ML19207C240
| ML19207C240 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 08/09/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TASK-04-01.A, TASK-4-1.A, TASK-RR NUDOCS 7909100540 | |
| Download: ML19207C240 (12) | |
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August 9,1979 Docket No. 50-155 Mr. David Bixel Nuclear Licensing Administrator Consumers Power Company 212 West Michigan
- 9. venue Jackson, Michigan 49201
Dear Mr. Bixel:
RE: TOPIC IV-1.A - OPERATION WITH LESS THAN ALL LOOPS IN OPERATION Enclosed is a copy of our revised safety assessment of Topic IV-1.A, Operation With Less Than All Loops In Operation. This revision supersedas the evaluation issued by our letter dated August 17, 1978.
This revision completes our assessment of Topic IV-1.A which will be used as input to the integrated review of the Big Rock Point P1 ant.
However, it should be noted that the acceptability of this topic evaluation is contingent upon your agreement to (1) include in a procedure for N-1 loop operation a statement that the bypass and isolation valves in the inactive loop be closed during N-1 operation,
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(2) lysically lock-out power to the inactive pump, and (3) incorporate the MAPLHGR limits for N-1 loop operation in the Technical Specifications.
If there are any errors in the facts of this revised assessment, please supply corrected information and your response with respect to items (1) through (3) above within 30 days of the date you receive this letter.
Sincerely,
.EA w E
'w Dennis L. Ziemann / hief Operating Reactors Branch #2 Division of Operating Reactors
Enclosure:
Revised Assessment for Topic IV-1.A c, enclosure:
- ee next page
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Mr. David Bixel August 9, 1979 cc Mr. Paul A. Perry, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 40201 Hunton & Williams George C. Freeman, Jr., Esquire P. O. Box 1535 Richmond, Virginia 23212 Peter W. Steketee, Esquire 505 Peoples Building Grand Rapids, Michigan 49503 Charlevoix Public Library 107 Clinton Street Charlevoix, Michigan 49720 K M C Inc.
ATTri: Mr. Richard E. Schaffstall 1747 Pennsylvania Avenue, N. W.
Suite 1050 Washington, D. C.
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Enclosure SYSTEMATIC EVALUATI0ft PROGRAM Topic IV-1.A Operation With Less Than All Loops In Service Plant:
Big Rock Point Discussion The majority of the presently operating BWRs and PWRs are designed to permit operation with less than full reactor coolant flow. That' is, if a PWR reactor coolant pump or a BWR recirculation pump becomes inoperative, the flow provided by the remaining loop or locps is sufficient for steady state operation at some definable power level, usually less than full power.
Plants authorized for long term operation with cne reactor coolant pump out of service have submitted, and the staff has approved, the necessary ECCS, steady state, and transient analysis. The remaining PWR and BWR licensees have Technical Specifications which require reactor shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if one of the operating loops laecomes inoperable.and cannot be returned to operation within the time period.
In a letter dated August 17, 1978, Consumers Power Compa,ny (the licensee) was sent draft evaluations of eight essentially co.upleted Systematic Evaluation Program (SEP) Topics. We requested that the licensee review and verify that the information was factual and that all documentation cited was current. Topic IV-1.A, Operation 33 fe.DO
. With Less Than All loops In Service, was one of the eight essentially completed for Big Rock Point. The assessment stated that authori-zation for N-1 loop operation is provided (Technical Specification 4.1.2(b)); however, there is no supporting ECCS analysis to justify this mode of operation.
By letter dated October 24, 1978, the licensee responded to our August 17, 1978 request and provided comments concerning the correctness of our assessments. With regard to Topic IV-1,A, the licensee took exception to our assessment and concluded that opera-tion with less than all loops in service at Big Rock Point was justified. This conclusion was based on an analysis performed by General Electric Company (GE) in early 1977. The licensee derived and implemented operating limits for Big Rock Point based on this analysis.
The October 24, 1978 letter from the licensee contained three enclosures:
(1) Consumers Power Company statement concerning N-1 loop operation, (2) the GE Single Loop LOCA Analysis, and (3) a document entitled Operation of the Big Rock Point Reactor With One Loop Out of Service - Impact on MCHFR Limits. contained 3 attachments:
(1) the General Electric Analysis, (2) Addendum A General Electric's Answers to Consumers Power's questions, and (3)
Addendum 3 "APLHGR Limits for Exxon Fuels.
33CC.10
. Evaluation Several factors have to be considered when evaluating N-1 loop operation:
(1) the impact on normal operation (i.e., are there adecuate thermal margins _when one considers the effect of antici-pated transients), (2) the potential effect on accidents which are analyzed (principally the LOCA and locked rotor accident), and (3) the potential for a new accident (in this case, a coldwater accident caused by the startup of the inactive pump).
