ML20054L507
| ML20054L507 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 06/29/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| References | |
| TASK-15-07, TASK-15-7, TASK-RR LSO5-82-06-121, LSO5-82-6-121, NUDOCS 8207080190 | |
| Download: ML20054L507 (12) | |
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June 29,1982 Docket No. 50-29 0505-82-06-121-Mr. James A. K&y Senior Engineer - Licensing Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701
Dear fir. Kay:
SUBJECT:
YANKEE - SEP TOPIC XV-7, LOSS OF FORCED COOLANT FLOW, REACTOR COOLANT PUMP ROTOR SEIZURE, REACTOR COOLANT PUMP SHAFT BREAK By letter dated June 30, 1981, you submitted a topic assessment on the above topic.
In response to our request of December 14, 1981, you subtitted additional infomation on March 9,1982. The staff has reviewed your asses-smcnt in the enclosed safety evaluation report.
It is the staff's position that you should provide justification that simultaneous coastdown of all pumps is not an event of moderate frequency for the Yankee Nuclear Power Station or demonstrate that no fuel failure would occur and should also demonstrate that the radiological consequences of the most limiting loss of flow event are acceptable.
The issue of classification as a moderate frequency event is also related to Topic XV-4, " Loss of Non-Emergency Power to the Station Auxiliaries."
As noted in our evaluation, there is a difference between fuel failure and loss of core coolability. Clad damage that results in release of the fission in the gap is defined as fuel fillure and is assumed to occur when the gcd M {3}
gases DNB ratio drops below 1.3.
The 2700*F temperature limit in WCAP-9500 was for core coolability, i.e., absence of rubble or fragmentation.
This evaluation will be a basic input to the integrated safety assessment for pg o'.
your facility unless you identify changes needed to reflect the as-built con-G.S IQY ditions at your fadtlity. This assessment may be revised in the future if
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Sincerely, original cicned byf AD-
- DL GL nas 8207080190 820629 Dennis M. Crutchfield, Chief 6
/82 PDR ADOCK 05000029 Operating Reactors Branch No. 5 P
PDR Division of Licensing
- See previous concurrence
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Docket No. 50-29 LS05 Mr. James A. Kay Senior Engineer - Licensing Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701
Dear Mr. Kay:
SUBJECT:
YANKEE - SEP TOPIC XV-7, LOSS OF FORCED COOLANT FLOW, REACTOR COOLANT PUMP ROTOR SEI2URE,RREACTOR COOLANT PIMP SHAFT BREAK By letter dated June 30, 1981, you submitted a topic assessment on the above topic.
In response to our request of December 14, 1981. you submitted additional information on March 9,1982. The staff has reviewed your asses-sment in the enclosed safety evaluation report.
It is the staff's position that you should provide justification that simultaneous coastdown of all pumps is not an event of moderate frequency for the Yankee Nuclear Power Station or demonstrate that no fuel failure would occur. The licensee should also demonstrate that the radiological consequences of the most limiting loss of flow event are acceptable.
The issue of classification as a moderate frequency event is also related1to Topic XV-4, Loss of Non-Emergency Power to the Station Auxiliaries.
As noted in our evaluation, there is a difference between fuel failure and loss of core coolability. Clad damage that results in release of the fission l
gases in the gap is defined as fuel failure and is assumed to occur when the DNB ratio drops below 1.3.
The 2700*F temperature limit in WCAP-9500 was for core coolability, i.e., absence of rubble or fragmentation.
l This evaluation will be a basic input to the integrated safety assessment for your facility un1bss you identify changes needed to reflect the as-built con-ditions at your facility. This assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this subject i
are modified before the integrated assessment is completed.
t Sincerely, I
l AD:SA:DL ORB #5 Dennis M. Crutchfdhld, Chief Rainas l
DCrutchfield Operating Reactors Branch No. 5 6/ /82 6/ /82 Division of Licensing Fncineneet
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Yankee Docket No. 50-29 Mr. James A. Kay Revised 3/30/82 i
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]I CC Mr. James E. Tribble, President l
Yankee Atomic Electric Company 7
25 Research Drive Westborough, Massachusetts 01581 Chairman Board of Selectmen Town of Rowe g
Rowe, Massachusetts 01367 Energy Facilities Siting Council 14th Floor One Ashburton Place Boston, Massachusetts 02108 U. S. Environmental Protection Agency Region I Office 1
ATTN:
Regional Radiation Representative JFK Federal Building Boston, Massachusetts 02203 Resident Inspector Yankee Rowe Nuclear PcNer Station c/o U.S. NRC L
Post Office Box 28 l
Monroe Bridge,' Massachusetts 01350 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I 631 Park Avenue King of P, russia, Pennsylvania 19406 1:
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SYSTEMATIC EVALUATION PROGRAM
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TOPIC XV-7 (a)
YANKEE NUCLEAR POWER STATION -
TOPIC:
XV-7, Loss of Forced Reactor Coolant Flow Including Trip of Pump 4
Motor and Flow Controller Malfunctions I.
