ML20148H113
| ML20148H113 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/20/1977 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | ALTERNATIVE ENERGY COALITION |
| Shared Package | |
| ML20148H104 | List: |
| References | |
| NUDOCS 8011170186 | |
| Download: ML20148H113 (2) | |
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, umTro STATES NUCLEAR REGULATORY COMMisslON WASHINGTON, D. C. 20555 I t, / '
July 20, 1977 l
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, Docket No. 50-29 i I
l Secretary
' The Alternative Energy Coalition . l l
31 Federal Street
- Greenfield, Massachusetts 01301
Dear Secretary:
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Thank you for the draft copy of a press release in which the Alternative
' Energy Coalition ( AEC) of Greenfield states its demand for a Franklin :
County investigation of concerns relating to the current refueling opera-tions and conditions leading up to the recent strike of guards at the Yankee-Rowe nuclear power station.
We have noted the AEC's concerns in these matters and offer the following l
clarifying comments without prejudice to the investigation demanded by
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the AEC as indicated in the draft press release.
In accordance with the Nuclear Regulatory Commission (NRC) requirements, Yankee Atomic Electric Company (the licensee) promptly notified the NRC's regional Office of Inspection and Enforcement on June 9,1977, that
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shutdown of the Yankee-Rowe reactor had been initiated that date and indicated the reasons for the earlier than anticipated shutdown.' On June 22,1977, the licensee submitted to the NRC the required written follow-up report on this occurrence in the prescribed fonnat of a Licensee Event Report (LER). A ccpy of this LER is enclosed. The specific conditions resulting in the earlier shutdo m of Yankee-Rowe, and the actions required of the licensee to obtain NRC approval to return the reactor to operation with the new core loading are summarized in our June 22, 1977 minutes of meeting with the licensee's staff. A copy of the minutes are also enclosed. ,
We would like to point out that the enclosed documents are part of the public record and are on file under Docket No. 50-29 at the local public document room at the Greenfield Public Library.
During the recent strike of guards at Yankee-Rowe an NRC inspector was dispatched to the site to assess the impact of the strike on plant security. The inspector confirmed that regulatory requirements for maintaining plant security had been maintained. The matters relating to the guards demands which apparently caused the strike, as indicat5d in your press release, are outside NRC's jurisdiction.
We therefore cannot comment on, these aspects of your concerns.
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The Alternative Energy Coalition July 20,1977
-l i I trust that our commentssare responsive to the concerns expressed l in the AEC's draf t press release.
Sincerely, W ,
Y or S't ello Ar. Director ,
Division of Merating Reactors !
Office of/fuclear Reactor Regulation
Enclosures:
- 1. Licensee Event Report
,i dated June 22, 1977
- 2. Meeting Minutes dated June 22,197i
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S O P G fit sAlF"N40 The Alternative Energy Coalition ( AEC) G. demandh't as A
Frank.lin County investigaticn of conditions at the Yankee RoweFkovc4 h4A:rtfW Uuclear Power Station.4MT_G:5, '.~he AEC demands -to know '
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why the Rowe YAKEEE operato<rs lied about the conditions sur-rounding their current refuelin e operations, and also wants to get a core co=plete story of the conditi'ons z.n l'eading up to the strike ogf Suards there. The AEC also m: pressed 4 p, M0 A'
/ %d its support for that strike /g;lt+,-4.4 h i 2,.
J Specifically,....
1.cccrding to three articles priric d in the Recorder 6/11, 15, 20, there was cuch confusion as to the reality of the situatien. Faving checked tith the ~.e corder,1 the p1 n'
?.R. n and hcird ~,ecerder st;;crents frc: the Union, the
.C and the utility-:~ew Enc Towcr -still all 7 's tre nct answered 7 the !.20 feels there is but cne choice.
~he investicaticn is eccential to cY.pitin thy :
- 1. Jtility spe:cescen licd 6/11 as to why the nuhe was shut down early. FR can always explain lies ence they have .t;scd,
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J., . ', 1, L!9 but truthful cocounts are the only acceptable method. ' ?E tan Eerb 1.utio said cn 6/23, 6/4 wac the date they were using for shutdown for a Jon,I tire but the :?.C caid " the 2. owe plant was 21.ut dcun for refueling at least 4 wee::s ccrly because of calculation errore? . . .'r r.icinfor stien fed into the computcrs. Co . . officials ccknowledged refueling var schedule d for July.
