ML20054G698
| ML20054G698 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 06/18/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-15-01, TASK-15-1, TASK-RR LSO5-82-06-058, LSO5-82-6-58, NUDOCS 8206220218 | |
| Download: ML20054G698 (15) | |
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June 18, 1982 i
Docket flo. 50-213 LS05-82 058 Mr. W. G. Counsil, Vice President Nuclear Engineering and Operations Connecticut Yankee Atomic Power Company Post Office Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
SUBJECT:
HADDAM NECK - SEP TOPIC XV-1, INCREASE IN FEEDWATER FLOW, DECREASE IN FEEDUATER TEMPERATURE, INCREASE IN STEAM FLOW In your letter dated September 30, 1981, you submitted a safety assessment report on the above topic. The staff has reviewed your assessment and our conclusions are presented in the enclosed safety evaluation report.
As noted in the evaluation. of the increase in feedwater flow event, it is the staff's position that you should either demonstrate that the operator would have ample time to diagnose and mitigate the transient or provide an automatic trip of the main feed pumps and the reactor on high steam generator level.
The enclosed safety evaluation will be a basic input to the integrated safety assessment for your facility.
The assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed, l
Sincerely, l
Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 i
Division of Licensing
Enclosure:
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Dear Mr. Counsil:
SUBJECT:
HADDAM NF.tK - SEP TOPIC XV-1. INCREASE IN FEEDWATER FLOW,
-DECREASEtIN FEEDWATER TEMPERATURE, INCREASE IN STEAM FLOW s
In your letter dated September 30, 1981, you submitted a safe #ty assessment report on the above topic. The staff has reviewed your assessment and our conclusions are presented in the enclosed safety evaluation report.
As noted in the ev-'"ation, further information is requested concerning the increase in '
ir flow event.
The enclosed safety evaluation will be a basic input to the integrated safety assessment for your facility. The assessment may be revised in the future if your facility design is changed or if HRC criteria relating i
I to this topic are modified before the integrated assessment is completed.
Sincerely,
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Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Liernsing
Enclosure:
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OFFICIAL RECORD COPY uscm an-ass.m j unc ronu ais oo aci nncu ou
s Mr. W. G. Counsil cc Day, Berry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 i
Superintendent Haddam Neck Plant RFD #1-Post Office Box 127E East Hampton, Connecticut 06424
. Mr. Richard R. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Board of Selectmen Town Hall Haddam, Connecticut 06103 State of Connecticut 0Ffice of Policy and Management ATTH:
Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06115
'U. S. Environmental Protection Agency Region I Office ATTN:
Regional Radiation Representative JFK Federal Building Boston, Massachusetts 02203
' Resident Inspector Haddam Neck Nuclear Power Station c/o'U. S..NRC East Haddam Post Office East Haddam, Connecticut 06423 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsylvania 19406
HADDAM NECK PLANT SEP TOPIC XV-1 INCREASE IN FEEDWATER FLOW I.
INTRODUCTION An increase in feedwater flow can result from the excessive opening of the feedwater control valves, overspeed of a feedwater pump, or st'arting of a second feedwater pump at low reactor power.
The sharp increase in feedwater flow results in excess heat being removed in the steam generators.
This causes a power increase through reactivity feedback when colder primary coolant reaches the reactor core.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility.
+-
Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include safety limits whi.ch protect the integrity of the physical barriers l
which guard against the uncontrolled release of radioactivity.
l The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 10 " Reactor Design" requires that the core and associated coolant, control and protection systems be designed with appropriate margin to assure that 0001.0.0 INCREASE IN FEEDWATER FLOW
--on.
M specified acceptable fuel design limits are not exceeded during
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including the effects of anticipated operational occurrence GDC 15 " Reactor Coolant System Design" requires that-the reactor associated protection systems be designed with sufficient margin to and the design conditions of the reactor coolant pressure boundary are not during normal operation, including the effects of anticipated' oper'ationa exceeded occurrences.
GDC 26 " Reactivity Control System Redundancy and Capability" reactivity control systems be capable of reliably controlling reactivity W
changes to assure that under conditions of normal operation, includin pated operational occurrences, and with appropriate margin for malfunct r
r such as stuck rods, specified acceptable fuel design limits are not excee III.
RELATED SAFETY TOPICS y
Various other SEP topics evaluate such items as the reactor protectio The effects of single failures on safe shutdown capability are considere a
Topic VII-3.
g IV.
REVIEW GUIDELINES
'The review is conducted in accordance with SRP 15.1 1
, 15.1.2, 15.1.3, and 15.1.4.
The evaluation includes review of the analysis for the event and
. identification of the features in the plant that mitigate the consequences o the event as well as the ability of these systems to function as requi' red The extent to which operator action is required is also evaluated.
V.
EVALUATION Two cases were considered:
failure of all four feedwater control valves to the full open position at full load and startup of a second feedwater pu from 50% power.
These transients were analyzed by a detailed analog simula-tion of the reactor plant including pressurizer, reactor coolant system steam generators and reactivity effects.
