,b t hod o f shu t t i nn down the rev tori e.e., 1 rip <iuto",itic rundown, or !
rnanuall y con t r ol led dol i bor aic t.hu t d:,wn i hhnually controlled deliberate slutdown. ; l.
- 3. Duration of outai,et t.p qg Genorator off 1ine: 5/15/75 2215 , ,
(. Generator or. 1 nc: 5/18/75 0046
' 17 ~ ' <
l 4 Pl an t S t a t us du r i ng t ho ou t oco ( e. p,. . co l d t.hu t d.'wn o r ho t s t andb e ): Cold slutdown, t I ; 4 t l S. Corroetive Action taken to prevent r cipe t i t i on, if approprlate: r Not applicable. ;
! I i
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0002 8 J ging oprota i'p i i l' da !..n nit dN ii coo
- 1. Cause of the outage:
The pulsation snubber in the pressure sensing line to the Mechanical Pressure Regulator became plugged, causing a loss of pressure control stability. This resulted in an increase in neutron fltu to the scram trip setting.
- 2. f.bthod of thutting down the rea tor t e.r. T r ip automt ic r ur dawn, or r trivall y cont r ol led del i t.or at e thu t Lu ,:
Reactor scram due to high nWtron fitu.
- 3. D.r at ion e t outare: ,,,,,
3 , ,, , ,
~
Generator eff l i r <o : S/28/75 1551 , t Ger.o r ato r c n 1 i r.e 5/29/75 14 9 .,- 4 Plant 5t a t us da r i ri. t he au t ore ( e.g. , cold chai'< *, or hat
- i a ad! , ): .
llot shutdowTi. S. Corrective A:t ion t at.en to prever t re;otitn n. if Npropriate: The pulsation snubber was replaccd. Cleaning of the snubber has been added to the preventive maintenance procedure. II - 6
.W1 -
Rev. 2 7: 1 M. :. J::C_i r _~1 m Y i
' ! ~ '-
Sa m Peric=cl, Includin I I Corr t:. n n T i.c.1 t a !'rev cit y;en.. :e c;p11, .Mcj rI Pre-sy t cm :.e .:.... a Tr,.C.- f r 7_;ctor Co w .... 1
s lin ': " r :~. ~ .m i' 'if !."I2Cf4C)-
._ __.u. _ __ _ _ _ _ - _ _ . _ . __ _ .
Neutron :fonitoring f Unknown RMt Channel 8 Failed Fuse Replaced in + 5 volt DC Power Supply. Systen-RBIChannel; Upscale Initiating R3't Bypassed until Repairs Coc:pleted then S i f Rod Blocks Returned to Service. Fanergency Diesel Unknown Excessive Setpoint Replace 3 Pressure Switch. Air Ta:is P:x ped Generator 811 ; Drift-Cce: pressor to Required Pressure Prior to Initiating _ Diesel, #11 Air i j Loading Point Shifted Repairs. Cmpressor . i Causing Low Pressure Alams ., l?bdified Off-Gas l Components Replaces as Required. Randaa Failures in the Six 11 and 21 Alams t
!HoldupSystca , OutIct Analy:crs Received. !
' Outlet ifydrogen j Analy crs j l Pressure Relief Loose 'ihemocouple Teminal Loss of Tc=perature !
Themoccuple Teminal Secured j Syste::-Relief Valve !!anitoring on Discharge
' 'll' Piping "
I ! iControlRod Under '.essel Connector Loss of ' Full-In
- l Repaired Conna. tor Position Infomation Da . aged During Plant Signal, Preventing l Systen - CR 26-15 ' Outage . Refueling Bridge l l
; ; Ilovement Over Core j t
' Neutron ionitoring Randera Component Failure . Failed Downscale- ' Replaced Failed Operational A,:plifier on Syste: - UT! I Downscale Alam UT! A:plifier Card aa! Returned to Nor :al Channel 20-293 I l Received. Operation
- , I l \
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- . . . - . . ~ . - - . . . - -
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4001 Rev. 2 .-
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; I ; c:r e ,. c : :On :: Pe-for:r1. IncIuding I
cor rat?. . _tn:. Traca to Prevent Sy. m . er ; IOpc:it r fif rpplic:Me) and P:c-r: s dr- l : .. crit i m En ;; Previle far 2utor 1 Cn~~v nnt l 'a' n 1 i '. I
.~' -
W ' N t'; P P3ir (if VPliCOIC)- _.__ !, _ _ . - Ii Neutron !bnitoring Under Vessel Connector ! IfiN Output Drifted i j Systm - IFM Repaired Under Vessel Connector Damaged During Plant j Downscale - Downscale
- Channel 36-29A ! Outage t Alam Received NeutronMonitoringlUnderVesselConnector LPPJ4 Output h'ent Repaired Under Vessel Connector Systes - IFrN Draged During Plant Upscale - Upscale Channel 44-213 i. Outage '
Alam Received
. t *-' t Neutron Monitoring } Randm Component Failure
- LIT 4 Upscale Alam Systen-IPR't Channelj Replaced Failed Zener Diode on LFRM i
44-29A ' Received Even Thotgh Amplifier Card Output ?.omal Neutrm Monitoring ' Breakdown of Detector Excessive Noise - SIN 1 Systca-SIM Channel , Shielding Replaced SRM Detector. Rod Block Initiated 24 Counts Nould Not Go to ,l while Repairs were in Progres I 2cro with Detector ! Nithdrawn. ; i
- Reactor Protection ! Unknown Excessive Setpoint
! Replaced Pressure Switch; 1/2 Scran 4 Systca-Reactor High'
} Pressure Scram Drift - Tripped at too' Manually Initiated While Repairs w re in
; Lcw a Pressure Channel A t { Progress
, t 4 j l . I i i , t 'J a 4 . I i i j __.
y
- g
, 4001 Rev. 2 mlT.TIXV;C SL'*1W I l g Surary of ' hark Perfor.vl, IncItrlin; Cos tective Action Taken to Prevent Systea or Pectiti .m (if applicable) an.1 Pre-Cause of the Rec.Ct. raj N ft.*
- crftions Tuien to Provide for Rextor
! Co=ponent L 1 function ... %f.; 0;o c: k . Sa? ty D.1 ring Eepair (if applicable;. i s
- CRD Unknown Caused Rod to Drift in ; Disassembled CV-126 on CRD 18-39 and i
1 Insert direction ' installed new seat. 4 g Reactor was in cold shutdown condition.
- SBGT Open Relay Coil Caused both indicating l Replaced the coil on centrol relay CR-15. ,
I lights for rom heater : I to indicate simultan- The redundant system was operational. l cously. l 2
! - I
' I C CRD The Seat "O" Ring was cut Nitrogen leakage from Reaoved a;xi replaced the #111 valve, t CRD accumulator. accunulator gas charging valve on CRD w Module 26-23. j The reactor was in cold shutdown condition.
#12 Diesel Not known at this time. .None. l Recoved heads fran the cooling water heat l
- Generator , exchangers. Cleaned, inspected and tested j
i the tubes. Plugged a total of 81 tubes j I l The event is under investigation as to e
, i cause and remedy. .
i
- Safety Relief Dirt under the pilot valve ; Very slight Icak of 8R The reactor was in cold shutdown conditio l Valve Disc
- pilot valve l "emoved G" *122 top for works inspection frcx2aai safety testing.relief valve l j l t Lapped the pilot valve and machined the j preload spacer 3
2 .
- Reactor was in cold shutdown cmditicn.
- }
- i
( . i i, l i 1 __ -. . - - . - - - -- ---- - - - - - - - - - - - - ~ ' - - - - - - - ~ " - - ~ ~ - - ' ~ -- ' ~ ~ ~ - - ~ " ~
*001 .-
Rev. 2
- ni:hT'a':.E 91TnRY i
S: ary of Nork Perfor:cd, Including Curr xtive .k tian Taken to Prevent repetit:en iif applicahic) and Pre-Sy: tan er Cc r > o f th - :.m ' rifn t ; caut2e c ~c .L. r. to Prcride for Rcactor Cc:panent IM fm;t ic > en U b :aticn Safety Purin;; Papair (if appliccMe). i
- p _ _._. _ .- ~ _ ____ .
RHR Service hn Slight leak thrwgh Replaced broken adjustment screw and valve Xater vent valve.systou flow seat on the discharge vent valve. rc=ained within specified limits. Redundant equipment was operable. Seismic 'U' Ring at the manifold Loss of Hydraulic Fluid- Disassembled HSM-10 snubber 85S-21. Restraint ; block was cut.
- snubber inoperabic. Cleaned and installed new parts kit.
s s j Rebuilt the snubber with certified parts. e Piping Restraints jSand in the cylinders ' Slight loss of ifydraulic' Disassembled 2 HSSA-10 Snubbers. 'SS3M & j ; fluid-snubbers operable 85537 cleaned and installed new parts kits.
! lReacterincoldshutdowncondition.
i IW Suppression ploisture Loss of vacutza breaker ' Cleaned and lubricated position indicating i Chamber Vac , i ralundant position switch for 10 2382J. Breaker indication.
! : Scaled the cover gasket i .
j !
! Reactor was in cold shutdown condition.
RHR , Valve Body distortion ! Did not meet leakage Disasscnbled check valve AO-10-4M, j jtest requirements. , Cleaned lapped the disc and seat. 4
; tReassmbled with a new gasket.
l l Carefully lapped the disc and seat. I , j i 1
; : Reactor was in cold shutdown corxlition.
i e
- I s
s
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. - ~ . _ . _
. v. -
u . - 4 Wil - Rev. . .-
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-0 ?::.~a m.;?, Incl.li g j
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. - . . : ct . : .. - * ,; 7. '.L*:r
_ s , . . , . . - .- ... 7 g7 .., 3 g 39;,), b .- .... _. __ i i RHR i The head bolts were not - Very small Icak Renoved bottara heads from i11 RHR system I i torcued correctly by ' heat exchanger. Stoned the tube sheet i Vendcr. gasket surfaces. Replaced heads with I 5 new gaskets. Torqued the head bolts to the correct value Reactor was in cold shutdown condition.
. RCIC . The disc was tight to the Did not meet Icalage uDisassembled check valve RCIC-9. Machined u, ' hinge am. test requirements. 140 mils from the disc back.
iMachined the back of- the disc to have the {
' correct disc to hinge arm clearance.
jReactor was in cold shutdown car.lition.
; Reactor Recirc- Stress Corrosion None
; ulation ' Replaced sepacnts of recirculation discharge yalve 4" bypass lines containing linear indi-
; cations detccted by ultrasonic test.
Feactor was in cold shutdown and lines i
~ kere isolated.
t ICIC 1 Relay Coil Opened I.oss of Valve Control peplacedcoilonundervoltagerelayin Power. RCIC inoperable. controller for valve ?-O 2078, RCIC turbine 1stca:a supply valve. l
- l
. 'HPCI operability was demonstrated.
t i s I
-m _ - _ _ _ _ _ - _ _ - _ . _ . _ . - _ - _ . . _ _ _ _ _____---_____w _m-*-+,werT-TMT
- +
- _.-
4001 - Rev. 2 , Till*.*IT'CM RM1\RY _ Sun..ary of Nork Perforned, Incle.linn l Corrective Action Taken to Prevent Repetitlen (if applicable) aal Pre- j System or Cause of the Pasults and Effect cautions Taken to Previde for P.cact ;r j Component Malfunction na Safe 0;o at ic-. . Safety D; ring Repair (if applicable;. RhTIJ I Short cirajit in control Valve inoperable.
- Replaced the motor starter for 30 2399, relay Redundant valves RNCU discharge isolation valve.
operable. ' Refueling Limit switch am locknut "B ridge-Orcr-Core" Adjusted limit switch ara and tightened 1 i Bridge ca:ne loose. refueling interlock locknut using Loc-Tite on threads. l e i inoperabic.
!i 1 RCIC !.DV 2076 Limit and torque switch Valve did net close on Reaoved limit and torque switch assc=bly. !
,_, malfunction-peaaiSly manual actuation - Cleaned old grease fran gears. Reset
= aggravated by grease redundant valve linit and torque switches.
, deterioration operable. :
c' Cleaned the gear assanblics and regreased f with high te=perature grease. ; 1 . Reactor was in cold shutdown condition. {
+
Core Spray Crud on the disc and Did not meet Icakage Disassembled, c1 caned, inspected aai t seat surfaces. test reg 2iremaits. re-assembled with new gaskets check valves ; 9 t
?O 14-1M G B. l Reactor in cold shutdown conditio.h.
I I ! RCIC Valve body distortion Did not meet Icakage Disasseabled check valve RCIC-10. Lapped { test requirements. the disc and seat. Reassaabled with new l gaslet. l l: Reactor was in cold shutdown condition. i l 1
- t i >
l i
- i e
f i s ,. . . _ . . - - . . - i L i i
- - ~
. .: a ,
4001 I' i Rev. 2 . i
- i in1?TraNCI~ MrinRY t I !
i 5 :rarv of Xerk Perfam-d. Incicfir, { Curacctive Action Taken te termt : R.p :iti (if applicablei and he-System or Cause of the ib 9 d ts r.-G r i R + , f ca riunr faien to Provide 'or Rext ;r Componcnt ralfunction ,
. ; hf.: 0;a at ic.. Sfety Daring P.epair [if spplitabit t
[ t HPCI Bent hinge am Did not racet Icakage Disassembled check valve HPCI-9. Lapped I { test requircraents. the disc and sea't. Straightened the hinge ! am and reasseabled with new gasket. ; Welded a strongback on the hinge am. ! l Reactor was in cold shutdown condition. !
,_. Main Steam Foreign material damaged Did not meet Icakage ! ~
Drain Welded a new 3" gate valve in place of i
. valve seat. test requirencnts. ,
outboard steam line drain valve 10 2374 I l w Installed a new valve. Reactor was in cold shutdown condition. Diesel Gener- Oil pump seal failure Small oil leak Installed a new lube oil circulating ator pu=p on #11 diesel. Installed spare pump ard rebuilt the warn i i pu=p. t l i Reactor was in cold shutdown condition. f 3
- i. t f.
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. )
. l IV. Chang,es, Testing and thperiments i
The following sections include a brief description and sinrnary of the safety evaluation for those changes, tests and experiments which were carried out without prior NRC approval, pursuant to the requirunents l of 10CFR 50:59(b). j 1. ADDITION OF PRESSURE AVERAGING MANIFOLD 04 M\lN SrF#1 LINE (SRI-147) ! Description of Cinnge A pressure averaging manifold was installal on the main steam lines inrnaliately upstream of the turbine stop valves. The manifold consists of a header located above and connected to each main steam line by sloped risers. It provides an average pressure signal for the press-ure control systm, and allows main tur iine stop valve testing at a higher power IcVel. Previously the pressure control system utilized a single pressure tap on the A steam line. ' Strimary of Safety Evaluation The main steam line pressure averaging manifold conforms to the original coje and specifications. The manifold consists of schedule 160 piping and valves with a design pressure of 1110 psig at 582*F. A 150% (1665 psig) hydrostatic test was perfonned on the manifold fol!owing installation. The installation utilizes the existing main steam line low pressure switch connections, hence a manifold break in the RUN mode will initiate a safe closure of the Group 1 isolation valves, i 2.
!:ODIFICATION OF INN 7R FILTIR ASSIMBLY ON TWJ C0f(IROL ROD DRIVE _
Description of Clange The inner filter assam 'es on two CRD's have been mWified such that the inner filter is att ched to the top of the stop piston. The mcdification runoves the inner filter frm the flow path during a scram so that a plugged tilter will not affect scram inser tion times. Simnary of Safety Evaluation The intent of the inner filter to protect stop piston seals and CRD j' internals is not affected by this molification. General Electric has i adopted the stop piston mounted inner filter assunbly for their hiel 7RB144B CRD. T' l! 3. DRW7:LL-TO TORUS VACLRI BREAKER TEST SOIINOID VALVES INSTALLATION INSIDE
~
1TDPiDkT(TfT4-45) Description of Change The drywell-to-torus vacuum breaker solenoid valves were moved from l l iV - 1 1 l l
i a location cutside of the toms to a location inside the torus. This change eliminates the possibility of Icakage fran the contain- . ment via the vacun breaker air operator air lines. The air supply line to the solenoid valves located inside the torus utilizal an existing penetration. Caps were weldal on the spare penetrations which resulted frcn the design change. Simrnary of Safety Evaluation The modification is consistent with the original design codes and specifications. The penetration, instnanent line isolation valve and piping within the QC loundry meets design seismic requirunents. No new failure modes are created. The isolation valve fails closed on loss of air or power. NOTE: 1ho air operatal isolation valve is expected to be insta11al during the fall 1975 refueling outage. A locked closcd manual valve is presently providing conttiinment isolation. 4 RDOYAL OF REACIDR 11iEDPLff RlNOUT PIO11rrION CIRWITRY (ff74 51) Description of Change The reactor feedpap runout protection circuit was ranoved frun the feedwater control syston to eliminate the potential for false and undesirable feedwater flow transients. Sunrnary of Safety Evaluation AEC acceptance criteria for ECCS are satisfied without taking credit for feedwater addition. The reduced po'.ential for transients caused by failure or inadvertent initiation of the mnout protection circuit is felt to cutweigh any advantages that the circuit provides.
- 5. INSTA!JATION OF 00faLT FREE SYfLE PROBE (M75-3)
Description of Change A cobalt free sample probe was installed on the feedwater systun down-stream of high pressure boater E 15A. The probe will be used for detemining the cobalt levels in the feedwater. Stranary of Safety Evaluation The probe is designed to withstard the maxinn expected fluid drag forces and vibrational stresses due to vortex shedding. All pipe fittings, valves and tubing were selectcd in acconlance with General Electic, Northern States Power and ASIN specifications for nuclear service. The sample line isolation valve assanbly is designed to withstand the maximum static, dyntunic and vibrational loading. IV - 2 l l
Vc WA DIO A C 11Vi LTFLUE NT R( Lf ASES FOR
...................................12/30/74 THkOUGH 6/30/75 A. f45LCU$ fifLbENIS l l ;l i
- 1. ACPLE GA$ RELFA$F$
A. ILI AL GROSS F ADIDACilVliY (CURIES) jab Fid NAP APR MAY JUN TOTAL 42?(. 20013. 13734. If 868. 22861. 21742. 100250. B. PAAIMUM G AOS$ R ADIDAC TIVITY Rf LE ASE lh ANY G
.............................................Nr HOUk PfRICO (UCl/5FCI JAL FEB MAh APR MAY JUN TOTAL 14t43. 28490. 10347. 9081. ??309. 11943. 32309.
C.
........ ,.................................IES) 101 A1 C.POSS FA0lOACilvlTY BY NUCL10E (C UF J AL FEB MAR APR MAY JUN TOTAL AE133 1737 6060. 10 f 17. 15132. 17716. ! Fees.
xE135 3tt. 9914. 381. 70171. 19 22, 24 1C725 7 KkB5M 75 1205. 34. KR84 3. 4. 5. 1327
- 42. 1361. 9. 7.
KRb7 55. B. 10. It 3 8. 459 9 15 17 AE138 0,43. 204. 19 574 170. 234 291 251. 2009. dR90 7 B. AE139 19,
- 6. 8. 12. 10. 51.
23 19 24. 35. i KoE9 191. 223. 31. 1 51. xt137 186. 238 344 302. 249. 290 242. 1483.
' AE 135M 311. 449 394 1935.
14 17. 14 18. 27. ??. KR83M 2 71. 36. 114 AE133F
- 7. 3. 4 3. 324 127 290. 64 14.
I,XE 131 P 16. 13. 5253 14 30. 15. 12, 12. 1. L RR 63 217 474 1760. 83. 829 3935. 1957 9141. [' L. PtCCfbT Pf f f.CHNI C AL $PECIFICAllOt ANh0AL AVERA i I
...............................................Gf F ELF A SI L lMIT
! JAN FEB NAR APR MAY JUN i ........ ........ ......_. ........ TOTAL 1.75 4.59 3.00 3.92 5.17 4.95 3.86 Y-1 e --4 .- -- r-.---es -i- +--- >g - - , + -
, 20 10EINE RfLEASES '
i A. TOTAL IODINE kADIUACTIVV BY NUEL!OE ICURIES) JAN FEB MAk APR ftAY JUN ' TOTAL 3 1-131 0.427 O.047 0.115 0.188 0.462 0.291 1.530 1-133- 0.233 0.220 0.435 0.690 0.976 0.652 3 195 1-135 0 001 0.266 0.346 0.563 0.743 0.400 2.377
- b. PEkCE .
NT OF TECHNICAL SPECIFICAll0h ANhUAL ~VF RAGE RELE ASE L IMI T JAN FEB MAR APR MAY JllN TOTAL 37.44 11. 19 34.88 63.89 121.72 87.98 60.14
- 3. P AR TICUL ATE R ELE A SE S A. GROSS FADIDALTIVITY IBET A AND GAMPA) (CURIE 51 JAN FEH MAR APR MAY JUN TOTAL 2.82E-02 2.39"-02 3.60E-02 8.60E-02 9.90E-02 6.35E-02 3. 3 7E -01
- b. GF OSS ALPHA RADIOACTIVITY (CURIES!