2 c/.t i thic ' ab .c r:11 c 0 0'. rr ince j witr tre ccmputcr" W Yr
! re'T rs 6/15 to" f,1:ul;;icn r rrors" dis ecu re d by Y; .;:c t c'_. c.c nh
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)db .tho had " cade an ' incorrect assumption in preparing an accident analysis" . Are we to rely on assumptions? '.l hat E.". ACT LY w a s f e d incorrectly to the computer?
- 3. 'hy would Yankec chut down early and loac cven 507.' capacity just beni 1
I because a cocputcr was fed ticinformation? Thisinformationwasnecdedl I
later to submit to the URC prior to start up once refv' ling is cocplc t e . By the way, refueling process takes apx 7 wks and costs J 2 -3 01111c n .
- 4. A strike is currently on at Rowe but when firct interviewcd, I
the ::ew Eng Fower Co could not explain this clearly to rcporters.
5 ~lc demand honest and coarlete c':planaticns froc all nuc1 car Gsetal facilities at cach cvent- normal ofi abnoraal. If it were within our jurisdiction, .these mysteries would already have been in', c;G tigated. 1he public can act wisely if treated with respect. Nuclear I
technolo75 can be cxplained j unlecs the utility is fearful of the truth and of event they consider bad publicity.
'/lc k av e s e nt copies of this critique and our demands to
- cw Ins Icwcr and ::RC. The Altcrnative Encrgy Coalition 31 :edcr:1 Et 3rccnficid :/. ass 01301 l
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YANKEE AT0hilC ELECTRIC C0&1 pally e- , , . g.
Rowe,Mossachusetts 01367 ,
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Ya.=E.N uxse -' June 22, 1977
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l Mr. J. P. O'Reilly, Director WN U S. Nuclear Regulatory Cor.T.ission TN c Office of Inspection and Enforcement Regicn I
- 631 Park Avenue King of Prassia, Pennsylvania 19406 )
I Subj ect: Reportable Occurrence 50-29/77- 30/01T E.rrors. in Accident Analysis
Dear Mr. O'Reilly:
4 In accordance with Technical Specifications, Section 6.9.4.a, the attached Licensee Event Report is hereby submitted.
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Very truly yours, I
$fh H. A. Autio Plant Superintendent MiE/nid
Enclosure:
cc: [40) Director, Office of Inspection and Enfcrcement U.S. Nuclear Regulatory Comission . ..
Washington, D.C. 20555
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Director, Office of Management Infomation 6 Program Control U.S. Nuclear Regulatorf Comissien Washingten, D.C. 20555 pAAW Po///4'o 92 G7
, f (1.ER 77-30/1T) 7 1' UCENSEE EVENT REPORT
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CONTROL BLOCK:l l l l l l l (PLEASE PRINT All. REQUIRED INFORM ATION)
. i 6 UCENst NUMetA YPs b l M ! A lY lK lR l114l l O_ l 0 l-l0 !010 l0 l0 l-l 0 l 025l l4 26 l1 l1 l1 l Il 30 10 l1 l 31 32 7 89 15 ,
DOCKET NUMet A tYENT DATE RCPCRT DATE CATICCRT l0l6l0l817l7l TYP $
l l Tl [L_j l 0 l 5 l 01-10 10 12 19 l 1016121217l71 QCON7 l l 60 61 68 69 74 75 80 7 8 57 SS 59 EVENT DESCRIPTION
@l Apearent errors in the accident analvSiS recuired the plant first to reduce power to80j 7 89
@- l 50% on June 8,1977 then to order a clant shutdown on June 10, 1977. ErrorSinthe_] b0 7 89
@l nc-idant e-nivcie have e evinnelv cccitr ed. The nnelve{s Will be corrected for l
60 7 89 g i Core XIII submittal. (LER 77-30) 60 l
7 89 I o 1. 60 7 89 mus SLI4V F VOLAIDN coto $cce CCMPoNINT COCl O is IF 1 LF_j l ZI Zl ZlZ lZ lZI (L l l Zl 919 19 l lN l 12 17 43 44 47 48 7 89 10 11 ,
CAUSE DESCRIPTION
@l 7 89 See Attached Sheet.
