Text

Forwards Response to NRC Bulletin 96-001, Control Rod Insertion Problems
	ML18151A620
Person / Time
Site:	Surry, North Anna
Issue date:	04/08/1996
From:	Ohanlon J; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	96-135, IEB-96-001, IEB-96-1, NUDOCS 9604100177
	Download: ML18151A620 (56)
	v • d • e

(

I I I~

.a e

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 8, 1996 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555-0001 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY Serial No.96-135 NL&OS/GDM R1 Docket Nos. 50-280 50-281 50-338 50-339 License Nos. DPR-32 DPR-37 NPF-4 NPF-7 NORTH ANNA AND SURRY POWER STATIONS UNITS 1 AND 2 NRC BULLETIN 96-01 CONTROL ROD INSERTION PROBLEMS On March 8, 1996, the NRC issued Bulletin 96-01, entitled "Control Rod Insertion Problems." The bulletin requests licensees to address four requested actions for their

-- licensed facilities and to provide a written response describing both the actions taken and planned.

We have revie"!\\fed the bulletin requirements and have initiated actions to ensure that any anomalous control rod behavior would be promptly identified and corrected. We have also evaluated the operability of the control rods at North Anna and Surry Power Stations and have determined that they are fully operable throughout their respective operating cycles.

Furthermore, control rod drag testing was already performed for North Anna Unit 1 during the recently completed refueling outage. We will also perform control rod testing for Surry Unit 2 during the upcoming Spring 1996 refueling outage and for North Anna Unit 2 during the Fall 1996 refueling outage. Surry Unit 1 and North Anna Unit 1 are not currently scheduled to be shut down for the remainder of 1996.

However, consistent with bulletin requirements, control rod testing for any unit would be performed should they be shut down for reasons other than the currently scheduled outages, and the appropriate conditions are met.

We will continue to monitor and participate in the ongoing industry efforts to address potential control rod insertion concerns, and we will evaluate any identified issues for applicability to our stations.

Our response to the bulletin for North Anna and Surry Power Stations is provided in the enclosed report and includes the status of our implementation of the NRC requested actions. Verification of the operability of the control rods at North Anna and Surry is 10003.%

960~4~1nono~17777~9a6L.0~4~0~8:;--~~~~I PDR ADOCK 05000280 I

G PDR

.-*~

provided in Attachments 1 and 2 of the report, respectively. Attachment 3 of the report provides core map information and end-of-cycle burnup for each rodded fuel assembly as requested by the bulletin.

Should you have any questions or require additional information, please contact us.

Very truly yours,

~P.074~

James P. O'Hanlon Senior Vice President - Nuclear Enclosure cc:

U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia Mr. R. D. McWhorter NRC Senior Resident Inspector North Anna Power Station Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station I_

e COMMONWEAL TH OF VIRGINIA )

)

COUNTY OF SURRY

)

e The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. P. O'Hanlon, who is Senior Vice President -

Nuclear, of Virginia Electric and Power Company.. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this ~ day of ~~r** L

, 1996.

My Commission Expires: --~fr\\~~--------* 19.9_k 6

Notary Public

RESPONSE TO BULLETIN 96-01 INCOMPLETE CONTROL ROD INSERTION REC'D W/LTR DTD 4/8/96... 9604100177

-NOTICE -

THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE INFORMATION &

RECORDS MANAGEMENT BRANCH.

THEY HAVE BEEN CHARGED TO YOU FOR A LIMITED TIME PERIOD AND.

MUST BE RETURNED TO THE RECORDS & ARCHIVES SERVICES

SECTION, TS C3. PLEASE DO NOT SEND DOCUMENTS CHARGED OUT THROUGH THE MAIL. REMOVAL OF ANY PAGE(S) FROM DOCUMENT FOR REPRODUCTION MUST BE REFE.RRED TO FILE PERSONNEL.

