ML19211C160
| ML19211C160 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck, Millstone File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 12/21/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-10, TASK-RR TAC-42410, NUDOCS 8001110004 | |
| Download: ML19211C160 (13) | |
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i UNITED STATES g
NUCLEAR REGULATORY COMMISSION y
E WASHINGTON, D. C. 20555 a
49.....,o "ECEVBER 2 1 M Cocket No. 50-336 and 50-213
!'r. W. G. Counsil, Vice President Nuclear Engineering & Operations
!iortheast Nuclear Energy Company Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, Connecticut 06101 l
Cear Mr. Counsil:
i
SUBJECT:
AUTOMATIC :MTIATION OF AUXILIAY FEEDWATER SYSTEMS AT HADDAM NECX AND MILLSTONE UNIT NC. 2
- n recent corresponderce, you have transritted proposed designs, using 7
control grade conponents, which would autoratically initiate the auxiliary feedwater systers at your facilities upon the loss of main feedwater flow.
This submittal was in response to Short-Tem Recommendation 2.1.7.a, i'
' Auto Initiation of the Auxiliary Feedwater System", as clari' led in i
cur letter of October 30, 1979 which was addressed to all operating nuclear power plants.
'We are reviewing your :roposed design against each of the seven positions stipulated in Short-Tem Recommendation 2.1.7.a.
In response to this recommendation, you nave raised the issue of the applicability of current analysis of a main steam line break or main feedwater line break assuming early initiation of auxiliary feedwater flow with a failure to limit flow to the affected stean generator.
In question is whether the change in j
assumptions would increase the calculated containment pressure or the likelihood of return to power.
These questions are believed applicable for either manual or automatic initiation of the auxiliary feedwater system. You are requested to resolve this concern by submitting an analysis within twen+y (20) days after receipt of this letter (telecopied on date signed). The enclosure to this letter provides a list of questions and i~nformation you should address as appropriate.
As a result of this concern and pursuant to our letter of October 30, 1979, you should not implement automatically initiated AFWS flow until we have completed our review and issue an approval.
However, to resolve this matter as expecitiously as possible, you should continue with the
- rocurement of equipnent and proceed with the installation to the extent
- ossible without activating the i.. 2nat.ic-start system or adversely affecting the manual-start AFWS.
1738 119 8 0 01110 Crop
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_ t You are requested o pro:ose Technical Specifications for the AFWS modifications.
i Sample Technical Soe:ifications are enclosed for your consideration.
In addition, you will need to revise romal and emergency operating procedures as recuired by this modification and train the plant operations people as required by these procedures. Particu:ar attention to the means of controlling the bypass capability of the au ocatic AFWS turbine start signal is recomended.
Si ncerely, l
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Robert W. Reid, Chief Operating Reactors Branch !4 h
Division of Operating Reactors I
Enclosure:
Sample TS Pages cc: w/ encl osure r
See next page L
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1738 120 ee I
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Northeast Nuclear Energy Company cc:
William H. Cuddy, Esquire Day, Berry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 Waterford Public Library Rope Ferry Road, Route 156 Waterford, Connecticut 06385 Northeast Nuclear Energy Company ATTN:
Superintendent Millstone Plant Post Office Box 128 tlaterford, Connecticut 06385 Northeast utilities service Company ATTN:
Mr. James R. Hirre1 wright Nuclear Engineering and Operations P. O. Box 270 Hartford, Connecticut 06101 Anthony Z. Roisman, Esq.
Natural Resources Defense Council 917 15th Street, N.W.
Washington, D.C.