INTRODUCTION 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 accompanying fuel damage could then result in ex-ceeding the specified acceptable fuel damage limits during the transient.
The transients that are expected 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 requires that each applicant for an opera-ting license provide an analysis and evaluation of the design and perfonn-ance 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.
The loss of forced reactor coolant flow is one of the postulated transients used to evaluate the adequacy of these struc-tures, systems and components with respect to the public health and safety.
Section 50.36 of 10 CFR Part 50 requires the Technical Specifications in-clude safety limits which protect the integrity of the physical barriers that guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled r.eac tors.
The staff acceptance criteria are based on meeting the relevant requirements of the following regulations:
A.
General Design Criterion 10 (Ref.1), as it relates to the reactor coolant system being designed with appropriate margins to assure that specified acceptable fuel design limits are not exceeded during normal l
operational occurrences.
I B.
General Design Criterion 15 (Ref. 2), as it relates to the reactor
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coolant system and its associated auxiliaries being designed with appro-priate margin to assure that the pressure boundary will not be breeched i
j during normal operations including anticipated operational occurrences.
I C.
General Design Criterion 26 (Ref. 3), as it relates to the reliable l
control of reactivity changes to assure that specified acceptable fuel design limits are not exceeded, including anticipated operational occur-i J
rences. This is accomplished by assuring that appropriate margin for malfunctions, such as stuck rods, is accounted for.
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The specified criteria necessary to' meet the relevant requirements of GDC 10,15 and 26 for incidents of moderate frequency are:
1 Pressure in the reactor coolant and main steam systems should be
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a.
maintained below 110% of the design values.
b.
Fuel cladding integrity shall be maintained by ensuring that the min-imum DNBR remains above the 95/95 DNBR limit for PWRs and the CPR remains above the MCPR safety limit for BWRs based on acceptable cor-i j
relations (see SRP Section 4.4).
I An incident of moderate frequency should not generate a more serious j
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l plant condition without other faults occurring independently.
d.
An incident of moderate frequency in combination with any single j;
active component failure, or single operator error, shall be considered and is an event for 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 must be assumed for all rods for which the DNBR or CPR falls below those values cited above for cladding integrity unless it can be shown, based on an acceptable fuel damage model (see SRP Section 4.2), that fewer failures occur.
There shall be no loss of function of any fission product barrier other i
i than the fuel cladding.
s bj III. RELATED SAFETY TOPICS Various other SEP topics evaluated such items as the reactor protection 1
system. The effects of single failure on safe shutdown capability are con-sidered under T6pic VII-3.
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VI.
REVIEW GUIDELINES h
1 The review is conducted in accordance with SRP Sections 15.3.1 and 15.3.2.
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The evaluation includes reviews of the analysis for the event and identifica-L tion of the features in the plant that mitigate the consequences of the event l!
as well as the ability of these systems to function as required. The extent j
to which operator action is required is also evaluated.
Deviations from the criteria specified in the Standard Review Plan are identified.
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V.
EVALUATION l
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The Yankee reactor coolant pumps (RCPs) are designed without flywheels. Due to this unique design feature, a postulated loss of power to a RCP motor will J
lead to a very rapid flow coastdown in the affected loop. Hydraulic perform-ance for loss of power to a single pump is thus similar to a RCP seizure event.
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The licensee, in his letters dated June 30, 1981, and March 9,1982, provided the results of an analysis for the subject topic. The Yankee plant design Scludes the following reactor trip circuits which provide the necessary protection against this event:
(1) a low main coolant flow trip; (2) a trip l
on high or low current to two or more pumps during four loop operation or, l
(3) a trip on high or low current to one or more pumps during three loop j
operation.
Currently, three loop operation is not permitted for the Yankee Nuclear Power Station.