.. NOTICE -

- e VIRGINIA POWER NORIBANNA UNITS 1 AND 2 SURRY UNITS 1 AND 2 RFBPONSE 10 NRC BULIEllN 96-01 INCOMPLEIE CON1ROL ROD INSERllON

TABLE OF mNTENTS 1.0 INIRODUCTION 3

2.0 RESPONSE TO REQUESTED ACTION (1)............................. 4 2.1

North Anna Response............................. **............

4 3.0 4.0 2.2 Surry Response............................................

4 RESPONSE TO REQUESTED ACTION (2).............................

3.1.1 North Anna Control Rod Operability Report........................

3.1.2 North*Anna Control Rod Operability Statement......................

3.2.1 Surry Control Rod Operability Report............................

3.2.2 Surry Control Rod Operability Statement..........................

4 5

5 6

6 RESPONSE TO REQUESTED ACTION (3).............................. 6 4.1 Control Rod Test Plan for Mid-Cycle Outages During Calendar Year 1996...

7 4.1.1 t:i1~9~~-~~l~ -~~n~~~ ~o~. ~~~ ~~~~~~~~ ~~

-~~~~..

7 4.2 Control Rod Test Plan at the End of Operating Cycles During Calendar Year... 1996................................................

8 4.2.1 ~cl~fi~ -~~~~o~ ~o~. ~~~ ~:~~~~~ ~~ ~~ -~~~ ~~ ~ ~~a~~~...

8 5.0 RESPONSE TO REQUESTED ACTION (4).............................

9 5.1 North Anna and Surry Response................................

9

5.2 Test Reporting............................................

9 6.0 CYCLEANDFUELASSEMBLYDATA..............................

9 7.0 ADDITIONAL PLANS............................................

9

,ATTACHMENT 1 - North Anna Control-Rod Operability Report ATTACHMENT 2 - Surry Control Rod Operability Report ATTACHMENT 3 - Cycle and Fuel Assembly Data Page 2

e 1.0 IN1RODUCTION The NRC issued Bulletin 96-01, "Incomplete Control Rod Insertions," pertaining to events that occurred at South Texas Unit 1 and Wolf Creek whereby several control rods failed to fully

* * **insert following a reactor trip. In addition to these two events which occurred during unit operation, two new control rods could not be removed with normal operation of the control rod

**handling tool from the fuel assemblies in which they were temporarily.*stored at North Anna.

The Bulletin lists four Requested Actions:

(1)

(2)

(3)

(4)

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 ( e.g., boration of a pre-specified amount).

Promptly determine the continued operability of control rods based on current information. As new information becomes available from plant rod drop tests and trips, licensees should consider this new information togetlier with data already available from Wolf Creek, South Texas, and North Anna, and other industry

.. experience, and make a prompt determination of control rod operability.

Measure and 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 clrag forces for all rodded fuel assemblies.

a.

b.

Rods failing to meet the rod drop times in the Technical Specifications.,

shall be deemed inoperable.

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 Regulations.

For each reactor trip during calendar year 1996, verify that all control rods have promptly fully inserted (bottomed) and obtain other available information to assess the o_perability 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.

The Bulletin further requires the following responses to be submitted to the NRC:

(a)

(b)

(c)

Within 30 days of the date of the Bulletin, submit a written report certifying that control rods are determined to be operable, actions taken for Requested Actions (1) and (2) above, and the plans for implementing Requested Actions (3) and (4) above.

Within 30 days of the date of the Bulletin, submit core maps of rodded fuel assemblies indicating fuel type (materials, grids, spacers, guide tube inner diameter) and current projected end-of-cycle burnup-for each rodded assembly for

,the current cycle. -Also, when available, provide tlie.same.. information for the next cycle.

Within 30 days after completing Requested Action (3) for each outage, submit a report that summarizes the data and aocuments the results obtained. This is also

* 'applicable to Requested Action ( 4) when any abnormal rod behavior is observed.