20005 Mr. John T. Shedlosky Nuclear Regulatory Commission, Region I Office of Inspection and Enforcement 631 Park Avenue King of Prussia, Pennsylvania 19406 1738 12i i
Mr. W. G. Counsil CC Day, Berry & Howard U. S. Environmental Protection Counselors at Law Agency One Constitution Plaza Region I Office Hartford, Connecticut 06103 ATTN:
EIS C'00RDINATOR JFK Federal Building Superintendent Boston, Massachusetts 02203 Haddam Neck Plant RFD #1 Post Office Box 127E East Hampton, Connecticut 06424 Mr. James R. Himmelwright Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Russell Library 119 Broad Street Middletown, Connecticut 06457 Board of Selectmen Town Hall Haddam, Connecticut 06103 Connecticut Energy Agency ATTN: Assistant Director Research and Policy Development Department of Planning and Energy Policy 20 Grand Street Hartford, Connecticut 06106 Director, Technical Assessment Division Office of Radiation Programs (AW-459)
U. S. Environmental Protection.
1738 122 Agency Crystal Mall #2 Arlington, Virginia 20460
Enclosure REQUEST FOR INFORMATION AUTOMATIC INITIATION OF THE AFWS AFFECT ON MAIN STEAM LINE BREAK ACCIDENT ANALYSIS A.
Return to Power 1.
Provide the results of analyses of main steam line breaks that are the most limiting with respect to fuel failure resulting from return to power. Analyses should be presented covering:
a.
Break inside containment b.
Ereak outside containment Avaliability or loss of offsite powcr c.
Justify caitting on analysis for any of the above.
- 2. Provide the time secuence of all actions and events occurrinF during each of the postulated steam line break transients.
These events and actions should include:
a.
Reactor scram b.
Turbine trip c.
Steam line isolation d.
Feedwater isolation e.
ECCS actuation f.
Auxiliary feedwater actuation and control 9
Safety /relier valve actuation (prirary and secondary systems) h.
Operator actions (define credit for operator action) 1.
Initiation of onsite power (if required).
3..
For each of the above, identify the initiating signal, the protection system that initiates the action, and the extent of the action ending with the time the elenent (i.e., MSIV, turbine stop, turbine control, turbine bypass, etc.) reaches its new condition. The above events are to reflect the expected response of the plant and systems.
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4.
Identify and justify any equipment that does not meet Regulatory Guides and IEEE-279 requirements.
5.
Provide a list of potential single failures that could affect each of the above actions and show how the analyses presented consider the worst single failures from a fuel failure standpoint. Note that norral control systems should not be considered to function if their action would bi beneficial with respect to fuel failures.
6.
Provide the followirg information as a function of time:
a.
Minimin DNBR b.
Cladding temperature if DNBR lbnit is exceeded c.
Feedwatae flow into faulted and nonfaulted steam generators (main and auxiliary) d.
Steam generator liquid mass, beat transfer area covered, heat transfer rate, and pressure e.
Break flow rate f.
Other steam release rates in secondary systems 6
g.
Primary system pressure h.
Pressurizer level i.
Hot channel flow rate j.
Core inlet and outlet temperature k.
Pressurizer safety / relief valve flow rate 1.
ECCS flow rate.
The analysis should be carried out until the effects of delayed neutrons and moderator feedback have turned around and the suberiticality marFin is increasing.
Note the DNBR calculations must reflect the initial plant perturbations due to moderator and ' pressure ~ decrease and loss of offsite power (if appropriate).
Also discuss how the effects of a stuck rod are censidered when calculating DNPRs afte~r the rods have been inserted.
If fuel damage occurs (i.e., violation of DNPR), provide fraction of fuel that failed and offsite dose calculations. Also provide and,
justify DNB correrlations used in the analyses.
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. B.
Containment Pressure Provide the following infomation to show that the containment pressure will be acceptable following a main steam line break.
1.
Review your current analysis of this event, and provide NRC with the assupmtions used during this analysis. Particular emphasis should be placed on describing how AFS flow was accounted for in your original analysis.
(Reference to previously submitted information is accep-table if identified as to page number and date.) Any changes in your design which would impact the conclusions of your original analysis should be discussed. We are particularly concerned with design changes that could lead to an underestimation of the containment pressure following a MSLB inside containment.
2.
Provide the following information for the reanalyses performed to detemine the maximum containment pressure for a spectrum of postulated main steam line breaks for various reactor power levels for the pro-posed AFS design.
a.
Specify the AFS flow rate that was used in your original containment pressurization analyses. Provide the basis for'this assumed flow rate.
b.
Provide the rated flow rate, the run out flow rate, and the pump head capacity curve for your AFS design.
c.
Provide the time span over which it was assumed in your original analysis that AFS was added to the affected steam generator following a MSLB inside containment.
d.