JAF FEB MAR APR HAY JUN TOTAL 3 59E-06 1 32E.05 1. 25 E.05 3 29E.05 6.26E.05 1.27E.05 1 38E-Oi C. TOTAL GROSS P AD10 ACTIVI TY F EL E ASED OF NUCLIOE S WIT H HALF LIFE GRE ATER THAN 8 OAYS (CURIES) JAN TES MAR APR II A Y JUN TOTAL 1.1kE-u2 h.33E.03 ' .16 E- 03 h.46E-03 1. 80E.T 1.02E-02 5. 2C E.00 i D. PERCENT OF T ECHNIC AL SPECIFIC ATION ANhUAL AVER AGE RELE Ast L IMIT
- JAN TEB MAR APR MAY JUN TOTAL 2.42 1.29 1.07 L.36 6.34 3.28 2.65 i
V.2 { l
B. LIQU10 EFFLUENTS NO L10Ul 0 EF FLUENT RELE ASES WFRE MADE OUPING THI S REPORTING PER IOD. C. SQL 10 WASTE SHIPPED
- 1. TOTAL AMOUNT OF SOLIO WASTE SHIPPE0 ICUBIC FEET)
JAN FEB MAR APR MAY JUN TOTAL 918. 800. 1456. O. 920. 720. 4814. l
- 2. TOTAL ESTIM ATED AC TIVITY ICURIES)
I JAN FEB MAR APR MAY JUN TOTAL O.4 310.5 SC8.1 0. 0 667.1 828.9 2315.0
- 3. DA TES AND DI SPOSI T ION 1/29/75 NUCLEAR ENGINEERING COMPANY, SHEFFIELD, ILLINDIS 2/17/75 NUCLEAR ENGINEERING COMPANY, SHE F F I ELO , ILLIN015 2/17/75 L JC L E A R ENGINEERING COMPANY, SHEFFIELO, ILLINCIS 2 /18/ 75 NUCL E AR ENGINE ER ING COMP ANY, SHE F F I EL D , ILLINOIS 2 / 22/ 75 NUCL EAR ENGINEER ING COMP ANY, SHEFFIELO, ILLINOIS 2/22/75 NUCLEAR ENGINEERING COMPANY, SHEFFIELO, ILLINotS 2/25/75 NUCL E AR ENG INEEQ ! hG COMP ANY, SHE FF I EL O , ILLINOIS 3/ 3/75 NUCLEAP ENGINEERING COMPANY, SHEFFIELD, ILLINGIS 3/ 3/ 75 CHEM N UC LE A R 8NC., BARNWELL, SOUTH C AROL IN A 3/ 5/75 NUCL EAR ENG INEER ING COMP ANY, SHE F F I EL D , ILLINOIS 3/ 7/75 NUC L E AR ENGINEERING COMPANY, SHEFFIELD, ILLINDIS 3/ 23 / 75 NUC LE AR ENGINEER ING COMP ANY, SHE F F I EL O , ILLINOIS 5/ 20 / 75 NUC L E AR ENGINE ER ING COMP ANY, SHEFFIELD, ILLINOIS 5/ 21 /75 CHEM NUCLE AR INC., BARNWELL, SOUTH CAROLINA 5/ 22/ 75 NUC L E A R ENGINEERING COMPANY, SHE FF I ELD , IL L IN0 I S 5/27/75 NUCL E AR ENGINEEk I NG COMPANY, SHE F F I EL D , ILLIN0tS 5/29/75 NUCL E AR ENG INEER I AG COMP ANY, SHE F F I EL D , ILLIN0IS 5/ 30/ 75 NUCLEAR ENGINEERING COMPANY, SHEFFIELD, ILLINOIS 6/ 2/75 CHEM NUCLE AP INC., BARNWELL, SOUTH CAROLIN A 6/ '/'5 CHE M N UC l.E A R INC., BARNWELL, SOUTH C AROLIN A 6, '73 CHEM N UC LE A R INC. , B ARNWELL , SOUTH CAROLINA 6/ a/ 5 CHEM N UC L E A R INC., BARNWELL, SOUTH C AR OLIN A o / lc / (5 NUC L E AR ENGINE ER I NG COMP ANY, SHE FF I EL O , ILLINOIS V-3
4 s f* VI. RADIATION ENVIRONMENTAL MOMITORING The following report on the Radiation Environmental Monitoring Program is presented in a format tailored to the requirements set forth in the current Technical Specifications. This format is also a consistent
.\
transition to future reporting (Regulatory Guide 4.1 and draft Guide 4.X). Data presented in this report is primarily in aummary , and average form to allow more effective evaluation. However all of the individual sample analysis data has been organized and tabulated in a reference docu-ment. This data document is available for review at the Northern States Power Co. general of fices and information
, from it can be copied and furnished on request.
p. l VT-1 l l [ -- -
t i Table 1 provides a summary of the Radiation Environmental Monitoring Program for the Monticello Nuclear Generating Plant as required by Technical Specification Table 4.8.1.* The highest activity monitoring location for each sample medium is included in Table 2. The analytical results for each medium are summarized in the following sections with the six-month average activity levels for each location and medium presented in Table 3. Table 4 provides an explanation for each required sample that was not collected or analyzad. With the exception of possibly milk, none of the sample media indicated any elevated radioactivity attributable to plant operation during the period of this report. Milk Milk, collected during the first month of each quarter f rom eight f arms five or scre mile. f rom the plant site, and monthly f rom f our f arms near the plant, was ar.alyzed for iodine-131 and other gaLma emitters. During the grazing season (May 30 to October 15), weekly samples f rom the four nearby farms were analyzed for iodine-131. All iodine-131 determinations were made utilizing a highly sensitive ion-exchange resin technique and gamma spectrometry (Nal). Quarterly, a strontium-90 determination was performed on a sample from each farm. With the possible exception of the iodine-131 levels observed in milk from the nearby farms, no contribution from the plant operations could be detected. The iodine-131 levels detected in milk from the four nearby farms are higher than those observed elsewhere during this six month period
- Revised 1/15/75 - Change #12.
1 VI-2
I
'but are consistent with levels observed during the summer of 1974 at other locations. Sampling of the four nearby f arms began this year.
Aquatic Radioactivity Gross beta and gamma emitter concentrations in the aquatic environment were determined by analysis of water, bottom sediments, algae, insects and aquatic vegetation from the Mississippi River and 2 lakes. Analyses were also performed on fish taken from the Mississippi River, r.nd topsoil from fields irrigated with water taken from the Mississippi River downstream of the plant. Tritium concentrations were determined on monthly composites of the river water and on the monthly lake water samples. Strontium-90 concen-trations were determined on the insect and algae samp1*a but per Technical Specifications change #12, this analysis is no longer pe-formed on samples of river water, lake water, well water, bottom sediments, and aquatic vegetation. No increased concentrations of radioactivity attributable to plant ; I operation were detected in any of the samples. Ground Water Ground water samples, collected f rom two welle on the pla ', site and rive other wells, did not show any measurenble increase in radioactivity due to plant operation. Jerrestrial Deposition Surface deposition of radioactivity was measured by analysis of monthly precipitation and semi-annual field vegetation and topsoil samples. The iodine-131 concentration detected at the plant site is consistent with values observed earlier. Cross beta and strontium-90 concentrations, as well as the concentration of other gamma emitters in all samples were consistent with pre-operational data. VI-3
_ _ _ _ . .. . _ . _ . . _ _ _..._____.__._.__._._.-.____m _ . _ . . _ . _ _ ____._ g, 3 Airborne Radiciodine and Particulate Radioactivity ,, Air samples were collected weekly at eight locations surrounding the l plant site. Iodine-131 was collected using carbon impregnated air filters 7 exposed in tandem with the particulate filter. The eight carbon filters were analyzed simultaneously for iodine-131 using gaina spectrometry (Nal). If iodine-131 was detected in the composite, the filters were analyzed separately. The individual particulate filters were analyzed for gross beta activity and a monthly composite of all filters was analyzed for gamma emitters using gamma spectrometry (GeLi). All gamma and beta emitter concentrations were consistent with data of previous years. Gamma Radiation The external radiation dose was measured at the fourteen locations required by the Technical Specifications as well as sixteen other sites using calcium fluoride thermoluminescent dosimeters (TLD's) that were housed in an aluminum-lead-tin shield which compensates for the detection over-response occurring from low-energy gamma radiation. To provide information on radiation levels that may be attributable to plant operation, the locations were grouped as " reference" or " control" (over 2.0 miles awcy from the Monticello stack), " indicator" (less than 2.0 miles from the Monticello stack), and "Monticello" (within the city of Monticello, Minnesota). The six-month accumulated gamma dose is 27.2 mr for all reference locations and 27.7 mr for all indicator locations yielding a not six month dose of 0.5 mr. VI-4
J The highest dose was measured at location #6. The six-inonth measured net dose for this location was 4.3 mr above the average fot all reference locations. This level is not considered to be attributable to plant operations because statian 6 is located 12.4 miles from the plant. These dose values show that the TLD network did not measure radiation i levels attributable to plant operation during the period of this report. See Table 3. page VI-11 for detailed TLD data. See Figure 1 for a graph of net gamma exposure vs. off-gas release rates from the plant. Film Badge Results i During the report period, two film badges in two different exposure periods showed readings above the minimum "M" (less than 10 millirem). These two badges were located at sites approximately five miles from the plant boundary and based on TLD data, the readings from these badges are considered spurious. Small Game Animals in accordance with Technical Specifications change #12, January 15, 1975, small game animals have been included in the environmental monitoring program and are included in this report for the first time. No concentrations of radioactivity attributable to plant opera
- ion were detected in the indicator or reference samples.
( > t VI-5
I I 1 1 Table 1
\
SUMMARY
OF RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM Monticello January through June,1975 Number of Radiation Number of Number of Locations Above Attributable to ,'
, Medium Locations Samples. Background Plant Operation River Water 3 75 0 0 Lake Water 2 4 0 0 Well Water 7 14 0 0 Precipitation 2 12 0 0 (Terrestrial Deposition)
Lake & River Bottom 4 4 0 0 Sediment Plankton, Algae, or Insects 4 6* 0 0 Aquatic Vegetation 4 2* O O Clams 2 0* Fish 2 6* 0 0 Milk (Red ons 1-4) 8 16 0 0 Milk (Nearby dairy farms) 4 32 ** ** Topsoil 6 6 0 0 Vegetation 3 3 0 0 Edible Cultivated Crops 3 0' Air Samples (Particulate) 8 200 0 0 Air Samples (Airborne Iodine- 8 201 0 0 131) TLD's 30 658 0 0 Small Game Animal 2 2 0 0 OSee Table #4 COSee Narrative Section > l vi-6
- '~'
Table 2 SUIDIARY REPORT OF RADIOACTIVITY IN Tile L:;VIR0hME!.T llIGilEST ACTIVITY LOCATION FOR EACil MEDIUM .
!!onticello (Jan. 1 - June 30, 1975) crivity Distance Direction High Lou Average Analysis fledf a, _
Miles Degrees Units pCi/1 20 (4 8.1 Gross Beta iiver Water (In unstream) c 292 22 20 21 Lake Water (Locke Lake) 5.6
" 6.5 1.2 249 10 3 '.' ell Water (Schultz) "
40.0 310 pCi/m2 13,600 1390 6200 Precipitation (Terrestrial Deposition) (MDil) " " pCi/g 51 - 51 l Bottom Sediment (Mississippi - - River-Downstream) " " l 292 24 - 24 Aquatic Insects (Locke Lake) 5.6 90 140 - 140 Algae (Big Lake) 4.6 d A 5.6 292 65 - 65 4 quatic Vegetation (Locke Lake) " " 8.4 6.5 7.5 Fish (Mississippi River- - Downstream) pCi/1 1.6 (0.18 0.42 1311 Fulk (Shovelain) 3.0 250 40 9 19.4 137Cs Milk (Shovelair) 3.0 250 13.0 335 18 14 16.0 90S r Milk (Dwinger) pCi/g 52 - 52 Gross Beta Topsoil (Dechene) 4.7 118
" 72 72 280 Field Vegetation (Field #3) 1.1 11.1 306 pCi/m3 0.20 0.017 0.087 Air Particulates (Station #1) 12.4 133 mr/4 weeks 8.2 6.6 31.5 Gamma TLD's (St ation #6) pCi/g 0.19 -
0.19 137Cs Small Gaoc Animal (Flesh) (On - - Plant S ite)
, =
- n. .
'- . ~
Table 3
SUMMARY
REPORT OF RADIOACTIVITY IN TrlE EWIRONME!T' Monticello 6 Month Average Activity Levels Sample / Location Gross Beta Tritit.2 . 3-137 River Water pCi/1 pCi/1 pCi/l Upstream of Plant (1000 ft.) 8.1 2 0.7 540 t 80 1.1 1 0.2 Downstream of Plant (1000 ft.) 1.82 0.7 350 2 80 1.1 1 0.2 St. Paul Water Intake 8.1-1 0.7 470 2 80 1.0 1 0.2 Lake Water pCi/1 pCi/1 pC1/1 Big Lake (Mitchell) (4.6 mi.0 90*) 2014 500 1 130 2.9 2 1.3 Locke Lake (5.6 mi. O ?92*) 21 1 4 400 1 130 (2.0 Well Water pCi/1 pCi/1 ,pCi/1 Gauthier (1.3 mi. 0 130*) 4.2 1 3.1 660 1 140 s1.9
$ Schultz (1.2 mi. 0 249*) 6.5 1 3.4 (185 (2.2 E Swanson (1.5 mi. 0 277*) 3.8 1 2.6 620 1 140 (1.7 Trunnel (0.3 mi. 0 214*) (5 320 1 140 1.8 1 0.9 l Plant Well #1 (On-site) 2.8 1 2.0 280 1.120 (2.2 Plant Well #2 (On-site) 61 3 (185 (2.0 City of Monticello (3.2 mi. 0 128*) 51 3 (190 (2.3 Fish (Flesh) pCi/gm pCi/gm Upstream (1000 ft.) '
7.5 1 0.1 0.034 1 0.005 Downstream (1000 ft.) 7.5 1 0.1 0.051 1 0.019 Bottom Sediment pCi/gm pCi/gm Big Lake (4.6 mi. 0 90*) 47 1 3 2.2 1 0.1 Locke Lake (5.6 mi. 0 292*) 14 1 2 1.6 ! 0.1 l Mississippi River, Upstream (1000 ft.) 37
- 2 0.032 1 0.010 Mississippi River, Downstream (1000 f t.) 51 t 3 0.019 1 0.002 Aquatic Vegetation pCi/gm pCi/gm Big Lake (4.6 mi. 0 90*) 44 1 3 0.63 1 0.07 Locke Lake (5.6 mi. 0 292*) 65 i 3 0.10 t 0.03
.aj _ l ..~1
Table 3 (Cont.) SUP01ARY REPORT OF RADIDACTIVITY IN Tile FliVIRO!;MS.'T Monticello 6 Month Average Activity Levels Sample / Location Gross Beta Strontium-90 Cesium-137 Iodine-131 Field Vegetation pCi/gm pCi/gm pCi/gm 69 z3 0.087 e 0.024 (0.20 Field #1 (0.8 mi. @ 100*) (0.29 Field #2 (0.7 mi.@ 198*) 58 z3 0.033 e 0.025 , 72 s3 (0.024 (0.35 l Field #3 (1.1 mi. 0 280*) pCi/1 pCi/1 pCi/1 Milk Dwinger (13.0 mi. @ 335*) 16 1 12 s 2 (0.24 Kirchenbauer (11.5 mi. @ 323*) 12 2 1 5.5 1.7 (0.27 Kotilinek (5.6 mi.0 230*) 4.7 1 0.6 8.2 1.5 (0.28 Vandergon (8.3 mi.0 247*) 4.0 2 0.5 8.4 e 1.4 (0.31 Holland (8.1 mi. @ 199*) 3.1 2 0.5 4.4 2 1.6 (0.20 Hopkins (7.6 mi. 0 193*) 5.1 1 0.5 7.2 s 1.3 (0.21 5.2 1 0.6 8.0 t 1.6 (0.22 Becker (10 mi. @ 130*) (0.18 4.02 0.5 7.4 1 1.5
< Vetsch (9.4 mi. @ 128*) 0.35 0.06 q* Nelson (2.4 mi. @ 269*) 7.2 2 0.6 8.6 2 1.0
- Olson (2.5 mi.924*) 6.81 0.6 8.8 t 1.1 0.28 1 0.05 i
Peterson (2.3 mi. @ 111*) 4.42 0.4 10.9 t 1.0 0.42 2 0.07 Shovelain (3.0 mi. @ 250*) 10. 2 t 0.7 19.4 s 1.1 0.42 e 0.05 Topsoil pCi/gm pCi/gm Vegetation Field #1 (0.8 mi.0100*) 5123 0.34 2 0.02 Vegetation Field #2 (0.7 mi.0198*) 51 z 3 0.22 2 0.02 Vegetation Field #3 (1.1 mi. 0 280*) 48 t 3 0.27 t 0.02 Dechene Potato Co. (4.7 mi. @ 11.8*) 5223 0.17 2 0.02 Dechene Potato Co. (4.9 ni. @ 115') 51 3 0.091 0.012 Ewing Potato Co. (5.7 si. @ 13 0 ) 51 z 3 0.04 3 0.02 Air Particulates & Airborne Iodioe-131 pCi/m3 ,pci/m 3 0.087 2 0.002 'O.02 Station #1 (11.1 mi. @ 3u6') i 0.02 Station #2 (3.8 mi. @ 334*) 0.080 ! 0.002 0.0092e 0.0021 Station #3 (6.4 mi. @ 290*) 0.055 1 0.001 0.083 0.002 0.011 t 0.002 Station #4 (0.6 mi. @ 100*) IO.016 Station #5 (10.4 mi. 0 106*) 0.055 t 0.001 (0.016 Station #6 (12.4 mi. @ 133*) 0.077 2 0.001 0.084 1 0.002 (0.013 Station #7 (8.8 mi. @ 39*) 0.0072 t 0.0015 Station #8 (10.3 mi. @ 227*) 0.043 1 0.001 I .
Table 3 (Cont.) .-
SUMMARY
REPORT OF RADIOACTIVITY IN THE ENVIRONMENT Monticello 6 Month Average Activity Levels Sample / Location Cross Beta Tritium Strontium-90 Iodine-131 Cesium-137 other l 's Precipitation (Terrestrial deposition) pCi/m2 pCi/m2 2 pCi/m2 pCi/m2 pCi/m2 On-site Station (0.5 mi. 9 105*) 4600 1 100 350 1 80 4kCi/m
! 2 78 15 94 i 8 1180 1 60 Minnesota Dept. of Health (40 mi. 0 310*) 6200 1 100 300 1 80 54 1 2 16 i 8 9116 1450 1 50 Algae pCi/g pCi/g pCi/g pCi/g Big Lake (4.6 mi. 0 90*) 140 1 10 0.93 1 0.15 2.0 1 0.5 9.7 1 2.4 Locke Lake (5.6 mi. 0 292*) 85 2 3 0.56 1 0.26 1.9 2 0.2 6.6 2 1.0 Aquatic Insects pCi/g pCi/g pCi/g pCi/g Big Lake (4.6 mi. 0 90*) 22 1 2 0.63 i 0.43 7.6 2 1.1 6.5 2 3.0 Locke Lake (5.6 mi. 0 292*) 24 1 2 (0.32 3.7 1 0.7 6.0 1 2.8
< Mississippi River-Upstream 21 1 2 0.45 1 0.22 1.3 1 0.6 1.8 2 1.8 [ Mississippi River-Downstream 21 1 2 (0.15 1.9 1 0.7 0.7 1 0.7 o Air Particulate (Composite) pCi/m3 pCi/m3 Eight Station Analysis 8.5 10.7 E-4 2.010.1 E-Small Game Animals pCi/g pCi/g Indicator (On Plant" Site) Flesh 0.19 1 0.04 0.54 1 0. Liver 0.12 1 0.04 0.14 2 0. Reference (12 mi.0 "258*) Flesh 0.12 1 0.05 0.33 1 0.- Liver 0.13 1 0.05 0.41 1 0. s 5
l- [ Tabic 3 (Cont.) 1
SUMMARY
REPORT OF RADIOACTIVITY IN THE ENVIRONMENT Monticello 6 Month Average Activity Levels Semple / Location Gamma Dose
- Thermoluminescent Dosimeters ar/6 months .
