ML19331C310
| ML19331C310 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 07/13/1980 |
| From: | Heider L VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| IEB-80-17, WVY-122, NUDOCS 8008140531 | |
| Download: ML19331C310 (8) | |
Text
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VERMONT YAN KEE NUCLEAR POWER CORPORATION SEVENTY SEVEN GROVE STREET B. 4.1.1 Ru rl.AND, VERMONT 05701 WVY 80-100 REPLY TO:
ENGINEERING OFFICE TURNPIKE RO AD WESTBORO. M ASS ACHUSETTS 01581 TELEPHONE 417-344-909I July 13, 1980 United States Nuclear Regulatory Commission Office of Inspection and Enforcement Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 Attention:
Mr. Boyce H. Grier, Director
Reference:
(a) License No. DPR-28 (Docket No. 50-271)
(b) USNRC Letter to VYNPC dated July 3, 1980; IE Bulletin 80-17 (c) General Electric Letter, F. P. Felini to all BWR Plant Superintendents dated July 7, 1980 (d) VYNPC Letter (WVY 80-98) to USNRC dated July 8, 1980
Subject:
Response to IE Bulletin 80-17; Failure of Control Rods to Insert During Scram
Dear Sir:
As a result of a situation at Brown's Ferry Unit No. 3, wherein a number of control rods failed to fully insert following a manual scram attempt, all General Electric Boiling Water Reactor licensees were requested to take certain actions.
Reference (d) documented our response to Bulletin Item #1; l
the testin,. required by Items 2 and 3 was completed on July 12 and 13 and the l
results of that testing will be submitted within the timeframe specified in the Bulletin. Our responses to Items 4, 4e, 5, 6a, 6b, 6c and 7 are as I
follows:
Item 4:
"Within 10 days, complete a review of emergency procedures by the licensee and the NSSS vendor to assure that, for scram, operator actions include:..."
Re spons e :
A review of operating procedures has been completed.
Revisions have been made to the procedures to include the NRC recommended actions.
l 800814o 5 3 i
Unitad Stetes Nuclect Riguletcry Commission July 13, 1980 Attsntion:
Mr. Bcyca H. Grier, Dirsctor Pega 2 Item 4e:
" Review the Brown's Ferry occurrence with all licensed operators and train them in the procedures to recognize and mitigate the event.
Verify that preliminary training of operators is completed within ten days of the date of this Bulletin..."
Response
All available operators have received the required preliminary training within the prescribed timeframe.
For those operators that were absent during the time period, the necessary training will be provided prior to their assuming their shif t responsibilities.
Item 5, 6a & 6b:
5
" Review and develop surveillance procedures such that scram discharge volume is monitored daily for residue water for 6 days and, if results are acceptable the interval may be extended to 7 days 6a Prompt notification (within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) of any of the above systems when it is less than fully operable and when it is restored to service.
Operability of both pumps in the SLCS is required for full operability.
Surveillance tests and preventive maintenance less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> need not be reported.
6b Operate all the available suppression pool cooling whenever the suppression pool exceeds the normal operating temperature limit"
Response
Station operating procedures have been revised, or new procedures generated, to address the requirements of Items 5, 6a and 6b.
Item 6c:
" Perform a 50.59 review to increase SLCS flow to the maximum consistent with safety (2 pumps, unless unsafe)."
Response
General Electric has recently performed a generic review of the Standby Liquid Control System (SLCS).
In Reference (c), GE identifies four I
potential unreviewed safety questions involved with simultaneous SLC pump operation; these concerns are summarized as follows:
l 1.
" Currently, the SLCS is designed for single pump operation. The system selector switch precludes operation of two pumps simultaneously. Any design change to enable simultaneous operation must be reviewed to determine any impact on plant reliability.
2.
Running of two pumps simultaneously could reduce the available NPSH to below that required by the pump design.
New NPSH calculations should be performed prior to making system modifications for simultaneous two pump operation.
United Sect:0 Nuclect Ragulatory Commission July 13,' 1980 Actsntion:
Mr. Boyce H. Grier, Dirsctor Pega 3 3.
It is recommended that operational data be obtained for 2 pump operation which demonstrates that the pump discharge accumulators are effective in dampening the positive displacement pump surges.
Two pump simultaneous operation will increase the flow rate in the SLCS which will increase the frictional pressure drop in the discharge piping.
