ML20101N299
| ML20101N299 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 04/03/1996 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| IEB-96-001, IEB-96-1, VPNPD-96-019, VPNPD-96-19, NUDOCS 9604080334 | |
| Download: ML20101N299 (7) | |
Text
.
s
' Wisconsin 1
l Electnc POWER COMPANY l
l 231 W Mchigart PO Box 2046. Mdwoukee. wt 53201-2046 (414)221 2345 VPNPD-96-019 10 CFR 50.54(f)
April 3,1996 U.S. NUCLEAR REGULATORY COMMISSION ATrN: Document Control Desk i
Washington, DC 20555-0001 Gentlemen:
DOCKETS 50-266 AND 50-301 30-DAY RESPONSE TO NRC BULLETIN 96-01 CONTROL ROD INSERTION PROBLEMS P_OINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 O
In accordance with 10 CFR 50.54(f), we submit the following 30-day required response (" Required Responses" I and 2) to NRC Bulletin 96-OL Within 30 days of completing the control rod testing conducted during the present Unit I refueling outage, we will submit a response pursuant to " Required Response" 3 of the Bulletin.
- 1. EF. PORT CERTIFYING THAT CONTROL RODS ARE OPERABLE We have neviewed test data and operational history of all rodded fuel assemblies in our existing cores. Considering the normal control rod behaviors, the available shutdown margins, and the industry experience described in the Bulletin, we have concluded that cach control rod in both units has been operable and remains operable. This operability determination is based on:
- Acceptable Rod Drop Times During Testing and No Adverse Trends Over Fuel Lifetime
- Acceptable Rod Recoil Characteristics During Testing
- Acecptable Rod Drag Forces During Testing
- Acceptable Performance During Rod Stepping Tests
- Acceptable Rod Characteristics During Recent Trips and Rod Shullies Acceptahic Rod Drop Times During Testing and No Adverse Trends Over Fuel Lifetime - We have reviewed existing Beginning of Cycle ('BOC) rod drop data of both units for the last five years. In addition, we performed special End of Cycle (EOC) rod drop testing for Unit 1 on March 30,1996. We found no drop times in excess of the Point Beach Technical Specification limit (2.2 seconds). Also, we found no significant differences between EOC rod drop times and 130C rod drop times; indicating no correlation between fuel assembly burnup and the associated rod drop times.
As identified in the attached core map, the Unit 1 EOC core had nine (9) rodded fuel assemblics with burnups greater than
-10.000 MWD /MTU, with one assembly exceeding 52,000 MWD /MTU. Considering that Westinghouse Owner's Group has identified high fuel burnup (>40,000 MWD /MTU) as a potential contributing factor to incomplete control rod insertion to be investigated, we compared beginning oflife (burnup equal to 0) rod drop times to the final EOC drop times for a naiple of high burnup fuel assemblics. This comparison demonstrated that the drop times for these assemblics did not change over the life of the assembly, nOOO97 9604080334 960403 I
{DR ADOCK 05000266 PDR A subsWmofXiswnsm Exy(hrim&m
U.S. Nuclear Regulatory Conunission April 3,1996 Page 2 Acceptable Rod Recoil Characteristics During Testing - We have resiewed existing BOC rod drop data of the last five years to evaluate the rod recoil characteristics of both units. In addition, we have reviewed rod recoil data from the special EOC rod drop testing for Unit 1. The EOC Rod Position Indicator (RPI) traces were compared to the BOC traces to verify the presence of rod recoil. We have determined that all rods have exhibited acceptable recoil characteristics; indicating that all rods have reached the bottom of the dashpot region with excess energy.
Acceptable Rod Drag Forces During Testing - Beginning-of-Cycle rod drag force testing is performed routinely on both Point Beach units aner the control rods are relatched during an outage. All test results have satisfied Westinghouse guidelines. There have been no problems noted with recent performances of this test.