REFERENCE Station #1 (11.1 mi. 0 306') 25.6 Station #2 (3.8 mi. 0 334') 23.9 Station #3 (6.4 mi. 0 290') 25.8 Station #5 (10.4 mi. 0 106') 29.1 1 1.8 Station #6 (12.4 mi. 0133') 31.5 Station #7 (8.8'mi. 0 39*) 28.2 Station #8 (10.3 mi. 0 227') 29.4 Station #42 (2.7 mi. 0 345') - 25.3 1 0.5 Station #43 (2.6 mi. 0 10') 24.9 2 3.5 Station #44 (2.9 mi. 0 27*) 25.8 1 0.3 Station #45 (3.5 mi. 0 344') 27.0 Station #47 (8.1 mi. 0 357') 27.1 1 0.7 Station #48 (8.7 mi. 0 10') 29.0 Station #55 (4.7 mi 0 126') 29.4 INDICATOR Station #4 (0.6 mi. 0 100') 26.4 2 0.8 Station #13 (0.6 mi. 0 242*) 26.5 5tation #14 (O.4 mi. 0 155') 26,3 1 0,8 { Station #15 (0.8 mi. 0 100') 28.5 Station #16 (0.9 mi. 0 321') 26.9 Station #17 (0.7 mi. 0 250') 27.6 2 0.9 Station #18 (1.7'mi. 0 110') 30.9 2 1.2 Station #20 (0.8 mi. 0 65') 24.2 1 1.0 Station #24 (0.6 mi. 0 208*) 29.0 1 0.9 Station #25 (0.8 mi. 0 157') 28.5 1 0.9 I Station #26 (0.8 mi. 0 140') 29.8 2 0.7 Station #27 (0.9 mi. 0 104*) 27.2 2 0.9 Station #50 (1.8 mi. 0 132') 28.9 2 0.7 MONTICELLO Station #51 (2.7 mi. 0 132*) 28.9 1 0.7 Station #53 (3.4 mi. 0 130') 28.8 1 0.8 I' ' L, Station #54 (3.0 mi. 0 122') 29.2 1 0.5 i
- Includes background VI-11
l s ! Table 4
SUMMARY
REPORT OF RADIDACTIVITY IN THE ENVIRONMENT Honticello Samples Not Collected Medium Location Period Reason Algae Mississippi River Upstream Semi-annual Not available due to and Downstream high water Aquatic Vege- All Locations 1st quarter Not available due to tation ice conditions Mississippi River Upstren 2nd quarter Not available due to and Downstream high water Clams All Locations Both quarters Not available due to ice conditions and high water Fish Mississippi River Upstream ist quarter Not available due to dangerous ice condition Air Filters Station #1 3/11/75 to 3/25/75 Sampler malfunction 6/10/75 to 6/17/75 " station #3 3/18/75 to 3/25/75 " " 6/24/75 .to 7/1/75 station #6 4/15/75 to 4/22/75* Personnel Error Station #7 1/14/75 to 1/21/75 Sac:pler malfunction Station #8 6/24/75 to 7/1/75 " " TLD's Station #2 1/22/75 to 2/18/75 Bad reading 5/14/75 to 6/10/75 Lost in field Station #8 4/16/75 to 5/13/75 " " " Station #15 2/19/75 to 3/18/75 " " " station #18 " " 1/22/75 to 2/18/75 " 2/19/75 to 3/18/75 5/14/75 to 6/10/75 4
- Particulate Filter Only VI-12
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CONTROL NO: 9?N9 FILE: __REPOUS F11.E FROM: Northern States Pwr Co DATE OF DOC DATE REC'D LTR TWX RPT OTHER Minneapolis, Mn t 0 Mayer 6.-J9-75 9-2 75. XXX ORIG CC OTHER SENT AEC POR vr TO: Mr Giambusso none signed SENT LOCAL PDR XX PROPINFO INPUT NO CYS REC'D DOCKET NO; CLASS UNCLASS CXXXXX 1 50-263 DESCRIPTION: ENCLOSURES: Ltr trans the following: Semi- Annual Operating Report #9 for the period January 1 to June 30, 1975.................... M NOT REMOVE
~ (40 cys enel ree'd)
ACKNOWLEDGE " ** PLANT N AME: Monticello FOR ACTION /INFORMATION 9-2-75 ehr
- NOTE: 3 cys Or THIS BUTLER (L) SCHWENCER (L) ZIEUIANN (L) REG AN (E) d W/ Copies W/ Copies / WMCopics W/ Copics PACKAGE ARE FOR DIST 10 STOLZ (L) DICKER (E) LE AR (L) EPA WITil THE FOLLOWING CLARK (L)
W/ Copies W/ Copies W/ Copics W/ Copies llRE AKDOWN : 2 CYS EPA P ARR (L) VASSALLO (L) KNIGHTON (E) SPELS !!DCS ; 1 CY EPA REGION. W/ Copies W/ Copies W/ Copies W/ Copies KNIEL (L) PURPLE (L) YOUNGBLOOD (E) W/ Copies W/ Copies W/ Copies W/ Copies INTERNAL DISTRIBUTION REG ~ TECH REVIEW DENTON LIC ASST A/T IND . DR SCHROEDER GRIMES ,, R. DIGGS (L) BW4TTiXAN OGC, ROOM P 506A #MACCARY GAMMILL H. GE ARIN (L) SALTZMAN GOSSICK/ST AF F KNIGHT p K AST NE R E. GOULBOURNE (L) MELTZ CASE PAWLICKl BALLARD P. KREUTZER (E) GIAMBUSSO Sil AO SP ANG LE R J. LEE (L) P_L ANet. BOYD STELLO M. RUSHBROOK (L) MCDONALD MOORE (L) HOUSTON ENVIRO S. REED (E) CHAPMAN DEYOUNG (L) NOVAK MULLER M. SERVICE (L) DUSE (Ltr) SKOVHOLT (L) 4 0SS DICKER S. SHEPPARD (L) E. COUPE
/IPPOLITO KNIGHTON M. SLATER (E) PETERSON GOLLER (L) (Ltr)
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/Fil E & REGION I2) BENAROYA f _Qe.u d q. V. WILSON (L) WIGGINTON MIFC/PE (ad) VOLLMER HAROTSS R. INGRAM (L) ,,, ll AN AU ER STEELE / J. COLLIUS M. DUNCAN (E)
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9 AUXILIARY SYSTEMS (This section will be provided in a supplement to this SER.) ABWR DSER
._.._ . _ _ ____ .._m _ . .. ___ ._ _ ____._.__._= _ __ _ __ _ __ _ _- _ ._.._ _ _ _ _ _ .__.._ _ . _.
i i 10 STEAM AND POWER. CONVERSION SYSTEM (Thin section will be provided in a supplement to this SER,) i i l l
\
l ( I I i 10-1 ' ABWR SER I __ _. _ - _ _ . _ _ . . _ _ . _ _ _ . _ _ _ _ _ _ . . _ _ . . _ _ _ _ _ _ . , _ _ _ . . , . . . - . , . _ . _ _ _ , . . . _ _ _ _ . . . ~ . , _ , . _ , .
11 RADI0 ACTIVE WASTE MANAGEMENT {This section will be provided in a supplement to this SER.) i n ABWR DSER
i 12 RADIATION pyoygcy3ay (This section will be provided in a supplement to the SER.) I ( i i I i r f I l l I I l 12-1 ABWR DSER
13 CollDUCT OF OPERATIOllS (This section will be provided in a supplement to this SER.) l 13-1 l ABWR DSER
14 INITIAL TEST PROGRAM (This section will be proVided in a supplomont to this sgg,) J l 14-1 ABWR DSER
l I I f 15 TRANSIENT AND ACCIDENT ANALYSIS (This section will be provided in a supplement to this SER.) l l
\
l i i l 15-1 ABWR DSER l l_-
- . - , . . . -_, .. . , , , _ - , _ , . . . , _ = . _ , _ . _ _ . . _ _ . , _ . _ . _ . . _ _ , _ _ - _ _ . _ . _ . _ , . _ _ , _ , . . _ . _ , _ . , _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ . . , _ _ . . _ . . . , _ _ _
l 16 TEclINICAL SPECIFICATIONS (This section will be provided in a supplernent to this sER,y 16-1 ABWR DSER
. - .___ ..._ ...-, . _ . . ~ _ _ _ _ - .
17 QUALITY ASSURANCE 17.1 Quality Assurance Durina the Denian Phase 17.1.1 General The quality assurance (QA) program for the design phase of the ABWR is described in Chapter 17 of the ABWR SSAR. Chapter 17 refarences the General Electric Company (GE) QA topical report,
" Nuclear Energy Business Operations Quality Assurance Program Description," May 1987, NEDO-11209-04A, Revision 7, which the staff reviewed and found acceptable. Chapter 17 also provides additional QA information specifically applicable to the ABWR. GE is responsible for designing the ABWR and submitted the SSAR. The staff assessed GE's description of the QA program for the design phase of the ABWR to determine if it complies with the requirements of 10 CFR Part 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,"
and with applicable QA-related regulatory guides listed in Table 7.1. The basis of the staff's review was SRP Section 17.1. 17.1.2 Organization The structure of the organization responsible for the design of the ABWR and for the establishment and execution of the design-phase QA program is shown in Figure 17.1. The line organizations have been assigned specific QA responsibilities, including both internal audits and audits of suppliers to ensure compliance with the QA program. Audits conducted by GE's Nuclear Quality Assurance (NQA) organization are superimposed on these audits. 17-1 ABWR DSER
I I l The General Manager of Nuclear Operations is responsible for i ensuring that (1) the intent of GE's nuclear quality policy is ' reflected in its nuclear products and services, (2) a system is in place to independently assess the performance of organizations that affect the quality of these products and services, and (3) a system is in place to resolve issues that could affect GE's ability to satisfy its nuclear quality policy and other quality-related commitments. 1 NQA is a staff organization responsible for establishing the nuclear quality policy and procedures that are 1: sued by the Vice President and General Manager of GE Nuclear Energy. NQA is also responsible for (1) auditing the various line organizations involved in the nuclear business and ensuring conformance of these organizations' procedures and practicos with applicable corporate and nuclear quality-related policy and procedures, (2) ensuring integration of the organizations' quality planning into an > offective QA program, (3) participating in management review boards that operate independently of the design verification by the line organizations, and (4) specifying how the line organizations are to comply with the nuclear quality policy and procedures. For the ABWR design, NQA is responsible for coordinating and integrating the QA program as it relates to engineering and management of the project. A quality council aids NQA in fulfilling its responsibilities. The council's responsibility is to ensure total quality system coverage, uniformity, consistency, and continuity and to eliminate system deficiencies. The Manager, NQA chairs the quality council. Members of this council, as shown in Figure 17.1, are the managers i responsible for QA in each of the major nuclear organizations. The council provides these managers direct access to top-level l I l l 17-2 ABWR DSER l l t .
management and provides a forum for the review of quality problems I and corrective actions. l l l The line organizations are responsible for planning and implementing the QA functions performed within their areas of responsibility so that each organimation's QA program complies with the nuclear QA policy and procedures established by NQA. The individual QA managers report to their department level management and have the organizational independence and authority to identify quality-related problems; initiate, recommend, or provide solutions pertaining to conditions adverse to quality; and verify l implementation of such solutions. GE and its major technical associates, Hitachi and Toshiba, designed the ABWR. The lead responsibility to produce each speci-fication (through the major purchasing specifications) and drawing is assigned to one design organization within GE Nuclear Energy, Hitachi, or Toshiba. The content of each of these documents is reviewed and approved by GE engineering personnel, and GE is responsible for the design and the supporting calculations and records for the ABWR. GE Nuclear Energy engineering organizations are responsible for the ABWR design and design control by 4 (1) ensuring incorporation of applicable regulatory requirements, codes, standards, criteria, and design bases into the design (2) ensuring incorporation of project design requirements into the design (3) translating the design information onto the appropriate design documents 17-3 ABWR DSER
(4) verifying the design adequacy either through independent design review, the use of t;1ternative or simplified calculational methods, or the performance of a suitable testing program (5) coordinating design activities cmong interfacing design engineers and design organizations (6) reviewing, approving, issuing, and distributing design documents under a controlled document system (7) controlling design changes and changes to design documents in accordance with documented procedures (8) providing for the retention, storage, control, and retriev-ability of design record documents (9) taking corrective action as necessary to correct design errors and to improve the design control function 17.1.3 Quality Assurance Program GE etructured its nuclear QA program to satisfy Appendix B to 10 CFR Part 50 and the provisions of the NRC guidance shown in Table 17.1. This QA program is used for the design of the ABWR. The program is implemented by means of written policies, proceduras, and instructions. These documents control quality-related activities in accordance with the requirements of Appendix B to 10 CFR Part 50 and with applicable regulations, codes, and standards. The GE Nuclear Energy QA organizations are responsible for ensuring that procedures and instructions provide 17-4 ABh'R DSER
for meeting the QA requirements. In addition, QA personnel conduct reviews and audits to verify the effective impicmentation of the program. GE's nuclear QA program requires that implementing documents encompass detailed controls for (1) translating codes, standards, regulatory requirements, technical specifications, engineering requirements, and process requirements into drawings, specifica-tions, procedures, and instructions; (2) developing, reviewing, and appreving procurement documents and changes thereto; (3) prescribing all quality-related activities by documented instructions, procedures, drawings, and specifications; (4) issuing and distributing approved documents; (5) purchasing items and services; (6) identifying materials, parts, and components; (7) performing special processes; (8) inspecting and/or testing materials, equipment, processes, and services; (9) calibrating and maintaining measuring and test equipment; (10) handling, storing, and shipping items; (11) identifying the inspection, test, and operating status of items; (12) identifying and dispositioning nonconforming items; (13) correcting conditions adverse to quality; (14) preparing and maintaining QA records; and (15) auditing activities that affect quality. Training ar.9 experience requirements are defined for each position in the GE Nuclear Energy organization. In addition, GE provides for indoctrinating and training personnel performing activities affecting quality to ensure that appropriate proficiency is achieved and maintained. The indoctrination and training are carried out through documented procedures, on-the-job training, personal contacts, and meetings. The training also ensures that personnel responsible for quality-related activities are instructed as to the purpose, scope, and implementation of the quality-related manuals, instructions, and procedures. 17-5 ABWR DSER
1 1 The ABWR design and changes to it are formally verified. Design verification is a process for an independent review of designs against design requirements to confirm that the designer's methods and conclusions are consistent with requirements and that the resulting design is adequate for its specified purpose. Design verification is performed and documented by persons other than those responsible for the design, using the method specified by the design organization. Designs are verified by one or more of the following methods: design review, qualification testing, alternative or simplified calculations, or checking. Team design reviews are continuing reviews of design, selected by engineering management, to evaluate design adequacy that includes concepts, the design process, methods, analytical models, criteria, materials, applications, or development programs. When appropriate, team design reviews are used to verify that product designs meet functional, contractual, safety, regulatory, industrial codes and standards, and GE Nuclear Energy requirements. The selection of the design review team depends on the product design and the type of review. Each team's technical competence encompasses three broad categories: (1) those with broad experience on similar products; (2) those with specialized technical expertise such as in heat transfer, materials, and structural analysis; and (3) those with a functional expertise such as QA, manufacturing, engineering, and product service. For the ABWR design, the lead design organization prepares the document and circulates it internally for engineering review, approval, and design verification. Evidence of verification is entered into the design records of the responsible design organization. Each document is distributed to the design organizations of the other parties for their review and approval 17-6 ABWR DSER
of technical content,and design interfaces. All comments resulting from this proccas are resolved. After resolution of the comments, the design verification is reviewed and, when necessary, updated to ensure that changes did not invalidate the original verification. After final agreement is reached, the document is finalized by the lead design organization, circulated to the other parties for their approval signatures, and then issued. Changes to ABWR documents are handled similarly. Differences between international and domestic designs are identified in a controlled list for future design action and i application. GE Nuclear Energy's QA organizations are responsible for establishing and implementing the audit program. Audits are performed in accordance with preestablished written checklists by qualified personnel not having direct responsibilities in the areas being audited. Periodic audits are performed to evaluate all aspects of the QA program including the effectiveness of implementation. The QA program requires the review of audit results by the person having responsibility in the area audited to determine and take corrective action where necessary. Follow-up audits are performed to determine if nonconformances and deficiencies have been effectively corrected and the corrective action precludes repetitive occurrences. Audit reviews, which indicate performance trends and the effectiveness of the QA program, are reported to responsible management for review and assessment. [ 17-7 ABWR DSER
17.1.4 Conclusion on the basis of its detailed review and evaluation of the QA ' program description contained in Chapter 17 of the ABWR SSAR, the staff concludes the following: (1) The organizations and persons performing QA functions have , the required independence and authority to effectively carry out the QA program without undue influence from those directly responsible for cost and schedule. (2) The QA program describes requirements, procedures, and controls that, when properly implemented, comply with the requirements of Appendix B to 10 CFR Part 50 and with the acceptance criteria in Section 17.1 of the Standard Review Plan (NUREG-0800, Rev. 2). Accordingly, the staff concludes that the description of GE's QA program complies with applicable NRC regulations. 17.1.5 Implementation During its review of the QA program described in the ABWR SSAR, the staff audited the implementation of the program at GE's ' offices in San Jose, California. The report of this audit is in the Commission's Public Document Room, the Gelman Building, 2120 L Street NW, Warshington, D.C. On the basis of the san:.le of design activities audited, which included Hitachi and Toshiba documents requested by the staff and trr.nslated into English, the auditors concluded that the design QA programs implemented by GE, Hitachi, i and Toshiba meet the applicable requirements of Appendix B to 10 CFR Part 50 and are acceptable for designing the ABWR. l l 17-8 ABWR DSER
__ . -__ . _ _ _ - - . . _ . ~ . _ _ . . _ - . . . . - . - TABLE 17.1 Quality Assurance Regulatory Guido Commitments Applicable Number Revisi SD Date 1.8 1 9/75 1.26 3 2/76 1.28* 3 8/85 1.29 3 9/78 1.30 0 8/72 1.37* O 3/73 1.38* 2 5/77 1.39 2 9/77 1.58* ** 1.64* ** 1.74 ** 1.88* ** 1.94 1 4/76 1.116* 0-R 6/76 1.123* ** 1.144 ** 1.146* ** (
- NRC accepted the GE Nuclear Energy positions given in the quality assurance topical report, NEDO-11209-04A, Revision 7, May 1987.
** Superseded by Revision 3 of Regulatory Guide 1.28.
1 i 1 l l l 1 l 17-9 , ABWR DSER l
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18 CONTROL ROOM DESIGN Rgyygg (This section will be provided in a supplement to this seg,y l I ) 10-1 ABWR DSER
19 RESPONSE TO SEVERE ACCIDENTS (This section will be provided in a supplement to this SER.) J l l l i 19~1 ADWR DSER
i 20 COMMON DEFENSE AND SECURITY (This section will be provided in a supplement to this SER.) l 20-1 ABWR DSER
21 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS l l (This section will be provided by utility applicants who references the ABWR design. ) i i l l I l i 21-1 l ABWR DSER a_. - . _ - . . - - . - - - - - . . - . . .. .- _ . - - - -. -.
_ _ _ _ _ . _ . _ _ _ _ _ . _ . _ . _ _ . . _ _ . - _ _ . _ . . _ . _ _ _ . _ _ ________._..__.__...__._.______._..__m__. ___ 22 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (This section will be provided in a suppic.iuent to this SER.) ABWR DSER
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I e 23 CONCLUSION (This section will be provided in a supplement to this SER.)
+
ih t l l l 1 I I 23-1 1 ABWR DSER ' l 1 I. , -- , - _ - _ , ,_ , _ , _ _ _ _ _ _,
l APPEllDIX A CllRONOLOGY - To be provided in the Final SER. A-1 ABWR DSER
_ _ . . . .__. ,__ _ _ _ _ _ _ _ _ . _ _ . , _ . _ _ ~ _ _ . _ . . _ _ . _ _ . _ . _ _ _ _ . _ _ . - - _ . . . . . . _ _ _ _ _ _ . . _ . _ _ _ t i APPENDIX D + NRC REVIEW GUIDANCE e s k d h a
-1 ABWR DSER
f' "'( UNIT st'sT AT s s
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'* AUG 1 1M1 Project ho, 671 Mr. Ricardo Artigas, Manager Licensing & Consulting Services General Electric Company Nuclear Energy Business Operations 175 Curtner Avenue, Mail Code 652 San Jese, Celtiernia 95125
Dear Mr. Artigas:
SUSJECT: ADVANCED BOILING WATER REACTOR LICENSING REY 1EW BASES As you know, for the past several months the staff has been developing, wi'th coordination and input from GE, certain licensing review bases for the staff review of the forthcoming Advanced Boiling Water Reactor (ABWR) Final Design Approval / Design Certification application. These bases address the review process and selected technical issues. In certain key areas, Comission
- policies and staff positions are still under development. The Licensing Review Bases represent our understanding of certain approaches which GE has proposed and comitted to follow in the ABWR design and license application in ,
crder to pemit the- review to proceed efficiently until final Comission positions and staff requirements are defined and implementad. A copy is enclosed for your infomation and use. The staff believes that these GE proposals and comitments are adequate to start the review of the ABWR SAR l upon submittal. , Please let me know if you have any questions. .. 4 Sincerely.
