ML20137S799
| ML20137S799 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 09/24/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20137S777 | List: |
| References | |
| TAC-42418, NUDOCS 8509300508 | |
| Download: ML20137S799 (6) | |
Text
- _ _
'o, UNITED STATES 8
NUCLEAR REGULATORY COMMISSION l
o 3
l wassiwarow, p. c. 20sss
\\,*... /
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 94 TO FACILITY OPERATING LICENSE NO. DPR-46 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298
1.0 INTRODUCTION
By letters dated September 20 and 23,1985, the Nebraska Public Power District (the licensee) requested emergency changes to the Cooper Nuclear Station (CNS) Technical Specifications to (1) permit reactor operation with one recirculation loop out of service, (2) to include General Electric Company's (GE) Service Information Letter (SIL) No. 380, Revision 1 recommendations regarding thermal-hydraulic stability concerns for dual loop and single loop operations, and (3) to incorporate administrative changes dealing with updating references and deleting blank pages.
Presently, the CNS operating license requires the reactor to be in cold shutdown within the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if a recirculation loop becomes inoperable and cannot be returned to service. The licensee previously i
requested authorization for unlimited single loop operation of CNS.
l Subsequently, Tennessee Valley Authority's operation of Browns Ferry Unit 1 L
(a boiling water reactor similar in design to CNS) in the single loop mode of operation at 59% power lead to concerns related to thermal-hydraulic instability. GE, in SIL No. 380, Revision 1, addressed these concerns by providing the boiling water reactor licensees generic guidance for actions which suppress thermal-hydraulic instability induced neutron flux oscillations..The licensee has proposed Technical Specifications in 1
accordance with the guidance provided by GE in SIL No. 380, Revision 1.
Specifically, the proposed changes requested by the licensee consist of (1) deletion of the license condition restricting the single loop operation; (2) for single and dual loop operation, incorporating requirements in the Technical Specifications to detect thermal-hydraulic instabilities
[
induced neutron flux oscillations and specifying operator response to the detected instabilities; and (3) updating of some references and deletion of l
some blank pages. The change noted in (1) above involves a revision of the Technical Specifications for Average Power Range Monitor (APRM) flux scram 1
trip and rod block settings, an increase in the safety limit Minimum 4
l Critical Power Ratio (MCPR) value, and a revision to the allowable Average Planar Linear Heat Generation Rate (APLHGR) values.
l 2.0 EVALUATION We have evaluated the licensee's proposal to pennit unlimited operation of
(
the CNS with one recirculation loop out of service, incorporate the GE SIL j
No. 380, Revision 1 guidance regarding thermal-hydraulic instabilities and implement some administrative changes in the CNS Technical Specifications.
i F
8509300508 850924 i
PDR ADOCM 05000298 t
P PDR
-. d
)
i
' I t
l 2.1 Single Loop Operation We have reviewed the licensee's analysis of accidents and transients which are judged to be affected by operation of the CNS in the single operating loop mode. The accidents and transients which are of concern relate to j
inadvertent variations in the coolant flow through the core and design bases of the fuel performance safety limits. The events evaluated include j
One Pump Seizure Accident, Idle Loop Startup Event, Rod Withdrawal Error j
Event, and loss-of-Coolant Accident.
One Pump Seizure Accident The licensee states that the one pump seizure accident is a relatively mild event during two recirculation pump operation. Similar analyses were perfomed to determine the impact this accident would have on one recirculation pump operation. These analyses were performed using NRC approved models for a large core BWR/4 plant. The analyses were conducted from steady-state operation at the following initial conditions, with the i
added condition of one inactive recirculation loop. Two sets of initial conditions assumed were:
a.
Thermal Power = 75% and core flow = 58% of rated j
b.
Themal Power = 82% and core flow = 56% of rated These conditions were chosen because they represent reasonable upper limits j
of single loop operation within existing Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and Minimum Critical Power Ratio limits at the i
4 l
Same maximum pump speed. Pump seizure was simulated by setting the single i
l operating pump speed to zero instantaneously.
(
The anticipated sequence of events following a recirculation pump seizure i
which occurs during plant operation with the alternate recirculation loop j
out of service is as follows:
i a.
The recirculation loop flow in the loop in which the pump seizure l
occurs drops instantaneously to zero.
b.
Core voids increase which result in a negative reactivity insertion and a sharp decrease in neutron flux, c.
Heat flux drops more slowly because of the fuel time constant.
i d.
Neutron Flux, heat flux, reactor water level, steam flow, and feedwater flow all w hibit transient behaviors. However, it is not l
anticipated that the increase in water level will cause a turbine trip i
and result in scram.
1 It is expected that the transient wili terminate at a condition of natural circulation and reactor operation will continue. There will also be a small decrease in system pressure.