Acceptable Performance During Rod Stepping Tests - We have reviewed the data from Point Beach Unit I and 2 " od stepping" tests to help determine acceptable control rod performance in both units. Rod stepping tests consist of moving the control rods out and in, and observing the current traces supplied by the rod control system power cabinets to the Control Rod Drive Mechanism (CRDM) coils. These tests are typically performed during startup, following each refueling. These tests are performed with the rods withdrawn five steps, which is within the dashpot region of the control rod guide tubes (the region where most control rod sticking problems have occurred at other plants). We believe that an adverse signature may predict drag forces that could lead to a stuck rod. Current traces were resiewed and found normal, indicating no control rod sticking or excessive drag forces during control rod step testing.
Acceptable Rod Characteristics During Recent Trips and Rod Shuffles - For both PBNP units, we resiewed post-trip l
data from the last five fuel cycles. There were no reports of incomplete rod insertion during any of these events. In addition to reactor trips, other control rod manipulations such as control rod shuflies have been conducted without abnormal l
rod insertion. There have been no instances of sticking control rods during recent performance of these activitics.
Adequate Shutdown Margin Even If An Entire Shutdown Bank Stuck Above The Dashpot Region - For normal operating conditions, current core designs typically have at least 1% Ap of excess shutdown for the most restrictive condition, which occurs at EOC. If the entire most reactive bank of control rods (Control Bank A) were to stick at 40 steps withdrawn (which is above the dashpot region), the amount of reactivity added would only be about 374 pcm, or 0.374% Ap. Therefore, the required shutdown margin is maintained for sticking rods assuming the worst case normal l
conditions.
For accident conditions, the least excess shutdown margin occurs at EOC for the steam line break accident, where there is typically less than 200 pcm of excess shutdown margin. In the analysis of this accident, it is assumed that the most reactive I
control rod is stuck full out. The reactivity of this stuck rod is usually at least 650 pcm, and often much greater. It is not credible to assume that the most reactive rod will stick fully out and that a group of rods will fail to fully insert, therefore only one of these failures need be considered at a time. Since the worth of the fully out stuck rod is greater than the worth of rods which failed to fully insert, the excess shutdown margin will be greater with a failure of rods to fully insert that with a fully out stuck rod.
- 11. ACTIONS TAKEN IN RESPONSE TO "REOUESTED ACTIONS" 1 AND 2 l
Requested Action 1: Promptly inform operators of recent events (reactor trips and testing) in which control rods did not fully insert and subsequently provide 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.
I Response: Licensed operators of Point Beach are routinely trained for rod control casualties. During Requalification Cycle 1 (January / February 1996), licensed operators received training on rod control casualties and the associated abnormal operating procedures. Additionally, entry into the emergency operating procedures for rod control anomalies was covered.
These topics are a component of our Probabilistic Safety Assessment (PSA) based long range training plan.
I
U.S. Nuclear Regulatory Commission April 3,1996 Page 3 Specifically with regards to NRC Bulletin 96-01, all crews have been made aware of the components of the bulletin. Each licensed operator and operations trainer has been required to self-study a detailed required-reading package addressing the identified issues. This training includes the use of procedures for responding to an event in which the control rods do not fully insert upon reactor trip. All licensed operators and trainees in our six (6) operating crews have completed the self-study / required-reading guides. Additionally, each operating crew will receive a simulator session which specifically addresses post-trip stuck rod scenarios. We plan to complete these training scenarios by July 1,1996. Any additional experience gained from the Bulletin activities will be incorporated into the licensed operator training courses.
Beauested Action 2: Promptly determine the continued operability of control rods based on current information. As new information becomes available from plant rod drop tests and trips, licensecs should consider this new information together with information already available from Wolf Creek, South Texas, North Anna, and other industry experience, and make a prompt determimtion of control rod operability, i
Response: As discussed previously, we have completed the operability determination of PBNP control rods based on current information. We will be closely monitoring our test data and revising test criteria in consideration ofindustry experience.
To ensure that appropriate post-trip data is recorded, evaluated, and reported, we are planning a permanent revision to our post-trip procedure (ECL-5," Post-Trip Reviews"). As a member of the Westinghouse Owner's Group, we will be closely monitoring industry experience in this area and assessing our control rod operability accordingly.