. An Thomas E. Murley, Director Office of Nuclear Reactor Regulation i
i
Enclosure:
l As stated i
- ~ . - - , , . . - . . . . , - , _ _ , , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l Enclosure GE Advanced Boiling Water Reactor Licensing Review Bases August, 1987 tm
1 INTRODUC710N GE intends to subnit an application for final design approval (FDA) and design certification (DC) for the Advanced Boilinc Water Reactor (ABVR), Initial portionsoftheSafetyAnalysisReport(SAR)willbesubmittedbeginninginthe fa)) of 1987. Both the NRC staff and GE believe that staff's safety review of the SAR will proceed more smoothly if certain licensing review bases are established before the review starts. These bases are intended to address aspects of the review process and certain technical issues that have caused difficulties in past reviews of standard plant designs. The Standard Review Plan (SRP) is the basic staff document which will be used in the ABWR review. The Licensing Review Bases addressed in this enclosure provide st;pplernentary guidance on regulatory issues and areas which are either not addressed at all, or not covered in detail, by the SRP. In many cases, these are areas where the staff's positions and requirements are evolving. These Licensing Review Bases contain no new regulatory requirements. In certain key areas where Comission policies and staff positions are still under development, both the staff and GE have comitted to impiement acceptance criteria which, if satisfied by the ABWR standard design, would result in a licensable design. Should, however, - '1tial new infomation becor>e available that results in new requirements . promulgated by the NRC, they will be addressed during the course of the A6WR review. The staff supports the efforts of the Department of Energy (00E) and GE to obtain design certification (DC) of the ABWR. Once the design has been certi-fied, it could be referenced by a number of applicants for use on a number of different sites without further review except for matters which cannot be reviewed or accepted until a specific f a:;111ty is constructed. These matters could be specifically identified in the staff Safety Evaluation h p6rt (SER). When an applicant references the pre-approved design, the staff would conduct a ccnpliance review to confim that the plant was built in accordance with the DC. The design would be certified for the period specified in the Comission's Policy Statement on Standardization, with an option for renewal. GE has agreed that all Generic and Unresolved Safety Issues relevant to the ABWR will be resolved for the ABWR design before a Final Design Approval (FDA) is issued. Af ter an FDA is issued, new issues will be considered for back-fitting under the provisions of 10 CFR $0.109, or under other applicable Comission regulations. GE is to provide a Safety Analysis Report (SAR) for the entire nuclear island ' design. The SAR will meet all applicable Comission regulations and contain enough information for the staff to complete its safety review.
I 3.1 Scope and Centent cd the ABWR SAD l The scope of the ABWR standard design, as illustrated in Figure 1 1, is,a nuclear isltnd. The SAR is to includt all of the infomation necessary for the staff to cenplete its SER. This includes interfaces (design, construction, testing and operational) betmeen the ABWR standard design and the remainder of plant. GE ray later expand its submittal to include some or all of the remainder of plant. The ABWR standard design employs an advanced boiling water reactor enclosed in a steel lined reinforced concrete containment vessel integrated with the reactor building which, in turn, forns the secondary containment boundary. The reactor building houses the equipment associated with auxiliary systems (such as energency core cooling, residual heat removal and reactor water cleanupsystems). It also houses the fuel handling and storage and the diesel generators, which have traditionally been housed in separate buildings. A separate control building is located adjacent to the reactor building. The control building includes the control room, change rooms and plant superviser's office and provides plant access control. Because GE wishes to obtain an FDA and DC for the ABWR design before any ' applicant, site, architect / engineer or equipment suppliers are identified, it is necessary that GE provide the necessary level of detailed information to enable the staff to complete its review without preempting ccepetitive bidding on any future project that references the certified design. The technical inferr.ation fer the ABWE standard design portion of the submittel must teet the requirements of 10 CFR 50.34(g) and the guidance in Regulatory Guide 1.70 Revision 3, appropriate to the degree of design available for standard designs. The corresponding contents of the SAR are listed in Section 8.4.1. Tests, inspectiers, analyses, and acceptance criteria necessary for an appli-cant to assure that the designs are properly implerented in the plant will also be defined in the.SAR. The applicant will later demonstrate compliance
. trith this design and implementation information.
Section S.4.1 lists the design docufnentation that GE intends to submit to support the AEWR standard design. GE does not plan to submit these types of supporting documents for the renainder of plant for the DC effort. 2.2 Scepe and Content of Future Applications Referencino the ABWR When the approved ABWR standard design is referenced in an application, the a staff's revien of r.atters related to the approved design need consider only whether the interfatt requirem&nts have been satisfied in the referencing application (the applicant's Final Safety Analysis Report (FSAR)). Specif- , ically, for those areas in the remainder of the plant and the site envelope where the ABWR SAR has specified interface requirements, the applicant will have to demonstrate compliance with them. h'o further review of the referenced design will be required when the site envelope parameters fall within the ; desien envelope er.d the interf ace requirements are met. 2 SCHE M E as The schedule for the FDA review of the ABWR design is shewn in Table 2-1. The schedule for the subsequent design certification rulemaking phase depends on, among other factors, the type of rulemaking proceeding selected by the Comis s ion. The range of dates shown is the staff's current best estimate of the design certification duration.
Figure 1-1 Typical ABWR S!te Plan Nuclear Island Nucicar Steam Supply System Primary Containment T " buIldt g Secondary Containment Emergency Core Cooling Sys. Residual IIcat Removal Sys. Emergency Diesel Generators Essential Elcetrical Power Control Radwante Fuel Storage & IIandling ~ butIdtna factitty Equipment Control Room Plant Access Control Emergency Response Center Reactor butIdtng Remainder of Plant - Turbine Building Radwaste Facility j Condensate and Diesel Oil Storage
. ...mm -
Tatu, 2-1 AEtiR FDA Revin Schodule h g 7 C.utula t i ve I Review Elapsed line review Element Sebecule' H e r.t h s " I Chapters 4, 5, 6, 15 9/f7 3/89 19 (Reactor, Reactor Coolant System. Engineered Safety Features, Accident Analyses) Chapters 1, 2, 3, 17 3/B5 9/89 25 (General Description, Site Char-acterization Design of Structures. Co penents. Equiprent, and Systees, OA) Cracters 7-9, 11-14, 16 6/88 12/89 28 (l&C, Electric Power Auxiliary Systers, Radioactive Waste, Radiation Pretection, Conduct of Operations, Initial Tests and Operation Technical Specificatiens) - Chapters 10, 18 1/89 12/89 26 (Steam and Pomer Conversion. Emergency Planning) PRA and Failure Moces & Effects Analysis (FNEA) 1/89-12/89 2B Integrated Review / Final SER 3/29 2/90 30 ACRS Review 9/67 4/90 32 Proposec Decision Date for FDA 9/90 37 Design Certification 49 61
- Time fro?. sutrittal of chapters to issuance of drsft SERs
" From beginning of HEC review-3 CONTENT OF AFPLICATION 3.1 Safety Analysis Repert Femat The ABA SAR and all subsecuent 5AR amendments are to be organized in accorcance with Regulatory Guide 1.70, Revision 3, and the SRP in effect on March 30, 19E7. The application will include the infomation specified by 10 CFR 50.33, 50.34, and Appendia 0 to Part 50. GE will comply with the provisions of 10 CFR 50.34(g)(1)(11).
3.1 Use of Netric Units Because the ABW has been designed for international applicatient, the SAR ray use enetric units in describing equipment dimensions and perfomance. However, the values in the SAR are to have their corresponding English units included, in parentheses, next to the metric values (e.g., camieur fuel ciadding temperature during an accident is 1204 *C (2200'F)). GE intends to include a table of conversion f actors between metric and English units at the front of each vo1U9e of the SAR. 4
- . - - - - ~. - .- - - -
1 1 I i [ La Dett Interchange romat Because of the iecreased use of Computers in the licensing process, GE plans
- to provide a copy of the SAR on a diskette suitable for use on en 16H (or compatible) Personal cceputer (except for drawings and graphs that are not spenable to such portrayal). GE also is to provide the requisite number of hard copies of the SAR specified in 10 CFR 50.30(a), (c)(1) and (3).
4 INCORFORATION OF FUTURE !$$UIS I As stated in its Severe Accident Policy Statement (see Section 7 below), the < l Comission expects all new power plant designs to address all Unresc,1ved Safety I 1ssues (U515) and all medium and high priority Generic Safety issues (G5!s). NUeEG-1197,
- Advanced Light Water Reactor Program", December 1986, presents these issues and their status as of July 1,1986. GE is to identify wHeh l 1ssues are applicable to the AEWR design and address them. These issues will l include both applicable issues identified in AUREG 1197 and any new generic ;
- issues raised up to the time of FDA issuance. It is the intention of the staff i 3
that there will be no open items regarding the resolution of U5!s or G5!s or . < other plant features for the ABWR st the time of the FDA decision. j
!ssues introduced after en FDA is issued would be analyzed and resolved in .
- accordance with the backfit requirements of 10 CFR 50.109, except to the extent that the DC rulemaking provides otherwise.
5 STAFF REVIEW PROCEDURES The staff will follow its review procedures in the SRP, supplemented and ' modified as fo110ws: t (1) The ABWR SAR is to be submitted chapter by chapter, over a period of about : 16 months. Correspondingly, the staff SER will also be issued in draft ; fom, in sections in accordance with the schedule shown in Section 2. The draft SER sections will be made publicly available. J (2) At the completion of the review of the individual SAR chapters, the staff will perforn an integrated review of the application. This review will ' complement the Probabilistic Jtisk Assessment (PRA) review, in that it will be en overall assessment of the design. The staff will issue a composite final SER in accordance with the schedule described in Section 2. (3) It will be irportant to carefully document open or unresolved issues that may be identified early in the review process, but which cannot be resolved until the completion of later chapters. Each draf t SER section ' will contain a description of such issues. In addition, with the submittal of each chapter of the SAR, GC is to provide an updated check-list which identifies outstanding issues and the future chapter (s; in which resolution is anticipated. ( ; (4) Each draft SER will contain a target schedule for closing outstanding SER f issues that is compatible with the target FDA decision date. l 5 l
6 ACRS FAR71CIPAT10N One step in the design review of a standard plant is the independent review by the Acvisory Ccerittee on Resctor Safeguards (ACRS). The ACR5 review of the ABWR design certification procesh started before sutnittal of the first chapters of the ABWR SAR. An initial briefing of the ACR$ by the staff and GE took place early in 1957. Periodic reviews will address the safety aspects of the design on matters selected by the ACR5. The ACRS review is scheduled to continue through April 1990, when the ACks will be requested to issue a letter report on its review. The staf f will leep the ACR5 infomed of the progress of the review by for. warding to it copies of the SAR chapters as they are submitted along with copies of the draf t SERs as they ere issued, in addition, the staff will meet with the ACRS. es needed, to discuss the draft SERS. 7 SEVERE ACCIDENT POLICY STATEMENT (SAPS) 7.1 Introduction On August 8, IgES, the Comission issued a Policy Statement on Severe Ace'idents
,(50FR32138, " Policy Statecent on Severe Reactor Accidents Regarding Future . Designs and Existing Plants.* and NUREG 1070, *hRC Policy on Future Reactor Designs *). The policy statement provides criteria and procedures for the licensing of new plants, and sets goals and a schedule for the systecatic examinatien of existing plants. The Comission encourag i the development of nen designs that night realize safety improveeents sad stated that it intends to take all reasonable steps to reduce the chances o. Securrence of a severe accident and to mitigate the consequences of such an accident, should one occur. The Ccm ission's licensing criteria for new plant designs are spec-ified in the policy statement.
The Commission also recognized the need to provide defense-in-depth by striking a balante between accident prevention and consequence mitigation, through a t>etter understanding of containment perfomance, with the understanding that new performance criteria for containment systems might need to be established. It also recognized the importance of such potential contributors to severe
- accident risk as human perfomance and sabotage, and detemined that these issues should be carefully analyzed and considered in the design and operating procedures for a nuclear facility. Specific discussions of each of the policy statement licensing criteria follow.
7.2 Construction Pemit/Manufacturino License Rule GE will comply with all applicable tcmission regulations, including those listed in 10 CFR 50.34(f) applicable to the ABWR. except 10 CFR 50.34(f)(2)(i) and 10 CFR 50.34(f)(3)(iv). Any future applicant that references the ABWR design must satisfy 10 CFR 50.34(f)(2)(1) by providing simulators. With regard to 10 CFR 50.34(f)(3)(iv). GE has stated that the ABWR design has specific festi".es that function to mitigate the consequences of severe accidents wi'.hin the offsite dose objectives discussed in Section 7.5 (Severe Accident P :rforrance Goals). GE intends to provide justification to demon-strate inat a dedicated containment penetration is not required in the ABWR design. The staff will consider this justification as part of its review. 6
7.3 Fesolution of U515 and G$is See Section 4, above. , 7.4 trebabilistie Disk Assessment (PFA) GE has proposed certain criteria and methodolegies relevant to PRAs in the following sections. These criteria and methodologies will be used by the staff as the bases for their review of the ABWR unless new criteria and trethodologies are promulgated by the hRC, 7.4.1 Scope GE has comitted te provide a level 3 PP.1 for the AEWR design, as defined by the *PRA Procedures Guide,* hVREG/CR 2300. The level 3 scope PRA includes the f olleming eleeents: (1) An analysis of the plant design and operation focused on the accider.t secuences that could lead to a core melt, their basic causes, and their frequencies , (2) An analysis of the physical processes of the accident sequences and the response of the containment (3) An analysis of the transport of radionuclides to the environment and an assessment of potential public health consequences founding analyses of external eventt that can be quantified (e.g., seismic, internal fires, internal floods, torrados) are to be included in this ovaluation. 7.4.2 Methodology The PF.A is to be based on a methodology that originated with the approach taker,in the Reactor Safety Study (WASH 1400), and that has been developed and syster.atized through applications in numerous plant specific studies. The general procedures have been documented in NRC NUREG reports, such as the *PRA Procedures Guide" (NUREG/CR 2300) and the 'Probabilistic Safety Analysis Procedures Guide" (NUREG/CR 2815). GE intends to utilfie the IDCOR developed Modular Accident Analysis Program (FAAF) wtitch has been modified by GE for utilization in BWR analyses. If technical disagreements surface between the NRC methods (such as the Source Tere Code Package) and MAAP. specific sensitivity studies will be performed. GE is to use the CRAC.!! code or other suitable model acceptable tosthe staff to compute the potential consequences of fission product relcases. , 7.4.3 Reference by L'tility Applicants The FRA is intended to be applicable to all sites within the ABVR envelope. It is contemplated that applicants will not have to prepare or submit plant-i spec;fic pus before an operating license is issued, but that the GE PU would be updated by the licensee within 2 years after a plant is licensed. 7
\
I 7.5 Severe Accident Perforn.ance Goals This section describes the goals for severe accident performance criteria. to chich GE has coctritted for the ASWR design, consistent with existing st4ff ' regulations and policy statements pertaining to severe accidents and defense. ' in. depth through a balance between accident prevention and consequence mitiga. tion. The staff will utiltre these goals as the basis for their review unless new criteria are promulgated by the NRC. 7.5.1 M ention of core Dameoe GE intends to demonstrate by anelyses that the likelihood of core darage will have a megn value of less than one in one hundred thousand reactor years (i.e., 1.0
- 10* ), including both internal and external events. (The staff will determine the adequacy of this goal and the analyses.)
7.5.2 H tigetion of Core Damage GE has stated 'that the ABVR design will provide protection against containn+nt failure if a severe accident occurs and results in core damage. GE inten'ds the containnent capabilities to include:
- a. Measures to reduce the probability of early containment failure for dominantsaccident sequences
- b. Measurer to accomodate hydrogen generated from the reaction of the ecuivalent of 100t of the rirconiur. in the active fuel clad, consistent with 10 CFR 50.34(f), as provided for by the Severe Accident Policy Statement
- c. Highly reliable heat removal systems to reduce the probability of containment failure by loss of heat removal
- d. Rekiable r+ans to prevent hydrogen deflagration and detonation, consistent with 10 CFR 50.34(f), as provided for by the Severe Accident Policy Statement.-
7.5.3 Offsite Consequences for Severe Accidents GE has ccmittet to meet the following goals: (1) The espected rean frequency of occurrence of offsite doses in excess of 25 ! Rem beyond a half mile radius fry the r I million reactor years (i.e., 1.0 x 10*6)eactor , considering both is tointernal be lessandthan once per external events (2) The containment design is to assure that the containeent conditional f ailure probability is less than one in ten when weighted o(er credible core damage sequences i
- e
E ADDITIONAL T[CHNICAL 155L'f 5 8.1 Intre4vetion - l The ABVc design will incorporate several features that are novel or which have been usec in relatively few other nuclear power plants in the United States. I In accition, because standardized plant reviews are conducted before actual l facility applications are n.ade, these reviews cannot address every aspect of a j facility. This section is intended tc address some of the issues which arise ' free these circumstances and that have caused difficulty in previous standard plant reviews. F.2 Physical Security B 2.2 Basis f or Requirerents The basis for the requirements will be as defined in 10 CFR 73.55,
'Recuirements for Physical Protection of Licensed Activitie: in Nuclear Power F.eactors Against Radiological Sabotage,* and other applicable portions of 10 CFR Fart 73. ,
8.2.2 Acceptance Criteria The ABVD. SAR is to include enough infortation to demonstrate the existence of adequate physict1 barriors to protect vital equipment in accordance with 10 CFR 73.55(c), 'Fhysical Barr trs.' and to identify access control points to all vital areas in accordance > 10 CFR 73.55(d), ' Access Requirenents.* The ABA SAR will net provide de is but is to identify design reovirements to be satisfied by an applicant for me following sections of 10 CFR 73.55 (the applicant inust then address all remaining requirements): (b) Physical Security Organtration k) Detection Aids (f) Corrnunication Requirements (g) Testing and Haintenance - (h) Response Requirertents The design requirerents are to include reference to existing *NRC documents such as Regulatory Guide 6.44, ' Perimeter Intrusion A1 erin Systems and NUREG 0908,
" Acceptance Criteria Evaluation of Nuclear Power Reactor Security Plans", as well as to industry standards such as IEEE.692 1986, 'lEEE Standard Criteria f or Security Systees for Nuclear Power Generating Statier.s."
8.3 Site Envelope Parameters ar.d Soil.* truet,ure Interaction Analysis Selected site envelope parameters and erP s for soil-structure interaction analysis proposed by GC are provided ' /. is A to this enclosure. E.4 A B'it Desien 9
I l 4.4.) Completeness of Design
' The AfWR SAR is to provide essentially cosplete design inforestion. The tem
- essentially complete
- is defined as follows:
(3) The SAR will define the major design components and include the results c,f sufficient engineering to identify, as appropriate:
- s. design basis criteria l
- b. analysis and design methods r
- c. functional c'esign and physical arrangement of auxiliary, 90F, and NS$$
j systems
- d. pier.t physical arrangements sufficient to accomodate system and .
components j
- e. functional and/or perfomance incifications for coeponents and materials sufficiently detailed to be w ie a part of associated procurement ,
specifications
- f. acceptance / test requirernents
- g. risk essessment methodology 1
.c (2) Design decurentation for systems, structures, and components should include as appropriate:
- a. design basis criteria
- b. plant general arrangements of structures and components, including piping system layouts C. process and instrumentation diagrams, electrical system layouts, and ,
major conduit and cable tray layouts
- d. control logic diagrams ,
e, systemfunctionaldescriptionsandsupportingstudiesandanalyses , i
- f. component and procurement specifications, including acceptance criteria anc test requirements
- g. construction and installation specifications, including acceptance criteria and test requireeents
- h. program for the assurance of quality
- i. design related aspects for the emergency plans J. supporting design docunentation such as site envelope data and calcula-tions sufficient to support the level of design detail noted above
- k. design related aspects of the physical security program 10
.. . _ _ _ . _ . _ _ _ - . - _ _ __. _,- _ . _ __ _ _ _ _.~.__ _
i I
- 1. an ALAR,A radiation protection plan ,
l .
- m. accident analyses ,
i n. Technical Specifications ,
- c. risk analysis >
In the limited cases where design infomation is not available GE is to i provide infortration on trethods, procedures, and perfomance criteria. G[ also 15 to define those related tests, inspections, analyses, and acceptance criteria that are necessary to assure that the desiges are prt>perly implemented in the plant. These tests, inspections, analyses, and acceptance criterie are intended to be implemented and verified in a series of reviews by the applicant during construction and ore operation. The staff will monitor the perfomance . of these reviens and irrplementation of the design through its inspection program. The degree of design detail necessary for providing an essentially complete design is to be that detail that is suitable for obtaining specific equipment or construction bids and to demonstrate confomance to the design safety limits and criteria.