. Page 3

e This report provides the required responses listed in Item (a) above. Specifically, this report will indicate what actions were taken'to*promptly inform operators-of the incidents identified in the Bulletin,/rovide the evaluations that were performed to determine that the control rods at North Anna an Surry are operable, and identify the plans for testing the control rods at North Anna Units 1 and 2 and Surry Units 1 and 2 Included as Attachment 3 to' this report* is the required response to Item {b }*above.

This report will also serve to commit Vir~a Electric and Power Company (Virginia Power) to provide a written report to the NRC within 30 days after completing Requested Actions (3) or ( 4) as specified in the Required Response ( c) above.

2.0 RESPONSE 10 REQUES1ED ACDON (1)

Requested Action (1) requests licensees to promptly inform operators of recent events (reactor trips and testing) in which control rods did not fully insert ancl subseguently provide necessary training, includmg 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).

2.1.. Nonh Anna Respome Licensed operators have been provided awareness training related to the control rod insertion problems identified in Bulletin 96-01. Operators were advised of the already 2.2

*established procedures *should a similar incident occur during operation at North Anna.,

It is currently planned to provide licensed operators additional industry training on the control rod insertion problems during the next Licensed Operator Requalification Program training cycle. This will include the use *of simulator scenarios to ensure_ operators use

established procedures to cope with events in which control rods do not fully insert following a reactor trip. This training is currently scheduled to be completed in May.

1996.

Suny~nse Licensed erators have been informed of the stuck rod events that occurred at Wolf Creek, So Texas Project, and North Anna through required reading.

Additional simulator training will be provided to licensed operators on multiple stuck rod events during the next Licensed Operator Requalification Program (LORP) training cycle.

This training will contain a review of plant events from NRC Bulletin 96-01, provide information on fuel assemblies with burnup values that may increase the possifiility of RCCA binding, and allow teams to respond to events in which control roas do not fully insert following a reactor trip. The traming is currently scheduled to be completed in May 1996.

3.0 RESPONSE 10 REQIJES1ED ACDON (2)

Requested Action (2) requests licensees to promptly determine the continued operability of control rods based on current information. As new information becomes available from plant rod droP. tests and trips, licensees should consider this new information together with data already available from Wolf Creek, South Texas, and North Anna, and other industry experience, ana.

make a prompt determination of control rod operability.

Page4

e 3.1.1 North Anna Control Rod Operability Report A report which determines that the control rods at North Anna Unit 1 *and North Anna Unit 2 will remain operable is included as Attachment 1 to this report.

3.1.2 *

North Anna Control Rod Qperabili~ Statement On March 1, 1996, prior to the issuance o NRC Bulletin 96-01, Virginia Power drafted a report to document-the experiences with*control rod insertion problems at,ollier utilities, the event at North Anna, and actions taken by Virginia Power to support the startup and operation of North Anna 1 Cycle 12 (N1C12) and the continued operation of North Anna 2 Cycle 11 (N2Cll). This report concluded that the control rods will remain operable and that operat10n ofNiC12 and N2Cl 1 to the end of reactivity and to the design bumup limit is acceptable based on the following:

Rod cluster control assembly (RCCA) drag test data indicate that the 17Xl 7

'VANTAGE-SH fuel assemblies previously irradiated in North Anna 1 do not exceed the criteria established for RCCA drag for average assembly bumups up to 50,000 MWD/MfU. North Anna drag test data do exhibit a significant increase in RCCA drag force at assembly bumups greater than 45,000 MWD/MfU, but the magnitude of the RCCA drag force does not exceed limits established by the fuel venoor.

Two RCCAs could not be removed with normal operation of the RCCA handling tool in the spent fuel pool from two discharged fuel assemblies ( discharged in September 1994 from NlCIO) on February 21, 1996. The two RCCAs were temporarily placed into these assemblies in Janl}fil)' 1996 for staging p!llJ)oses to be removed ouring the insert changeout phase dunng the upcorinng refueling outage. The RCCAs were fully inserted into these two fue1 assemblies. Also,..

these two assemblies were rodded assemblies during operation of Cycle 10.. There were no reported problems with the oper~tion of RCCAs during Cycle 10, and there were no reported difficulties in removing the RCCAs from tliese assemblies during the Cycle 10 to Cycle 11 refueling outage.