The licensee indicates that two out of four RCPs are directly connected to the power supply from the turbine generator.
It is likely that these two pumps will continue to operate for 30 to 60 seconds during generator coastdown when offsite power is lost. On this basis, the licensee has classified the complete loss of flow event as a condition IV event. The staff categorizes this event as a moderate frequency event. The licensee's analysis has conservatively assumed that the electrical power supplies to all four pumps are simultaneous-1 ly lost during this complete loss of flow event. For the cycle XI analysis, the DNBR dropped below 1.3 in 2.16 seconds. The results of the licensee's fuel heatup analysis indicated that 1.25 percent of the fuel would fail during this transient using a clad temperature limit of 1100 F as a basis of determin-i ing the amount of fuel failure.
In Reference 1....the licensee. Stated that the results f'or core Xf/, presented in Reference 3. indicated that no fuel damage was expected foll'o' wing a icinplete loss of pumping power. However, the DNBR for this event was not presented.
Our review identified deviations from two acceptance criteria of the SRP.
SRP Section 15.3 states that for events of moderate frequency, no fuel failures
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should occur, and SRP Section 4.2 requires fuel failure to be assumed whenever DNBR drops below the 95/95 confidence level, (i.e.,1.3) unless it can be the shown, based on an acceptable fuel damage model, that no fuel damage results.
Staff evaluation of the licensee's model is provided in part (b) of this topic.
The licensee has provided the results of the complete loss of flow analysis to 4
demonstrate that the peak reactor coolant pressure is less than 2200 psia.
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VI.
CONCLUSION I
The staff concludes that the consequences of a postulated loss of forced l
reactor coolant flow event meet the requirements set forth in the General Design Criterion 15 with respect to integrity of the primary system boundary during a
This nomal operations including anticipated operational occurrences.
l conclusion is based on the fact that the peak reactor coolant pressure does not i
l exceed 110% of the reactor coolant system design pressure.
1 In the staff evaluation in Section V above, we have indicated that the results of the licensee's analysis do not meet the acceptance criteria of SRP 15.3.1.
l We recommend that the licensee provide justification that the simultaneous coastdown of all four pumps is not an event of moderate frequency. This justi-fication should include information on the inertial characteristics of the turbine to support the 30-60 second delay discussed above.
Further, as discussed in part (b) of this topic, the licensee should demonstrate, using an acceptable fuel damage model, that the radiological consequences of any loss of flow event are acceptable.
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j VII. REFERENCES j
1.
J.A. Kay letter (FYR 81-95) to D.M. Crutchfield (NRC), dated June 30,1981.
2.
J.A. Kay letter (FYR 82-30) to D.M. Crutchfield (NRC), dated
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March 9,1982.
I 3.
J.A. Kay letter (FYR 81-52) to D.M. Crutchfield (NRC),
Core XV Refueling Proposed Change No. 173, dated March 26, 1981.
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1 SYSTEMATIC EVALUATION PROGRAM i
TOPIC XV-7 (b)
YANKEE NUCLEAR POWER STATION 1
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TOPIC:
XV-7, Reactor Coolant Pump Rotor Seizure and Reactor I
Coolant Pump Shaft Break I.
INTRODUCTION
.j The events postulated are an instantenous seizure of the rotor and break j
of the shaft of a reactor coolant pump.
Flow through the affected loop i
is rapidly reduced. The sudden decrease in core coolant flow while the j
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 for the rotor seizure event. However, the shaft break
- j event permits a greater reverse flow through the affected loop later during i
the transient and, therefore, results in a lower core flow rate later in j
time. This topic is intended to cover both of these accidents.
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II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for an operating license provide an analysis and evaluation of tre design and performance of structures, systems and components of the facility with the objective of 1
assessing the risk to public health and safety resulting from operation of the facility.
The reactor 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 and safety, i
Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to j
include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled releasa of radioactivity.
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h The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimun requirements for the principal criteria for water-cooled reactors, y
l GDC 27 " Combined Reactivity Control System Capability," requires that the reactivity control systems, in con.ianction with poison addition by the emergency core cooling system, has the capability to reliably control reactivity changes to assure that under postulated accident conditions, and with appro-L, riate margin for stuck rods, the capability to cool the core is maintained, n
GDC 28 " Reactivity Limits." requires that the reactivity control systems be i
designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support j
structures or other reactor pressure vessel internals to impair significantly j
the capability to cool the core.