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@ lE l l0 l8 l1 ] l RA l_ lC l l NA l 12 13 44 45 40 60 7 8 9 10 ktNto C ELNst AMOUNT Cf AC*MTV LOCATCN OF AtLEASI 8 LZ_J LU . l NA l lRA l 10 11 44 45 80 7 8 9 PERSCNNEL EXPCSURES MJM Ein Tv P! CISCAiPTCN 1 10 l 0 l 0 ] [Z_j l RA -l 13 60 7 89 11 12 PERSCNN EL 'INJUR:ES NuMata C( $ C R'PT CN 1 1010101 l NA eO ll 7 89 11 12 ,
CFFSfTE CONSECUENCES l
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1 CAUSE DESCRIPTION:
In the course of performing the LOCA analysis for the Core XIII reload submittal, a small break scenario was discovered which appeared more i limiting than any analy:ed for Core XII. In Core XII analyses, it was postulated that the most limiting small break would be one which would cause a very slow blowdown while at the same time allowing direct ECCS spillage into the containment. This break was correctly assumed to be a complete severance of a safety injection line near the Reactor Coolant System (RCS) injection point such that blowdown of the RCS was through a 2.25" I.D. thermal sleeve. In the Core XII analysis, the hydraulic resistance of the therral sleeve was analytically included in the ECCS piping consistent with the assurption of a postulated break near the injection point. In subsequent Core XIII analysis, a new break was identified in which the resistance of the thermal sleeve would not be present in the injection line. (Specifically, the break would have to occur in a small length of piping (1 to 2 feet) upstream of the location of the previously assumed break.] The absence of the flow resistance produces a higher spill rate and a lower ECCS headet pressure than would be calculated with the presence of the flo'. resistance. Lower ECCS header pressure in turn delays the time
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of ECC injection. - - '
It was recognized that the "renoval" of flow resistance in the broken injection line had potential significance for Core XII which was then in coastdown. An immediate Appendix K blowdown analysis was performed (assuming that the plant was at 67% power) to analy e the consequences of the newly discovered break scenario. The results of that analysis indicated that the removal of the thermal sleeve's resistance reduce'd the pressure (and, therefore, increased the time) of ECCS injection to an extent that Appendix K PCT criteria appeared not to be met.
The plant was derated to 50%. Analysis at tnat power level indicated that ECCS injection appeared possible, but sufficient uncertainty with the calculations of header pressure existed to warrant perfoming an orderly plant shutdown. This decision reflected the need for flow tests to verify analytical prediction of pressure drops and pung flow
. deliverability.
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=~ -' t NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20556 DOCKET NO 50-29 DATE: JUN 2 21977 LICENSEE: Yankee Atomic Electric Company (YAEC)
FACILITY: Yankee-Rowe
SUMMARY
OF MEETING HELD ON JUNE 17, 1977, FOR BRIEFING ON MATTERS RELATING TO ECCS PERFORPNiCE EVALUATION FOR YANKEE-ROWE On June 17, 1977. representatives of YAEC met with the NRC staff to report on matters relating to ECCS perfomance at Yankee-Rowe.
A list of attendees is attached.
Important highlights of YAEC's presentations and commitments made during the meeting are summarized below. A copy of YAEC's handout which ,
illustrates significant aspects of the presentations are also attached.
On June 9,1977, YAEC shutdown Yankee-Rowe following its discovery of modeling errors in the ECCS performance analysis being done in ,
preparation for obtaining NRC approval to operate Yankee-Rowe with the next Core XIII. YAEC decided on early shutdown because of difficulties to resolve the analytical uncertainties in the Core XII ECCS performance analysis and to provide more time to accomplish the '
necessary work in preparation for Core XIII startup.