The two assemblies in the spent fuel pool which exhibited excessive resistance to removing RCCAs stored in them were specifically not included in the RCCA drag test measurement program due to concerns with RCCA handling. However, the RCCA breakaway force was measured as the RCCAs were pulled free from the assemblies. The breakaway forces in these two assemblies were approximately 110 pounds in one and from 140 to 170 pounds in the other. Wlule these forces exceed the Westinghouse F-Spec limit for drag in a fuel assembly, these forces alone would not be sufficient to prevent the control rod from fully inserting in the core with the additional weight of the drive shaft attached.

Shutdown margin calculations performed for N1C12 and N2Cll indicate that

excess shutdown margin exists even in the unlikely event that all rods in fuel.

assemblies with bumup greater than 45,000 MWD/MfU were to not fully insert (assumed to be 18 steps withdrawn).

A safety evaluation applicable to Nl C12 and N2Cl 1 indicated that an unreviewed safety question does not exist even in the unlikely event that all rods in fuel assemblies with bumup greater than 45,000 MWD/MfU were to not fully insert

( assumed to be 18 steps withdrawn).

Page 5

e The conclusion that N1C12 and N2Cll can operate to their respective design bumup limits is based upon the data obtained from the RCCA drag tests performed at Nortli Anna in February 1996 and current industry experience. However, Virginia Power will evaluate additional data as it becomes available prior to exceeding 45,000 MWD/MfU in a rodded fuel assembly to ensure

*this conclusion is still valid.

3.2.1 Suny Control Rod ~rabilitr Report A report wliich determines ~t the control rods at Surry Unit 1 and Surry Unit 2 will remain.,

operable is included as Attachment 2 to this report.

3.2.2 Suny Control Rod Qperabilitr Statement In response to NRC Bulletin 96-01, Virginia Power drafted a report that was issued on March 22, 1996 which concluded that the control rods in Surry Units 1 and 2 will remain operable in light of the information _provided in NRC Bulletin 96-01 and that operation of Surry 1 Cycle 14 (S1C14), Surry 2 cycle 13 (S2C13), and Surry 2 Cycle 14 (S2C14) to their respective design bumup limits is acceptable based on the following:

There have been no reported problems with control rod insertion at either Surry 1 or Surry 2 similar to the events that occurred at Wolf Creek and South Texas 1.

Both Surry 1 and Surry 2 have considerable operating experience with rodded fuel assembly bumups exceeding 40,000 MWD/MfU. This includes experience with rodded fuel assemblies exceeding 45,000 MWD/MTU up to 56,000 MWD/MTU.

There have been no problems with control rod insertion in any of these assemblies.

In addition, there have been no r~orted problems with control rod insertion at

other plants with similar 15X15 fiiel designs.

The maximum rodded assembly burnup in the current Surry 2 cycle (Cycle 13) is expected to be less than 44,000 MWD7MfU. This cycle is expected to end in -.

May 1996. The maximum rodded assembly burnup m the subsequent Surry 2.

cycle will be approximately 49,000 MWD/MTU at the maximum cycle design

bumup.

Similarly, the maximum rodded assembly burnup in the current.Surry. I cycle (Cycle 14) is approximately 46,500 MWD/MTU at the maximum cycle design burnup. All these rodded assembly burnups are well within the experience base for operating with rodded assemblies at Surry.

A review of beginning-of-cycle control rod drop time results for recent cycles confirmed that there were no anomalous trends in control rod drop times.

The available shutdown margin for Surry Units 1 and 2 was evaluated even in the unlikely event that control rods in fuel assemblies with burnups greater than 45,000 MWD/MTU fail to fully insert and remain withdrawn at 18 steps. This evaluation concluded that there is excess shutdown margin available in S1C14, S2Cl3, and S2Cl 4 should these specific control rods fair to fully insert.

