ML19211C166
| ML19211C166 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/21/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| References | |
| NUDOCS 8001110015 | |
| Download: ML19211C166 (12) | |
Text
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Jg UNITED STATES
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NUCLEAR REGULATORY COMMISSION 3.g E
WASHINGTON. D. C. 20555 l
v CE EMBER 21 @3 Docket No. 50-317 and 50-318 Mr. A. E. Lundvall, J r.
Vice President - Supply Baltimore Gas & Electric Company P. O. Box 1475 Baltimore, Maryland 21203
Dear Mr. Lundvall:
SUBJ ECT: AUTOMATIC INITIATION OF AUXILIARY FEEDWATER SYSTEMS AT CALVERT CLIFFS, UNITS NOS.1 AND 2 In recent correspondence, you have transmitted proposed designs, using control grade components, which would automatically initiate the auxiliary feedwater systems at your facilities upon the loss of main feedwater flow.
This submittal was in response to Short-Term Recomendation 2.1.7.a.
" Auto Initiation of the Auxiliary Feedwater System", as clarified in our letter of October 30, 1979 which was addressed to all operating nuclear power plants.
We are reviewing your proposed design against each of the seven positions stipulated in Short-Term Recomendation 2.1.7.a.
In response to this reconnendation, you have 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 steam generator.
In question is whether the change in 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 twenty (20) days after receipt of this letter (telecopied on date signed). The enclosure to this letter provides a list of questions and information you should address as appropr'ite.
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 expeditiously as possible, you should continue with the procurement of equipment and proceed with the installation to the extent possible without activating the automatic-start system or adversely affecting the manual-start AFWS.
1738 155 T8 0 01 1 10 0/5
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You are requested to propose Technical Specifications for the AFWS modifications.
Sample Technical Specifications are enclosed for your consideration.
In addition, you will need to revise nomal and emergency operating procedures as required by this modification and train the plant operations people as required by these procedu res. Particular attention to the means of controlling the bypass j
capability of the automatic AFWS turbine start signal is recommended.
i Si ncerely, A,, )
N j;3 Robert W. Reid, Chief Operating Reactors Branch #4
'Jivision of Operating Reactors
Enclosure:
Sample TS Pages cc: w/ enclosure See next page 1738 156 e
Ia:tincre Gas & Electric Company cc:
.'s es A. Biodison, Jr.
Mr. R. " : agiass, flanager
' ea.eral Counsel Quality Assurance Department Fi sna E Builaing Roor 922 C-as & Electric Building Charles Center P. O. Box :475 Bal timore, Maryland 21203 Bal ticore, "aryland 21203 George F. Trowbridge, Esquire Shaw, Pittnan, Potts and Trowbridae 1500 ft Street, N.U.
Wasnington, D. C.
20036 Mr. R. C. L. Ol son Baltinore Gas and Electric Company Room 922 - G and E Building Post Of fice Box 1475 Baltimore, Maryland 21203 fir. Leon B. Russell, Chief Engineer Calvert Cliffs Nuclear Power Plant Baltinore Gas and Electric Company Lusoy, fiaryland 20657 Bechtel Power Corporation ATTil:
Mr. J. C. Judd Chief Nuclear Engineer 15740 Shady Grove P.oad Gaitnersburg, Maryland 20760 Co-bustion Engineering, Inc.
ATTH: fir. P. W. Kruse,flanager Engineering Services Post Office Box 500 Winasor, Connecticut 06095 Calvert County Library Prince Frederick, fiaryland 206.78
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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.
Break outside containment Avaliability or loss of offsite power c.
Justify cmitting an analysis for any of the above.
- 2. Provide the time secuence of all actions and events occurring durinF each of the postulated stean line break transients.
Bese 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 centrol 9
Safety / relief valve actuation (primary 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 element (i.e., RSIV, turbine stop, turbine control, turbine bypass, etc.) reaches its new condition. De above events are to reflect the expected response of the plant and systems.
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-2 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 frem a fbel failure standpoint. Note that norral control systems should not be censidered to function if their action would be' beneficial with respect to fuel failures.
6.
Provide the followirg infortration as a function of time:
a.
Minimin DISR b.
Cladding terperature if Df0R limit is exceeded c.
Feedwter flow into faulted and nonfaulted steam generators (main and auxiliary) d.
Steam generator licuid mass, beat transfer area covered, heat transfer rate, and pressure e.
Ereak flow rate f.
Other steam release rates in secondary syster.s g.
Primary system pressure h.
Pressurizer level i.
Hot channel flow rate j.
Core inlet and outlet temperature k.
Pressurizer safety /relier 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 margin is increasing.
Note the DIER calculations inust 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 considered when calculating Df0Rs after the rods have been inserted.
If fuel darnage occurs (i.e., violation of DfBR), provide fraction of fuel that failed and offsite dose calculations. Also provide and,
justify DNB correrlations used in the analyses.
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1738 159
h.
B.
Containment Pressure Provide the following infomation to show that the containment pressure
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will be acceptable following a main steam line break.
I 1.
Review your current analysis of this event, and provide NRC with the assupmtions used during this analysis. Particular emphasis should be f
placed on describing how AFS flow was accounted for in your original r
analysis.
(Reference to previously submitted information is accep-
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table if identified as to page number and date.) Any changes in your design which would impact the conclusions of your original analysis 7i should be discussed. We are particularly concerned with design i
changes that could lead to an underestimation of the containment pressure following a MSLB inside containment.
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2.
Provide the following information for the reanalyses performed to determine 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 renerator 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 infomation 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
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this information would become available, and the time it would take the operator to complete this action. Define credit for operator action.
If temination 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 fuentifies 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 160
. 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.
1738 161
TABLE'3.3-3 (Continued)
EliGINEERED SAFETY FEATURE ACTUATION SYSTEM IllSTRUMENTATION MINIMUM TOTAL N0.
CllAtlNELS CilANNELS APPLICABLE FUtiCTI0liAL UNIT OF CilAfitiELS TO TRIP OPERABLE MODES ACTION 9.
EMERGEllCY 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.
F.. ilwa ter 4/FDW line 2/FDW line 3/FDW line 1, 2, 3, 4 B*
l~ low-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.
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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, restort 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 tae tripped channel are also placed in the tripped condition sithin 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.
1738 1(33
TABLE 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 Appiicable b.
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Level-Low gpm c.
Feedwater Flow gpm
-Low d.
Steam Generator psia psia Pressure-Low c.
Safety Injection (see Safety Injection Setpoints)
M N
U CD y
a
TABLE 3.3-5 (Continued)
ENGINEERED 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 Feedwater System '
1
- /
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3.
Steam Generator Level-Low Emergency Feedwater System 3
- /
4.
Feedwater Flow-Low Emergency Feedwater System 1
- /
i NOTE: Response time for Motor-driven Emergency Feedwater Pumps on all Safety Injection signal starts 1
- Diesel generator starting and sequence leading delays included.
- 31esel generator starting an'd sequence loading delays not included.
Offsite power available.
1738 165
TABLE 4.3-2 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CllANNEL CllANf0 CllANNEL FUNCTIONAL NODES IN W:llCil FUNCTIONAL UNIT CllECK CAliBRATI0ft TEST SURVEILLANCE REQUIRED 1.
EMERGEllCY FEEDWATER I
a.
Manual InfLiation N.A.
N.A.
H 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)
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- 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.
uco Ch 6