ML20039C211
| ML20039C211 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 12/11/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20039C209 | List: |
| References | |
| NUDOCS 8112290062 | |
| Download: ML20039C211 (5) | |
Text
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o UNITED STATES 8" 3m ',i NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20555
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w SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMEN 0 MENT NO. 60 TO LICENSE NO. OPR-36 MAINE YANXEE ATOMIC POWER COMPANY MAINE YANKEE ATCMIC POWER STATION
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DOCKET NO. 50-309 INTRODUCTION:
In c. letter to the Maine Yankee Atomic Power Company dated July 10, 1981, the USNRC granted approval of certain Technical Specification (TS) changes made as part of the Maine Yankee Cycle 6 Reload. This letter stated in part that approval for continuation of operations in Cycle 6 beyond 4000 MWD /MTU would require that the revised steam line break analysis be performed and submitted. This analysis was transmitted to NRC as an attachment to a letter frcm Maine Yankee dated October 29, 1981. The October letter stated that the results of the analysis indicated the need for certain TS changes.
A letter from Maine Yankee to NRC dated November 18, 1981 requested the additional TS changes. The staff and our technical assistance consultants at Argonne National Laboratory have reviewed these submittals and additional documents requested from Maine Yankee in support of their analysis.
For several previous reload cycles Maine Yankee has analyzed the main steam line break transient with the FLASH-4 computer code. This code has been reviewed and approved for use by NRC. However, because of design modifications to the main feed system and additional safety requirements imposed by NRC, Maine Yankee determined that FLASH-4 was no longer capable of analyzing this transient, and in a letter to NRC dated December 23, 1980, they announced their intention to use the RETRAN computer code. Generic review of the RETRAN computer ccde is being conducted by the staff and our consultants at Argonne National Laboratory. Although this review is not complete, enough progress has been made so that along with the material submitted by Maine Yankee and the staff's experience with RETRAN for other reloads, there is an adequate basis for the review of this transient.
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Several changes were made to the Maine Yankee power plant during the outage for reload Cycle 6.
A steam driven main feedwater pump was added to replace the two electric pumps for operation between 50% and 100% power. Additional controls were added to the feedwater isolation system. The plant modifica-l I
tions and the NRC requirements,.i.e., the effects of feedwater regulating l
valve failure and tripping the reactor coolant pumps, have a significant effect on the main steam line break transient.
In order to support this new analysis using the RETRAN. code and the more sophisticated and i'
complex mcdel of their reactor power plan, Maine Yankee submitted a kh O 00
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4 number of documents for our review. The validity of the nodalization used by Maine Yankee to model the reactor plant and the accuracy of the RETRAN code are demonstrated by a comparison of calculations made by RETRAN and FLASH-4 of the previous cycle steam line break transient. This was reviewed by the staff and our consultants at Argonne. The two codes agree within an acceptable margin for the principal parameters of the transient.
Addition of the steam driven main feedwater pump and consideration of feedwater regulating valve failure requires a very detailed modeling of the feedwater and condensate systems. The details of the final model for the Maine Yankee Plant and the results of their analysis of the main steam line break transient were transmitted to NRC October 29, 1981. The staff and our consultants at Argonne reviewed this submittal and held a meeting with representatives from Maine Yankee at Argonne November 23 and 24,1981, 4
to allow Maine Yankee to answer questions.on their models and methods, 2
and to discuss the results of their calculations.
MSLB EVALUATION:
A number of variations of the main steam line break transient were analyzed i
by Maine Yankee to account for different assumptions about break location, initial plant conditions, and single failures. Although an enormous number of combinations of these parameters is possible, the licensee identified eleven cases which bound the spectrum of plant responses to the main steam line break. The staff agrees that these cases include those which are most limiting. Two new requirements imposed by NRC have a significant effect on the main steam line break transient:
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the requirement that the operator trip the reactor coolant pumps after receiving a safety injection actuation signal. For all cases analysed, this increased the severity of the MSLB transient; and 2.
consideration of the failure of a main or auxiliary feedwater regulating valve to close.
The minimum shutdown margin for all cases analyzed was for f ailure of the main feedwater regulating valve to close along with tripping the reactor coolant pumps 30 seconds after receiving the safety injection actuation signal. The most limiting case for zero power conditions was the failure of one high pressure safety injection pump to start. The proposed change to the TS Section 3.6.B is necessary to insure that at least one high pressure safety injection pump will start.