Proper functioning of the pump discharge relief valve requires confirmation of this mode of operation. Also, the events for which the system is to be available require definition so that the associated reactor pressure can be determined.
In addition, if the relief valve opens, SLCS flow will recycle resulting in reduced system effectiveness."
As an initial step in performing an independent review we have evaluated GE's concerns.
Our preliminary calculations indicate that if two SLC pumps are run simultaneously, no hydraulic problems (i.e., NPSH, relief valve lif ting) are expected as long as the reactor is below approximately 1160 psig. At pressures above 1160 psig pressure, we agree with GE's statement regarding the uselessness of operating the second pump due to relief valve lif t. We agree with GE that the effectiveness of the accumulators in dampening pulsations from two pump flow cannot be determined analytically and would require actual test data.
Secondly, the safety design bases and safety evaluation contained in the Vermont Yankee FSAR were reviewed. A statement in the safety evaluation specifies that boron injection rate be limited to the range of 8 to 20 ppm per minute.
The upper limit injection rate was established to assure "that there is suf ficient mixing so the boron does not recirculate through the core in uneven concentrations which could possibly cause the nuclear power to rise and fall cyclically." There are certain situations when the upper limit may be exceeded with two pump flow; however, we have been unable to quantitatively determine the effect on core stability.
Also, it should be remembered that the existing Standby Liquid Control System did undergo extensive review and was specifically installed for situation's where no control rods could be inserted into the reactor core to accomplish shutdown and cooldown in the normal manner, a situation cuch more degraded than that which occurred at Brown's Ferry.
For the above reasons, we have concluded that the SLCS should not be modified to allow two pump operation.
Item 7:
"For plants without ATWS related RPT, perform an analysis of the net safety of derating such that, in the event of an ATWS, calculated peak pressures do not exceed the service level "C" Limit ( /1500 psig) by taking into consideration the heat removal capability of safety valves, isolation condenser, bypass to the main condenser and other available heat removal systems."
,_y
Unit d Statoo Nucle r Regulctory Commicoien July 13, 1980 Attsnticn:
Mr. Boyca H. Grier, Director Pega 4
Response
Attached are the results of a transient analysis performed by General Electric of 1) a generic bounding case for MSIV closure with scram of all rods in a_1800 sector of the core, and 2) a plant specific case for turbine trip with bypass with no scram.
We have been made aware of discussions held between General Electric and the NRC during which it was agreed that these analyses adequately address the Bulletin requirement. We share General Electric's belief that basing decisions relative to plant safety on a complete failure to scram does not properly reflect the occurrence at Brown's Ferry and therefore, fully support the GE/NRC agreement. Furthermore, because of the differences in design between the Vermont Yankee and Brown's Ferry control rod drive scram systems, the adequacy of the Vermont Yankee system as demonstrated by testing done to date, and our commitment to install the recirculation pump trip modification during the September refueling outage, we feel that continued full power operation is justified.
We trust that the above supplied information is satisfactory; however, should you desire additional information, please contact us.
Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION Y
L L. H. Heider Vice President RJW/kaf Enclosures COMMONWEALTH OF MASSACHUSETTS)
)ss COUNTY OF WORCESTER
)
Then personally appeared before me, L. H. Heider, who, being uuly sworn, did state that he is a Vice President of Vermont Yankee Nuclear Power Corporation, that he is duly authorized to execute and file the foregoing request in the name and on the behalf of Vermont Yankee Nuclear Power Corporation, and that the statements therein are true to the best of his knowledge and belief.
A.< daff Lib i.
i Renee M. Kossuth Notary Public My Commission Expires May 18, 1984 y
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7-I' RESPONSE TO IE BULLETIN 80-17
(
ATWS WITHOUT RPT FOR VERMONT YANKEE
[
Introduction This document provides,.the results of the evaluation of anticipated transients without scfam (ATWS) without recirculation pump trip (RPT) as required by Item 7 of IE Bulletin 80-17.
Based on discussions with the NRC, an assessmint of a full ATWS in plants not having RPT implemented is required as part of an analysis of the net safety of derating plants such that calculated peak vessel pressures do not exceed the assumed
" Service Level C" limit of 1500 psig considering all available heat removal systems.
This evaluation was provided to Yankee Atomic Electric Company for Vermont Yankee by the General Electric Company.