4.0 RESPONSE 10 REQUFSTED ACDON (3)

Requested Action (3) requests licensees to measure and 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 times in the Technical Specifications shall be deemed inoperable.

Page 6

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 Regulations.

4.1

. Control Rod Test Plan for d-cle es Duri Calendar Year 1996

'An outage*of sufficient duration was*de ed by the Westin ouse Owners Grou_P. as an outage that allows time to 9cr+/-Tuierly setup and test the rods bv establisht!Nrocedures without restraining blant restart. This e "tion was tentatively accepted by NRC s at a March 25, 1996 meeting etween Westinghouse, WOO, and the NRC staff.

Virgiaja Power has reviewed the industry data pertaining to the control rod insertion problems identified in NRC Bulletin 96-01. Based on the evaluat10ns which support control rocl operability for North Anna and -Surry (Attachments 1 and 2, respectively),* Virginia Power plans to establish a burnup threshold of 35,000 MWD/MfU for rod drop time tests. Therefore, Vir~a Power will perform mid-cycle testing of control rod drop times if there is an outage of sufficient duration and rodded assembly burnup has exceeded the burnup threshold of 35,000 MWD/MfU.

However, at all times during cycle operation and prior to the time when rodded assembly burnup exceeds 35,000 MWD/MfU, North Anna and Surry plant o_Rerators will be instructed to trip the control rods into the core should it become necessary to perform a normal shutdown. Rod position indication will be recorded following the tnp. Consistent with our response to Requested Action ( 4) below, in the event that anomalous control rod behavior 1s evident following the trip, control rod drop time tests will be performed re88!dless of any test contingencies (i.e., the rodded assembly burnup thresliold or the mmimum cycle burnup increment which is discussed below).

This plan is supported by extensive industry data which indicates the control rod insertion problem occurs m high burnup fuel assemblies. The drag tests performed at North Anna.

mdicated that drag forces began to increase at fuel assembly burnups of approximately. 45,000 MWD/MfU. Also, there is no evidence of control rod insertion problems similar to that which occurred at Wolf Creek and South Texas 1 occurring in Westinghouse 15X15 fuel. Surry has

* ~pera!ed "1th rodded asse~bFes exceeding 55,000 MWD/MfU ~* no d<?monstrated control rod.

tnsert1on difficulty. By tnppmg the rods mto the core when a umt 1s requrred to be shut down, and performing rod drop t1me tests when rodded assembly burnup exceeds 35,000 MWD/MfU, the control rod drop behavior will be subjected to surveillance for the entire cycle with particular emphasis beginning toward the middle of the cycle when rodded assembly, burnup exceeds 35,000 MWD/MfU.

Control rod drop time tests would not be required for either North Anna or Surry if less than 2500 MWD/MfU has elapsed since the last control rod drop time tests were performed. The 2500 MWD/MfU burnup increment was proposed to the NRC staff by the Westinghouse Owners Group at a meeting between Westinghouse and the NRC staff on March 25, 1996. The NRC staff accepted the 2500 MWD/MTlJ minimum burnup increment between rod drop time tests.

The burnup increment would not apply to the stipulation of tripping the rods into.the core should it become necessary to shut a unit down.

If abnormal control rod insertion. behavior should be observed, control rod time tests would be performed regardless of either the burnup threshold or the 2500 MWD/MfU burnup increment.

4.1.1 1.)

Sfi:ific Mid-Qrcle Omtml Rod Test Requirements During Calendar Year 1996 s ould it become necessary to perform a nonnal shutdown at any time during cycle

operation, the control rods will be tripped into the core. Rod position indicat10n will be record~ follo~g the trip. In the eyent that ~omalous control rod behavior was evident

followmg *the tnp, * *control rod drop t1me tests will be performed regardless of any test Page 7

,:;11

~:'~:

2.)