Discuss the design provisions in the AFS used to terminate the AFS flow to the affected steam generator.
If operator action is required to perform this function., discuss the information that will be available to the operator to alert him of the need to isolate the auxiliary feedwater to the affected steam generator, the time when this information would become available, and the time it would take the operator to complete this action. Define credit for operator action.
If termination of AFS flow is dependent on automatic action, describe the basic operation of the auto-isolation system. Describe the failure modes'of the system. Describe any annunciation devices associated with the system.
e.
Provide the single active failure analyses which specifically identifies those safety grade systems and components relied upon to limit the mass and energy release and the containment pressure response. The single failure analysis should include, but not necessarily be limited to:
partial loss of containment cooling systems and failure of the AFS isolation valve to close.
f.
For the single active failure case which results in the maximum containment atmosphere pressure, provided a chronology of events.
Graphically, show the containment atmosphere pressure as a function of time for at least 30 minutes following the accident.
For this 1738 125
. case, assume the AFS flow to the broken loop steam generator to be at the pump run out flow (if a run out control system is not part of the current design) for the entire transient if no automatic isolation to auxiliary feedwater is part of the current design.
g.
For the case identified in (f) above, provide the mass and energy release data in tabular form. Discuss and justify the assumptions made regarding the time at which active containment heat removal systems become effective.
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TABLE'3.3-3 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION HINIMUM TOTAL NO.
CilANNELS CilANNELS APPLICABLE FUNCT10fiAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION 9.
EMERGENCY.FEEDWATER a.
Manual 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 A
per FDW line per FDW line per FDW line b.
Steam Generator 4/SG 2/SG 3/SG 1,2,3,4 B*
Level-Low c.
Fe.dwater 4/FDW line 2/FDW line 3/FDW line 1,2,3,4 B*
Ilow-Low d.- Steam Generator 4/SG 2/SG 3/SG 1,2,3,4 B*
Pressure-Low e.
Safety Injection (See Safety Injection initiating functions and requirements)
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- The provisions of Specification 3.0.4 are not applicable, u
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3.0.4 Entry into an OPERATIONAL MODE or other specified applicability condition shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with.ACTIO.N statements.
ACTION STATEMENTS ACTION A With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and.in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
ACTION B With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:
a.
The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
b.
All functional units receiving an input from t$1e-tripped channel are also placed in the tripped condition within~1 hour.
c.
The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.2.1.
I738 l28 F
TADLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES FUNCTIONAL UNIT TRIP VALUE ALLOWABLE VALUES.
9.
EMERGENCY FEEDWATER a.
Manual Not Applicable Not Applicable b.
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Level-Low gpm c.
Feedwater Flow gpm
-Low d.
Steam Generator psia psia Pressure-Low e.
Safety injection (see Safety Injection Setpoints)
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TABLE 3.3-5 (Continued)
E'iGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1.
Manual Emergency Feedwater System Not Applicable 2.
Steam Generator Pressure-Low Emergency Feer.' water System '
1
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3.
Steam Generator Level-Low Emergency Feedwater System 5
- /
4.
'Feedwater Flow-Low Emergency Feedwater System 1
- /
NOTE: Response tire fcr Motor-driven Emergency Feedwater Pumps on all Safety Injectior. signal starts 1
- Diesel generator starting 'and sequence loading delays included.
" Diesel generator starting and sequence leading delays not included.
Offsite oower available.
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TABLE 4.3-2 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CllANNEL CllANNEL CilANNEL FUNCTIONAL MODES IN WllCil FUNCTIONAL UNIT CllECK CALIBRATION TEST SURVEILLANCE REQUIRED 1.
EMERGENCY.FEEDWATER I*)
a.
Manual Initiation N.A.
N.A.
M 1, 2, 3, 4 b.
R H
1,2,3,4 Level-Low c.
R H
1,2,3,4 Flow-Low d.
R H
1, 2, 3, 4 Pressure-Low e.
Safety Injection (See Safety Injection surveillance requirements)
- Manual actuation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual safeguards actuation shall receive a CilANNEL FUNCTIONAL TEST at least. Once per 31 days.
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