YAEC described the progressive upgrading of the ECCS which was originally installed at Yankee-Rowe during 1960. Presently, the ECCS includes three 50 percent pumping trains (3 High Pressure Safety Injection and 3 Low Pressure Safety Injection Pumps) capable of being powered from redundant onsite emergency diesel generators. One ECCS accumulator ;
provides rapid response to large ruptures in the reactor coolant pressure boundary. Flow from the accumulator begins when the reactor coolant system (RCS) pressure drops below the pressure in the accumulator with :
the concurrent opening of several swing check valves in the injection flow path. The injection flow path separates into four safety injection lines (Yankee-Rowe is a 4-loop reactor) each connected to an RCS cold j leg by a themal sleeve. Each safety injection line (nominal 4 inch) !
has a 4 inch check valve and a 4 inch motor operated valve upstream of !
the check valve. A 3 inch motor operated valve is downstream of the check valve. Existing instrumentation permits monitoring of flow in each safety injection line. The original functional requirements for the 21/2 inch 1.D. therrral sleeve no longer exists. Prior to operation with Core XII YAEC intended to use the motor operated valves to isolate a break in a safety injection line downstream of the check va l ve . A break upstream of the check valve would not result in depressur-ization of the RCS. Because of single failure implications, YAEC was required to operate Core XII with power rerreved from the motor operated valves in the safety injection lines and the valves in the open position.
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JUN 2 21977 YAEC has previously determined in its ECCS performance analysis for Core XII that a break in a safety injection-line at the location of the 21/4 inch 1.D. thermal sleeve would be the most limiting small break (resulting ,
in highest clad temperature for the spectrum of snall breaks). During the Core XIll ECC performance analysis efforts YAEC discovered that if a break were assumed in the short 4 inch pipe section downstream o; the 4 inch check valve in the safety injection line, this would result in a higher pean clad temperature than for the break location at the thermal sleeve. j While the RCS blowdown characteristics would remain the same (blowdown '
would still be through the flow resistance of the thernal sleeve), the spill of accumulator and pumped injection water su the containment floor (previously assumed through the 21/4 inch thermal sleeve) would be significantly greater because of the lower flow resistance at the location i l
of the 4 inch break. .
To determine the impact of the modeling error on past operations with Core ;
XII, YAEC performed best estimate calculations using non-conservative assumptions. YAEC stated that its calculations indicate that a break at the 4 inch section downstream of the 4 inch check valve in a safety injection line would not have resulted in unacceptable peak clad temperatures, To correct the analysis error prior to operation with Core X111, YAEC proposed to restore power to the motor operated valves in the safety injection lines and to assume in the ECCS analysis for Core XIII, isolation of the broken safety injection line (in the 4 inch section downstream of the check valve) within 15 minutes into the accident.
To enhance the performance capability of the ECCS, YAEC had previously proposed modifications involving the addition of an injection delay feature to the ECLS accumulator subsystem. This proposal is presently under staff review in conjunction with its review of YAEC's Core XIII refueling evaluada n. A model change for the large break analysis involving an altern te definition of End of Bypass (E0BY) has also been submitted by YAEC. The staff has found this model change to be acceptable for use in the Core XIll ECCS performance analysis. YAEC l also intends to propose a model change for the small break analysis ;
involving the use of r heat transfer correlation that more accurately describes heat transfer at low flows.
With regard to YAEC's proposal to reinstate power to themotor operated valvesin the safety injection lines to permit valve closure fcr preserving accumulator inventory, the staff commented that considerable support would have to be provided to justify operator action (to identify and isolate the broken line). The staff suggested that as an alternative to <
relying on operator action, YAEC should give thorough consideration to flow l balancing by changing the flow resistances as necessary 50 that the system flows would more closely natch the ECCS performance that had previously been considered acceptable.
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At the conclusion of the meeting YAEC withdrew its initial proposal to reinstate power to the motor operated valves in the safety injection l lines and committed to the following actions for obtaining NRC j approval '
i for operation of Yankee-Rowe with Core X111.
. To provide the increased perinanent flow resistances in each safety injection line by replacement of the 4 inch check valves with a 21/2 inch check valve or by other appropriate means as detemined to be suitable and practical. Provide descriptions and bases fe r J i
the modifications.
- To proceed promptly with the planned ECCS performance verification tests which in part will provide data for determining the added flow resistances needed in the safety injection lines. .