)
111. PLANS FOR IMPLEMENTING "REOUESTED ACTIONS" 3 AND 4 Eeauested Action 3: Measure and evaluate at cach outag of suf11 dent duration during calendar year 1996 (end of cycle, maintenance, etc.), the control rod drop times and rod recoil d:aa 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 be: tom or exhibiting high drag forces shall require prompt corrective action in accordance with Appendix B to 10 CFR 50.
Response: For each 1996 outage of sufficient duration, we plan to measure and evaluate the BOC and EOC control rod drop times and rod recoil characteristics for all control rods. When the reactor vessel head is removed, we plan to measure and evaluate BOC drag forces for all rodded fuel assemblics. These tests are planned for the Unit I spring refueling outage (UiR24) and a similar testing regimen is planned for the Unit 2 fall refueling outage (U2R23). Within 30 days of completing the last test of each outage, we will report the outage tew results to the NRC. As discussed previously, recent EOC rod drop tests for PBNP Unit I were acceptable.
In addition, we plan to conduct testing as a WOG test facility. Following completion of the present Unit I outage, Westinghouse will assist us in conducting End-of-Life control rod drag force testing on some of the highest burnup fuel assemblies in the Spent Fuel Pool.
Reauested Action 4: For cach reactor trip during calendar year 1996, verify that all control rods have promptly fully inserted (bottomed) and obtain other available information to assess the operability and any performance trend of the rods.
In the event that all rods do not fully insert promptly, conduct tests to measure and evaluate rod drop times and rod recoil.
Response: For cach 1996 reactor trip, we will verify that all control rods have promptly and fully inserted (bottomed), as currently required by procedure. We plan to revise our post-trip procedures as necessary to clarify the acceptance criteria l
for " rod bottom" Control rod operability will be assessed after each reactor trip. Operability criteria will be primarily l
based on prompt rod insertion upon demand. In the event that any control rod does not fully and promptly insert, we will conduct rod drop tests and rod recoil tests on each control rod in that core. These test data, in combination with control rod exercise test data, will be evaluated to determine control rod operability prior to re-starting the reactor.
o l
4 i
U.S. Nuclear Regulatory Commission i
April 3,1996 Page 4 IV. CORE MAP OF RODDED FUEL ASSEMBLIES A core map of rodded fuel assemblies for each Point Beach Unit is attached. This attachment indicates fuel type (materials, i
grids, spacers, guide tube inner diameter), and current and projected end of cycle burnup of each rodded assembly for the
)
l current cycle. When it becomes available, we will provide the same type ofinformation for the next cycle.
1 l
)
Please contact us if there are any questions.
Sincerely, I
Bob Link Vice President Nuclear Power Subscribed and sworn to before me
' on this 3rd day of April,1996, t N u Lki.
6'U2 h ar@blic, State of Wisconsin My commission expires 10 9h GDA
)
Attachment ec:
NRC Regional Administrator NRC Resident Inspector Public Senice Commission of Wisconsin Public Senice Commission of Wisconsin (Attn: Paul Kitzenbel) l i
s a
i PBNP UNIT I CYCLE 23 CURRENT AllD PROJECTED
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16.005
,16,661 b M1 tim 104 M16fR136 D
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N.778 29.000 E0171R146 AA28m86 AA18m02 S018/R124 E
'l 3 5,I79. l 27.798 27.748
' ! I5.197 l E19 598 -
59.947 M,006
'E16,605!
0025/R100 0326/R$1
!I2,'919
{l3 $$
F
_11u!_E __
__ E _ti m __
J DB271RI42 B02ilRi06 m
l 12,814 113.064!
- 14.000
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[s s.224 j
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29.805 30.294
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(15,28 7 L N6.1 M IL TsMsC I i6.tip7 BQ28/RBS.
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ft 2,902 l-82587 t
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14.179' l
l M
FMFL ASSEIIBLY # RECA ID fMFL TYPE.
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$[E AllAEHED SHEET 2
PROJECTED EUL BWHEIIP lCOLSES BASED WII THit NUIIRER)
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- PRIMECif D EDL BWIINIIP CALCULAIED BY EMIRAPULAINIE f t SRUAllY PBSURN DAIA IIllipuGN INE FS f uLL POWER DAY 5 E.MPECTED IN MARb81 125ti IKl/22/95
9 PBNP UNIT 2 CYCLE 22 CURRENT AelD PRCAIECTED* EOL BURNUP (FOR RODDED FAAs ONLY)
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j 3,618, p 17.480 17.900 y 3,844 i
'12.488-26.885 27.546 12.883 AA70lR602 AA561R107 AAL6/R126 AArWR113 F
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- :3.886 ::
I.~ 8.221 :
11.031!