We and our consultant at Argonne have reviewed the models and assumptions used by Maine Yankee in their steam line break analysis and we have determined that the models are appropriate and the assumptions are sufficiently conservative. One important example of the conservative nature of the analysis is the calculation of reactivity addition from coolant temperature and boron addition. Although emergengy. core cooling water is assumed to be at 400F which maximizes the positive reactivity addition due to temperature, the boron concentration is assumed to be the technical,specificatica minimum,
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) and this corresponds to a temperature of 1200F. For the cases where reactor coolant pumps are tripped, the core reactivity is calculated based on the faulted cold leg temperature only (no mixing with warmer water from other legs). This made it necessary to calculate the boron concentra-tion external to RETRAN. The method used to perform this calculation has
' N also been reviewed and is considered acceptable.
Among other conservative assumptions in this analysis are the feedwater j
component arrangement which maximizes flow and minimizes temperature, the application of maximum uncertainties in the conservative direction to moderator, doppler and control rod worths, and minimum decay heat levels.
r In order to quantify the effect of these conservatisms the staff requested that Maine Yankee rerun the most limiting case with best estimate values for control rod position, control rod worth (except the highest worth rod remains stuck out), doppler, moderator and boron reactivity, decay.
heat, and core average temperature. The maximum reactivity for this transient t
was -3.27% delta rho for the hot full power case and -4.76% delta rho for the hot zero power case.
The staff and our consultants at Argonne National Laboratory have reviewed the analyses and effects of main steam line breaks during various modes of plant operation. The accident which resulted in the most limiting transient was determined and evaluated using a mathmatical model which was reviewed and found acceptable by the staff for this analysis. The parameters used as input to this model were reviewed and found to be suitably conservative.
The results of the analysis of the spectrum of steam line break accidents showed that for the worst case, with the most limiting single failure and the control rod of greatest worth stuck out of the core, the reactor did not return to criticality during the transient. Based on this, the staff concludes that the Maine Yankee Cycle 6 reload submittal is acceptable with regard to steam line break accidents.
TS CHANGE EVALUATION:
The steam line break analysis results indicated the need for certain TS changes to assure that the core does not return critical following the most i
limiting steam line break at the end of cycle (E0C), with the most reactive rod assumed to be stuck out of the core.
The licensee proposed a TS change to require that two high pressure safety injection pump subsystems be operable whenever the reactor coolant system boron concentration is less than that required for hot shutdown condition. This is necessary to increase the boron concentration to assure the reactor remains subcritical during the cooldown following the MSLB accident. We have found this change acceptable.
The licensee has proposed to change the requirement for the available shut-down margin, with one CEA withdrawn, from a constant 3.2 percent reactivity to a graph showing minimum required shutdown margin vs. actual RCS boron concentration. This graph would'reVuire a minimum shutdown margin of 4
4.21 percent reactivity at E0C (more.resctrictive than the present 3.2.
percent). At the beginning of Cycle (BOC), approximately 848 ppm boron,
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the graph requires 1.39 percent reactivity to preclude return to power in the event of the SLB accident. The justification for the lower shutdown margin requirement at BOC is from the MSLB analysis, i.e., the reactivity gain up]n RCS cooldown is less because the moderate temperature coefficient is smaller at B0C. The licensee proposed a variable reactivity from B0C to EOC to preclude unnecessary reactor shutdown; for example, should an electri:al drive problem occur on a CEA. This option may not be available with a fixed minimum shutdown margin as currently exists.
We have reviewed this change, and conclude that sufficient minimum shutdown margin will be available to preclude return to criticality throughout cycle operation, in the event of a SLB transient. Therefore, we conclude this change acceptable.
Near the EOC life when the RCS boron concentration is less than or equal to 100 ppm the revised TS requirements place the CEA at upper electrici limit rather than the previous requirement of 4 steps inserted from their electrical limit. This was done to gain additional reactivity at E0C.
The previous requirement for CEAs to be inserted 4 steps was established to reduce CEA guide tube wear. The CEA guide tube wear was subsequently reduced by the insertion of sleeves in the CEA guide tube design. This has previously been approved for Cycle 6 operation, see our Safety Evalua-tion supporting Amendment No. 58 issued July 10, 1981. Therefore, we agree that the rods may be withdrawn to their electrical limit. We find this TS change acceptable.
We have reviewed the new Power-Dependent Insertion Limit (PDIL) graph contained in the TS. The new PDIls are more restrictive than the previous PDILs.
In addition, the PDIls are reoriented for easier comparison with control room instrumentation. We conclude that the PDIL change is acceptable, i
Additional wording revisions were proposed to the TS for improved clarity.
We conclude that these administrative changes are acceptable.
8 Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made l
this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement or negative declaration and environ-mental imoact appraisal need not be prepared in connection with the issuance of this amendment.
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Safety Conclusion We have conciuded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Connission's regulations and the issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of the public.
Date: December 11, 1981 1
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