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- To submit detailed information in support for the planned model change for the small break analysis involving the pool boiling heat transfer coefficient.
- To provide the Core X111 ECCS perforTnance analysis with the approved evaluation models and acceptable model changes. The analysis will include two large breaks and one small break with the safety injection dealy feature and the added flow restrictions in the Core XIll configuration.
YAEC also committed to submit the confirmatory Core X111 ECCS analysis for the entire break spectrum shortly af ter obtaining NRC approval for Core X111 operation.
YAEC stated that because of the anticipated heavy summer demand for electric power, startup with Core XIll is scheduled for August 1,1977.
Therefore , YAEC asked for prompt staff review of its submittals . We indicated that in order for us to be responsive, YAEC must time its submittals I of . remaining items so as to allow at least two weeks for staff review . l In this connection, we pointed out that we consider the small break rrodel l change to be the critical path item in our Core XIII review. Therefore, it is necessary for YAEC to make this submittal as soon as possible but not later than two weeks from the date of this meeting.
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L t. ,& dJ L . t z - .p Alfred Burger, Project danager Operating Reactors Branch #1 i Division of Operating Reactors
Enclosures:
- 1. List of Attendees
- 2. YAEC's Handout cc w/encis:
See next page
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MEETING WITH YANKEE ATOMIC ELECTRIC COMPANY C_0NCERNING YANKEE-ROWE Li3T OF ATTENDEES NRC A. Burger D. Haverkamp W. Lazarus K. Herring
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- 7. . v. ramal S. Rhow N. Anderson K. Parczewski P. DiBenedetto K. Jabbour D. Tondi F. Nolan R. Woodruff YAEC J. Thayer J. Consolatti W. 5:ymaczak A. Ladieu J. Chapman T. Keenan ,
J. Turnage j A. Husain R. Grube R. Shone P. Rainey M. Ebert 1
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4 Meeting Sumary for -
JUN 2 21977 Yankee Atomic Electric Company Docket NRC PDR LOCAL PDR ORB #1 Reading NRR Reading E. G. Case V. Stello K. R. Goller D. Eisenhut A. Schwencer D. Davis G. Lear R. Reid L. Shao B. Grimes W. Butler R. Baer '
Project Manager Attorney, OELD 01&E (3)
Licensing Assistant Each NRC participant T. B. Abernathy J. R. Buchanan l
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HEETI:lG AGE!:DA l
l Yankee Atomic Electric Company .
l and Nuclear Regulatory Cocmission June 17, 1977 9:00 /Ji Bethesda, Maryland NAME TD2 I. Introduction . . . . . . .. . . . . . . . . . . . R. M. Grube 9:00 - 9:10 R. P. Shone 9:10 - 9:30 II. Rowe ECCS Description . . . . . . . . . . . . . . . .
A. History B. Current Conficuration III. LOCA Analysis . . . . . . . . . . . . .. . . . . .
J. C. Turnage/ 9:30 - 10:15 A. Husain A.- Core XIII
- 1. Large Break
- 2. Small Break B. Core XII Irp.'.icati ons P. A. Rainey 10':15 - 10:30 IV. ECCS Perfor:ance Verificatien Tests . . . . . . . .
Break Restoratien of Power to Safety Injection Valves . . R. P. Shone / 10:45 - 12:15 V.
- 7. D. Baxter A. Syster. Histerv B. Philosothv ef Procosed chance
- 1. pperater Action
- 2. Single Failure-Valve Installation C. Electrical Circuitry Chances
- 1. Spuricus Valve Motion i
- 2. Keylock Switches ,
Su=sary . . . . . . . . . . . . ...... . . . . T. D. Keenan 12:15 - 12:30 VI.
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j VII. Sub-group Discussions (as needed)
VIII. NRC and/cr YAIC Caucus (as needed)
IX. Conclusiens t
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SUMMARY
I. SEQkJENCE OF EVENTS LEADING TO SHUTDOWN A.. LOCA analysis associated with Core XIII revealed certain modeling errors.
B. Reanalysis of present Core XII configuration was done to determine modeling error impact on operation.
C. Conclusion was that shutdown was warranted due to analytical uncertainties and to maximize time available for Core XIII work.