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FIIEE ASStaleLY # REEA IB FUEL TYPE.
Y CURREMI gueuen" taASED De FEBRUAAY PSOUNE DATAl SEE AIIAEIIED SHEf f 2
FIISJECTED Els1 punBUP lCELORS BASES Su IHIS IMMIBERI MIE.
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(% DeK 4% DM esK.s2K c2eK
- PROJECIES EBL BWAIIMP CALOMATED BY EXIRAPOLAIINE flBAUARY PBOURII DAIA IHROUEN INI Me IIAININE IIS tUll POWER DAYS 1253 EF22/95
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POINT BEACH NUCLEAR PLANT : UNIT 1 AND 2 l
FUEL TYPE:
WESTINGHOUSE, PWR, RECONSTITUTABLE,14 X 14, OFA MATERIALS:
TOP NOZZLE 304 SST BOTTOM NOZZLE 304 SST SPRING SET INCONEL - 718 i
GUIDE THIMBLE ZIRCALOY - 4 INSTRUMENT TUBE ZIRCALOY - 4 GRID ASSEMBLY INCONEL - 718 (TOP AND BOTTOM GRIDS)
GRID ASSEMBLY ZIRCALOY-4 (MIDDLE GRIDS) i ADAFTOR PLATE 304 SST NOZZLE INSERT 304 SST l
l LOCK TUBE 304 SST GUIDE TUBE INNER DIAMETER:
ABOVE DASHPOT 0.492" BELOW DASHPOT 0.4465" 1
l
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.:.. :s 0*/09,96 08:28 FAI 412 643 5018 w BEAVER VALIIY m
MllR-08-1996 20827 NRC IEpiR VALLEY 412 E83 2 W r.
OMB flo. 3150-0012 NRC8 95-02 f./. /.!
NUCLEAR REGULATORY CG903SION 4
[#
UNITED STATFS p
0FFICE OF NUCLEAR REACTOR RECULATION
~
WA$HINGTON, DC 20555-0001 March 8, 1996 NRC SULLETIN 94.Cl CONTACT. R00 IHsERTION pro 8LEMS Addreasses
,j This bulletin is being sent to all holders of pressurized-water reactor (PWR) operating licenses (except these licenses amended to passession only status).
It ta expected that recipients will review the information for applicability to their facilities and ccasider actions, as appropriate, to avoid sim11st problems.
However, action is only requested from PWR licensees of Westinghouse-designed plants.
Pursene The U.S. NRC is issuing this bulletin to acconiplish the following:
(1)
Alert addressess to problems encountered during recent events in which control rods failed to comp 1stely insert open the scrati signal.
(2)
Assess the operab111ty of control rods, particularly in nign burnup fuel assemblies.
Backaround South Texas Project On December 18, 1995, with South Texas unit I at 100 percent power. a pilct wire monitoring relay actua,W. ciused a atin transformer lockout, which resulted in a tuttine trip and a reactor trip, ithile verifying that control rods had insertad fully after the trip, operators poted that the rod bottom lights of three control rod assemblies were not lit: the digital rod positten incication for each rod indicated six steps withdre.wn. A step is equivalent to 1.59 cm 15/5 inch), and the top of the dashpot begins at la steps.
Scration of the reacter coolant system was cccurring, with the charging pump
)
suction al10ned to the refueling water storage tank. One rod did drift into the fully inserted rod bottoa position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and the other two rods were manually inserted later. During subsequent testing of all contral rods in the affected banks, the rod position indication for the same three locatinns, as well as a new location, indicated six steps withdrawn. As compared to prior rod drop testing, no significant differences in rod drop times were r.oted before reaching the upper dashpet area for any of the control rods. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the red drop tests, two of the rods drifted to the rod bottom position and the other two were manually inserted.. All four M0?!37991
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PfR 9 '56 17:06 l
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'tve-eG-1996 28'18 g m VW 412 W 29e3
% 3 NRCS 96-01 starch 8, 1996 Page 2 of 6 control rods were located in KLR fuel assamblies that bore in their third cycle, with burnup greater than 42,580 megawatt days per matric ton uraniwn (W0/MTU).