References 1 and 2 indicate 0002.0.0 INCREASE IN FEEDWATER FLOW
that heat flux is slightly increasing and reactor care c:olant inlct tempera-ture slightly decreasing.
These two effects tend to compensate each other with respect to the DNB ratio.
The reactor coolant pressure increase is negligible, approximately 2% of normal reactor coolant system pressure, and the DNBR is maintained at approximately the initial steady state value of 1.73 following manual reactor trip.
For the case initiated from full power, a manual reactor trip is assumed after 3 minutes in response to the high steam generator level alarm.
For the case with startup of the second pump, manual trip is assumed in forty seconds.
Credit for operator action is not usually given before ten minutes.
- However, delaying the reactor trip to ten minutes would not result in 'a DNB ratio below 1.3 since the parameters are changing slowly and,as discussed above, tend to compensate for each other.
In addition, the reactor protection system would automatically trip the reactor if the event continued long enough that reactor parameters were approaching safety limits.
Since the heat flux decreases rapidly after the reactor trip, DNBR is not limiting for the excess feedwater incident.
Nevertheless, Haddam Neck does not have an automatic reactor or turbine trip on steam generator high level.
The licensee assumes that the transient is.
terminated by the operator who manually trips the reactor on steam generator-high level alarm.
The feedwater regulating valves are controlled in the AUTOMATIC mode by the steam generator water. level controller.
In response to increasing steam generator level, the regulating valves would be signaled to close down.
Following turbine trip, the regulating valves would normally open to provide feedwater to accomodate the loss of heat sink.
When core average temperature drops below the no-load reference temperature, the valves close to prevent overcooling.
A high steam generator level of 69% would override the above control logic and signal closure of the valves.
0003.0.0 INCREASE IN FEEDWATER FLOW l
Operator action to terminate feedwater flow by closing the feedwater isolation valves is the backup to the control system.
The staff considers ten minutes as the appropriate time frame for operator action.
If the flow increase is not terminated, the generator will begin to fill and could eventually overflow into the steam lines.
If the weight of the water resulted in rupture of the main steam line, the consequences could be more severe than the design basis steam line break.
The licensee has indicated that for a 10% increase in feedwater flow from full load it would take 30 minutes to fill the generator, assuming no trip and no operator action.
From the standpoint of generator overfill, however, the startup of a second pump may be worse since it results in,a larger increase in feed flow and a greater steam / feed mismatch.
The licensee is requested to provide a basis that, for the worst increase in f
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feed flow event, operator action at ten min'utes is sufficient to prevent possible rupture of the steam line.
VI, CONCLUSIONS We conclude that the consequences of not terminating the increase in feedwater flow transient before the steamline is flooded have not been adequately evaluated by the licensee.
If the flooding would result in a steamline rupture the ensuing reactor coolant system overcooling may be more^ severe ~
than during the worst case steamline break accident analyzed by the licensee.
~
It is our position that the licensee should provide sufficient information to demonstrate that the operator would have ample time to diagnose and mitigate the transient, otherwise, automatic trip for the main feed pump and the reactor on high steam generator level should be provided.
0004.0,0 INCREASE IN FEEDWATER FLOW
c VII.
REFEREtiCES 1.
W. G. Counsil to D. M. Crutchfield letter dated September 30, 1981.
2.
plant Design Change tio. 21 - October, 1967, Monthly Report to AEC.
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0005.0.0 If4 CREASE Iti FEEDWATER FLOW
HADDAM NECK PLANT SEP TOPIC XV-1 DECREASE IN FEEDWATER TEMPERATURE I.
INTRODUCTION A reduction in feedwater temperature can result from the loss of a feedwater heater or the accidental starting of the Auxiliary Feedwater S'ystem (AFWS).
A decrease in feedwater temperature will cause a decrease in the reactor coolant temperature, an increase in reactor power due to the negative moderator tempera-ture coef ficient, and a decrease in the reactor' coolant system and steam generator pressures.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis an.d evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, including " determination of the margins of safe.ty during normal operations and transient conditions anticipated during the life of the facility.
Section 50.36 of'10 CFR Part 50 requires the Technical Specifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 10 " Reactor Design" requires that the core and associated coolant, control and protection systems be designed with appropriate margin to assure that 1
DECREASE IN FEEDWATER TEMP.
U O
specified acceptable fuel design limits are not exceeded during normal operation, including the effects of anticipated operational occurrences.
GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational occurrences.
GDC 26 " Reactivity Control System Redundancy and Capability" r.equires that the reactivity control systems be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.
III. RELATED SAFETY TOPICS Various other SEP topics evaluate such items as the reactor protection system.
The effects of single failures on safe shutdown capability are considered under Topic VII-3.
IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.1.1, 15.1.2, 15.1.3, and t*
~~
15.1.4.
The evaluation includes review of the analysis for the event and identification of the features in the plant that mitigate the consequences of the event as l
well as the ability of these systems to function as required.
The extent to which operator action is required is also evaluated.
V.
EVALUATION The decrease in feedwater temperature event was reanalyzed in Reference 1.
