ML20126F617
| ML20126F617 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 03/11/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126F614 | List: |
| References | |
| NUDOCS 8103190175 | |
| Download: ML20126F617 (7) | |
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,g UNITED STATES y
.. y g NUCLE AR REGULATORY COMMISSION 3p i, j WASHINGTON. D. C. 20555 N
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMEitT N0. 64 TO FACILITY OPERATING LICENSE NO. DPR-28 VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POER STATION DOCKET N0. 50-271 I.
Introduction By letter dated February 12,1981 (reference 1), Vermont Yankee Nuclear Power Corporation (the licensee) requested changes to the Technical Specifications (Appendix A) appended to Facility Operating License No. DPR-28, (reference 2) for the Vermont Yankee Nuclear Power Station (VY).
The proposed changes per-mit the licensee to conduct stability and recirculation punp trip tests at Vermont Yankee during Cycle 8.
The objectives of the tests are to obtain data for licensing support of Vermont Yankee stability performance and for qualification of its stability and operational transient models.
The stability tests will be performed at one point under minimum pump speed and three points under natural circulation conditions.
The approach will be to perform tests with vessel and core pressure perturbation introduced through the turbine control system at steps of 10 psi.
The resulting neutron flux response of the core will be measured and used to determine a core transfer function.
The recirculation pump trip test will be performed af ter the stability test at minimum pump speed is completed and before the natural circulation tests.
The total actual time of the tests will be less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> within a time span of seven days.
The Technical Specification changes are required in order to allow the operation under natural circulation during the tests. The tests also require bypassing of any affected trip functions while the test instrumentation is being installed and removed.
The APRM ficw-biased rod block line and scram and rod block
, monitoring setting are to be raised, and special MCPR and MAPLHGR limits are to be used during the tests.
II. hal ua tion The test procedures are presented in Reference 3.
In order to ensure that the stability test will be within acceptable limits, two levels of criteria have been established as follows.
Level I criteria, if exceeded, require that the test be suspended immediately uritil corrective action can be taken.
The Level I criteria are that the APRM response to pressure perturoation must be within +20S of the rated power, the decay ratio 810s190F W
. must not exceed 1.0, the offgas and reactor coolant _ solubles must remain within the site administrative limit, and if the decay ratio reaches 1.0, the APRM oscillation must not exceed + 15% of the rated power.
Level 2 criteria are the expected test results which, if exceeded, the responsible test engineer i
must evaluate the situation and determine if the test should be suspended.
The criteria are that the APRM response is within + 155 of the rated power l
and the decay ratio is less than 0.5 The tests are to be conducted in orderly fashion so that one test result provides guidance for the next test.
The mininum pump speed test point, VPTl, will be performed first to allow determination of the sensitivity between minimum pump speed and natural circulation on stability of the systems.
Data from lower power test points, VPT2 and VPT3, at natural cir-culation condition will be extrapolated to establish a power level of VPT4, which is the least stable condition and will be linited to 70% of the rated nower and a decay ratio of less than 1.0.
Since VPT4 is above the design APRM flow-biased rod block line and very close to the APRM flow-biased scram, higher settings of the rod block and scram are necessary in order for the test to be conducted without an inadvertent scran during the test.
Sper.ial LOCA and MCPR limits will be used during the test in lieu of the APRM flow-biased scram.
The limits are 9stablished through the reanalysis of the Cycle 2 reload safety analysist4) using the natural circulation condi gn at V?T4 with the General Electric (GE) standard reload analysis method 1
The analysis describes local rod withdrawal error transient and core-wide transients such as load rejection without bypass, loss of 100 F feedwater l
heating, inadvertent HPCI start and feedwater control failure.
The results of the analysis are reported in Reference 3.
The rost severe t.CPR is 0.19 for the inadvertent HPCI start transient.
An additional 0.05.is added to the i
- CPR to account for lack of operating experience at the hich power natural l
i circulation condition.
With the acceptance criteria for the design-limit l
MCPR of 1.06 for Ex8,1.07 for 8x8R and P8x8R bundles, the overall operating-limit MCPR's will be 1.30,1.31 and 1.31, respectively, for 8x8, 8x8R and PSxER bundles.
These MCPR limits will be used during the natural circulation l
tests to replace the flow factor Kf which is normally used as a multiplier of the operating-limit MCPR for the reduced flow.
Since the "-G sets are off at natural circulation conditions and special test procedure will ensure that only one pump will be started at natural circulation conditions, protection l
against a flow increase transient is provided.
i The flow-biased APRM rod block which normally provides the LOCA margin required for operation at reduced flow is replaced by a special PAPLHGR limit.
The MAPLHGR limit will be 805 of the limits specified in the Technical Specification.
The recirculation pump trip is to be conducted between the stability test points VPTl (at minimum pump speed) and VPT2 (lower power test point at natural circulation ).
Since the reload analysis for natural circulation conditions, which is php least stable on power-flow map, has shown a decay ratin of less than 1.0 d /, the puc.p trip tost should result in no adver:e safety ef fect.
r
i 3-l i
t As for the evaluation of the stability test results, VV _is scheduled to record i
the paranetors listed on table 1.
However, the staff feels it is importans to record a jditional data in order to provide conplete and convincing test res ul ts. Tiie following parameters should be recorded i
(1) water level in the reactor vessel; I
(2) core inlet temperature; (3) steam flow at the turbine admission valves used to generate the test i
perturbations, and f
(4) steam pressure upstream of these sane valves.