The licensee concludes that MCPR for the pump seizure accident for the large core BWR/4 plant was determined to be greater than the fuel cladding l
L I
2
1 s
-3 integrity safety limit; therefore, no fuel failures were postulated to occur as a result of this analyzed event. These results are applicable to CNS, and were obtained with the staff approved methodology. We, therefore.
l agree with the licensee's conclusion that fuel cladding integrity safety margins will not be reduced.
1 Idle Loop Startup The idle loop startup transient was analyzed, in the CNS Final Safety Analysis Report (FSAR) for dual loop operation.
For single loop operation, the licensee proposed to increase the rated condition steady-state MCPR 4
~
limit by 0.01 to account for increased uncertainties in the core total flow and Traversing In-core Probe (TIP) readings. The staff found the MCPR increase of 0.01 to be acceptable. The MCPR will also vary depending on flow conditions. This leads to the possibility of a large inadvertent flow increase which could cause the MCPR to decrease below the Safety Limit for i
a low initial MCPR at reduced flow conditions. Therefore, the required I
MCPR must be increased at reduced core flow by a flow factor, K derived by f
assuming both recirculation loops increase speed to the maximum permitted by the scoop tube position set screws. This condition maximizes the power increase and hence the MCPR for transients initiated from less than rated conditions. When operating on one loop the flow and power increase will be less than associated with two pumps increasing speed, therefore, the K factorsderivedfromthetwo-pumpassumptionareconservativeforsingfe i
loop operation.
Rod Withdrawal Error The rod withdrawal error at rated power is given in the FSAR for the initial core and in cycle dependent reload supplemental submittals.
These analyses are performed to demonstrate that, even if the operator ignores all instrument indications and the alam which could occur during the j
course of the transients, the rod block system will stop rod withdrawal at a minimum critical power ratio which is higher than the fuel cladding integrity safety limit. The proposed correction of the rod block equation and lower initial power for single loop operation will assure that the MCPR safety limit will not be violated, One pump operation results in backflow through 10 of the 20 jet pumps while i
flow is being supplied to the lower plenum from the active jet pumps.
Because of this backflow through the inactive jet pumps the present j
rod-block equation and APRM settings must be modified. The licensee has proposed modified rod block equation and APRM settings in the Technical t
Specification for one pump operation, and the staff has found them acceptable.
loss-of-CoolantAccident(LOCA)
The licensee has contracted General Electric Company (GE) to perform single loop operation analysis for CNS LOCA. The licensee states that evaluation i
l
_ _ = --
l l
. j j
of these calculations (that are performed according to the procedure outlined in NED0-20566-2, Rev.1) indicates that a multiplier of 0.84 (7 X 7 1
i fuel), 0.84 (8 X 8 fuel), 0.77 (8 X 8R fuel) (Ref.:
NEDE-24258, May 1980))
l l
should be applied to the MAPLHGR limits for single loop operation of the CNS.
We find the use of MAPLHGR multipliers as indicated will be adequate to j
offset LOCA consequences in the single loop operation mode. The MAPLHGR i
factors are, therefore, acceptable.
j 2.2 Themal-Hydraulic Stability in Dual and Single Loop Operation We have evaluated the licensee's proposed Technical Specification changes to j
assure that the changes provide adequate detection and suppression of potential thermal-hydraulic instabilities.
J GE recently presented the staff with stability test data which demonstrated J
the occurrence of limit cycle neutron flux' oscillations at natural circulation and several percent above the rated rod line. The oscillations i
were observable on the APRMs and were suppressed with control rod q
insertion.
It was predicted that limit cycle oscillations would occur at the operating condition tested; however, the characteristics of the 1
observed oscillations were different from those previously observed during i
other stability tests. Namely, the test data showed that some LPRM j
indications oscillated out of phase with the APRM signal and at amplitude j
as great as six times the core average. GE has prepared and released a service information letter, SIL No. 380, to alert the BWR owners of these l
new data and to recommend actions to avoid and control abnormal neutron flux oscillations.
1 The General Electric recommendations were reviewed by the staff and found to be prudent recomendations which provide adequate detection and 4
suppression of potential thermal-hydraulic instabilities as required by General Design Criteria (GDC) 10~and 12. The staff compared these i
recommendations with the CNS Technical Specifications for operation with a l
recirculation loop out of service and found that the proposed changes are in confomance with the SIL No. 380, Revision 1 recomendations and are acceptable to the staff.