The results indicate that the loss of the high pressure feedwater heaters causes the roct cevere decrease in feedwater temperature, since the loss of any of the 2
DECREASE IN FEEDWATER TEMP.
low pressure heaters would be partially compensated by the high pressure heaters.
The loss of both high pressure heaters will reduce the feedwater temperature 66 F.
The corresponding change in the reactor core coolant inlet temperature is approximately 20 F.
Inadvertent initiation of the AFWS'has an even smaller effect as it would decrease the feedwater temperature by less than 20 F, with a corresponding small decrease in reactor core coolant inlet temperature.
The DNBR is maintained very close to the steady state value of 1.73.
The consequences of the transient is bounded by the event of increase in feedwater flow.
i VI.
CONCLUSIONS
)
As part of the SEP review for Haddam Neck Plant, we have reviewed the licensee's analysis of decrease in feedwater temperature event according to the criteria of SRP Section 15.1.1, 15.1.2, 15.1.3, and 15.1.4.
We conclude, based on our evaluation, that the consequences of this event are bounded by increase in feedwater flow and the analysis for the transient are acceptable.
VII. REFERENCES 1.
W. G. Counsil to D. M. Crutchfield letter dated September 30, 1981.
1 l
l r
l 3
DECREASE IN FEEDWATER TEMP.
I t
V HADDAM NECK PLANT SEP TOPIC XV-1
~
INCREASE IN STEAM FLOW I.
INTRODUCTION An increase in steam flow may be initiated by the opening of the turbine control valves, atmospheric steam dump valve, and/or the steam bypass to condenser valves.
The increased steam flow removes more heat from the primary system resulting in a decrease in primary system temperature, steam generator pressure, and primary system pressure.
Reactor power increases as a result of the moderator temperature coefficient.
Protection against fuel damage is provided by reactor trip from high neutron flux, variable low pressure, or high steam flow.
A variable low pressure trip signal is generated whenever the pressurizer pressure is lower than a calculated variable low pressure setpoint.
The set point is continuously calculated from measurements of reactor coolant temperature and core power to prevent a DNBR of le.ss than 1.3 and a. core outlet void fraction greater than 32 percent.
The Reactor Control System for Haddam Neck Plant is designed to accommodate 10%
step load increases and 5% per minute ramp load increases without a reactor trip.
Reactor Control is based on equilibrium reactor coolant average temperature.
This temperature is controlled by a programmed reference value which is a function of turbine load or a fixed manual setpoint.
The turbine load limiter is designed.to limit the maximum load demand to 100% of rated power as protection against excess loading by operator error or by turbine speed control malfunction.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the 0001.0.0 INCREASE IN STEAM FLOW
objective of assessing the risk to public health and safety resulting from operation of the facility, including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility.
Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Part 50) est,ablish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 10 " Reactor Design" requires that the core and associated coolant, control and protection systems be designed with appropr'iate margin to assure that specified acceptable fuel design limits are not exceeded during normal operation, including the effects of anticipated operational occurrence.
GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be designed Nith sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded
.during normal operation, including the effects of anticipated operational ~
occurrences.
a GDC 26 " Reactivity Control System Redundancy and Capability" requires that the reactivity control systems be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.
III. RELATED SAFETY TOPICS Various other SEP topics evaluate such items as the reactor protection system.
The effects of single failures on safe shutdown capability are considered under Topic VII-3.
0002.0.0 INCREASE IN STEAM FLOW
IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.1.1, 15.1.2, 15.1.3, and 15.1.4.
V.
EVALUATION The full opening of all steam bypass valves is the limiting case for the increase in steam flow incident.
The event was reanalyzed in References 1 and 2.
The analysis was performed from 80% power as that is the h,ighest power level at which the opening of all bypass valves would not cause actuation of the steam line isolation valves which would result in an immediate reactor trip.
A 35% steam flow increase, i.e., up to 115% of rated flow, was used with the maximum negative moderator temperature coefficient and a minimum negative doppler coefficient.
The results of the analysis indicated that the' minimum DNBR calculated was 1.92.
The reactor stabilized at approximately 99% power and no increase in the primary pressure was predicted.
VI.
CONCLUSION
~
As part of the SEP review for Haddam Neck Plant, we have reviewed the licensee's analysis of the increase in steam flow event according to the criteria of SRP Sections 15.1.1, 15.1.'2, 15.1.3, and 15.1.4.
We conclude that the initial conditions are acceptable and, since the peak pressures and minimum DNBR reached during the transient are within the acceptable limits of the SRP, the plant is adequately protected from this event.
VII. REFERENCES 1.
Plant Design Change No. 21 - Haddam Neck Plant, October, 1967, monthly report to AEC.
i 2.
Description and Safety Analysis Including the Effects of Fuel Densification of the Connecticut Yankee Reactor, Cycle V, November 1973.
0003.0.0 INCREASE I'N STEAM FLOW
3.
W. G. Counsil to D. M. Crutchfield, letter dated September 30, 1981.
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0004.0.0 INCREASE IN STEAM FLOW