.. E.4.2 Prograe for the Assurance of Quality in Design The design process and resultant design documents must meet the quality assurance (OA) reovirements delineated in Appendix B of 10 CFR Part 50, as addressed in Section 17.1 of the SRP. GE must submit justification, acceptable to the staff, for any deviations from Appendix B.
B5 Interf ace assu ptiens Affectine Safety Deteminations for the Nuclear 1siere The nuclear island scope of the ASWR reduces the number of interf aces between the nuclear island and the remainder of the plant. GE is to provide a list of the assumptions relied upon to rake safety deteminations for the nuclear island design. This listing is to identify the nuclear system, the instrumen-tation and contro1 recuirements, reliability assurptions and specific performance criteria. GE is to use the results of the PRA to indicate which interfaces are particularly sensitive to deviat.icms. B6 Instrurentation and Controls GI has co ritted to use standards and criteria and provide infomation pertaining to instrumentation and controls for the ABWR as discussed in Appendix B to this enclosure,
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i B.7 Water Cheristry Guidelines The maintenance of proper water chteistry in BVR cooling systems is essential to the prevettien of stress corrosion cracking of austenitic stainless steti piping and to the minirization of plant radiation levels due to activated corresion products. GE has comitted to using at least the following documents in this aree: 11
, _ - - . ... . -.. -. .-, . . . _ . . - _ - . _ . _ . - . . . . _ . _ - - - _ . . - _ _ - _ _ _ _ _ _ - _ _ _ . _ _ , _ . . _ ~ . - , - _ , . _ , - . -
(1) EPRI ht 3589 5RrLD, 'BWR Water Chemistry Guidelines," April 192! (2) EFR1 NF-4500 5R-LD, ' Guidelines for Pemanent Hydrogen Water Cherrist,ry Installations," Harch 1986; revised. November 1986 (3) EFR1 NF 8474, 'BWR Radiation Field Control Using Zinc Injection Fassivation,* March 1986 B.8 Maintenance and Surveillance GE is to provide in the SAR the reliability and naintenance criteria that a future applicant must satisfy to ensure that the safety of the as built facility uill continue to be accurately dese-ibed by the certified design. The SAR is to include the Ley assumptions of the PRA and other PRA licensing comitrents. 2.9 M51Y Allowable Lea 6 ace and Related Dese Calculations GE has comitted to an ABWR standard design that will provide a ncn-safety relatec nain stear isolation valve (H51V) leakage processing pathway consistent eith these evaluated in NUREG 1169, ' Resolution of Generit !ssue C-8," August 1986. The allowable M51V leakage is to be detemined based on the calculated total dose (using methodologies consistent with NUREG 1169) from all leakage sources and the exposure guidelines of 10 CFR 100,11. In addition, leakage for the final installed M51V test is to be less than 50t of the value allowed to account for equipment degradation during the design lifetime. B.10 Safety Goal Policy Stetement On August 4,1986, the Comission published a policy statement on
- Safety Goals for the Operation of Nuclear Power Plants * (51 FR 25044). This policy statement focuses on tte risks to the public from nuclear power plant operations. Its objective is to establish goals that broadly define an acceptable level of radiological risk.
Although'the implementation requiremerts for the Safety Goal Policy Statement are still being developed by the staff, GE has comitted to severe accident perforr.ance standards and criteria that are intended to assure compliance with those eventual requirements. :.. 9 FINAL DESIGN APPROVAL .- The stHf rey issue an FDA after it and the ACRS complete their reviews of the final design. The FDA means that an entire nuclear power plant design or major portion thereof is acceptable for incorporation by reference in individual applications for construction pemits, operating licenses, and ennufacturing licenses. The staff and the ACRS intend to use and rely on the approved final design in their reviews of those applications, However, an approved final design is subject to litigation in individual licensing proceedings on those applications. An FDA is a prerequisite for a design certification. 12 )
30 DESIGN CERT]FICATION 10.1 Introdvetion . The CoTission currently is considering staff-proposed revisions to its 1978 l pelicy statement on standardi2ation of nuclear power plant designs. The i Comission also is developing proposed regulations that will address licensing ' reform and standardization and provide a regulatory framework for impleeentation of the standardization policy, including Comission certification of standard l designs by ruleraking. Since design certification is the ultimate goal of the ASKD program, and since the focus of the proposed policy statement an,1 regulations is reference system design certification, the essence of these proposals, and GE's comitment to them, is surnerized here. It should be noted, however, that the Comission has not yet acted on these proposals and that they are subject to change. The staff-proposed revisions to the policy statement encourage the use of , standard plant designs in all future license applications. The staff believes that the use of standard plant designs can benefit public health and safety by: (1) Concentrating the resources of designers, engineers, and vendors cn particular approaches (2) Stimulating standardized programs of construction practice and quality assurance (3) Improving the training of personnel (4) Fostering enere effective maintenance and imp' roved operation The staff believes that the use of such standardized designs can also perrit r> ore ef fective and efficient licensing and inspection by the NRC. , 10.2 Desien Certification Conceot The design certification concept, as described in the staff's proposed standardization policy statenent, provides for certifying a reference system , design (such as the AEWR) through rulemaking. In this process, the Cctrission would certify a design af ter the staff issues an FDA and a rulerating pro-ceeding is ccepleted. The desig6 certification means that the portions of the nuclear power plant design that have been reviewed are acceptable for incor-poration by reference in an individual Itcense application. The conclusions of the certification rulemaking would be used and relied on by the staff, the ACRS, the hearing boards, and the Cornission in their reviews of applications that reference the design. The certified design would not be subject to litigation in individual licensing proceedings, except as provided in 10 CFR 2.755. Under the staff propcsal, the Ctruission could certify the standardized ABWR l design for referencing by applicants for a period of 10 years. Renewal of the I design certification could be granted for an additional period of up to ten l years unless the Comission found that the design would not comply with the j Cctrission's then-current regulations. Applicants could reference the certified ABWF design in applications fcr cps and OLs docketed during the period beginning with the docketing date of the FDA application and ending 13
at the expiration da9e of the design certification. However, no CP or OL could be issued for an application referencing the ABWR design until the FDA is issued. . 10.3 Coepleteness of Scope and Desian Detail The ABWF application fer design certification is to include a plant design that is essentially co plete in both scope and level of detail. The scope of design developed to support the design certification process is addressed in Section B.4 The ABWR application for design certification also is to demonstrate compliance eith the licensing criteria for new plant designs set forth in the Comissien's Severe Accident Policy Statement (Section 7). The ABWR SAR is to address the tests, analyses, and inspections that are necessary to provide reasonable assurance that the plant will be built and operated within the specifications of the certified design. For those individual aspects of the design where safety-related structures and components differ from those of existing designs, empirical information is to be included as part of the application for design certification." , The ABWR SAR is to include information that will pernit construction verification and compliance. This will permit reviews during the construction and startup phases of the plant and will eliminate the need for further design reviews on those portions of the plant that have been certified, except to verify that interface requirements have been met. 10.4 Changes to Aeproved and Certified Designs The staff believes that standardization will be best achieved if changes to approved or certified designs are kept to a minimum, hevertheless, there are situations in which changes may be needed or jesirable. It is the staff's intent that af ter issuance of the design certification, the Comissior would require backfittir.g only when it determines, using the standards in 10 CFR 50.109 and the results of the DC rulemaking, that a substantial increase in the overall protection of the public health and safety would result. GE may request modifications to an approved or certified design by applying for an amendment to the design approval or certification in accordance with the proposed regulations. 20.5 Rulemaking Appendix 0 to 10 CFR 50 provides the opportunity for the Comission to approve the ABWR reference system design in a rulemaking proceeding. The regulations that are currently under developnent will specify the procedures to be used for the rulemaking. In general, however, upon receipt of a request from GE, a notice would be published in the Federal Recister annctuncing the request for a design certification for the A5WR, The notice would set out the matters at issue, as specifically as possible, and the possible hearing procedures that could be used if the Commission decided to hold a hearing, and would request that all persons wishing to participate in a hearing notify the Cc m ission within a stated period of time. As a condition to participating in a herring, however, intervenors could be required to ctate the issues they wish to have considered at the hearing and to comit to providing expert testimory 14
J on those issues. Written corW r.ts could be invited from those not intenciag to parut4 rate in the hearings. As a result of responses to the notice, or en its own iritiative, the Cortission could then hold hearings on the proposed rulemaking.
]f a heering were held, the Comission could use a number of for1 rats, fro" the simple hearing and recording of testimony to interchanges among the parties and a lirited right of cross emamination. Because such rulemaling procedures go ,
beyctd the notice-and-co rtent requirements for rulemaking, the Comissier has . broad ciscretion to establish hearing procedures best suited to the retters at issue. Af ter try hearing, the Comission could review the cceplete record of the rulee.aling, including both the hearing record and any other written i come nt s . The notice of final rulemaking would have to include responses to written coments and the resolution of issues considered at a hearing, l The views of the ACRS would be sought and considered. The ACR$ would review the design before the rulemaking, and the results of the ACRS review would be made available when the proposed rule is announced. - 10.6 Renemel of Certifications Under the staff's proposal, the Comission could certify the standardized i AEWR design for referencing by applicants for a period of 10 years. Adcitier. ally, before the expiration of the design certificetion GE could apply for certification renewal. The design certification could be renewed for en additional 10-year period, provicto the design certplies with the Cctrission's ' regulations in effect at the time of the renewal application. l l 15 .
APPENDIX A SITE ENVELOPE PARAMETER 5 AND S0IL STRUCTURE ' INTERACTION ANALYS15 8 SITE ENVELOPE PARAMETERS This Appendix addresses some of the important site envelope parameters preposec by GE for the ABWR. Additional parameters are to be provided in the SAR. All of the proposed parameters and analyses will be considered by the staff during the review. Utility applicants must verify that their proposed facilities lie within the site envelope parameters assumed by GE in its safety analyses and approved by the staff. ho further analysis or actions will be required by an applicant when the site-specific pararreters fall within the design envelope. If site or interface parameters fall outside the design envelope, utility applicants must provide justification for the deviations. 2 50llSTRUCTUREINTERACTION(551) ANALYSIS
- 2.1 Scope GE will perfom soil-structure interaction (551) analyses of the reactor building design for a range of site conditions within the site envelope design parameters defined in Table A-1. GE will satisfy the staff acceptance criteria for the design ground motion and $51 analysis methods as specified in Revision 2 of the SRP, Sections 2.5.2, 3.7.1, and 3.7.2.
2.2 Pethodolocy GE intends to employ the state of-the art corrputer program. SA551 (Systee for Analysis'of Soil Structure Interaction) code. SAS$1 is a linear analysis program using the finite elenent approach. Solutions for a complete soil-structure system are sought in the frequency dec.ain employing the ccrrplex response technique. Problem fomulations are based on the flexible volume substructuring method in that the complete soil structure system is partitioned into the foundation and the strvefure. In compliance with the duality (ftnite elecent and half space) requirement, limited cases are to also be analy:.*d using the CLA551/ASD cceputer code, which is an improved version of the CLASSI family of computer codes. CLA551/ASD is a linear analysis program using the structure approach based upon continuum tr+chanics for half-space. A-1
Table A.) Envelope Of Selected Plant Site Design Parameters Applicable To ABWR NA11P'M Ge0VND WaTEP LEVELi 2 feet below grade . j FRtc1PITAT!0N (for roof design): 1 Naximum rainfall rate: 10 in/hr ! Maximum snom load 50 lb/sq ft DE51Gk TEMPERATURES: ! Artient: ! It Exetedance Yalues Maximur: 100'T dry bulb /77'F coincident wet bulb : Minimum: 10'F 1 01 Esceedance Yalues (historical limit) Maximum: 115'T dry bulb /82'F coincident het bulb Minimum: -40'F i Peak Emergency Cooling Water Inlet: 95'F ; Condenser Cooling Water Inlet 100'F i
)
SEl$MOLOGY: SSE PM: 0.30g* SSE Response Sptetra: per Regulatory Guide 1.60
$$E. Time History: Envelope SSE Response Spectra trTREW.E WIND: Basic Wind Speed: 110 mph" / 130 rph "
S0lt PROPERTIES:" " , Minimumbearingcapacity(demand): 15 ksf Minimum shear wave velocity: 1000 fps - Liquification potential: None at plant site resulting from OBE and SSE Free field, at plant grade elevation. ,
$0 year recurrence intervall value to be used for design of !
non safety related structures only. 100 year recurrence interval; value to be used for design of safety related structures only. l Values of bearing capacity and shear wave velocity are included in this table to ensure wide application of a standard mat-type foundation l design. The design must be evaluated parametrically against ranges of l possible soil properties to verify wide application. l t (
.,- . _ ~ _ _ _ . . - _ . . _ _ _ _ _ _ _ _ _ _ . , _ _ _
2.3 yey Desige Farareteos ! (1) Design Ground Motion . Tor the purpose of standard piant design, GI plans to use a value of 0.39 for the peak ground acceleration (PGA) for the Safe Shutdown farthquale ($$E) in accordance with Table A 1. (2) Centrol Motion Location , Ths control motion (or design ground motion) is to be defined ' at the finished grade in the free. field. This is consistent with Revision 2 of the $RP, Section 3.7.), and with the reep7+endations in kVRIG/CP 0054, ' Proceedings of the Workshop on Soil Structure Interaction", June 1986, for the application of ground motion defined by Regulatory Guide 1.60 for standard design ground spectra. (3) Design Ground Spectra , The design ground spectra are to be the Regulatory Guide 1.60 standard design ground spectra, normalized to the design peak ground acceleration. (4) Design Time History , A single set of three artificial time histories (two horizontal ' components and one vertical coeponent) are to be considered. They are to satisfy the spectra enveloping requirement of the Regulatory Guide 1.60 spectra. The power spectral density (PSD) function of the two horizonte1 coetonents are to be calculated and compared with the target PSD specified in Revision 2 of SRP, Section 3.7.1. (5) So'il Conditions To support the ABWR all soil envelope design concept GE plans to ! analyze a reasonably large number of cases by considering the variations of soil site conditions within the site envelopes of Table A 1. - The site conditions are to be selected based on those considered in the G[iSAR 11 design and adjusted to acconnodate the ABWR design. These site conditions cover a wide range of soil deposit depths, shear wave velocities, and water table locations. 1 9 9 N ~ A3
(() Strain Dependent Soil Properties GE plans to use strain dependent soil properties for shear riodulus and as defited in the GES$AR !! design. In accordance material with hUREG/CR-116 "Recom' ended Revisions to NRC Seismic Design Criteria, darping,1, May 1980, the strain. compatible shear rodulus will be lir.ited to 40t of its low strain value and r,aterial derping will be limited to 15% of critical. The effects of pore pressure are to be talen into account by varying the water table location. P (7) Floor Pesponse Spectra Pe:L Broadening The floor response spectra to be generated for subsystem seismic design are te be the enveloped spectra for a wice range of potential site conditions. The uncertainties associated with soil properties, therefore, 4 will be automatically accounted for in the enveloped spectra. The only ' 4 - other sources of uncertainty that may exist in the calculated floor response spectra are those associated with r.odelling approximations and structural material properties. To account for those uncertainties that ray result in variations in structural frequencies, GE considers it *
- sufficien; to broaden the peaks at the structural frequencies of the enveloped coeputed flocr response spectra by +101. This is consistent with tre GES$AR !! design. ,
P d e s 6 l A-4 l l
APPENDIX B INSTRLMENTATION AND CONTROLS , 3 lhTRODUCT10N The instrumentation and control (180) systems of the ABVR are to use state-of the art fiber optics, multiplexing, and computer controls. Staff guidance in this area has net been developed, however G[ has ccer*itted to the standards and criteria currently specified in the SRP, and use of the documents and > criteria identified below. GE is to provide the infomation listed below so the staff can detemine the acceptability of the ABWR !&C systems. In lieu of actual test or qualification reports for equipment that will be selected later by GE or a utility applicant, GE is to provide a detailed description of the testing or Qualification standards to be used to assure that the ecuipment that is ultimately selected will perfom as intended. 2 MULTIPLEX!$G SYSTEMS , In the SAR, GE is to: , (1) Provide a complete list of components (pumps, valves, etc.) whose actuation, interlock, or status indication is dependent on the proper operation of each Class 1E multiplexer. (2) for the components cited above, describe the means of remote or local control (other than by cutting wires or jumpering) that my be employed should the multiplexer fail. (3) Describe the multiplexer pre operational test program.
. (4) Des'cribe the test and/or hardware features employed to demonstrate fault tolerance to electro-r.agnetic interference.
(5) Describe the interconnection, if any, of any Class lE multiplexer to non Class 2E devices such as,the plant computer. (6) Describe the online test and/or diagnostic features that my be erployed, including any operator alarts/ indicators and their loca tions,. (7) Describe the multiplexer power sources. (8) Describe the dynamic response of the multiplexers to momentary interruptions of AC power. , (9) Describe the applicability of the plant Technical Specifications to multiplexer operability. (10) Describe the hardware architecture of all multiplexer units. B-1 l t
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(83) Describe the firmnare" archite8ture. (22) Previde an esplicit discussion of how the systems confom to the provisions of IEEE-279, Section 4.17. ~ l (33) Frevide an explicit discussion of how the systems confom to J[ff.279, paragraph 4.7.?, as supplemented by Regulatory Guide 2.7$ ar,e ](([.364 (14) Provide confimation that systee level failures of any multiplerer system detected by automated diagnostic techniques are indicated to the operators consistent with Regulatory Guide 1.47. (25) Provide an explicit discussion of the susceptibility of the multiplexer systems to ele:tromagnetic interference. 3 ILECTRICAL ISOLATORS GE has corritted to provide the following on isolation devices: (1) For each type of device used to acceeplish electrical isolation, a ' description of the testing to be perfomed to demonstrate that the device
.. is acceptable for its application (s). The test configure ion and how the retirum credible faults applied to the devices will be inL uded in the description. ' . . .
(2) Identification of the data that will be used to verify that the raximu-credible faults applied during the test are the taximum voltage / current to which the device could be exposed, and to define h w the r.4ximum voltage / current is deternined. (3) Ider.tification of the data that will L+ used to verify that the r.eximu-credible fault is applied to the output of the device in the transverse mode (between signal and return) and other faults are considered (i.e., openandshortcircuits). (4) A definition of tAe pass / fail acceptance criteria for each type of device. "
~
(5) A corritment that the isol'at1on devices will comply with all environmentel qualification and seismic qualification requirements. (6) A description of the measures taken to protect the safety systees from electrical interference (i.e., electrostatic coupling, EN!, comen mode, and crosstalk) that may be generated. (7) Infomation to verify that the Class 1E isolation devices are powered f rcn. a Class IE power source (s). - B-?
(8) A comparison of the design crith the guidance in NUREG/CR 3453/ EGG.2444,
- Electronic !selators Used in Safety Systems of U.S. Nuclear Power Plants," March 1986. ,
(9) A cecparison of the design with the guidance in draf t Regulatory Guide EE502 4, ' Criteria for Electrical Isolation Devices Used in Safety Systems for huclear Power Plants'. 4 FIEER OPTIC CABLE The staff is working with [G8G to develop comprehensive guidance on this subject. The guidance will be based on the existing ![El: cable standards, such as 1EEE 323 and IEEE 364, on the ANS! standards for fiber optic cables which are listed at the end of this Appedx), and the results of the [ GAG work.