It is also desirable, for some recorded parameters, to read not only the de j
values, but noise conponents.
For purposes of providirg stochastic conpari-son analyses, the noise conponents of the reactor vessel pressure signal, the core inlet temperature, core inlet flow rate, and as aany APR!1 and LPRM out-puts as possible should be recorded both during the test and for a t least r
ten minutes prior to the beginning of each of the tests.
For the neutron i
noise data, a minimum of one APRM signal, plus 4 LPRf1 signals all from the j
same chanriel, plus an additional LPRM signal from a nearby channel are requi red. Noise compon its fron any or all of the other output pt rameters j
listed above would be valuable, but are not required for the stability evalu-ation.
If the parameter recording is done digitally, a minimum sa apling rate l
of 10 Hz is needed for core stability analysis.
j In order to perform audit calculations, the raw test data, with appropriate signal identification and scaling factors will be desirable.
The following data should also be provided, i
i A.
Data syecific to each test at steady state cnnditicos prior to the i;ejinning of the test:
i 1.
Pcwer:
Thermal power generated in the ccre.
Map of the power generated in each fuel assembly.
Map of the vertical i
power distritsution in each characteristic bundle type according to the position
{
i of the neighboring control rods.
2.
Flow:
Mass flow rate entering the reactor core.
Fraction of the core nass flow rate which goes to the bypass region.
!' ass flow rate in each j
of the recirculation loop drive flows.
Feedwater rass flow rate.
Steam j
flow rate.
i i
3.
Tenperature:
Feedwater terperature. Core inlet tenperature.
l 4
Pressure:
Core inlet pressure.
Core outlet pressure.
Stean i
separator exit pressure.
Vessel outlet pressure.
Pecirculation pump inlet j
pressure. Downcomer pressure at the jet pump entrance.
Jet punp throat cressure.
j i
5.
ater level in the reactor vessel.
l l
l
t !
i 6.
Map of control rod position and degree of insertion.
)
B.
Plant data common to all proposed-tests.
(1)
Co're description: Fuel assembly dinensional and material data; core map including control rod and LPRM locations and support plate orifices; I
reference design data including fraction of the thermal power which is transferred to the coolant by convection and fraction of the thermal power which is deposited in the bypass coolant flow by radiation.
(2)' Recirculation pump data:
Characteristic equatioris relating pump head and pump torque to mass flow rate and pump speed.
Also, the moment of inertia of the pump and pump efficiency.
4 (3)
Friction coefficients:
At the core entrance orifice; at the core exit; between the core exit and the steam separator exit; between the steam separator exit and the steam dryers exit; at the jet pump suction entrance; at the drive flow nozzles; at the recirculation loop piping; at the jet pump diffusers; at the lower plenum; at the upper dome plenum.
Also, the friction factor nultiplier in the core fuel bundles.
(4)
Length of area flow ratios (L/A) for the steam separators as a function of steam quality, for the recirculation loops, the jet pumps and 4
the downcomer region.
5)
Neutronic parameters:
Sets of 2 group cross-sections (i.e. 1 a.
,J k,(D,1.R) as a function of local bundle void fraction, burnup, gadolinion, content and control rod position for each type of fuel bundle.
Also the doppler reactivity coefficient as a function of burn-up, void fraction and average fuel temperature.
(6)
Fuel gap conductance.
The staff has reviewed the procedure, Technical Specification changes and safety analysis report of the proposed stability and recirculation pump trip tests at VY.
Based on the above evalution, the staff concludes that the thernal hydraulic acceptance criteria, where MCPR must be greater than 1.06 and decay ration must be less than 1.0, will not' be violated during the tests.
Therefore, the Technical Specification change for VY is acceptable.
III.
Envi ronmental Considerations We have determined that the amendment does not authorize a change in effitent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant i
t from the standpoint of environmental impact, and pursuant to 10 CFR 651.5(d)(4)
^ hat an environmental impact statement, or negative declaration and environ-tmental impact appraisal need not be prepared in connection with the issuance of the amendment.
IV.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the pro-or consequences of accidents previously considered and does not bability involve a significant decrease in a safety margin, the amendment does not involve a significant hazards con:ideration, (2) there is reasonable assur-ance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activ' ties will be conducted in compliance with the Ccnmission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: March ll,1981.
_ _.... _ _ ~
1 I J
References:
l.
Letter from J. H. Heider (VYNPC) to Of fice of Nuclear Reactor Regulation of U. S. Nuclear Regulatory Conmission, February :2,1981.
2.
License No. DPR-28 (Docket No. 50-271).
3.
General Electric Company, " Vermont Yankee Nuclear Power Station Proposed i
Stability and Recirculation Pump Trip Test," 'iEDC-24279 80 NED 283, January 1981.
4.
General Electric Corpany, " Supplemental Reload Li:ensing Submittal for Vermont Yankee Nuclear Power Station Reload No.
7," Y1003J01A02, July 1980.
5.
General Electric Company, " General Electric Reload Application,"
NEDE-240ll-p-1.
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TABLE 1 Scheduled Paraneters to be Recorded During VY Stability Test LPRM readings APRM readings Jet Pump Pressure Differentials Jet Pump Loop Flows Core Pressure Differential Reactor Vessel Pressure Reactor Vessel Pressure Differential Core Exit Pressures Reactor Feedpump Flows Reactor feedwater Temperature Recirculation Loop Drive Flows Recirculation Loop Temperatures Total Core Flow (iiinimum Filtering)
Total Steam Flow EPR (Electrical Pressure Regulator)
Pressure Controller 7
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