One factor that can affect all three of these considerations is the effect of one loop operation on reactor coolant flow distribution.
Big Rock Point is a 2 loop, General Electric design, non-jet pump boiling water reactor. The coolant flows through two inlet nozzles (one per loop) which lie 72 degrees apart on the vessel lower head.
The flow entering through each nozzle impinges on a diffuser plate (one plate per nozzle). A flow diffuser baffle connected to the core support plate surrounds the fuel channel support tubes and causes the pressure at the inlet to the core support tubes to be relatively uniform. The fact that the vessel entrance re'gion acts as a plenum has been supported by test (" Core Performance and Transient Flo. 'esting - Big Rock Point Boiling Water Reactor",
GEAP-4496, July 1965, USAEC contract AT (04-3)-361). The test 330,11j.
showed that the frictional pressure drop between the vessel nozzles and the support tube inlets to be nearly 5 times the velocity head in the support tubes.
Instrumented fuel assembly measurements during forced-circulation tests (Figure 4-7 of the above reference) have shown relative assembly power to be insensitive to the number of loops in operation, further indicating that the relative flow to the assemblies is not substantially effected by the number of loops in operation. When considering the high losses due to flow resis-tance caused by the orifices in the assemblies, a small pressure difference in the lower plenum at the support tube entrance elevation should have a negligible effect on the core flow distribution.
Since the physical arrangement of the forced circulation systems at Big Rock Point, flow through the core, and supportive testing indi-cate that flow perturbations will not be introduced to the system, it is expected that the reactor will not discern the difference between one pump and two pump operation. The staff, therefore, concludes that uneven or asymmetric flow conditions will have a negligible affect on Big Rock Point during N-1 locp operation.
With regard to the effects of anticipated transients:
The licensee has provided (Enclosure 3 to the October 24, 1973 submittal) a discussicn on the effects of transients 0.1 the minimum critical heat flux ratio (FCHFR) when operating in the reduced flow con-fipration.
The licensee stated that the 3.0 'GR limit derived i
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for full flow steady state operation is valid for single loop operation. Our review of the codes used to predict the MCHFR limit support this conclusion. Although the critical heat flux correlation used by the licensee (synthesized ENC Hench-Levy) does not, per se, have a flow term which would directly support the licensee's state-ment that MCHFR is insensitive to flow chang.3; it does have a fluid quality factor. Since we have indicatea above that the reactor does not see the difference between one or two pump operation, the quality of the fluid does not change; therefore, the computed MCHFRs for two loop operation are bounding for the one lcop operation. Furthermore, as discussed in the Cycle 15 reload analysis, a transient MCHFR of 2.15 was established; to this limit was added additional margin to account for the worst case transients. The staff further added conservatism to the limit which yielded the steady state MCHFR of 3.0.
The total steady state MCHFR provides assurance that under the worst case transient the resulting7MCHFR will not be below the safety limit of 2.15. This margin of safety precludes operation of the reactor in a region conducive to fuel failure.
In addition to the operating restrictions of MCHFR, high neutron
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fiux and high reactor pressure trips are maintained within the same
- roximity for single loop operation as for two 1000 operation.
That is, the reactor orotective system is realigned to cause trips
' thin the s rs : lerance at the reduced power level as they would O.~$ Cl[3
. at the full power level (e.g., high flux trip - 120% of maximum allowed oper? ting power level, at reduced power, say 50%, in the event of a transient the reactor would trip if power reached or exceeded the 60". power level). The same reduction in setpoint would relate to the overpressure trip.
General Electric (GE) has performed for Big Rock Point an.ECCS-LOCA calculation at 102% of rated power, with one loop out-of-service, with the out-of-service loop isolated (pump suction, discharge, and bypass valve closed), and has compared the results to calculations for the two pump in service conditions. The calculatiens and comparisons performed demonstrate all effects of one loop operation that might significantly affect peak clad temperature (PCT). The analysis of the break spectrum revealed that the worst case break would result from a break in the recirculation discharge line of 2
the inactive loop (0.500 ft break size). This analysis predicts the PCT for this break size to be 2192 degrees F and a peak local oxidation fraction of 0.072. The Appendix K to 10 CFR 50.46 PCT limit is 2200 degrees F and the Big Rock Point oxidation upper limit is 3.17.
The calculated PCTs and oxidation fractions for all breaks a1alyzed for single loop operation are reported in Table 3 of Encicsure 2 to the October 24, 1978 submittal by the licensee.
