ML20053B202
| ML20053B202 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 05/25/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-15-07, TASK-15-7, TASK-RR LSO5-82-05-060, LSO5-82-5-60, NUDOCS 8205280167 | |
| Download: ML20053B202 (12) | |
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May 25,1982 1
Docket flo. 50-213 LS05-82 060 Mr. W. G. Counsil. Vice President fluclear Engineering and Operations Connecticut Yankee Atomic Power Company Post Office Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
SUBJECT:
HADDA!1 NECK - SEP TOPIC XV-7, REACTOR COOLANT PUMP ROTOR SEIZURE AND REACTOR COOLANT PUMP SHAFT BREAK In your letter dated September 30, 1981, you submitted a safety assessment report on the above topic. Your letter of March 1,1982, provided additio ial infomation. The staff has reytewed your assessment and our conclusions are presented in the enclosed safety ovaluation report.
As noted in the evaluation, reanalysis is required to show conformance with current criteria. The need to perform this analysis will be addressed in the integrated safety assessment.
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.
Sincerely, higinc101 :0 22tIf Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
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NRC FORM 318 (10-80) NRCM Cao OFF1C1AL RECORD COPY usogo; mi-m-,eo
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t Mr. W. G. Counsil
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cc Day, Berry & Howard Counselors at Law One Constitution Plaza
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Hartford, Connecticut 06103
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Superintendent Haddam Neck Plant RFD #1'-
I Post Office Box 127E East Hampton, Connecticut 06424
. Mr. Richard R. Laudenat.
s 3 Manager, Generation Facilities Licensings Northeast Utilities Service Company P. O. Box 270 s
Hartford, Connecticut 06101 4
1 Board of Selectmen
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Town Hall
'Haddam, Connecticut 06103 State of Connecticut Office of Policy and Management ATTN: - Under Secr.etary Energy Division 80 Washington Street l
Hartford, Connecticut 06115
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.U. S. Environmental Protection Agency '
Region I Office
' ATTN:
Regional Radiation Representative
.JFK Federal Building.
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Boston, Massachusetts 02203-1 Resident Inspector Haddam Neck Nuclear Power Station.
c/o'U. S..NRC East Haddam Post Office
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. East Haddam, Connecticut 06423 Ronald C. Haynes, Regional Administrator s
Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 s
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SEP TCPIC XV-7 (a)
HADDAM NECK PLANT
SUBJECT:
LOSS OF FORCED REACTOR COOLANT FLOW INCLUDING TRIP OF PUMP MOTOR AND FLOW CONTROLLER MALFUNCTIONS I.
Introduction i
A' decrease in reactor coolant flow occurring while the plant is at power could result in a degradation of core heat transfer.
A resulting increase in fuel temperature and accompahying' fuel damage could then result if specified acceptable fuel damage limits are exceeded during the transient.
A nu.T6er of.
transients that are expect'ed to occur with moderate frequency and that result in a decrease in forced reactor coolant flow rate are addressed in SRP 15.3.1 and SRP 15.3.2.
II.
Review Criteria
~Section 50.34 of 10 CFR Part 50 requ-ires'that each applicant for an operating license provide an analysis and eyaluation of the design and perfor5ance of structures, systems, and componcots.of facility With the objectives of assessing the risk to public health and safety.resulting from operation of the facility. The loss of forced reactor coolant flow is one of the postulated transients used to evaluate the adequacy of these stru'ctures, systems and components with respect to the public health and safety.
Section 50.36,of 10 CFR Part 50 requires the Technical Specifications to include safety limits which protects the integrity of the physical barr'iers which guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum The requirements for the principal design criteria for water-cooled reactors.
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staff acceptance criteria are based on meeting the relevant requirements of the following regulations:
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A.
Geiien al Design Ciiterion 10 (Ref.1), as it relates to the reector coolnnt system being c'esigr.ed with appropriate cargin to assure that specified acceptable fuel design limits are not exceeded during normal operations including anticipated operational occurrences.
B.
General Design Criterion 15 (Ref. 2), as it relates to the reactor coolant system and its associated auxiliaries being designed with appropriate margin to assure that the pressure boundary will not be breeched during normal operations including anticipated operational occurrences.
C.
General Design Criterion 26 (Ref. 3,) as it r' elates to the reliable control of reactivity changes to assure that specified acceptable fuel design limits are not exceeded, including anticipated operational occurrences.
This is accomplished by assuring that appropriate margin for.
malfunctions, such as stuck rods, are accounted for.
The specific criteria necessary to meet the relevant requirements of GDC 10,15 and 26 for incidents of moderate frequency are:
Pressure in the reactor coolant and main steam systems should be a
maintained below 110% of the design values.
b.
Fuel cladding integrity shall be maintained by ensuring that the minimum DNBR remains above the 95/95 DN8R limit for PWRs and the CPR remains above the MCPR safety limit for BWRs based on acceptable correlations (see SRP-Section 4.4).
T c.