II. PROPOSED MODIFICATIONS FOR POST-CORE XII OPERATION ,
A. Analytical modifications regarding heat transfer .
correlations.
B.,
System Modifications
- 1. Restore electrical power to eight safety injection
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valves. .
- 2. Add additional circuitry to effectively preclude i
spurious valve motion. l l
- 3. Add keylock switches to essentially eliminate the possibility of operator error.
- 4. Install safety injection valves in positions upstream of check valve in each injection line
- to provide redundant isolation capability remote from postulated break location.
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III. ' BASIS OF POSITION FOR RESTORING POh R TO VALVES AND ALLOWING I .
OPERATOR ACTION A. ' The restoration of power to the safety injection valves A
essentially restores the system to its operational mode l
l prior to Core XII, with the addition of protection for: i
- 1. Spurious valve motion 1
- 2. Operator error B. The time required for operator action - 15 minutes is a reasonable time frame within which one can be expected to act, is outside the "immediate action" category, and, in our judgement, is acceptable for licensing. This is i particularly true in view of the fact that:
The need for any operator action exists only for a small break of the size in question at a very specific location.
All breaks of larger size will be adequately responded to by the system independent of operator action. ,
C. The physical separation of the valves in question from the break location, including the existence of barriers, precludes any direct impact on the valves from the LOCA.
The conclusion readhed is that the valves, op'erators and wiring remain operable for the required time interval.
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ECCS DESCRIPTION i
HISTORY ;
FEATURES OF ORIGINAL SYSTF",(1960)(SEE ATTACHED SKETCH) i
- 1. TWO LOW PRESSURES HIGH VOLUME PUMPS. l
- 2. CHARGING SYSTEM CONSISTING OF THREE 33 GPM POSITIVE ;
DISPLACEMENT PUMPS PROVIDED HIGH PRESSURE INJECTION.
- 3. BACK UP POWER PROVIDED BY TWO OUTSIDE LINES.
- 5. OPERATOR ACTION REQUIRED TO CROSS OVER CHARGING FLOW AND !
TO STRETCH OUT SI WATER INVENTORY.
EARLY MODIFICATIONS
- 1. ADDED ONE INTERMEDIATE PRESSURE PUMP IN PARALLEL WITH THE LOW PRESSURE PUMP IN 1962.
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- 2. IN 1970 THE CAPABILITY TO PROVIDE LONG TERM POST ACCIDENT RECIRCULATION WAS PROVIDED. ,THIS SYSTEM FEATURED:
A. THE CAPABILITY TO WITHSTAND A SINGLE FAILURE OF ONE PUMP OR ONE ACTIVE VALVE.
B. THE CAPA31LITY TO INCLUDE THE SHUTDOWN COOLING HEAT EXCHANGER AND CLEAN UP OF THE ECCS WATER.
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C. 0PERATOR ACTION WAS REQUIRED TO INITIATE RECIRCULATION.
CURRENT CONFIGURATION FEATURES IN 1971 A MAJOR MODIFICATION TO THE ECCS WAS MADE. THIS SYSTEM FEATURES: -
l.. REDUNDANT ON-SITE EMERGENCY DIESEL GENERATORS.
- 2. THREE 50 PERCENT PUMPING TRAINS CAPABLE OF FURNISHIllG ECC WATER FOR THE FULL RANGE OF BREAKS.
'3. PROTECTION FOR S!flGLE ACTIVE FAILURE.
4.... INJECTION FLOW COMMENCES ON RCS DEPRESSURIZAT10ft 1.E.
MOV'S ARE PASSIVE.
- 5. PRESSURIZED ACCUMULATOR. ,
- 6. OPERATIOR ACTION IS GREATLY SIMPLIFIED AND IS REQUIRED IN 1
EARLY PHASE ONLY FOR THE BREAK OF THE SI LINE ITSELF.
FOR CORE XII THE SYSTEM WAS MODIFIED TO PREVEllT SPURIOUS '
FAILURES AND OPERATOR ERROR. EARLY PHASE OPERATOR ACTION WAS ELIMINATED. IN ADDITION'LONG TERM HOT LEG INJECTION WAS PROVIDED TO PREVENT BORON PRECIPITATION.