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. The analysis has demonstrated that a correction factor must be applied to the all-loops-in-service maximum average planar linear heat generation rate (MAPLHGR) limit for conservative one-loop-out-of-service operation. The MAPLHGR limits for double loop operation were calculated for the design basis accident (DBA) for the cycle 15 reload. New MAPLHGRs were calculated by the licensee considering 2
the worst case break (0.5 ft recirculation discharge line) and compared them with those for the two loop worst case break. The maximum offset in MAPLHGRs was 0.5 kW/ft. The licensee proposed to operate the plant in the N-1 loop configuration with MAPLHGR limits calculated by subtracting 0.5 kW/ft from the double loop MAPLHGR numbers. The analysis yielding approximately a 5% reduction in MAPLHGR applies to the GE fuel. Big Rock Point also employs fuel fabricated by Exxcn; however, an analysis for the benavior of the Exxon fuel in single loop operation has not been performed.
The licensee states that since MAPLHGR is insensitive to the number-of loops in service, since they changed only by at most 5% in single loop for the GE feel, MAPLHGRs can be conservatively derived by reducing the existing two-loop Exxon MAPLHGR limits by 10%.
Based on our review of the methodology employed to calculate the MAPLHGR limits associated with Cycle 15 and our review of the analysis performed for the single loop mode of 0::eratic.i, we conclude that the new limits are conservatively deri';e: and are therefore acceptable. However, since these limits are subject to
- ni ge ;ith each reioad review, we recommend tnat the single locp 000l1%
MAPLHGR limits be made part of the Technical Specifications and subsequent changes to them be evaluated by the staff in the same manner as any other change to safety limits.
Regarding the locked roter accident, Big Rock Point has not provided an analysis of the effects that this accident might have when operating in the N-1 loop configuration. However, since operation of the facility with less than all loops in service is a relatively low likelihood event (based on operating experience with several reactors of this design) we conclude that an event such as a locked rotor wnile in this mode is even more remote. Furthermore, the Systematic Evaluation Program in the course of reviewing design basis events will review the locked rotor accident in both the N loop and N-1 loop conditions. Thus, we conclude that in the interim this deficiency is acceptable for Big Rock Point.
With regard to the potential for a new accident the staff considered the potential for a cold water injection. accident caused by the startup of the inactive loop.
Staff criteria requires that an v alysis of this event be performed to determine the pot'ential con sequences. Technical Specifications prohibiting startup of an inactive loop are not considered by the staff to be an acceptable alternative to analyzing the event. However, in lieu of an analysis, U,].;': p,
-9 reducing the credibility of the event is an acceptable alternative.
Methods such as the use of temperature differential interlocks, which prevent opening a valve or starting a pump unless a predetermined minimum temperature differential exists between the active and inactive loop, or requiring the isolation valves to remain open when the pump is inactive, thereby maintaining the idle loop in thermal equilibri'um with the operating loop, are examples of effective measures to reduce the likelihood of the event which the staff would review for acceptability.
The analysis performed by the licensee to justify the operating MCFR and the limiting MAPLHGR in the N-1 loop configuration assumes that the inactive loop is completely isolated allowing no bypass flow through the inactive loop. Although operating with the isolation valves open will establish a thermal equilibrium between the loops and resolve the co'id water injection accident, the diversion of bypass flow (backflow through the loop) may impact the previously discussed limits in such a way that they can no longer be considered conservative without additional analysis. On this basis the staff cannot permit Big Rock Point to t wate with an idle loop (non-isolated) when in the N-1 configuration. Based on the above, we recommend that the licensee establish a procedure that administratively dictates the closing of the isolation and bypass valves in the inactive loop if operation is to continue. Section 12.8 of the Big Rock ~
Nint Final Hazards Summary Report presents the cold water accident irilysis perfor.ed by the licensee. A comoarison of this analysis g
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to the methods used in current criteria (Standard Review Plan 15.4.4) reveal some deviations. The SRP method is more detailed than that of the licensee in that it discusses the effects of the cold water accident on operating limits (MCFR), linear heat generation rate (LHGR), and the potential'for overpressurization. However, re-analysis of the cold water accident need not be performed if the potential for occurrence is substantially reduced or removed. Therefore, we recommend that the power to the inoperative pump be physically re-moved (locked out) during N-1 loop operation. This action will provide additional assurance that an inadvertent cold water injection accident will not occur.
It should further be noted that this admin-istrative operating restriction may be removed by performing a re-analysis of the cold water accident in accordance with SRP 15.4.4 and the acceptability of that analysis determined by the NRC staff.
We therefore conclude, based on our review of the docketed and submitted material, that operation with less than all loops in service is acceptably resolved. However, it should be noted that the acceptability of this topic (IV-1-A) is contingent upon the licensee's agreement to (1) include in a procedure for N-1 loop operation a statement that the bypass and isolation valves in the inactive loop be closed during N-l operation, (2) physically lock-out
- ower to the inactive pump, and (3) incorporate the MAPLHGR limits for 1-1 loop operation in the Technical Specifications.
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