An incident of moderate frequency should not generate a more serious plant condition without other faults occurring independently.
l d.
An incident of moderate freo,uency in ccabination with any single component I
failure, or single operator error, shall be c' nsidered and is an event for o
1 which an estimate of the number of potential fuel failures shall be provided for radiological. dose calculations.
For such accidents, the number of fuel failures assumed must be equal to the number of all rods 2
for..hich the C!dR or CPR falls below these values cited 2.bme for c1bdding integrity unless it can be shown, based on an acceptable fuel dtr. age nedel (see SRP Section 4.2), that fewer failures occur.
There shall be no loss of function of any' fission product barrier other than the fuel cladding.
III. Related Safety Topics Various other SEP topics evaluate such items as the reactor protection system.
The ' effects of single failure on safe shutdown capability are considered under Topic VIII-3.
IV.
Review Guidelines The review is conducted ia accordance with SRP Sections 15.3.1 and 15.3.2.
Tbe evaluation includes reviews of the analysis for the event and identification of the f.eatures in the plant that mitigate the consequences of the event as well as the ability of these systems to function as required.
The extent to which operator action is required is also evaluated.
Deviations from the criteria specified in the Standard Review Plan are identified.
V.
Evaluation
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The licensee, in letters dated September 30, 1981, and March 1, 1982, and l
Section 10.3.2 of the FSAR, provided the results of an analysis for the subject
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t topic.
The analysis indicated that a loss-of-coolant flow event can result from a riechanical or electrical failure in one or more reactor coolant pumps, 7
or from a fault in the power supply to these pumps.
If the reactor is at power at the time of the incident, the immediate effect of loss-of-coolant flow is a rapid increase in coolant temperature.
This increase could result in departure from nucleate boiling (DNB) with subsequent fuel failure if the reactor is not tripped promptly. [TheHaddamNeckdesignincludesthefollowingtripcircuits which provide the necessary protection against this event:
(1) undervoltage trips on each of the two RCP power supply buses, (2) individual trips on each j
of the four pump circuit breakers; and (3) individual low flow trips activated 3
by a dif ferer.tial pressuie 6cr oss the str ax. 4-r.erators in osch of the four reactor coolant loops.
The results of the licensee's analysis indicate that a postulated trip of all four pumps during four loop operation is the most
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limiting case with respect to a decrease in thermal margin ~.
For this event, a minimum D!GR of 1.3 is reached approximately four seconds after initiation of the transient.
The licensee has not'provided the results of an analysis to demonstrate that the peak reactor coolant pressure is within the allowable limit (110% of the design' pressure) during the postulated event.
However, in a letter dated March 1, 1982, the licensee has provided a qualitative comparison between certain plant design parameters for the Haddam ' Neck Plant and the Comanche Peak Steam Electric Station (CPSES).
The Haddam Neck Plant is designed with higher pressurizer relief capacity, higher RCS volume to power ratio, and icwer
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nominal operating pressure than the design of the CPSES.
Therefore, the results for this transient at CPSES (a peak pressure of less than 2550 psia) adequately bound the results for the Haddam Neck Plant.
The.)icensee has not provided the results of an analysis for this event in combination with.a single failure.
Howev.er, we could not identify any single failure which will lead to unacceptable results.
VI.
Conclusions The staff concludes that the Haddam Neck Plant design with regard to loss or decrease in forced reactor coolant flow is acceptable and meets the relevant requirements of General Design Criteria 10, 15, and 26.
This conclusion is based on the following:
The applicant has met the requirements of GDC 10 and 26 with respect to 1.
demonstrating that the specifibd acceptable fuel design limits.are not exceeded for this event.
These requirements have been met since the results of the analysis showed that the thermal margin limits (DNBR) are satisfied.
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2.
The applicent has met t!.c require <.ent of GDC 15 with reyect to de::onstrating that the reactor coolant pressure boundary limits have not been exceeded for this event.
This requirement has been met since the analysis has shown that the maximum pressure of the reactor coolant and main stcam systems do not exceed 110% of the design pr' essure.
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3.
The applicant has met the requirements of GDC 26 with respect to the capability of the reactivity control system to provide adequate control of reactivity during the event while including appropriate margin for stuck
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rods since the specified acceptable fuel design limits werenot exceede d.
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v SLP 1GPIC XV-7 (b)
HADDAM NECK PLANT
SUBJECT:
REACTOR COOLANT PUMP ROTOR SEIZURE AND REACTOR COOLANT PUMP SHAFT BREAK I.
Introduction The events postulated are an iristantaneous seizure of the rotor or break of t'he shaft of a reactor coolant pump.
Flow through the affected loop is rapidly reduced.
The sudden decrease in core toolant flow while the reactor is at power results in a degradation of core heat transfer which could result in fuel damage.
The initial rate of reduction of coolant flow is greater forr the rotor seizure event.
However, the shaft break event permits a greater..
reverse flow through the affected loop later during the transient and, therefore, results in a lower core flow rate later in time.
