ML20101P046
| ML20101P046 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/04/1996 |
| From: | Gabe Taylor SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| IEB-96-001, IEB-96-1, NUDOCS 9604090247 | |
| Download: ML20101P046 (8) | |
Text
_
S: th Carolina ElIctric & GIs CompIny ry J.T Jenkinsville, SC 29065 Nuclear Operations (803) 345-4344 SCE&G Ascamacompay April 4, 1996 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Gentlemen:
Subject:
VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12 NRC BULLETIN 96-01 CONTROL ROD INSERTION PROBLEMS This letter provides South Carolina Electric & Gas Company (SCE&G) response to requested actions as delineated in NRC Bulletin 96-01. The response and requested information is attached.
I declare that these statements and matters set forth herein are true and correct to the best of my knowledge,information and belief.
Should you have any questions, please call at your convenience.
Ve trul s,
/
Mt Gar)pTN CJM/GJT/ews Attachment c:
J. L. Skolds W. F. Conway(w/o attachment)
R. R. Mahan R. J. White S. D. Ebneter J. l. Zimmerman S. F. Fipps NRC Resident inspector J. B. Knotts Jr.
Dave Campbell (WOG Project Office)
DMS (RC-96-0094)
RTS (IEB 960001) l File (815.02) 00M 9604090247 960404 PDR ADOCK 05000395 O
PDR NUCLEAR EXCELLENCE-A SUMMER TRADITIONI
Documtnt Control Desk IEB 960001 RG-96-0094 4
STATE OF SOUTHCAROLINA TO WIT :
COUNTY OF FAIRFIELD Ihereby certify that on the
&wf day of /l*2 19 94,before me, the subscriber a Notary Pubhc of the State of South Carolina personally appeared Gary J. Taylor,,
being duly sworn, and states that he is Vice President, Nuc., ear Operations of the South Carolina Electric & Gas Company, a corporation of the State of South Carolina, that he provides the foregoing response for the purposes therein set forth, that the statements made are true and correct to the best of his knowledge, information, and belief, and that he was authorized to provide the response on behalf of said Corporation.
WITNESS my Hand and Notarial Seal
- u. [^
g NotaryPublic My Commission Expires Tde 26 2sf
/
Date 4
Document Control Desk
. IEB 960001 RC-96-0094 Page 1 of 6 South Carolina Electric & Gas Company Response to NRC Bulletin 96-01 NRC Bulletin 96-01 is applicable to the Virgil C. Summer Nuclear Station (VCSNS) l since the facility is a Westinghouse designed pressurized-water reactor (PWR) l potential impacted by recent industry events. The following response is provided to each o he requested actions contamed in the bulletin:
l e
Promptly inform operators of recent events (reactor trips and testing) in which control rods did not fully insert and subsequently provule necessary training, including simulator drills, utilizing the required procedures for responding to an event in which the control rods do not fully insert upon reactor trip (e.g.,
boration of a pre-specified amount).
l Operators at the Virgil C. Summer Nuclear Station (VCSNS) have been informed of the recent events durin shift meetings. Additionally, this NRC bulletin was added to the March requi readinct program to further ensure that all i
operators were aware of event detai s and actions to be taken by l
Westinghouse PWR licensees.
Nuclear Operations Training currently addresses the issue raised in the NRC bulletin in both classroom and simulator training. Classroom training for Emerclency Operating Procedure EOP-1.1, Reactor Trip Recovery, defines specif c operator actions to emergency borate the Reactor Coolant System by 1500 gallons (approximately 7.5% Boric Acid Tank level) for each control rod not fully inserted into the core when two or more rods are not fully inserted.
The actions required in response to two or more rods not fully inserted following a reactor trip were reinforced during simulator traming while l
performmg LOR-ST-063 in cycle 4 of LOR 95001. This training was completed in l
November and December of 1995.
i j
Additionally, the Nuclear Operations Training department wi!!:
1)
Develop a specific lesson plan for NRC Bulletin 96-01 to be presented during the licensed operator requalification training currently scheduled j
for April 29 through Ma 2,1996. This training will be completed during the refueling outage sc eduled to commence on April 15.
I 2)
Incorporate two or more stuck control rods into a reactor trip simulator scenario for aresentation during the licensed operator requalification program cyc e 8immediate following the upcoming refueling outage.
This simulator scenario will clude discussion to point out that an control rod position indication that is less than rod bottom bistabl is to be interpreted as not fully inserted.
3)
Include EOP-1.1 in the classroom portion of cycle 8 to reinforce the actions j
being taught in the simulator sessions.
- Document Control Desk IEB 960001 l
' RG-96-0094 Page 2 of 6 Promptly determine the continued operability of control rods based on current e
information. As new information becomes available from plant rod drop tests 4
and trips, licensees should consider this new information together with data already available from Wolf Creek, South Texas, North Anna, and other l
industry experience, and make a prompt determination of control rod j
operability.
3 Current rod drop data has been reviewed and it was confirmed that there is no i
indication of degradation. Rod drop times taken at the beginning of the current fuel cycle were all under the Technical Specification limit of 2.7 seconds from beginning of decay of stationary gripper coil voltage to dash aot entry.
i
~
Additionally, there has been no observed problems (indicative of c ragging) during the performance of the monthly RCCA stepping tests or manual 1
insertion of rods during the May 1995 maintenance outage.