Wolf Creek Plant on January 30, 1996 after a manual scrta from 80 percent power, five control rod asseetblies at t$e Wolf treek plant failed to insert fully. Two rods remained at 6 steps withdrawn, two at 11 ste s, and one at 16 steps. At Wolf Creek, a step is equivalent to 1.59 cw I/4 Inth] and the top of the dasheet begins at approximately 30 step 6.
res of tfie affected rods drifted to fully inserted within 20 minutes, one within 60 minutes, and the last one within 78 minutas.
The resulta also indicate that thart was some slowing down of affected rods befers they reached the dashpot.
After the scram, the licensee initiated esergency boration because al*; rods did not insert fully.
During subsequent cold red drop tests, the same five rods, plus an additional three rods, failed tc fully insert. All cf the affected rods were in 17x17 VANTAtt SH fuel assemblies, with turnup greater than 47,600 MWD /MTU.
North Anna Plant on February 21, 1996, during the insert shuffle in preparation for loading Worth Anna 1 cycle 12, two new control rod assemblies could not be removed with normal o,peration of the handling tool from the fuel assemblies in the spent fuel pool in which they were tegorarily stored. The control rod assemblies were rereoved usin with the bridge erane hoist.g the rod assembly handling tool in con.1 unction The two affected fuel assemblies were VANTAGE EH asscmblies, which had achieved 47,782 MWO/KTU and 49.613 WO/MTU turnup during two cycles of irradiation.
Discuse*en At both South Teres units, a 14-foot active fuel length core design is used.
Several differences between the standard it fcot active fuel design and the 14-foot design are as follows: the 14-foot fuel desion is approximately e
76.2 cm [30 inches) langer than the standard fuel assembly design, it has 10 mid grids compared to 8, and the dashpot region is 25.4 cm (10 inches) longer and comprises a double dashpat.
The centrol rod radial clearances abeve and in the deshpot region of the 14-fcot fuel assembly are similar to i
those af the standard design. The South Texas core contains three different 17x17 fuel types--Standard XL, Standard XLA, and VANTAGE 5H--a11 of which are designed sad fabricated by Westinghouse. This was the first operating cycle with VANTAGE SH fuel. The core also contains 57 silver-indtum. cadmium rods.
The four affected rods were found in twice-burned Standard XLR fuel assemblies.
During subsequent testing, the rod drop traces revealed no significant change in dashpet entry time; however, the affected roos did not show recoil on the rod drop trace. Recoil is a dampening affect that is normally seen in the traces as a result of contact of the control rod assenbly spider huD spring against the fuel assembly. When similar rods in Unit 2 were tested, the results revealed no adverse indications. One rod did show the "no recoil" effect but inserted fully into the core, r%A 9 *96 17 06 PAGE.CD-2
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.. -.i 03/09.ss 06:30 FAI 412 643 sole W SEA ER VAL 11Y Q gg 4 cvut-ge-1996 ae510
- AC eEAK'R MEY 412 643 2883 P.en NACB 96-03 l
March 8, 2006 Page 3 of 5 1
i At Wolf Creek, subsequent cold, full-flew testing of all of the sentrol red l
essemblies indicated that eight control rods, including the five control rods that did not fully insert following the reacter trip on January 30, 1995, did not ft,lly insert when tripped.
One control rod in core location H2 paused at i
96 steps streed et 90 steps, and slowly insarted to 30 steps over the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />., The control rod was then manually inserted. The seven other affected rode stepped at vertous heights in the dashpot region, five of which fully inserted within 22 minutes, One of the other two drifted to the bottom within 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, while the reesining red needed to be manually insertec.