This topic is intended to cover both of these accidents.
.. II.
Review Criteria
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Section 50.34 of 10 CFR Part 50 requires that each applicant for an operating license provide an analysis and evaluation'5f 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
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facility..The r,eactor coolant pump rotor seizure and reactor coolant pump shaft break are two of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health 1
and safety.
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.
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The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the-principal design criteria for water-cooled reactors.
1
a GDC 27 "Cc..bined Reactivity Centrol System C.; ability," requires (Imt ti.e reactivity control systems, in conjunction with poison addition by the emergency core cooling system, has the capability to reliably ccntrol
' reactivity changes to assure that under postulated accident conditions, and with appropriate margin for stuck rods, the capability to cool the core is maintained.
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GDC 28 " Reactivity Limits" requires that the reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can n'either (1) result in da.nage to the reactor coolant pressur'e boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core.
GDC 31 " Fracture Prevention of Reactor Coolant Pressure Boundary" requires that the boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fr.actures is minimized.
10 CFR Part 100.11 provides dose guidelines'for r'eactor siting against which calculated accident dose consequences may,be compared.
III. Related Safety Topics Various other SEP topics evaluate such items as the reactor protection 7
system.
The effects of single failure on safe shutdown capability are considered under Topic VII-3.
IV.
Review Guidelines The review is conducted in accordance with SRP 15.3.3, and 15.3.4.
The evaluation includes review of the analysis for the event and identification of the features in the p.lant..that mi.tigate the consequences of the event as well as O
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the :-bility of these systems to functicn as required.
The extent to &hich cptrator action is required is also evaluated.
Deviations from the criteria specified in the Standard Review Plan are identified.
V.
Evaluation The licensee has not provided adequate information to demonstrate that the
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results of this postulated accident scenario are acceptable in accordance with the staff criteria established in SRP 15.3.3/15.3.'4.
In response to staff
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requests for additional information, the licensee in a letter dated March 1, 1982 (ref. 3) has stated that the reactor coolant pump rotor seizure and shaft break events are not in the design b. asis of the Haddam Neck Plant, and therefore, specific information requested by the staff (Ref. 2) is not available.
However, the licensee has provided a qualitative comparison between the Haddam Neck design and the other plant designs which are currently,
under staff review to justify the acceptability of the Haddam Neck design.
With regard to the peak pressure during this postulated accident, the licensee has provided a comparison between certain plant design parameters for the Haddam_ Neck Plant and the Comancha Peak Steam Electric Station (CPSES).
The Haddam Neck Plant is designed with higher pressurizer relief capacity, higher RCS volume to power ratio, and lower nominal ope-r-ating pressure than the design of CPSES.
The CPSES results (peak pressure of 2590 psia) adequately bounds the results for the Haddam Neck Plant.
As a result of a qualitative analysis, the licensee has identified that the results of the shaft break event are bounded by the locked rotor accident with respect to thermal margin.
The licensee has indicated that based on locked T
rotor flow curves for similarly designed plants, a low flow trip would be exp'ected to occur at approximately one second following the rotor seizure, and a limited number of fuel pins may experience DNB.
However, the licensee asserts that exposure of a fuel pin to DNB conditions for several seconds will not result in fuel failure.
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,e staf f does rut eg (e with the licensee's fuel failure criteria.
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will assume fuel f ailure for all fuel rods that experience DNBR below the 95/95 D:GR limit.
Other mechanistic methods may be acceptable, but new positions recommending different cr.iteri'a should address cladding temperature, pressure, time duration, oxidation, and embrittlement.
.9 The licensee has not addressed the effects of the postulated loss of offsite power following reactor trip / turbine trip and a limiting single active failure during the transient.
Also the amount of fuel failure and radiological consequences for this accident scenario are not discussed in this submittal.
l V'I.
Conclusions The staff, concludes that the consequences of a postulated reactor coolant pump rotor seizur.e or broken shaft event meet the requirements set forth in the
. General Design Criteria 31 with respect to integrity of the primary system boundary to withstand the postulated accident.
This conclusion is based on the finding that the peak reactor coolant pressure does not exceed 110% of the reactor coolant:.syttem design pressure.
Based on the staff evaluation described in Section V. a.bove., we cannot' conclude that the Haddsm Neck Plant meets the requirements of GDC 27, 28 and 10 CFR 100 if analyzed in accordance with SRP sections 15.33$and 15.3.4.
The need to j
perform an analysis of this event using appropriate SRP assumptions and assuming fuel failures consistent'with an approved fuel failure model will be addressed in the integrated assessment.
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y REFEREf1CES l
I 1.
W.
G.
Counsit letter to D.
M.
Crutchfield, dated September 30, 1981 2.
D.
M..Crutchfield letter t o W.
G.
Counsite dated December 10, 1981 3.
W.
G.
Counsit letter to D.
M.
Crutchfield, dated March 1r 1982" 8
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