$ FROGRAP.ABLE DIGITAL COMPUTER SOFTWARE As a starting point, the following documentation is to be used by GE in the design and by the staf f in its riview:
(1) ANS1/IEEE-MS 7.4.3.2, ' Application Criteria for Programable Digital Computer Systems in Safety Systers of Nuclear Power
. Generating Stations,' 1982 (2) Regulatory Guide 1.152, " Criteria for Programable Digital Computer System Software in Safety Ritated Systems of Nuclear Power Plants," November 1985 (3) NUREG 0308, ' Safety Evaluation Report . Arkansas Nuclear 1 Unit 2,* November 1977 (4) NUAEG-0493, 'A Defense in Depth and Diversity Assessment of the RESAR-414 Integrated Protection Syster),' May 1985 (S) NUREG-0491s5'jafetyEvaluationReportofRESAR414,' February 1979 6 PROGRAW.AELE ITAL COMPLFIERiMRDWARE
, As a starting poiN, the following documentation is to be used by GE in the design and by the ataff in its review: (1) IEEE 603, '1EEE Standard Criteria for Safety Systees for Nuclear Power Generating Stations,9-1980 (2) NUREG-0308, " Safety Evaluation Report - Arkansas Nuclear 1. Unit 2,* November 1977 . l . (3) Regulatory Guide 1.153, " Criteria for Power, Instrumentation ~, and Control Portions of Safety Systems
- December 1985 (4) NUREG-0493, 'A Defense in Depth and Diversity Assessment of the RESAR-414 Integrated Protection System," May 1985 (5) NUREG-0491, *Ssfety Evaluation Report of RESAR-414,' February 1979 B-3
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POLICLY ISSUE - (Notation Vote) For: The Corrissioners from: James M. Taylor Executive Director for Operations
Subject:
EYOLUT10f1ARYLIGHTWATERREACTOR(LilR)CERTir!CATIONISSUES AllD THEIR RELAT10liSHIP TO CURRENT REGULA10RY REQUIREMENTS
Purpose:
To present thh staff's recomendations concerning proposed departures from current regulations for the evolutionary ALWRs. The staff reouests Commission approval of the positier.s as described in this paper. ! Ba c kground: In the April 21, July 31, and August 24, 1989 staff requiretrents nemorarida ($RHs), the Comission asked the staff to identify the issues and acceptance criteria used to judge the acceptat.111ty of future designst to identify where the staff proposes to go beyond the regulations or to be less restrictivel and to identify if tle Advanced l Poiling Water Reactor (ABWR) would meet the Corrission's l Safety Goal with or without a vent. The Comission asked that these issues be discussed in the context of certificattor, of the ABWR and the other evolutionary advanced light water reactor (ALWR)designseswellasthestaff'sreviewofthe evolutionary Electric Power Research Institute (EPRI) Re-quirements Document. Discussient Operating ex serience as well as a number of studies (e.g. PRAs) aave identified a number of issues signifi-cant to reactor safety. Based on this background the staff has identified the following list of issues as i fundamental to agency decisions on the acceptability of evolutionary ALWR designs. l (1 evolutionary LWR public safety goals (2 source tem l l
Contact:
O.C. Scaletti, NRR/DRSP 2-1104 C.L. Hiller, NRR/DRSP 8-1118
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l The Comissioners 2-anticipated transients without scram (( mid loup operation stetion blackout (( fire protection f'))cure-concrete interaction - ability to cool core ( intersystem LOCA hydrogen generation and control debris high pressure core melt ejection containment performance ABWR containment vent design ( equipment survivability ( cperating basis earthquake / safe shutdown earthquake L inservice testing of pumps and valves The resolutions proposed by EPRI and the LWR vendors, and the staff positions and recommendations regarding each of these i', sues are discussed in detail in the enclosure. In addition to these issues, each application for a Design Certification will have to propose technical resolutions for those Unresolved Safety Issues and medium and high-priorty Ceneric Safety Issues which are identified in NUREG 0933 and technically relevant to the design in accordence with 10 CFR 52.47 and the Severe Accident policy Statement. The Comission's approval of, or alternate guidance on, the proposed resolution of these issues is necessary for the staff's continued review of EPRI's ALWR Requirements Document,GeneralElectric's(GE's)ABWR,(Westinghouse's RESAR SP/90 and Combustion Engineering's CE's) System 80+ designs. Approval or guidance is particularly important to the staff's evaluations of the GE ABWR and the epa! ALWR Requirenents Document since these reviews have progressed the furthest. The certification review for CE's Systeen 60+ is just beginning. Westinghouse has indicated that they do not intend to pursue an FDA/ certification for the RESAR SP/90 at this time. Additional Commission approval or guidance on significant issues related to certification of CE's System 80+ and other future designs will be discussed with the Commission as part of the development of thelicensingreviewbases(LRB)forthesedesigns. This approach is consistent with recent Comission guidance in , an SRM dated December 15, 1989. It should be noted that t some of the issues presented in the enclosure are issues ! proposed by EFR1 which they refer to as plant optimization subjects. NRC approval of a plant optimization subject l l l
The Comissioners 3-would result in a resolution that is less restrictive than present regulatior.t Comission policy, or past licensing practices. For these reasons the optimizat' ion subjectr such as hydrogen control, sourc,e tem, and the relationship between the operating bases earthquake and the safe shutdowr, earthquake could have a major schedular fruact on the evolutionary LWR certification process. Tne staff has provided the respective applicant's proposed solutions as well as the staff's positions on these issues in the enclosure to provide a comparison, and to provide an indication of the diversity of proposed solutions under consideration by the staff. The staff recomendations identified 'n this paper have been developed as a result of (1) the staff's reviews of current designs and evolutionary ALWRs, (2) generation consideration reactor of creratingexperience,includingtheTN!-2 (3) results of the probabilistic risk assessrents accident [PRAs)of current-generation reacter designs and the evolutionary LWRs, (4) early efforts conducted in support of severe accident rulemaking, and ($) research conducted to address previously identified safety issues. Infomation related to the issues and staff positions discussed in this paper have previously been previded to the Comission in SECY-89 013, SECY-89-153, SECY 89 228 and SECY 89 341 and have been under11neo in the enclosure. The steff positions recomended in this paper are consistent with those previously taken in the staff's review of the ABWR LRB and in several ABWR related safety evaluation reports issued to date. The staff believes that pending detailed staff review, there is a high degree of, confidence that the ABWR would neet the positions identified in the enclosure. Therefore, Comission approval of the staff recomendations would close these policy issues for the ABWR and would pemit staff review to continue. The recomended positions are also consistent with those identified in the staff's draft safety evaluations relatd to certain chapters of the EPRI-ALWR Requirements Document. The staff is reviewing severe accident and certification issues addressed in the EPRl ALWR Requirements Docurent and the staff's final conclusions are awaiting Cemission approval of the positions described in this paper. Approval of the staff recomendations would allow for continuation of the staff review of the EPRI ALWR Requirerents Document in accordance with recent Comission guidance.
Conclusions:
The staff believes its conclusions and reco mtndations regarding these matters are in keeping with the Comission's
The Comissic,ners policy expectation that future designs for nuclear plants will achieve a higher standard of severe accident safety pe rf orms,nce. *
' The staff will infonn the Commission during its reviews if additional enhancements to existing requirements, beyond .
those identified in the enclosure, are determined to be necessary for evolutionary ALWR designs. _Coordinetion: The Office of General Counsel has reviewed this paper and has no legal objection but notes that any program for review of new reactor designs,which authorizes the NRC to isnpose requirernents beyond those needed to meet current Comission regulations raises the issue that, if the NRC staff can pose additional requirements for certification, other parties should be able to do so as well. The traditional way to avoid such problems is through rulemaking, but as indicated in SECY 89 311 the staff believes that design certification process Is a rore effective method of resolving severe accident issues than a generic severe accident rule or several individual changes. Further, OGC notes that questions regarding the
- desirability of additional severe accident mitigation rnessures still need to be addressed under NEPA._either in the design ecrtification rulemaking and hearing or in sorne preliminary rulemaking. t A copy of this paper has been provided to the Advisory Comittee on Reactor Safeguards.
Recorrendations: That the Comission (1) Approve the staff positions detailed in the enclosure, and (2) Note that if the staff determines other issues need to be addressed in accordance with Comission guidance, the staff will inform the Comission of its positions on these matters in a timely manner.
t
?
I The Comissioners 5- ' t (3) Note that the staff in accordance with the Staff I!equirenents Memoran,dum dated August 24 1989, plans to issue the draft SER on Chapter 5 of the EPRI Alk'R 3 Requirements Document. (4) Note that following Comission resolution of the II6TTcyissuesdiscussedinthispaper,thestaff plans to finalize and reissue the above draft $EF:. , i
- /' ,
i f-T , J s H. Tay1 xecutive Di ctor for Operations
Enclosure:
As stated I l. Commissioners' comments or consent thould be provided directly to the Of fice of the Secretary by COB Tuesday, January 30, 1990. : i Comission Staff Of fice comments, if any, should be submitted to the Commissioners NLT Tuesday, January 23, 1990, with an information copy to the Office of the Secretary. If the paper : is of such a nature that it requires additional time for analytical review and comment, the Commissioners and the Secretariat should be apprised of vhen comments inay be expected. DISTRIBUTIOut Commissioners OGC OIG GPA REGIONAL OFFICES '
- EDO l ACRS ACNW ASLBP ASLAP SECY l
ENCLOSURE EVOLUTIONARY ALWR CERTIFICATION ISSUES - l l
- 1. General Issues l
A. ALWR Public Safety Goal The EPR1 Requirenents Document proposes that the evolutionary ALWRs comply with the followin; public safety goals: (1) The frequency of core damage will be less than
, 1.0 X 10E-5 events /per reactor-year, and (Note: EPRI refers to this as a " quantitative investment protection goal")
(2) Uhole body dose at an assumed 0.5 miles site boundary must be less than 25 rem for events whose cumulative frequency exceeds 1 X 10E-6 per reactor-year. In the Licensing Review Bases (LRE) for the ABWR, GE has comitted to meet the following goals: (1)Demor.stratethatthelikelihoodofcoredamagewillhaveameanvalueof less than one in one hundred thousand reactor years (i.e.,1.0 X 10E-5). (2) The expected mean frequency of occurrence of offsite doses in excess of 25 rem beyond a half mile radius from the reactor is to be less than once per l millien reactor years (i.e.,1.0 X 10E-6), considering both internal and external events. (3) The containment design is to assure that the containment conditional failure probability is less than one in ten when weighted over credible core damage sequences. The staff is presently reviewing the LRB for the System 80+ design in which CE has proposed goals which are similar to the goals developed in the ABWR LRB. Since Westinghouse has no imediate plans of pursuing certification of RESAR SP/90, work on the LRB is presently not planned. However, similar to GE and CE, We tinghouse has stated, in meetings with the staff and the Advisory Comit'.ee on Reactor Safeguards, they are comitted to meeting the ALWR public sefety goals as well as the goa's in the Commission's Safety Goal Policy Sti.tement. The staff is reviewing the proposed /. LWR public safety goals to ensure they are consistent with the Comission's Safety Goal Policy Statement, which proposed both qualitative as well as quantitative safety goals for future reactor designs. The current regulations do not specify requirements in numerical terms of frequency of core damage or large release events. However, the Commission in its safety Goal Policy Statement, has proposed that the staff exumine a general performance Suideline that "the overall mean frequency of a large release of radioactivity to the environment from a reactor accident should be less than 1 in 1,000,000 per year of reactor operation."
SECY.89-102 recommended approval by the Comission of the use of the following quartitative objectives in its implementation of the Safety Goal Policy for future standardized plants: -
- 2. The rean core damage frequency target for each design should be less than 1.0 X 10E-5 event per reactor-year, and
+
- 2. The overall meen frequency of a large release of radioactive materials to the environnent from a reactor accident should be less than 1 in 1,000,000 per year of reactor operation where a large release is defined as one that has a potential for causing an offsite early fatality.
The staff concludes that the staff-proposed quantitative safety goals submitted in SECT-89-10? are consistent with the Commission's Safety Goal Folicy Statement. Accitional Comission guidance en the establishment of quantitative goals and im. plerentation of safety goal policy will assist the staff in its continuing assessment of the evolutionary ALWRs. Although the staff considers the goals defined in SECY-89-102 to be acceptable for evolutionary ALWRs, it should be noted that both 'he EFR) public safety geel and the AEWR public safetv goal are considerably more stringent than the large release guideline defint.d in 5ECV-b9-102. Although the staff has indicated it believes toe ALWR Public Safety Goal contains meritable coals for the industry to adopt, the staff has not completed its review of this issue 6nd is in the process of reviewing how EPRI implements these goals. B. Source Term As noted in SEry 89 341, the staff's methodology for determining compliance with the siting requirements of 10 CFR Part 100 has been based on the 1962
TID-14844" source term. This methodology, which involves calculation of offsite dose for comparison against Part 100 dose criteria (i.e. criteria for establishing the size cf the exclusion area and the low population zone), is widely acknowledged to utilire conservative assumptions. At the time this approach was developed, these conservatisms were considered appropriate and were based on uncertainties associated witn accident sequences and equipment performance; and as a means to assure that future plar.t sites would be essen-21a11y equivalent to sites approved up until that time. The conservatisms initially included in the methodology have been essentially retained up to this time.
I EPRI has stated that the evolutionary ALWR licensing design-basis requirements as well as design enhancements related to severe accidents should be based on the full body of current knowledge regarding accident source terms. They believe that the evolutionary designs should be evaluated based on a realistic treatment of fission product source terms, including the extensive research j that has been done on fission product behavior since TID-14844 was issued, and especially sir.ce the Three File Island accident in 1979. EPRI's view is that this approach will result in designs which are improved and provide enhanced i safety protection. EPRI has identified this as a plant optimization issue.
GE in 10has CFR indicated Part 100.thatHowever, the ABWR will rneet the offsite dose criteria established they propose to utilize updated information such as system performance and reliability information, developed since promulgation of Partoffsite the 100, tocose. justify so:ne departure from the current methodology 'for calculating The ABWR's current design includes a single stand-by gas treatment syr. tem (SGTS) charcoal filter bed, and no main steam isolation valve (MSIV) leakage control system (LCS). Previous BWR designs utilized redund6nt SGTS charcoal filter beds and, since 1976, rost have been equipped with a fiSIV-LCS. The staff's interpretation of the General Design Criteria (GDCs) would classify filters as active corponents and require redundancy to permit any dose reduction credit in calculating a Part 100 dose. Since 1976, MSIV-LCSs have been required in most BWRs to meet 10 CTR Part 100 for design bases accidents. Part 100 requires equipment used to mitigate consequences of design basis acciderts to be seismically cesigned (it identifies equipment necessary to mitigete the consequerices of accidents whose offsite consequences are comparable to the Part 100 dose guidelines as designed to withstand the vibratory motion of an SSE). Since non-safety grade equipment such as piping downstream of the MSIVs and the condenser are not seismically designed for SSE, credit for these systems has not been accorded in calculating offsite doses for Part 100 purposes. The staff is considering these deviations from the current methodology for demonstrating compliance with Part 100. The staff has concluded, based on current informatior and experies.ce, that some deviation from current practice, or exemptions from the regulations identified above, may be warranted in the review of evolutionary designs. Presently, the staff believes that no other deviations woulc bc necessary to demonstrate ABWR compliance with Part 100. The other evolutionary ALWR vendors (Westinghouse and Combustion Engineering) ( have indicated that their evolutionary designs will comply with 10 CFR Part 200 and that they will work with the NRC and EPRI to utilize more realistic source term information to assess design enhancements related to severe accidents. As stated in SECY 89-341, the staff is undertaking an examination of the implications of decoupling siting frorr plant design for future reactors. 1:nder this plan, re6ctor site characteristics would be reviewed separately fror the reactor without utilizing 'iource terms or dose calculations. This would recuire revision to Part 100 and other regulatory staff practices. The results plant of such license a study will establish appropriate guidelines for any future applications. In the interim, however, the staff recomnends that t the Commission approve the following approach f or evolutionary ALWRs: s Assure that ev lutionary designs meet the reoutrements of 10 CFR 100 Consider deviations to current methodology utilized to calculate Part i 100 doses on a case-by-case basis utilizing engineering siucgement irclucing upcated inf ormation on source term and equipnent re l iabili ty. I \ _ - --
4-Such deviations could_ impact plant design features, therefore, these deviations will be identified in the SERs that will be forwarded to the Commission for its information well in advance of issuance, as directed in the SRM pertaining to SECY 89-311 dated Decenber 15,19r9, Do not modify current siting practice, even though 1t is recognized that such deviations could result in calculated low population zones arTexclusion areas which are smaller than those that lave been approved for currently operating reactors. Continue to interact with EPRI and the evolutionary ALWR vendors to reach agreement on the appropriate use of updated source term information for severe accident perf orraance considerations.
- 11. Preventative Teature 1ssues A. AnticipatedTransientWithoutScram(ATWS) 1 The ATWS rule 10CFR 50.6? was promulgated to reduce the probability of an ATWS I event and to enhance mitigation capability if such an event occurred.
EPRI has indicated that its approach to resolving the Air issue is compliance with the ATWS rule. Design requirements beyond those whita would be required to neet the rule have not been proposed, j 1 The ABWR design includes a number of features that reduce the risk from an ATWS event. These features include a diverse scram system with both hydraulic and liquidelectric run-in cap (SLCS), and a recirculation pump trip In control system capability. abiliti addition, the scram discharge volume has been removed from the ABWR, eliminating some of the potential ATWS problems associated with the older BWR designs. While the ATWS rule requires an automatically initiated SLCS, GE has concluded that the diverse scran system and enhanced reliability of the reactor protec-tien system negates the need for an automatic SLCS. GE has agreed to provide a reliability analysis in order to support this position. The staff will review the analysis to determine if an exemption from 10 CFR 50.62, to approve manual SLCS, is justified. The staff analysis will be provided in a future safety evaluation report for the ABWR. Westinghouse has concluded that a diverse scram system is unnecessary for the RESAR SP/90 design due to 1) high reliability of the integrated reactor protection system (IPS), 2) a turbine trip and emerger.cy feedwater actuation i that is independent of the IPS, 3) Ability to manuall ' motor generators from the main control board, and 4)a yhighly trip the rod control negative moderator temperature coefficient. Westinghouse has committed to provide a detailed analysis to demonstrate that the consequences of an ATWS are acceptable at the time an FDA application is submitted. The CE System 80+ design includes a control-grade Alternatt Protection System l which provides an alternate reactor trip signal and an alternate feedwater actuation signal separate and diverse from the safety-grade reactor trip system.
I
-S-1 The staff believes, notwithstanding the Westinghouse position on diverse scram systems, that all future evolutionary ALWR designs should be required to provide a diverse scram system unless the LWR vendor can demonstrate that the consequences of an ATWS are acceptable. The ATWS rule presently requires a diverse scram system for all (CE, Babcock and Wilcox, and GE) LWR designs except Westinghouse PWRs. It had been determined that previous Westinghouse designs had adequate ATWS capebility and backfit could not be justified. The staff believes that evolutionary ALWR designs should provide diverse methods of inserting control rods to mitigate a potential ATWS and to ensure a safe reactor shutdown. The staff considers that diverse scram capability is a worthwhile measure of presention for all evolutionary ALWRs, especially when incorporated into the initial design. Imposition of a diverse scram system on the Westinghouse design would exceed the Commission's regulations. Therefore, the staff recommends that the Commission approve the ,
i staff position ~Ihat diverse scram systems be provided for evolutionary ALWRs. B. Mid-Loop Operation The staff is concerned that decay heat removal capability could be lost when a { PWR is shut down for refueling or maintenance end drained to a reduced reactor coolant system (RCS) or *mid-loop" level. For example, a significant problem has been the loss of residual heat removal (RHR) suction due to air-binding of the RHR pumps. This is usually caused by an uncontrolled low loop level and consequent air ircestion into the pump suction. The EFR1 Requirerents Docurnent specifies requirements consistent with measures i applicable to operating reactors as described by the administrative procedures identified in Gereric Letter 88-17, but does not specify design modifications to acdress the root cause of this event. Westinghouse has comitted to install a vortex breaker at the RHR hot leg connections to significantly reduce air entraintnent during mid-loop operation. This feature, in conjunction with other design features of the plant, should l greatly reduce concerns over mid-loop operation. CE has indicated that it taill adc'ress this issue through analysis, consideration of specific design features, and/or operational restrictions. Specific design resolutions for the System 80+ have not been provided. Mid-loop operation is not an issue with the ABWR, The staff expects improvements in instrumentation in many existing PWRs, but does not require specific modifications to the nuclear steam supply system (N555) to correct mid-loop problems. However, the staff believes tha.t physical modifications such as those proposed by Westinghouse, may be necessary to essentially eliminate any concerns with mid-loop operation for future evolu-tionary pressurized ALWRs. Mid-loop operation is not explicitly covered by current reculations, however imposition of such requirements would exceed current staff licensing practices. Therefore, the staff recomends that the Commission approve the staff oosition that evolutionary PWR vendors propose desicr, features to rnsure higi reliability of the shutdown decay heat removal syster.
i C. Station Blackout The station blackout rule (10 CFR 50.63) allows utilities several design alternatives to ensure that an operating plant can safely shut do.wn in the event that all ac power (offsite and onsite) is lost. The EpRI Requirements Document provides for inprovements in offsite power reliability, onsite power re11asility and capacity, and station blackout coping capability. EpRI is also proposing that a large capacity, diverse alternate ac power source (corbustion turbine generator) with the capability to power one complete set of n rmal safe shutdown loads be included in evolutionary ALWR designs. The RESAR Sp/90 emergency feedwater system includes two ac-independent and two de-independent turbine-driven pumps. The electrical design includes two full capacity emergency diesel generators. In addition, it includes a backup seal irgection pump powered by a small dedicated diesel generator which has enough capacity to also charge the station batteries. Westinghouse believes that this design will provide a 24-hour coping time which is sufficient to eliminate the need for the addition of an installed spare (full capacity) alternate ac power source. The System B0+ design includes two turbine-driven emergency feedwater pumps and two rautor-driven emergency feedwater pumps. The electrical design includes two full capacity emergency diesel generators and a diverse alternate ac power source. This alternate source of aC power is expected to be a control-grade combustion turbine with sufficient capacity and capability to power either one of the electrical divisions. In addition, the plant design has full load rejection capability and the capLbility to subsequently provide electrical power from the turbine generator. Each of the four safety-related instrument channels has a dedicated battery backup. Class IE electrical Divisions I and II, which include the two emergency diesel generators are each provided de power by an assigned pair of these batteries. The AEWR design includes three independent electrical divisions, each with high-pressure and low-pressure water injection ca) ability, each powered by a full capacity emergency diesel generator, and eac1 division capa.)1e of independently shutting down the reactor. Additionally, the ABWR design includes an alternate ac combustion turbine to back up the diesel generators. The design has a capability to survive a 10-hour blackout period utilizing the reactor core isolation cooling (RCIC) turbine and station batteries. Extended blacirout capabilities are also provided by the ac-independent water addition system. This systen allows for makeup to the reactor vessel following RCS depressurization by connecting a direct drive diesel fire pump or by cor.necting an external pumping source, such as a fire truck, to a yard standpipe. The staf' believes that the preferred method of demonstrating compliance with 30 CFR 50.63 is through the installation of a spare (full capacity) alternate ac power source of diverse design that is consistent with the guidance in Regulatory Guide 1.155, and is capable of powering at least one complete set of
rornal safe shutdown loads. Although an alternate ac power source is provided as an acceptable resolution to this issue in 10 CFR 50.63, staff imposition would exceed current Commission regulations. Therefore, the staff recomends that the Comr.ission approve imposition of an alternate ac source for evolutionary
%s .