The renseining 45 rods fully inserted when dropped, although a number of the rods i
did not exhibit the e acted nuserer of recoils. Of the total 53 control rod assemblies, the essen y at core location H2 (the only rod slowing outside the i
dashpot region) is a hafnium centrol red, while the remaining are silver-i i
Indium-cadelum control red assemblist. liowever, subsequent inspection of the I
i hafniuss rod did not indicate any adverse dimensional change. The licensee i
9 ratested all rods that stuck, as well as those rods that failed to recoil more than twice, and the results were stailar to the previous testing.
At North Anne, the two affecteo control rods were reteoved and were inserted ~
into a series of other fuel assemblies. No aeditional binding problems were a
i observed. However, difficulty was experiences when another control rod was d
inserted into the two affected fuel assemblies. On the basis of this result, the licensee determinec the cause of the einding problem to be related to the fuel assemblies and not the centrol rods. Subsequent control rod drag testing data indicated a correlat1Dn of control rod drag force to essembly burnup and l
'l a significant increase in drag force at assembly burnups greater than J
45,000 MdD/MTU.
4 These three events, as well as several siellar events at foreign reactors, raise concerns about the operability of control rods in high burnup fuel assemblies. Although most of the testing to date has demonstrated that the control rods have mached the dashpot region of the guide tube and that adequate shutdown margin has been maintained, thert have been indicaticns of 1
degraded rod drop times and a stdek rod well abeve the dashpot ragion. Thus, there is concern that these events say be precursors of more significant 4
control red binding problems in which required shutdewn snargins and rod drop times may be violated.
Ratuested Ac11ABA To ensure that the required shutdown margin is tsaintained during a reactor trip, all liesnsees of Westinghouse-designed plants are requested to take tra i
following actions:
(1)
Promptly infor1s operstors of recent events (reactor trips and testing) in which contral rods did not fully insert.and subsequently provide necessary training, including simulator drills, utilizing the required orocedures for responding to an event in which the control rods do not fully insert upon reactor trip (e.g., boratien of a pre-specified aamunt).
r1M 9 '96 17:07
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- sEAtu VALLZY f54HIB-1996 29:19 HRcIEf*4R N Y
% 005 412 643 2003 p.25 MRCS 56-01 March 8, 1996 Page 4 of 5 (z)
Promptly deterutna the continued operability of cent.rol rocs tasse on current information, As new information becomes available from plant red drop tests and trips, licensees should consider this new information together with data already eve 11able from Wolf Creek, South Texas.
North Anne, and other industry expertence, and eake a prompt determination of control red operability.
(3)
Measure and evaluate at each outage of sufficient duration during calendar year 1994 end of cycle, maintenance etc.), the control red drop times and rod r(ecoil data for all contro$ roda.
If kbprooriate i
plant conditions exist where the vessel head is removed, measure and i
evaluate drag forces for all redded fuel assemblies.
j Reds failing to meet the rod drop time in the technical i
s.
specifications shall be deemed inoperIble, b.
Rees failing to Dottom or axhibittr.g high drag forces shall require prompt corrective action in acccrdance with Appendix 5 to Part 50 of Title 10 of the fede of Federal Reculattens (10 CFR Part 50).
(4)
Fcr each reacter trip during calendar year 1994, verify that all centrol rods have promptly fully inserted Jbottomed) and obtain other available information to essess the operabildty and any performance trend of the
- tsds, in the event that all rods de not fully insert preestl tests to measure and evaluate rod drop times and rod recoil. y, conduct
-)
In sunenary, the first two actions requested by the bulletin ensure that all affected plants respond in a proactive manner te recent industry experience.
The second two requested actions support data collection that will parait the staff to more effsetively assess this issue and deterstne if rurther regulatory action is needed.
Reauired Reseenam Pursuant to Section illa, the Ateele Energy Act of 1954, as amended, and 10 CFR 50.54(f), all licensees of Westinghouse-cenigned plants must submit the following written information:
(1)
Within 30 days of the date of this bulletin, a report certifying that contret rods are determined to be operable; actions teken for Requested Actions (1) and (1) above; and the plans for implementing Requested Action (3) and (4).
1 (2)
Within Jo days of the date of this bulletin. a core map of roeded fuel asseabites indicating fuel type (materials, grids, spacers, guide tube inner diameter) and current and pro.iected end of cycle burnup of each redded assembly for the current cycle; when available, provide the same information for the next cycle.