FOR CORE XIll THE SYSTEM 15 BElf1G MODIFIED TO DELAY INJECTION DURING THE BLOWDOWN PHASE AND INCREASE FLOW RATES DURING ACCUMULATOR INJECTION.
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RESTORATION OF POWER TO SAFETY INJECTION VALVES HISTORY THE ECCS SYSTEM IN ITS PRESENT CONFIGURATION WAS ORIGINALLY DESIGNED TO PROVIDE THE CAPABILITY TO ISOLATE FLOW TO AN INDIVIDUAL RC LOOP. THIS WAS REQUIRED ONLY IN THE CASE OF A RUPTURE OF THE SI BRANCH LINE DOWNSTREAM OF THE CHECK VALVE.
THE CORE XII ECCS ANALYSIS DID NOT ASSUME' ISOLATION OF FLOW TO THE BREAK. BASED ON THE ASSUMPTION THAT ISOLATION ,
i WAS NOT ESSENTI AL, YANKEE PROPOSED TO PROTECT AGAINST OPERATOR ERROR AND SPURIOUS FAILURE BY REMOVING POWER FROM THE BRANCH LINE MOTOR OPERATED VALVES.
PROPOSED CHANGE o
YANKEE INTENDS TO ASSUME ISOLATION OF FLOW TO THE BREAK IN THE CORE XIII ANALYSIS IN THE CASE OF THE BRANCH LINE BREAK DOWNSTREAM OF THE CHECK. THEREFORE, RESTORATION OF POWER TO THE BRANCH LINE VALVES AND THE RE-RECOGNITION OF OPERATCR ACTION ARE REQUIRED.
YANKEE PROPOSES TO RESTORE FOIER TO CS-MOV-536, 537, 538, 539 AND SI-Mov-22, 23, 24, 25 WHICH WILL PROVIDE REDUNDANT CAPABILITY TO ISOLATE THE BROKEN BRANCH LINE, FROM THE CONTROL ROOM. RESTORATION OF POWER WILL PROVIDE PROTECTION AGAINST t
, , ,, , , __.,_s,~._.-- - . .. , . . _
l h
CPERATOR ERROR AND SPURIOUS FAILURE IN ACCORDANCE WITH THE INTENT'0F BTP-18.
ANALYSIS INDICATES THAT THE OPERATOR HAS 15 MINUTES TO IDENTIFY AND ISOLATE THE BROKEN BRANCH. YANKEE FEELS THAT OPERATORACTIONWITHINTHISTIMEFRAMEkSJUSTIFIEDBECAUSE:
- 1. -OPERATOR ACTION IS REQUIRED ONLY FOR'A BREAK OF THE SI LINE DOWNSTREAM OF THE CHECKS AND ..
- 2. NO OTHER SHORT TERM OPERATOR ACTION IS REQUIRED.
IN ADDIT 10N YANKEE PROPOSES TO PROVIDE IMPROVED RELIABILITY 0F THIS ISOLATION CAPABILITY BY EITHER OF THE FOLLOWING:
- 1. RELOCATE THE DOWNSTREAM VALVES OUTSIDE THE . LOOP, l.E.
1
- 2. INSTALL NEW REPLACEMENT VALVES OUTSIDE THE LOOP. ,
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. . .. 1 LOCA ANALYSES l 1
l CORE 13 e LARGE BREAK ANALYSI S
- ECC INJECTION DELAY
- ALTERNATE DEFINIT' ION FOR E0BY
- BREAK SPECTRUM STUDY j
- BURN-UP STUDY
' REFLOOD INSTAEiLITY FIX e SMALL BREAK P!ALYS1S ,
a POOL BOLLING HEAT TRANSFER
- 3REAK SPECTRUM STUDY
. - COFE 12 METHOD ,
- CORE 13 (WORST BREAK LOCATION)
. REMOVAL OF THERMAL SLEEVE RESISTANCE CORE 12 C IMPLICATIONS l l
+ RESutTS AT 57~ Arid 502 PCWER l
- l'ilS-MATCH OF CALCULATED AND COMPUTED RESULTS 'l
- IEST DATA RECUIREMENTS 0 3EST ESTIMATE .