D. Fire Protection The staf f has concluded that fire protection issues that have been raised through operatino experience and through the External Events Program must be resolved for evolutionary ALWRs. To minimize fire as a significant cuntributor to the likelihood of severe accidents for advanced plants, the , 1 staff concludes that current NRC guidance must be enhanced. Therefore, the evolutionary ALWP designers must ensure that safe shutdown can be achieved, assuming that all equipment in any one fire area will be rendered inoperable by fire and that re-entry into the fire area for repairs and operator actions is rot possible. Because of its physical configuration, the control room is excluced from this a)proach, provided an independent alternative shutdown _ capability tha t is p1ysically and electrically independent of the control room is incluced in the design. Evolutionary ALWRs must provice fire protection for redundar.1 shutdown systems in the reactor containment building that wil_1 ensure, to the extent practicable, that one shutdown division will be free of fire dariage. Aoditionally, the evolutionary ALWR designers must ensure thW smoke, hot gases, or the fire suppressant will act migrate into other fire areas to the extent thet they could adversely affect safe-shutdown capabilities, TTicluding operator actions. Because the layout of a nuclear plant is design-spectitc, plant-specific design details will be reviewed by tie staff on an individual basis. The staff will require a description of safety-grade provisions for the fire-protection systems to ensure that the remaining shutdown capabilities are protected, as well as derenstration that the design complies with the migratien criteria discussed above. The ALWR Requirements Document indicates that fire protection will be as , specified in 10 CFR 50.48 and Appendix R. It states that for equipment in the same general area, a 3-hour fire barrier will be utilized in lieu of physical l separation unless it is " impractical or less safe." However, no cuide11nes ! are provided in the Requirements Document as to the application of these criteria. The evolutionary ALWR designers have indicated that their fire protection designs' are consistent with the staff's proposed enhancerents. GE has provided its ABWR fire protection analysis which is currently under review by the staff. l Appendix R to 10 CFR part 50 was promulgated for plants that were in operatior prior to January 1, _ s9. Subsequently, PRAs performed on more than a dozen plants have showed that fire is a significant contributor to core damage. The staff believes that in keeping with the Comission's desire for enhanced safety for evolutionary ALWRs, fire protection requirements should. reflect experience f rom cperating reactors and the greater understanding of severe accider.ts that has been acquired since Appendix R was promulgated. Therefore, the staff recomends the Commission approve the use of the enhanced fire pro-l tection pcsition underlinec above f or evolutionary ALWRs. l l
l E. Inter %. tem LOCA Tuture evolutionary ALWR designs can reduce the possibility of a loss-of-coolant accident (LOCA) outside containment by designing (to the extent practicable) all systems and subsystems connected to the reactor coolant system (RCS) to an ultinate rupture strength at least equal to the full RCS pressure. For both EWRs and pl.'Rs, EpRI states that low-pressure systems which could be overpressurized by the RCS should be designed with sufficient margin to withstand full RCS pressure without structural failure. For SWRs, pressure isolation valve instrumentation and controls should be provided to (1) prevent opening shutdv cooling connectioris to the vessel in any loop when the pool suction valve, discharge valve, or s,! ray valves are > open in the sarse loop (2) prevent opening the shutdown connections to and from the vessel rhenever the RCS pressure is above the shutdown range, (3) automatically close shutdown connections when RCS pressure rises above the shutcown range, and (4) prevent operation of shutdown suction valves in the event of a signal that the water level in the reactor is low. For Ms, relief velves sized to protect against overpressure transients, should be provided on the RHR system. RHR sucticn valves should be provided with permissive interlocks to prevent opening if RCS pressure exceeds RHR design pressure. Westinghouse has indicated that, should the isolation valves of the RESAR SP/SL fail, the design pressure of the piping outside of the containment will be sufficient Emergency tostorage water withstand primary) tank (EWST . side pressure or will be vented to the CE has eliminated the low-pressure safety injection system and increased the design pressure o' the shutdown cooling system piping in the System 80+ design. With this higher design pressure, the shutdown cooling system is , expected te riaintain its integrity even when exposed to full reactor coolant system pressure. The AEWR has been designed to minimize the possibility of an interfacing system LOCA in the following ways. The low pressure systems directly interfacing with the RCS are designed with 500 psig piping which provides for a rupture pressure of approximately 1000 psig. In addition, the high/ low-pressure motor-operated , isolation valves have safety-grade, redundant pressure interlocks. Also, the motor-operated emergency core cooling system (ECCS) valves will only be tested when the reactor is at low pressure. All inboard check valves on the ECCS will be testable and have position indication. Additionally, design criteria used by GE require that all pipe designed to 1/3 or greater of reactor pressure i ' recuires two malfunctions to occur before the pipe would be subjected to reactor system pressure. The pipe designed to less than 1/3 reactor pressure requires at least three malfunctions before the pipe would be subjected to reactor system pressure. The steff concludes that designing, to the extent practicable, low-pressure systens to withstend full RC5 oressure is an acceptable means for resolving this issue. However, the staff believes that f or those systems that have '
'ot beer designec 10 %1thstano fiil RCS pressure, evolutionary ALWRs should L _ __ _ _ _ _ _ _ . _ _ . _ . -
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-g-provide (1) the capability for leak testino of the pressure isolation valves,
- 2) velve position inoication that is available in the control room when isolation valve operators are deenergized and (3) high-pressure alarms to warn control room operators when rising RCS pressure approaches the design pressure of attached low-pressure systems and both isolation valves are not closed. Ja>osition of these requirements exceed Commission regulations and cuidance; taeref ore, the staff recommends that the Commission approve these positions for evolutionary ALURs.
The staff notes that for sone low-pressure systems attached to the RCS, it may not be practical or necessary to provide a higher system ultimate pressure capability for the entire low-pressure connected system. The staff will evaluate these exceptions on a case-by-case basis during specific design certification reviews. III. Mitigative Feature I wies A. Hydrogen Generation ad ,'ontrol The Commission's Severe Accident and Standardization Policy Statements provide that future designs should address the provisions of 10 CFR 50.34(f). The Commission's stated policy has been codified in 10 CFR Part 5? to require the technically relevant provisions of 10 CFR 50.34(f) be met. Specificall order that containment integrity be maintained,10 CFR 50.34(f)(2)(ix) y, in recuires future designs to provide a system for hydrogen control that can safely a:commodate hydrogen generated by the egoivalent of a 100 percent fuel-clad rietal-water reaction. In addition the regulation requires this system to be capable of precludin exceeding 10 percent (by volume),g uniform or an inerted co,ncentrations atmosphere within the of hydrogen from containnent must be provided. The ALWR Requirements Document specifies that containment and combustible gas control systems should be designed to accommodate 75 percent in-vessel rirconium-water reaction of the active fuel cladding, and 13 percent containment uniform hydrogen concentration. It states that 75 percent cladding oxidation is believed to be a conservative upper limit on the amount of hydrogen generated in aoptimization an degraced-coreissue.situation including recovery. EPRI has identified this as The RESAR SP/90 design proposes to mitigate the effects of a 100 percent metal-cater reaction and to preclude uniform hydrogen concentration from exceeding 10 percent systems. (by volume) through the use of hydrogen igniter and hydrogen recombiner The System 80+ design proposes to be consistent with the recomendations of the ALWR Requirements Document resulting from staff review. The infomation will include justifications for the assumed extent of metal-water reaction and the allowable uniform hydrogen concentrations. TheABWRdesignmeetstherequirementsof10CFR50.34(f)(2)(ix)byutilizing a nitrogen-inerted atmosphere within its containment. Also, a hydrogen recembiner for design-basis accidents will be provided in the AEWR design.
l Aside from the issue of regulatory compliance and applicability, and due to the uncertainties in the phenomenological knowledge of hydrogen generation and corbustion, the staff concludes that compliance with the criteria of 10 CTR 50.34(f) remains appropriate for combustible gas control design in ALWRs. Research (discussed in ill' REG /CR-4551) indicates that in-vessel hydrogen generation associated with core-danage accidents may range from approximately 40-95 percent active cladding oxidation equivalent. The ersount of cladding oxidation is dependent on a variety of parameters related to sequence progression: reactor coolant system pressure, reflood timing and flow rates, as well as core-melt progression phenomena. Thus, a 75-percent-equivalent cladding reaction continues to be viewed at a reasonable design basis for hydrogen generation for severc accidents in which the reactor pressure vessel (r,rV) remains intact. However, it is the staff's view that ALWRs should previde r,rotection for hydrogen generation resulting from a wider spectrum of accidcrts, i.e., full core-melt accidents with RPY failure, in that context, it is also necessary to consider ex-vessel hydrogen generation as a result of core debris reacting with available water or core-concrete interactions. Calculaticns using the CORCON models indicate that if the core debris is cooled in relatively rapid fashion (1-2 hours), additional hydrogen generation will be less than that equivalent to a 25-percent cladding oxidation reaction, This relatively limited ex-vessel reaction is conditional on the existence of a coolable debris bed and the availability of sufficient water. If extensive cere-concrete interaction occurs due to the absence of cavity flooding, more hydrogen generation sbculd be considered. Corsidering the effects discussed above, the staff concludes that an equivalent 100 percent cladding oxidation reaction is an appropriate deterministic design criteria and a reasonable surrogate for the cortination of both in-vessel and ex-vessel hydrogen genera-Sion. Due to the uncertainties in the phenomenological knowledge of hydrogen generation and corbustion, it is still the staff's position that, as a minimum, evolutionary ALWRs should be designed to (1) accommodate hydro metal-water reacticn of the fuel cladding and (2) gen equivalent limit containment to 100-percent hydrogen concentration to no greater than 10 percent. Furthermore, because hydrogen control is necessary to preclude local concentrations of hydrogen below deton-able limits, and given uncertainties in present analytical capabilities, the staff concludes evolutionary ALWRs should provide containment-wide hydrogen control (e.g., igniters, inerting) for severe accidents. Additional advantages of providing hydrogen control mitigation features (rather than reliance on random ignition of richer mixtures) includes the lessening of pressure and temperature leadines on the containment and essential equipment. The staff recorrends that the Comnission approve the staff's position that the require-ments of 10 CFR 50.3al1 H 2)(ix ) remain unchanged f or evolutionary ALWRs. B. Core-Concrete Interaction - Ability To Cool Core Debris i ' In the unlikely event of a severe accident in which the core has melted through the reactor vessel, it is possible that containment integrity could be breached i if thE Colten core is not sufficiently cooled. In addition, interactions
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between the core debris and concrete can hydregen and other non-condensible gases, generate large quantities of additional overpressure failure of the containment. which could contribute to eventual . The EPR1 Requirements Document contains a number of design features that are intended to mitigate the effects of a molten core. To promote long-term debris coolability, the Requiremegts Document states that the cavity floor should be sized to provide 0.02 m /MWt. The Requirements Document also specifies that the cuntainment should be designed to ensure adequate water supply to the floor and that an alternate means of introducing water into the containment, independent of normal and emergency ac power should be provided. Passive schemes for providing flooding of tae floor areas,beneath the vessel are proposed and described in general terms for both BWRs and PWRs. The ; Requirements pocument also states that the steel shell or linar of the conteinment should be protected from core debris by at least 3 feet of concrete. Westinghouseindicg/MWtandthattheRESARSP/90designwillincludted criteria of 0.02 m e some t method (not yet defined) that would ensure automatic flooding of the lower cavity, using the in-containment refueling water storage tan t (RWST), in the
- ovent of a severe accident. Westinghouse is currently evaluating alternative designs to ensure compliance with that commitrent.
core-debris dispersal criteria of 0.02.m designAlso, CEhasalsoindicatedthattheSystem80)/NWt . will comply with the EPRI the in-containment
)
L refueling flooding. water storage tank will previde a source of water for lower cavity The ABWR design has a number of features that the staff generally agrees would mitigate the effects of a molten core. It is designed with a lower drywell flooder and a cavity space sufficient to be able to disperse core debris at an energy level of 0.02 m /MWt. The flooder consists of a number of temperature-l sensitive fusible plugs thet allow suppression pool water to enter the drywell i drywell.when cavity high tenperature resulting from core debris occurs in the lower The hori:ontal vents to the suppression pool will remain covered in the event of lower drywell flooding, ensuring that releases continue to be scrubbed through the suppression pool water. GE anticipates that any core-concrete interaction will be stopped when the suppression pool water quenches the molten core debris. By providing sufficient area to allow the core debris to spread to a shallow bed and by flooding the core debris, it is expected that the potential for extensive core-concrete interactions will be significantly reduced. In addition, even if limited core-concrete interactions continue, the-overlying pool of water will mitigate the consequences of these interactions by scrubbing the fission products and cooling the gases released from the core-concrete interaction.
The staff believes that an acceptable resolution to this issue can be provided by the evolutionary ALWR vendors if their designs provide sufficient reactor cavity floor space to ' enhance debris spreading, Lnd provide for quenching debris in the reactor cavity. Use of these criteria exceed current regulatory practice. 2 It should be noted that the specific cavity sizing criteria (0.02 m /MWt) proposed in the Requirements Docunent is still under evaluation by the staff. The issue of debris coolability is an area in which there is active ongoing experimental research including relatively large scale testing jointly sponsored by EPRI and NRC. Additionally, without assurance of core debris coolability, the level of protection afforded by a 3-foot thickness of concrete and the issue of vessel pedestal attack (atlation of concrete supporting the reactor vessel by the molten core debris) require further evaluation. The staff will continue to evaluate the issue of core debris coolability and the specific cavity sizing criteria (0.0?m'/MWt) proposed by EPRI as more data and information becomes available. The staff intends to assess the debris flooding schemes proposed by EPRI on a design-specific basis. The staf f recemends the Comission approve exceeding past regulatory prac-tice in resolving this issue. The statf recomends approval of the general criteria, stated above, that evolutionary ALWR oesigns; 1) provide suf ficient reactor cavity floor space to enhance debris spreading, and 2) provide for quenching debris in thc reactor cavity. Design specific approaches to resolve this issue will be evaluated by the staff on a case-by-case basis to ensure compliance with these criteria. C. High Pressure Core Helt Ejection One potential effect of a severe accident that could potentially result in containment failure is the phenomenon of direct containment heating (DCH). The staff is ccncerned that this phenomenon might occur from the ejection of molten core debris under high pressure from the reactor vessel resulting in wide dispersal of core debris and extremely rapid addition of energy to the containment atmosphere. To limit direct containment heating, the ALWR Requirements Document states that the cavity / pedestal /drywell configuration should be designed to preclude entrainment of core debris by gases ejected from a failed reactor vessel. It also states that a safety-grade RCS safety depressurization and vent system (SDVS) will be provided. The staff review has concluded that reactor vessel l depressurization capability and cavity design features to entrap ejected core debris constitute an acceptable approach to the issue of high-pressure core i melt ejection. I Westinchouse has indicated that the configuration of the cavity of the RESAR SP/90 containment will prevent core debris from entering the upper containment.
In addition, their ac independent depressurization system will reduce the probability of a high-pressure molten-core ejection from the reactor vessel. CE has indicated the System 80+ design includes an indirect cavity vent path, including a debris collection chamber, (which is configured to de-entrain solid core debris and minimize direct containment heating) and a large floor area to enhance core debris coolability. In addition, the design includes a safety grade depressurization system which minimizes the possibility of high-pressure molten-core ejection. The ABWR design incorporates a safety grade depressurization system and a suppression pool that surrounds the lower drywell cavity and thereby reduce the risk of high pressure core ejection and would prevent core debris from reaching the containment boundary and breaching its integrity. The steff concludes that, during a high-pressure core-melt scenario, a depres-surization system should provide a rate of RC5 depressurization to preclude molten-core ejection and to reduca RCS pressure sufficiently to preclude creep rupture of steam generator tubes. Primary systems of evolutionary ALVRs should have the capability to be depressurized after loss of decay heat removal. In addition, the staff concludes that the ALWR Requirements Document should include e requirement tFat reactor cavities be arranged in such a manner the high-pressure core debris ejection resulting from vessel failure will not impinge on the containment boundary. The staff concludes that ALWR desions should include a depressurt20 tion _ system anc cavity cesion f eatures to contain eiected core debris. Imposition of these recuirements exceed current Commission regulations. The staff recommends that the Comission approve this position for evolutionary ALWRs. D. Containment performance The containment function, i.e., maintenance of a strong leak tight berrier against radioactive release, is faced with distinct challenges as a result of a severe accident. These challenges may be roughly divided into two categories, energetic or rapid energy releases and slower, gradually evolving releases to the closed containment system. Examples of containment loadings that fall into the first category include high-pressure core melt ejection with direct containment heating, hydrogen combustion, and the initial release of stored energy from the reactor coolant system. Slow energy releases to the containment are typified by decay heat and noncondensible gas generation. Engir,eering practice in containment design calls for passive capability in dealing with energetic energy releases where practicable while long-term energy releases may be controlled by both passive means as well is through active intervention. In view of the low probability of accidents that would challenge the integrity of the containment, the staff concludes that the probability of failure of the ritigation systems (those systems which can reduce the consequences of a core damage accident), from the onset of core damage to loss of containment integrity resultire in an uncontrollable leakage substantially greater than the design basis leakage, should not exceed approximately 0.1. However, the staff intends
( te ensure that the contairment can deal with all credible challenges and does not intend to apply this conditional containment failure probability CCFP guide lir.e in a manner that could be interpreted to potentially detract from,(overa)ll safety. The staff will accept a CCFP of 0
- or a deterministic containment performance goal that offers comparable protection. For this reason, the staff concludes that the following general criterion for containment performance ALWRs ainscvere-accident during place of a CCFP. challenge would be appropriate fcr the evolutionary The containnent should Wntain its role as a reliable leak tight barrier by ensuring that containment stresses do not exceed ASME service level C limits for a minimum period of 24 hours following the one.et of core dama5e and that following this 24 hour period the containment should continue to provide a barrier against the uncontrolled release of fission products.
Maintaining containment integrity for a minimum period (e.g. 24 hrs) is baseo en providing sufficient time for the remaining airborne activity in the contain-rent (principally ncble gases and iodine) to decay to a level that would not exceed 10 CFR Part 100 dose guideline values when analyzed realistically, if controlled venting were to occur after that time. During this period, contain-nent integrity should be provided, to the extent practicable, by the passive capability of the conteinment (e.g., suppression pool). The itself and any related passive design features staff further believes that following this period, the containr'ent should continue to provide a barrier against the uncontrolled release cf fissior products. However, in keeping with the concept of allowing for intervention in coping with long-tena or gradual energy release, the staff believes that after this minimum perici, the containment design may utilize controlled, elevated venting to reduce t ie probability of a catastrophic failure of the containment. Alternatively, a d'. sign may utilize diverse con-tainment heat removal systems or rely on the :estoration of normal containment heat removal actions capability (e.g., 48 hours! if sufficient time is available for major recovery EPRI has indicated that the ALWR public safety criteria do not contain explicit criteria for conditional probability of containment failure or other mitigation features since the ALWR Steering Comittee believes that such criteria could potentially distort the balance in safety design and inhibit innovative improvements in core protection features. However, EPRI has not yet indicated their position on an alternate containment perforrance goal. Westinghouse has not yet comitted to a specific containment performance goal for RESAR SP/90 although it is expected that the mitigation features discussed by Westinghouse would lead to a CCFP of less than 0.1 for all credible accident scenarios.
CE expects that the Systern 80+ design will meet the CCFP goal of 0.1 when heighted over credible core damage sequences given the following assumptions.
-Credible core damage sequences are defined as all core damage event sequences with a frequency of greater than 1.0 X 10E-6 per reactor year.