1 As new information becomes available it will be evaluated to determine if i
there is an impact on the operability assumptions for the control rod system at i
VCSNS. The Westinghouse Owners Group (WOG) is currently performing root j
cause analyses on the noted industry failures and it is expected that all pertinent information from VCSNS testing and industry experience will be shared.
e Measure e.nd evaluate at each outage of sufficient duration during calendar year 1996 (end of cycle, maintenance, etc.), the control rod drop times and rod recoil data for all control rods. If appropriate plant conditions exist where the vessel head is removed, measure and evaluate drag forces for all rodded fuel assemblies.
a.
Rods failing to meet the rod drop time in the technical specifications shall be deemed inoperable.
b.
Rods failing to bottom or exhibiting high drag forces shall require prompt corrective action in accordance with Appendix B to Part 50 of Title 10 of the Code of Federal Reculations (10 CFR 50).
As requested, rod drop tests will be performed at VCSNS for all outages of sufficient duration in 1996. SCE&G considers an outage of sufficient duration i
to be a maintenance outage of greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or a forced (reactor trip) outage where the plant is not actively trying to restart following shutdown.
The only exception would be, as discussed between the NRC and WOG representatives on March 25,1996,where rod drop tests would not provide meaningful data (e.g., tests within 2,500 MWD /MTU burnup of the last test).
Rod drop tests performed at VCSNS measure the time interval from the first reactor trip breaker opening to the rod bottom bistable indicating full insertion of all rods. This test does not provide recoil data; however, the test actually measures a more conservative time than the time from beginning of decay of stationary gripper coil voltage to dashpot entry required by Technical Specifications. A review of all previous rod drop tests, including specific rod drop tests performed on the current generation of fuel assemblies (VANTAGE + and Performance +) with burnups up to approximately 25,000
Doqumsnt Control Desk lEB 960001
RC-96-0094 Page 3 of 6 MWD /MTU, has found no indication of insertion problems at VCSNS. It should be noted that the fuel design for VANTAGE + and Performance + fuel incorporates a number of different design characteristics from VANTAGE SH and Standard XLR fuel types. These characteristic differences may preclude similar occurrences to those experienced at the utilities referenced in the subject bulletin.
SCE&G supports the WOG position that recoil data is redundant to the preferred c rag tests and proposes to perform only the drag tests during the upcoming refueling outage scheduled to commence on April 15,1996. Current procedures and practices for rod drop testing performed at VCSNS do not mclude provisions for obtaining recoil data; therefore, to change procedures, practices, and test equipment immediately prior to a scheduled outage without i
pnor recoil data for comparison is not deemed prudent at this time. During the refueling outage, drag tests will be performed on all rodded fuel assemblies with greater than 40,000 MWD /MTU assembly burnup alonc with a representative sample of those rodded fuel assemblies with surnups in the range of 25,000 to 40,000 MWD /MTU. Rodded fuel assemblies with fuel assembly burnups less than 25,000 MWD /MTU will be excluded from drag tests based on rod drop test results and operational observations discussed above.
The results of the upcoming rod drop and drag tests will be evaluated by SCE&G to determine the condition of the control rods. Individual rod drop times will be obtained, as necessary, should there be any observed rod insertion problems during the rod drop test. Corrective actions will be initiated to correct any observed problems. The test results will be provided to the NRC within 30 days of completing the actions for each outage.
e For each reactor trip d uring calendar year 1996, verify that all control rods have promptly fully inserted (bottor.w<l) and obtain other available information to assess the operabili'v and any performance trend of the rods.
In the event that all rods do not fully insert and evaluate rod drop times and rod recoil.promptly, conduct tests to measure Actuation, currently requ,ing Procedure EOP-1.0, Reactor Trip / Safety I jectio VCSNS Emergency Operat ires operators to verify that all rod bottom li hts are lit following each reactor trip. This verification is made immediately f Ilowing each event. Post trip reviews of the Sequence of Events (plant computer) printout would also note times for reactor trip breakers opening and "all rods on bottom."
In the event that all rods do not fully insert promptly, SCE&G will perform rod drop tests, as necessary, to gather information to assess operability.
Within 30 days of the date of this bulletin, provide a core map of rodded fuel e
assemblies indicating fuel type (materials, grids, spacers, guide tube inner diameter)and current and pro assembly for the current cycle;jected end of cycle burnup of each roddedwhe for the next cycle.
Attached is a current core map for cycle 9 and a projected core map for cycle 10 along with pertinent information on the fuel assemblies used at VCSNS.
Document Control Desk IER 960001 RC-96-0094 Page 4 of 6 FUEL ASSEMBLY DESCRIPTION i
REGION ID 10A/10B 11 12 Assembly Type VANTAGE +
Performance +
Performance +
Fuel Rod Array 17x17 17x17 17x17 Clad Material Zirlo" Zirlo" Zirlo" I
Guide Tube Material (24)
Zirlo" Zirlo" Zirlo" t
Guide Tube Dimensions Upper Part OD (in.)
.474
.474
.474 Upper Part ID (in.)
.442
.442
.442 Lower Part OD (in.)
.430
.430
.430 i
lower Part ID (in.)
.397
.397
.397 Instrument Tube Material (l)
Zirlo" Zirlo" Zirlo" Instrument Tube Dimensions:
OD (in.)
.474
.474
.474
)
ID (in.)
.442
.442
.442 j
Grid Material Inner Structural (6)
Zr-4 Zirlo" Zirlo" End Structura1(2)
Inconel Inconel Inconel IFM (3 )
Zr-4 Zirlo" Zirlo Protective (l)
N/A N/A Inconel Nozzle Material Top Stainless Steel Bottom Stainless Steel a
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EOC BURNUF (MWD /MTU) - PROJECTED 15 1
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