(s) within so days after completing Requested Action (3) for each outage, a report that summarizes tie data and that documents the results obtained:
e i
MR 9*%
17:as PAGE.006
Gard by:. WOG PROJECT OFFICE 4123746144 03/11/96 7:56AM Job 88 Page 6/7 M 3h' M ld N'
03eosess es:31 FAI 412 043 5018 W DEANT.R VALLEY 2006 rWt-et-1996 20: 3 hec WJe4R AR. LEY a12 643 20h3 54 i
March 8, 1996 Page $ of 6 this to also applicabia to sequested Action (4) when any obnormal red beltavlor is observed.
Address the required written infometion to the U.S. Nuclear Regulatory commission, ATTMs Occument Centrol Desk, Washington, DC to355-0001, under oath cr afffrsation. In addition, submit a copy of the report to the appropriate regionel administrator.
Related Gaearic communicattent NAC Information Not. ice 94-11: ' Control Red Insertion Problems"
!.etter froe R.A. Newton (bo6) to Document control Desk; (NAC),
- Response e
l to NAC Questions Concerning Incomplete RCCA Insertion,' February 2J, l
1996.
anekrti otacunnion l
This bulletin is an information request under the prevision of 10 CFR
~
50.54 (f). The objective of the actions requested in this bulletin is to verify that licensees are complying with the current licensing basis for the
{
facility with respect to shutdown margin and control rod drop times. The l
l issuance of tne cultatin is Justified on the basis of the need to verify l
compliance with the current licensing basis with respect to shutdown margin.
l control rod drop times, and proper operator action when conttil rods art not l-prosiptly inserted into the reactor.
P2nanariladuction Act statemant This bulletin contains infomation collections that are subject to the l
Paperwork Redvetion Act of 1995 (44 U.S.C. 3501 et seq.). These infonnation i
collections were approved by the Offica of Management and Budget, approval nusiber 3150-0012, which empires June 30, 1997. The public reporting bunden for this collection of informatten is estimated to average 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> per i
response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. The U.S. Nuclear Regulatory Coeustseien is seeking public comment sa i'ne potential impact of the collection of information contained in the bulletin and on the following issues:
1.
le the proposed collection of information neesssary for the pmper performanca of the functions of the NRC, including whether the infomation will have practical utility?
2.
!s the estiente of burden accurate?
3.
Is there a way to enhance the quality, utility, and citrity of the infersation to be collected?
4.
How can the burden of the col".ection of informatioc.be miniestrad, including the use of automatal collecticn techniques?
l tim 9'%
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, by: WOG PROJEC7 OFFICE 4 2 746144 03/11/96 7:56AM Jo 88 Page 7/7 o,
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OUQ8,08 04:38 FAI OA3 843 sole W BEAST.R VALLEY f ::
i w -se-1996 2032e HE" N UnLLD' g ony 412 643 2ee3 m, g?
NRC8 96-01 March a. 1996 Page 8 of 6 1
Send cements on any espect of this collection of information including i
suggestiens for reducing this burden, to the Information and kecords i
Hanegement tranch, T-6 F33, U.S. Nuclear Regulatory Cosu:1 sion, Washington, DC 20555 0001, and to the Dest Officer, Office of Information and Regulatory Affairs, NEOS-10202 3150-00!!), Office of Management an0 Budget, Washirgton, DC 20503(.
The NRC may not conduct er sponsor, and a person is not required to respond to, a collection of information unless tt dispitys a currently valid OM8 control number.
If you have any questions about this matter, please contact one of the technical contacts listed below or the approprintt Office of Muclear Reactor Regulation (NRR) project manager.
I signed by i
Dennis M. Crutchfield, Otractor
~
i Division of Reacter Program Management Office cf Nuclear Reactor Regulation Tschefcal contacts:
Laurence Kopph IQUt
'301) 415-287 i
onterftet:l t k0nrc. gov
)
Mar aret Chttterton, NRR (30 ) 415-2649 Internet:mscifnec. gov Lead Project Manager: Kristine Thomas, NRR (301) 415-1382 Internet:hmt9nrc. gov i
Attachment:
List of Recently issued NRC Bulletins 4
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