External events which would cause both core damage and concurrently fail the containment and which have a frequency of less that 1.0 X 10E-5 per reactor year will not be considered.
-Containment failure is defined as a post core damage release resulting in a cose greater than 25 rem beyond one-half mile f rom the reactor.
The AEWR design currently includes a hardened wetwell vent for containtnent over pressure protection and is cormitted to a CCFP that is less than 0.1 when weighted over credible core damage sequences. In meetings with the staff EPRI has stated that they consider a containment vent to be philosophically t and institutionally undesirable and potentially unworkable. For additional informatier, see related discussions under ALWR Public Safety Goal and ABWR Containnent Vent Design. Defense in depth, a long standing fundamental principle of reactor safety, results in the concept that multiple barriers should be provided to ensure against any significant release of radioactivity. In its Severe. Accident Policy Statement, the Commission indicated that it "... fully expects that verders engaged in designing new (or custom) plants will achieve a higher standard of severe accident safety performance than their prior designs." The l Commission reaffirmed this policy in an SRM dated December 15, 1989 relating to l SECY-89-311. A defense-in-depth approach reflects an awareness of the neeo to I make conservative safety judgerrents in the fare nf uncertainties; in effect, not putting all the eggs in one basket. In that regbrd, the reactor containment boundary snould serve as a reliable barrier against fission product release for credible severe-accident phenomena / challenges. Special effort should be made to eliminate or further reduce the likelihood of a sequence that could bypass the conta ir. ment. The continued reliance on the traditional prirciple of containment I i of fission products following an accident is seen as a logical and prudent approach to addressing reasonable questions which will persist regarding the ( i ability to accurately predict certain aspects of severe accident behavior. In i order to ensure balance between prevention and mitigation, some criteria on t containment performance are appropriate. Accordingly, a general goal of l i limiting the conditional containment failure probability to less than 1 in 10 when weighted over credible core-damage sequences would constitute appropriate attention to the defense-in-depth philosophy. Alternatively, a deterministic - containment performance goal that provides comparable protection would be l appropriate. l Probabilistic risk assessment (PRA) is a very powerful tool that permits systematic integrated assessment of design strengths and weaknesses. However, l because very low frecuency scenarios (approrimately 1.0 X 10E-6 per reactor-year) I ' are being addressed, it is important to recognize the large uncertainties in the cuantification of these scenarios. The overall uncertainties in severe j accident behavior are driven largely by insufficient data for assessing ] comr.on-cause failures, difficulty in quantification of the potential for human i l
errors, and questions about completeness of analyses and uncertainties in phencrenological behavior. For this retson, the staff considers it acceptable to utilize a deterninistic containment performance criterion that would provide a level of containment performance comparable to that which could be demonstrated using a probabilistic containment failure goal of 0.1, given a severe accident. It is recommended that the Comission approv,e the staff's position to use a CCFP of 0.1 or a deterministic containment perf ormance coal that of fers
~
comparable protection in the evaluation of evolutionary ALWRs. I. ABWR Containnent Vent Design In Amendrent 8 of the ABWR SSAR submittal (July 28,1989) GE submitted a , sensitivity analysis of the ABWR PRA to determine the net risk ber.efit of a vent system. Basically, this system is a containment overpressure relief system and is designed to avoid gross containment failure resulting from postulated slow rising overpressure scenarios that could result from postulated riultiple safety system failures. These sensitivity analyses indicate that, with or without a vent system, the ABWR design meets the quantitative health ob,1ec-tives of the Commission's safety goal with a wide margin. The staff's detailed review of the ABWR risk analyses, including the sensitivity analyses on the vent system, is currently underway. Based on the review to date, the staff b(11 eves that the scope of n thods and data used in the AEWR PRA are sufficient and do not expect that the ABWR risk to exceed the Commission's quantitative health objectives with or without a vent system. The staff's safety goal implenientation plan also recomended that a subsidiary target related to plant performance be used. This target states that, for future plants, a mean core damage frequency due to inte'.nal events and external events be less than 1.0 X 10E B per reactor year of operation. The staff's review of Amendment 8 of the ABWR SSAR indicates that the overall core damagt frequency irem internal events (transients, ATWS events, and postulated LOCAs) and external everits (primarily from beyond design basis seismic events F and postulated fires) is about 6 X 10E-6 per reactor year. GE has determined thrt the proposed vent system has negligible impact on the :: ore damage frequency. The staff notes that GE has provided an additional means of decay heat removal (a third train of RHR and an ac-independent water makeup system which relies on the fire water system to supply water to the core and containment sprays in emergency situations) for the ABWR design to reduce the loss of containment heat removal frequency function, thus of reducing the sequences the benefit involving (on core damage frequency only) of t ABWR vent system for these types of accident sequences. The desirability of venting a BWP. containment to mitigate multiple-failure accidents far beyond the design basis has been accepted for some time. Since 1981 the BWR Emergency Procedure Guidelines (EPGs), developed by the BWR Owners Group and approved by the NRC for existing BWRs, have called for venting the containment wetwell air space. GE believes contain;aent
l 1 l overpressure protection represents a practical and beneficial feature to incorporate in the ABWR. The overpressure protection feature is essentially passive, relatively inexpensive in a new plant, provides insurance against the consequences and financial risks associated with end-of-spectrum accident scenarios, is consistent with the BWR EPGs, and appears to be consistent with the ALWR philosophy of robustness. I GE has established two severe accident goals in the risk analyses submitted to the staff. These goals were defined in the ABWR LRB. The first goal states that the frequency of a severe accident release resulting in a whole body dose of 25 rem beyond one-half mile from the reactor should not exceed 1 X 10E-6 per reactor-year. - goal. The second This design goal is basically the same as the EPRI ALWR design l goal defined in the ABWR LRB states that the conditional ; contairment failure probability should be less than one in ten (CCFP 0.1) : when weighted over credible core damage sequences. The staff and GE agree that the definition of containment failure is an uncontrollable leakace substantially greater than the design bas ~is resulting from loss of containment integrity following the onset of severe core damage. The ABWR design with the vent systen is expected to meet the above goals; however, staff review in this area is not yet complete. CE has performed an analysis utilizing this definition of containment failure to determine if the ABWR meets the CCFP goal of 0.1. The analysis indicates that the CCFP for the ABWR design, without a vent system, is equal to approximately 0.5 and does not meet the 0.1 goal, however with a vent system, the CCFP equals approxinately 0.06. l Based upon the preliminary review of the ABWR severe accident design, the staff has determined that, as far as overall risk impact is concerned the GE ABWR public safety goal is significantly more stringent than the ComIssion's quantitative health objectives. Also, the staff concludes that the public safety goal proposed by GE for the ABWR design is more stringent than the "large release guideline" as defined in the staff's proposed safety goal implementation plan. ! _Therefore, besee on the apparent enhanced level of safety provided by the i ASWR's severe accident design features, which include the over pressure protec-l tion system the staff recommends the comission approve its use in the ABWR design certification process._ F. Equipment Survivability With regard to the Comission's request concerning 'The measures to ensure 2 hat systems and equipment required only to mitigate severe accidents are i available to perform their intended function (e.g., environmental qualifica-tions)," the staff believes that features provided for severe-accident prctection (prevention and mitigation) only (not required for design basis accidents) need not be subject to (a) the 10 CFR 50.49 environmental qualifica-tion requirerents, (b) all aspects of 10 CFR Part 50 Appendix B quality assurance requirements, or (c) 10 CTR Part 50, Appendix A redundancy / diversity requirements. The reason for this judgment is that the staff does not believe that severe core damage accidents should be design basis accidents (DBA) in the traditional sense that DEAs have been treated in the past. l
- . _ - . _ - . _ _ -. = - _ _ _ .
-Sto Notwithstanding that judgment,' however, mitigation features must be designed 50 there is reasonable assurance that they will operate in the severe-accident environ.
ment for which they are intended and over the time span for which they are needed. In instances whtre safety related equipment, (which is provided for cesign bases accidents) is relied upon to cope with severe accidents situations; there should also be a high confidence that this equipment will survive severe accident conditions for the period that is needed to perform its intended func. tion. However it is not necessary for redundant trains to be qualified to meet this goal. During the rev m of a specific ALWR design the credible severe accident scenarios, the equiprnent needed to perform mitigative functions, and the conditions under which the mitigative systems must function, will be identifiec. Ecuipment surv h ttility expectatiora under severe accident conditions should include consideration of the circumstances of applicable initiating evtnts (e.g., station blackout, earthquakes) and the environment (e.g., prescure, temperature, radiation) in which the equipment is relied upon to function. The required system performance criteria will be based on the results of these design-specific reviews, in addition, the staff concludes that severe-accider.t mitigatien equipment for evolutionary ALWRs should be capable of being powered from an alternate power supply as well as from the normal Class IE onsite systems. Apperdices A and B to Regulatory Guide 1.155, " Station Blackout, provide guidance on the type of quality assurance activities and specificatiers which the sta'f concludes are apprcpriate for equipment utilized to prevent and citigate the consequences of severe accidents. The staff recuests that the Commission approve the staff pcsition that features provided only f or severe-accident protection need not be subject to the 10 UK 50.49 environmental cuelit ication recuirerer.ts.10 CFR Part 50, Appendix 5 cuality assurance recuirements, ano 10 CFR Part 50, Appendix A redundancy [ diversity reouirements. 8V. Non-Severe Accident 1ssues The folltsing issues, which are not normally considered through PRA analysis or not considered as severe accident issues for the evolutionary ALWRs, are brought to the Commission's attention because either the staff's positions or the vender requests differ from past practices. A. Operating Bases Earthquake (OBE)/ Safe Shutdown Earthquake (SSE) Presently,10 CFR Part 100 requires that the magnitude of the OBE be at least ore-half that of the SSE. It has been an industry wide experience that such a requirement leads to a design that is governed by the OBE requirements and produces unnecessary and inconsistent margins for the SSE loading. This requirement was included in the regulation when the staff did not have substantial experience with the seismic resistance of plants that incorporated OBE design at half the SSE value. Since then a number of research programs have been conducted including a large industry effort on testing and observation of actual earthquake experience of industrial facilities; consequently, the NRC funded Piping Review Committee has concluded that the OBE at existing plants are too high, therefore, controlling the design of some safety systems, and recoerended that the OBE be decoupled from the SSE. Certain interim measures, 1
such as allowing somewhat higher damping values for piping analysis, have been 1061 to talen partially implenent the Piping Review Comittee recortmendations (HUREG 1984 involvear)e. But the complete implementation of the recommendations would vision of 10 CFR Part 100, Appendix A. work, the effort on revision of this regulation has been postponed,Because it should of higher prio be noted that the Commission hat, in certain site specific cases, previously approved OBEs of less than one-half the SSE. EPRI has requested that NRC regulations be changed to reduce the magnitude of the OBE relative to the SSE as a basis for the design. All evolutionary ALWR verders agree with the request. GE has stated that they agree with EPRI in principle, however, the ABWR design uses an OBE that is one-half the SSE; therefore, this is a on-issue for the ABWR. optimization issue. EPRI has identified this as an The staff agrees that the OBE should not control the design of safety systems. _however, a staff positier. or, this issue to be applied generically to all f uture designs has not yet been f ully developed. For the evolutionary reactors, the staff will specific consider reouests to decouple the OBE from the SSE on a cesign-tesis. regulations, thereforeSuch thi a cecoupling would require an exemption to the Comission's statf desigr,-specific rtlief approach. recomends that the Commission approve this P. Inservice Testing of Pumps and Valves The ASME Code, Sectior. XI,
- Rules for Inservice Inspection of Huclear Power Plant Corrponents" has been used to establish past testing re ASME Code Class 1, 2, and 3 safety-related pumps and valves.quirements These for I requirements provide certain infermation on the operations 1 readiness of the components, but in general, do not necessarily verify the capability of the components to perform their intended safety function. It is the staff's
. judgement that the Code does not assure the necessary level of component operability that is desired for the evolutionary LWR designs. The staff believes that the following aspects of pump and valve testing and inspection are necessary to provide an adequate level of assurance of operability. The following provisions should be applied to all safety related pumps end vaTves and not limited to A5NE Code Class 1, 2. and 3 components.
-Piping desian should incorporate provisions for full flow testing (maximum design flow) of pumpsi and check valves.
-Designs should incorporate provisions to test motor operated valves under oesign b6 sis differential pressure.
-Check valve testing should incorporate the use of advanced ncn-intrusive techniques to address degredation and perforir.ance characteristics.
l l
-A program should be established to determine the frequency necessary for
! disasser.bly and inspection of pumps and valves to detect unaccept_cble deprecation which cannot be detected through the use of advanced non-intrusive techniques, i l [ l i
,,g ACTION - Murley NRR/
/po es 4,, UNITED STATES Beckjord. RES ;
' ',r NUCLE AR REGULATORY COMMISSION
[ y g . I wasmwatow.o.c.nosss Cys: Taylor ' June 26, 1990 Sniezek
\.....'c* Thompson st7[rN v" s. RES Beckner RES Scroggins. OC Scaletti. NRR CMiller. NRR MEMORANDUM TOR: James M. Taylor, Executive Director for operatic TROM: samuel J. Chi , c etary
SUBJECT:
SECY-90 EVOLUTIONARY LIGHT WATER REACTOR (LWR) CERTITICATION ISSUES AND THEIR RELATIONSHIPS TO CURRENT REGULATORY REQUIREMENTS This is to advise you that the commission as detailed below has approved in part and disapproved in part the staff's - recommendations in SECY-90-16. I. General Issues. A. ALWR Public Safety Goal. , The Commission (with Chairman Carr and Commissioners i Roberts, CurtQsandRemickagreeing) has disapproved the use of 10 per year of reactor operation as a core damage frequency for advanced designs. As noted in the SRM on SECY-89-102 (dated June 15 24 1990), the Commission supports the use of 10 per year of reactor operation as a core damage frequency goal. Although l the Commission strongly supports the use of the l information and experience gained from the current i generation of reactors as a basis for isnproving the safety performance of new designs, the NRC should not adopt industry objectives as & basis for establishing
! new requirements. However, if the staff in applying the criteria of 10 CFR Part 52 (and in view of the uncertainties associated with PRA's) concludes that l
additional requirements are needed, based on our experiences with prior designs, in order to provide NOTE: THIS SRM AND THE SUBJECT SECY PAPER WILL BE MADE PUBLICLY
, AVAILABLE 10 WORKING DAYS AFTER ISSUANCE OF THE SRM.
k '0 0fl. EDO 6el:, kdhD IJmi , E 2 $$
2 assurance that future designs will meet the Safety Goal Policy Statement, then the staff should provide those additional requirements to the Commission for consideration as they are identified. Commissioner Rogers approved the staff's use of 10-5,, an expected design target for ELWR designers and endorsed a requirement that applicants be able to demonstrate that they have taken reasonable steps to reach these targets. However, he does not endorse those goals as an absolute requirement for approval of any specific design. Consistent with the Commission's decision on SECY-89-102, the Commission approved the overall mean frequency of a large release of radioactive material to the environment from a reactor accident as less than one in one million per year of reactor operation. The Commission has not agreed on a definition of a large release and has requested a paper from the staff (See SRM from SECY-89-102). (RES) (Suspense: 9/28/90) 9000136
- a. Source Term.
The Commission (with all Commissioners agreeing) has approved the staff's approach to the source term with the addition of the following elements o on an expedited basis, incorporate appropriate changes to regulations, regulatory practices, and the review process resulting from source term ! research. (RES/NRR) - As appropriate - Pending Source Term Study Results. SECY paper due II. Preventative Feature Issues. irminently. fsjfscf A. ATWS. The Commission (with all Commissioners agreeing) has approved the staff position. However, if the applicant can demonstrate that the consequences of an ATWS are acceptable the staff should accept the demonstration as an alternative to the diverse scram system. Commissioner Curtiss further believes that the staff should retain the flexibility to accept designs with non-diverse scram logic in those instances where it is demonstrated to the staff's satisfaction that the reliability of the scram function is such that the risk from ATWS is insignificant. (NRR) l B. Mid-Loco Operation. l The Commission (with all Commissioners agreeing) has approved the staff's proposed position, with the ACRS recommendation of April 26, 1990, that four additional specific reguirements be considered for mid-loop operation. (NRR)
3 C. Station Blackout. The commission'(with all commissioners agreeing) has approved the staff's position that the evolutionary ALWR's have an alternate ac power source of diverse design capable of powering at least one complete set of normal shutdown loads. The staff should provide a clear definition of " diversity" so as to provide guidance on whether it means different types, different manufacturers, different models, etc. Commissioner curtiss noted that, in his view, the clarification ' should focus on limiting common mode failure pot 2ntial but need not go so far as to require completely different generator driver technologies (e.g. should not necessarily require both diesel and gas turbine driven generators). (NRR) D. Fire Protection. The commission (with all commissioners agreeing) has approved the staff's position on fire protection as presented in SECY-90-16 and supplemented by the staff's April 27, 1990, response to the ACRS comments. (NRR) E. Intersystem LOCA. - The Commission (with all commissioners agreeing) has approved the staff's position on intersystem LOCA provided that, as recommended by the ACRS, all elements of the low pressure system are considered (e.g. instrument lines, pump seals, heat exchanger tubes, and valve bonnets.) (NRR) III. Miticative Feature Issues. A. Hydrocen Generation and Control. The commission (with all commissioners agreeing) has approved the staff's position that the requirements of 10 CFR 50.34 (f) (2) (ix) should remain unchanged for evolutionary plants. The staff should seek additional technical information, as suggested by the ACRS, and if reconsideration is warranted the commission should be advised. (RES/NRR) B. Core-Concrete Interaction--Ability to Cool Core Debris. The Commission (with all Commissioners agreeing) has approved the staff's position. (NRR/RES) C. Mich Pressure Core Melt Eiection.
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The commission (with all commissioners agreeing) has approved the staff's position that the ELWR designs include a depressurization system and cavity design to contain core debris. The cavity design, as a 1
4 mitigating feature, should not unduly interfere with operations including refueling, maintenance, or surveillance activities. (NRR/RES) D. Containment Performance. The Commission (with all Commissioners agreeing) has approved, consistent with SECY-89-102, the use of a 0.1 CCFP as a basis for establishing regulatory guidance for the ELWRs. This objective should not be imposed as a requirement in and of itself. The use of the CCFP should not discourage accident prevention and the staff should review suitable alternative, deterministically-established, containment performance objectives providing comparable mitigation capability if submitted by applicants. Any such alternatives should be submitted to the Commission following staff review. (NRR/RES) E. ABWR Containment Vent Desian. The Commission (with all Commissioners agreeing) has approved the staff's recommended use of the containment overpressure protection system on the ABWR, subject to the results of the comprehensive regulatory review which should fully weigh the potential "downside" risks with the mitigation benefits of the system. Staff l should ensure that full capability to maintain control over the venting process is provided. (NRR) F. Eculerent Survivability. The Commission (with all Commissioners agreeing) has approved the staff's position. (NRR) IV. Non-Severe Accident Issue. A. Operatina Basis Earthcuake (OBE)/Rafe Shutdown Earthcuake (SSE). The commission (with all Commissioners agreeing) has approved the staff's position. (NRR) B. Inservice Testina of Pumos and Valves. , The Commission (with all Commissioners agreeing) has i approved the staff's position as supplemented in their April 27 1990, response to the ACRS comments. The 3 Commirsion notes that due consideration should be given to the practicality of designing testing capability, particularly for large pumps and valves. (NRR/RES) ! The Commission also agreed that in those cases where the staff i proposed requirements depart from current regulations, consideration should be given to incorporating these requirements into the regulations. (See SRM dated May 27, 1990, M90053A).
5
'.nally, the staff is encouraged to strive to sustain the level of attention and resources that have been devoted recently to the review process for the EPRI requirements document. The recent coraents of the EPRI representatives at the June 4, 1990 Commission briefing suggest that such a commitment, if sustained, can be most beneficial in assisting EPRI and the HRC staff in our respective efforts to reach a coraon understanding on the key technical issues. (NRR) cc Chairman Carr Coraissioner Roberts Commissioner Rogers Commissioner Curtiss Coraissioner Remick OGC ACRS IG ASLBP ASIAP l
l l t l l l l
APPEllDIX C UtiRESOLVED AND GENERIC SAFETY ISSUES (To be provided in a supplomont to this DSLR.) C-1 ABWR DSER
_ . . . _ - _ _ . _ . _ _ . . .. . . _ = _ . _ _ . _ _ . _ _ _ . . . _ . . . _ _ . . _ _ _ . - _ _ _ _ _ _ _ _ _ . _ _ - = _ . _ _ _._ ._ .m. APPENDIX D COMPLIANCE WITH 10 CPR 50.34(f) AND NUREG 0737 (To be provided in a supplement to this DSER.) t D-1 ABWR DSER
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APPENDIX E CONTRIBUTORS AND CONSULTANTS (To be provided in the Final sep,) M l-i l i 1 l E-1 l ABWR DSER 1
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