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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 j
(2) the probability of rapidly propogating fractures is minimized.
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10 CFR Part 100.11 provides dose guidelines for reactor siting against which calculated accident dose consequences may be compared.
III.
REVIEW GUIDELINES j
The review is conducted in accordance with SRP 15.3.3, 15.3.4.
The evaluation includes review of the analysis for the event and identification of the
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features in the plant that mitigate the consequences of the event as well as j
the ability of these systems to function as required. The extent to which operator action is required is also evaluated.
Deviations from the criteria
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specified in the Standard Review Plan are identified.
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IV.
EVALUATION In letters dated June 30, 1981 and March 9,1982, the licensee has stated that the minimum DNB ratio for either a seized rotor or shaft break event i
is in excess of 2.54.
The analysis is based on a single: pump seizure or shaft break without a direct reactor trip (a low main coolant flow signal
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from at least two loops must occur for direct reactor trip). The licensee has not explicitly addressed the effects of the postulated loss of offsite power following an indirect reactor trip / turbine trip and a limiting single failure during the transient.
However, the staff notes the following:
1 A.
The rotor seizure analysis did not take credit for reactor trip; the power / flow mismatch from coastdown (due to loss of offsite power) of the remaining pumps after the reactor / turbine trip would be less severe than for the complete loss of flow event where all pumps begin to coastdown
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prior to reactor trip.
B.
Since neither reactor trip nor other automatic protection features is relied upon to mitigate the rotor seizure event, the staff could not identify any single failure that would make the event more limiting l ;~
than the complete loss of flow event.
The licensee has stated that the results of a pump seizure or shaft break for both fuel performance and system pressure response are less limiting than the complete loss of flow event, discussed in Topic XV-7 (a) above. The staff agrees with this conclusion. That is, the peak pressure during this postulated accident is less than 2200 psia and the amount of fuel failure will be less than the complete loss of flow transient addressed in Topic XV-7 (a).
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j As discussed in XV-7 (a) above, the licensee detemined that 1.25 percent fuel failure would occur based on a 1100*F clad temperature limit.
The licensee, in Reference 2 stated:
"the 1100 F clad failure criterion used in this methodology is I
significantly conservative, compared to the current-day criterion j
for fuel coolability during Condition IV events.
In contrast, the NRC's recent safety evaluation of Westinghouse Corporation's j
Topical Report WCAP-9500, Reference 8, concluded that a criterion
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of 2700 F was an acceptable limit for coolability.
In comparison, clad temperatures for the Core XI reference analysis remained below 1200 F.
Approval of this limit for Westinghouse was based upon the relatively short period of time that fuel is typically calcula-i ted to experience DNB conditions for locked-rotor type events.
If this more realistic 2700'F criterion were applied to Core XV to j
determine fuel failure fractions, instead of the 1100'F criterion used for Core XI methodology, fuel failures of less than 1.25% would be predicted."
The staff concurs that the Yankee clad tempir'ature analysis results assure core coolability, and thus demonstrate conformance with GDC 27 and 28.
However, the staff does not accept the 1100 F clad temperature limit as a threshold for clad damage (i.e., fission gas leakage).
The licensee stated that the radiological consequences, would therefore l
be bounded by those for the rod ejection transient, in which 10% fuel However, the licensee's analyses have not demonstrated, failure is assumed.
using(an acceptable fuel damage model, that less than 10% of t fail i.e., clad damage releasing fission gas).
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s V.
CONCLUSIONS The staff concludes that the consequences of a postulated reactor coolant
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pump rotor sei;:ure or shaf t break event meet the requirenlents set forth in the General Design Criterion 31 with respect to integrity of the primary 1
i This conclusion systems boundary to withstand the postulated accident.
is based on the fact that the peak reactor coolant pressure does not exceed 100% of the reactor coolant system design pressure.
The staff further concludes that the requirements set forth in General Design Criteria 27 and 28 with respect to core coolability are met.
'I Based on the staff evaluation in Section V above and in Topic XV-7 (a), we l
do not have sufficient infomation to form a bases for staff conclusion with regard to the amount of possible fuel failure, and thus radiological Our l
consequences of this event for the Yankee Nuclear Power Station.
position is given above in part (a) of the topic.
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VI.
REFERENCES 1.
J.A. Kay letter to D.M. Crutchfield, dated June 30, 1981.
2.
J.A. Kay letter to D.M. Crutchfield, dated March 9, 1982.
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