- ASSUMPTIONS
- REsULTS CORE 13_
- - 0 CORE XIll ENVELOPING STUDY 0 EFFECT OF OPERATOR ACTION
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CORE 12 BEST ESTIMATE ASSUMPTIONS
- AVAILABILITY OF OFF-SITE POWER 0 3 LPSI & 3 HPSI AVAILABLE O CHARGING PUMPS AV,.4LABLE O MAIN COOLANT PUMPS RUNNING UNTIL CAVITATION ,
O STEAM DUMP ON HIGH SECONDARY PRESSURE UNTIL CONTAINMENT ISOLATED
+ IIO UNCERTAINTY ON AtlS DECAY CURVE
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[N }$f 39 3 DEPARTMENT OF THE ATT HNEY GENERAL
? 4 hk.,. f"f
- b) JOHN W. Mc CORM ACK GTATE OFFICE DUILDINo ? -
oN E ASH BURTON PLAC E. DO STo N 0 2108 l (j: g
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rmeis x. n etterr, July 22, 1977 AMO R N E Y DENEWAL Mr. Edson G. Case Director of Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 -
Dear Mr. Case:
We recently learned that the shutdown of the Yankee Atomic nuclear power plant in Rowe, Massachusetts, involves sub-stantially more than a normal refueling. In pursuing this matter, persons from this staff have spoken with Mr. Fred Burger, the Project Manager, and Messrs. Eldon Brunner , Burt Davis, and Bill Lazarus at N.R.C. in King of Prussia, Pennsylvania. They indic-ated their intention to send this office the available written material including meeting minutes and accident analyses reports.
This letter will confirm our request for that information.
It is my understanding, based on the conversations with .
these gentlemen, that Yankee Atomic was shut down during a
" coast-down," a few weeks before a scheduled refueling. The " prob-lem" has been described to us as an Emergency Core Cooling System condition that had not been considered in the evaluation of ECCS effectiveness. Apparently the omission was discovered during a routine re-fueling analysis.
Certain questions surrounding this incident continue to concern us. Primarily, I am interested in understanding the prob-lem at the Rowe plant from a layperson's perspective, how it hap-pened and what steps are being taken to correct it. I hope that you can clarify the situation for me by answering the following questions:
(1) What, precisely, is the nature of the situation which resulted in the plant shutdown? Is it a violation of, or nonconformity with, any N.R.C. regulations?
(2) Yankee Atomic was preparing for Core 13. Presumably, therefore, several pre-refueling analyses were prepared in the past.
Why did 'this situation go undiscovered until so recently ?
(3) What, precisely, is being undertaken to. remedy this situation? We were told that some computer tests are being done l
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-1 and that some piping is being-modified. Are the entire ECCS l and other sarety systems being reanalyzed in a search for other
" omissions?" Are other design or equipment modifications neces- j sary?
- (.4 ) Mr. Burger explained that Yankee Atomic shut down because it was the prudent thing to do but that it could have !
operated,'at leest until refueling, at a lower power. level. .
Yankee Atomic was at 68% capacity when it shut down. At what level could it have safely operated, staying within NRC regula- g tion specifications for heat rate?
(5) Will the plant be able to operate at its licensed
-power level after this problem is resolved, or will a license .
amendment be forthcoming?
. E ;ii (6) What is the relationship between the computer anal- F ysis and the piping modifications, both of which are appareatly being undertaken now? Is the computer analysis intended to test the effectiveness of the ECCS as modified? If so, how can it be performed prior to completion of the changes in the ECCS?
I thank you in advance for your attention to this -
matter.
Very truly yours, s A
LYN R. WEISS '
Assistant Attorney General Environmental Protection Division One Ashburton Place, 19th Floor Boston, Massachusetts 02108 (617) 727-2265 ERW:JK '
cc. Mr. Fred lurger 1 Project Manager Thomas Merrigan, .Esq.
Greenfield, Mass.
3 Mr. Eldon Brunner U.S. N.R.C. -
King of Prussia, Pa.
.