Discussion General Electric believes that basing decisions relative to plant safety on a complete failure to scram does not properly reflect the occurrence at Browns Ferry Unit 3.
It should be noted that the initial partial scram at Browns Ferry Unit 3 resulted in a power reduction from approx-imately 36% to less than 1%.
A conservative evaluation of the Browns Ferry 3 occurrence has been performed by GE for plants which do not have recirculation pump trip incorporated in their design.
These analyses indicate that the scram of 50% of the control rods will effectively mitigate the consequences of anticipated transients.
In light of the above discussion and in response to Bulletin 80-17 two ATWS transients are presented.
These transients are:
- 1) a generic bounding case for MSIV closure with scram of all rods in a 180* sector of the core, and 2) a plant specific case for turbine trip with bypass with no scram as required by IE Bulletin 80-17.
MSIV Closure A generic bounding case was analyzed in which end of equilibrium cycle core conditions were assumed and that only control rods in a 180* sector of the core are inserted during scram.
The control rods in the other half of the core were assumed to remain in the full power position.
General Electric believes this case bounds any possible non-detectable water accumulations in the scram discharge volume which are not detect-able with the current instrument configuration.
For this evaluation the control rods were separated into functional and non-functional 180* sectors of the core.
Under these conditions the reactor power was conservatively calculated to fall to 40% in the first 70 seconds.
A bounding analysis of the peak reactor pressure for the postulated half scram condition was performed for a MSIV closure in a plant with the following characteristics:
R sp:nse to IE Bulletin 80-17 Paga 2
(.
Initial Power Level 100%
Scram Worth
-3$
Void Coefficient
-11 $/%
Safety Valve Setpoint/ Capacity 1255 psia /16% NBR Relief Valve Setpoint/ Capacity 1110 psia /40% NBR
~
The results of this analysis show that the peak vessel pressure (without RPT) is less than 1460 psig at 47 seconds.
Based on the above itds concluded that for a conservatively defined partia.1 scram condition in plants without RPT and with combined safety and relief valve capacity of 56% NBR, the peak pressure is maintained well below 1500 psig.
The safety and relief valve capacity and reactor vessel size used in this assessment is small compared to operating BWR's which do not incorporate RPT, thereby maximizing the peak vessel pressure.
In additio.n, a conservative void coefficient was used.
Previous sensitivity studies hdve shown that this combination of parameters is a limiting case for operating BWR's without RPT and hence it can be concluded that this generic analysis indeed bounds the results which would be obtained for individual plants.
Turbine Trip With Bypass A plant specific analysis of the turbine trip with bypass transient for which no scram occurs has been performed for Vermont Yankee.
The input
(
parameters for this analysis are given in Table 1.
No credit is taken for heat removal systems other than the safety and relief valves, and/or the turbine bypass to the main condensor.
The results of this analysis show that the peak vessel pressure reaches 1064 psig in 60 seconds for full power operation.
The transient response of the system is shown in Figure 1.
I Conclusion Based on the above evaluation no plant derates are necessary to meet the l
1500 psig limit.
The conservative bounding MSIV half scram evaluation shows that the 1500 psig limit is not exceeded.
The plant specific i
analysis of turbine trip with bypass shows that the 1500 psig limit is not exceeded for the very conservative case of no scram.
Therefore, it can be concluded that continued operation of Vermont Yankee without ATWS l
RPT is not an unreviewed safety question and does not produce a safety l
hazard to the general public.
RTH:ggo/2-3 l
7/10/80
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i TABLE 1 8
Transient Input Parameters
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Power Level (swt) 1593 6
Rated Core Flow (10 lb/hr) 48.0 Rated Steam Flow (106'lb/hr) 6.4 7
Steam Dome Pressure (psig) 1020 Turbine Bypa'ss Capacity (% rated steam flow) 110 Number of Relief Valves N/A Setpoints (psig)
Capacity (% rated steam flow at setpoint)
Number of Safety Valves 2
Setpoint (psig) 1242 Capacity (% rated steam flow at setpoint) 28.2 Number of Safety / Relief Valves 4
Setpoint (psig) 1080 Capacity (% rated steam flow at setpoint) 48 Void Fraction (%)
40 Void Coefficient (-C/% Rg) 10.0 l
Doppler Coefficient (-C/'F) 0.29 RTH:ggo/4 7/10/80 l
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