ML19211C173
| ML19211C173 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/21/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | William Jones OMAHA PUBLIC POWER DISTRICT |
| References | |
| NUDOCS 8001110025 | |
| Download: ML19211C173 (12) | |
Text
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NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20555 CECEYBER 2 1 fd C ocket No. 50-285 Fr. W. C. J ones Civision Manager, Production Operations Craha Public Pwer District 1523 Harney Street Caaha, Nebraska 68102 Cear Mr. Jones:
SUBJECT:
AUTOMATIC INITIATION OF AUXILIARY FEEDWATER SYSTEMS AT FORT CALHOUN STATION, UNIT N0.1 In recent communications, your staff has indicated that a proposed design, using control grade components, which would automatically initiate the auxiliary feedwater systems at your faciTity upon the loss of main feedwater flow will be submitted in the near future. This submittal was in response to Short-Term Recommendation 2.1.7.a, " Auto Initiation of the Auxiliary Feedwater Systen", as clarified in our letter October 30, 1979 which was addressed to all operating nuclear power plants.
L' ke will review your proposed design against each of the seven positions stipulated in Short-Term Recomendation 2.1.7.a.
In response to this recocriendation, some licensees have raised the issue of the applicability of current analysis of a main steam line break or main feedwater line briak asseming es.-ly initiation of auxiliary feedwater flow with a failure to limit ilow to the affected steam generator.
In question is whether the change in assurations would increase the calculated containment pressure or the itkelihood 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 (telet.opied on date signed). The enclosure to this letter provides a list of questions and information you should address as appropriate.
As a result of this concern and pursuant to our letter of October 30, 1979, you should not implement autocatically initiated AFWS flow until we have completed our review and issued 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 168 800111 M 6
i You are recuested to propose Technical Specifications for the AFWS modifications.
Sample Technical Specifications are enclosed for your consideration.
In addition, you will need to revise normal and emergency operating procedures as required by this modification and train the plant operations. people as required by these procedures. Particular attention to the means of controlling the bypas capability of the automatic AFWS turbine start signal is recomended.
Sincerely,
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Robert W. Reid, Chief Operating Reactors Branch !4 Division of Operating Reactors
Enclosure:
Sample TS Pages cc: w/ enclosure See next page 1738 169
Omaha Public Power District ccw/ enclosure (s~):
Margaret R. A. Paradis LeBoeuf Lamb,. Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, D. C.
20036 Blair Public Library 1665 Lincoln Street Blair, Nebraska 68008 A
e 1738 170 e
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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.
Ereak inside containment b.
Ereak outside containment Availability or loss of offsite power c.
Justify omitting an analysis for any of the above.
- 2. Provide the ti:re secuence of all actions and events occurring durir.g 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 / relief valve actuation (priracy and secondary systems) h.
Operator actions (define credit for operator action) 1.
Initiation of onsite power (if required).
3.
Fer each of the above,' identify the initiating signal, the pretaci. ion system that initiates the action, and the extent of the action ending with the time the element (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.
1738 171 4.
Identify and justify any equipment that does not meet Regulatory Guides and IEEE-279 requirements.
S.
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. tbte that norral control systems should not be considered to function if their action would be beneficial with respect to fuel failures.
6.
Provide the following information as a function of time:
a.
tiinimin DNBR b.
Cladding temperature if DtIBR limit is exceeded c.
Feedwater flow into faulted and nonfaulted steam generators (main and auxiliary) d.
Steam generator liquid mass, heat transfer area covered, heat transfer rate, and pressure e.
Ereak flow rate f.
Other steam release rates in secondary systems g.
Primary system pressure.
h.
Fressurizer 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 iintil the effects of delayed neutrons and moderator feedback have turned around and the suberiticality margin 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 considered when calculating DNPRs after the rods have been inserted.
If fuel damage occurs (i.e., violation of DtIBR), provide fraction of fuel that failed and offsite dose calculations. Also provide and justify DNB correrlations used in the analyses.
1738 172 B.
Containment Pressure Provide the following infonnation 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 determine the maximum containment pressure for a spectrum of postulated main steam line areaks 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 infonnation 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 173
4-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.
17138 174 P
TABLE'3.3-3 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION HINIMUM TOTAL N0.
CilANNELS ClANNELS APPLICABLE FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE MODES ACTION 9.
tMERGENCY FEEDWATER a.
Manual 2 sets of 2 1 set of 2 2 sets of 2 1,2,3,4 A
per FDW line per F0W line per FDW line b.
Steam Generator 4/SG 2/SG 3/SG 1, 2, 3, 4 D*
Level-Low c.
Fe.alwa 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 D*
l'ressure-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, 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 />.
With the number of OPERABLE channels one less than the Total ACTION B 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 the-tripped channel are also 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 />.
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 176
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TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES FUf[CTIONAL UNIT TRIP VALUE ALLOWABLE VALUES 9.
EMERGENCY FEEDWATER a.
Manual Not Applicable Not Applicable b.
Steam Generator Level-Low c.
Feedwater Flow gpm gpm
-Low Pressure-Low psia psia d.
Safety Injection (see Safety Injection Setpoints)
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TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1.
Manual Not Applicable Emergen~ y Feedwater System c
2.
Steam Generator Pressure-Low Emergency Feedwater System 1
- /
3.
Steam Generator Level-Low Emergency Feedwater System 1
- /
4.
'Feedwater Flow-Low Emergency Feedwater System 1
- /
NOTE: Response time for Motor-driven Emergency Feedwater Pumps on all Safety Injection signal starts 1
- Diesel generator starting and sequence loading delays included.
- Jiesel generator starting and secuence leading delays not included.
Offsite power available.
1738 178
4 TABLE 4.3-2 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CllANNEL CilANNEL CilANNEL l'UNCTIONAL MODES IN WillCil 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 l~ low-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, U
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