In addition, on February 9, 1985 a single loop test was performed by l
Tennessee Valley Authority (TVA) on its Browns Ferry Unit I reactor during l
which themal-hydraulic stability decay ratios were measured. The main findings of the test were that the observed increase in neutron noise i
during single loop operation (SLO) is solely due to an increase in flow noise because the inlet flow to power transfer functions during two loop i
operation (TLO) and SLO are not significantly different when test plateaus with similar power and flow conditions are compared. The Browns Ferry Unit I reactor has been found to be stable in all modes of operation attained i
during the present tests. The most unstable test plateaus corresponded to minimum recirculation pump speed in SLO, which has the minimum flow and i
i 1
--,-,.,--,-.-_..--,.w.
__,~..r-.,-.~,-_----..,-_-,.,,r.-,,,,,,-,,_.-.-,.-_--,___.--,__m..._-..%,
.. maximum power to flow ratio. The estimated decay ratio at this plateau was 0.53.
The decay ratio decreased as the flow was increased during SLO (down to 0.34). This implies that the core-wide reactor stability follows the same trends in SLO as it does in TLO.
Finally, no local or higher mode instabilities were found in the data taken from local power range monitors (LPRMs). The decay ratios estimated from LPRMs were not significantly different than the ones estimated from the average power range monitors.
In conclusion, the measured decay ratios at Browns Ferry Unit I showed the plant to have adequate stability margin over a range of power / flow.
conditions which are of concern during single loop operation. Since the Cooper Nuclear Station maximum calculated decay ratio (.86) is similar to Browns Ferry Unit 1 (.87), and since it was shown that the stability characteristics of SLO are similar to TLO this test provides additional justification to allow single loop operation at CNS.
2.3 Administrative Changes The licensee has proposed to update some references and delete blank pages in the Technical Specifications. The staff finds the proposed administrative changes acceptable.
3.0 EMERGENCY CIRCUMSTANCES At approximately 0900 CDT on September 20, 1985, the Recirculation Pump B was tripped and Cooper Nuclear Station began single loop operation.
The recirculation pump was tripped after it was determined that a low insulation reading to ground existed in the recirculation pump motor-generator set windings. The Cooper Nuclear Station Technical Specifications contain a limiting condition for operation (LCO) which requires the reactor to be shutdown if a recirculation loop is out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee determined that approximately two weeks would be required to restore Recirculation Pump B to service. On September 20, 1985, the licensee infonned the staff of the conditions at CNS and its decision to file an expedited license amendment request which would permit single loop operation for an indefinite period and thus avoid reactor shutdown as a result of the LCO.
By letter dated September 20, 1985, the licensee proposed an expedited Technical Specification change which would remove the 24-hour LCO. The licensee also proposed to add surveillance requirements relative to thennal-hydraulic instability which would justify deletion of the LCO.
After discussion with the staff the licensee provided revisions to the original amendment application and documented these changes by letter dated September 23, 1985.
3.1 No Significant Hazards Consideration Determination The Commission's regulations in 10 CFR 50.92 state that the Commission may make a final detennination that a license amendment involves no significant l
e
=
! hazards considerations if operation of the facility in accordance with the amendment would not:
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
The infomation in Section 2.0 above provides the basis for evaluating this license amendment against these criteria. Since the requested operational mode is acceptable and the plant operating conditions, the physical status of the plant, and dose consequences of potential accidents are the same as without the requested change, the staff concludes that:
5 (1) Operation of the facility in accordance with the amendment would not significantly increase the probability or consequences of an accident previously evaluated because the types of accidents most likely to occur with single loop operation have been evaluated and formed to satisfy the Comissions regulations.
In addition, the amendment would add more restrictive limits and surveillance requirements to ensure that the consequences and probabilities would not be increased.
i (2) Operation of the facility in accordance with the amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated because all abnormal operating transients which could be initiated with single loop operation, such as an inadvertent startup of an idle recirculation pump or pump trip have already been analyzed in the FSAR and reviewed and accepted by i
the staff. The additions of thermal-hydraulic instability l
surveillance requirements involve nomal plant operating practices t
and, therefore, are not expected to create a new or different kind of i
accident.
(3) Operation of the facility in accordance with the amendment would not involve a significant reduction in a margin of safety because any I
decrease in margin resulting from single loop operation would be f
offset be the more stringent operating limits and surveillance requirements that are also added by the amendment.
Accordingly, we conclude the amendment to Facility Operating ~ License No.
I DPR-46, permitting single loop operation for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, l
involves no significant hazards consideration.
i 3.2 State Consultation j
i In accordance with the Comission's regulations, consultation was held with the State of Nebraska by telephone. The State expressed no concern either t
I I
s.
- from the standpoint of safety or of no significant hazards consideration determination, in view of the interim nature of the amendment and the compensatory measures.
4.0 ENVIRONMENTAL CONSIDERATION
S This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has made a final no significant hazards consideration finding with respect to this amendment. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that: '(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed n.anner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: George Thomas, George Schwenk and Ernest Sylvester Dated:
September 24, 1985
. - _.