3.)

contingencies (i.e., the rodded assembly bumup threshold or the minimum cycle bumup increment).

During an outage of sufficient duration, perform control rod drop time tests if rodded assembly bumup has exceeded 35,000 MWD/MIU and more than 2500 MWD/MIU (refers to. increment in cycle energy, no~ rodded assembly incremental bumup) have

- elapsed smce the last control rod drop tune tests were performed.., ---- -

2.a) 2.b) 2.c)

Verify that control rod recoil is observed on the timing test traces.

Rods failing to meet the rod drop times in the Technical Specifications shall be deemed inoperable.

Rods failing to bottom shall require prompt corrective action in accordance with established programs.

If appropriate plant conditions exist where the vessel head is removed, measure and evaluate drag forces for all rodded fuel assemblies.

3.a)

Rods exhibiting high drag forces shall require prompt corrective action in

. accordance with establislied programs.

4.2 Control Rod Test Plan at the End of Operati~ Qrcles During Calendar Year 1996 Virginia Power will perform the following control ro tests at the end of an operating cycle durmg calendar year 1996, subject to the 2500 MWD/MIU cycle bumup increment. Surry Unit 2 is expected to end Cycle 13 during the May 1996, and North Anna Unit 2 is expected to end _

Cycle 11 in September 1996.

4.2.1 Specific Control Rod Test Reterements at the End of an Operating Qrcle 1.)

At the end of an operating cyce during calendar year 1996, perform control rod drop time 2.)

tests-if more than 2500 MWD/MTU (cycle) have elapsed since the last control rod drop time tests were performed.

I.a) 1.b) 1.c)

Verify that control rod recoil is observed on the timing test traces.

Rods failing to meet the rod drop times in the Technical Specifications shall be deemed inoperable.

Rods failing to bottom shall require prompt corrective action in accordance with established programs.

Measure and evaluate drag forces for all rodded fuel assemblies. Drag tests may either be performed in-reactor or in-pool, as schedule or required tooling dictates.

2.a)

Rods exhibiting high drag forces shall require prompt corrective action in accordance with established programs.

4.3 Test Reporting A report summarizing the tests, data, and conclusions will be submitted to the NRC within 30 days after completing the tests. This reporting requirement applies to both mid-cycle rod drop time tests and end-of-cycle rod drop time tests. Reporting does not apply to occurrences when the rods were tripped into the core with no anomalous behavior.

Page 8

5.0 RESPONSE 10 REQUES1ED ACIION ( 4)

Requested Action (4) requests-'licensees, for each reactor trip during calendar year 1996, to verify that all control rods have promptly fully inserted (bottomed) and obtain other available infonnation 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.

5.1 North Anna and Suny Respome Station Operations and Engineering will be required to perform a review of control rod drop behavior following the trip. This review will be completed prior to restart. In the event that.

anomalous control rod behavior is evident, control rod drop time tests will be performed regardless of either the 35,000 MWD/MIU rodded assemoly burnup threshold or the 2500 MWD/MIU cycle burnup increment.

5.2 Test Reporting In the event that control rod drop time tests are required per anomalous control rod insertion behavior, a report summarizing the tests, data, and conclusions will be submitted to the NRC within 30 days after completing the tests.

6.0 CYUE AND FUEL ASSEMBLY DATA Cycle and fuel data for Surry and North Anna fuel cycles operating in 1996 are provided as to this reportm accordance with Required Response (2) of Bulletin 96-01.

Specifically, data for each of the four cycles currently in operation is provided (Surry 1 Cycle 14, Surry 2 Cycle 13, North Anna 1 Cycle 12, and North Anna 2 Cycle 11 ).

Data for Surry 2 Cycle 14, which will not begin operation until June, is also included. Similar infonnation will be provided for North -Anna 2 Cycle 12 when it becomes available later this.*

year.

Note*that both-Surry and North Anna use silver-indium-cadmium control rods. The original-....

Surry control rods were replaced with Westinghouse Enhanced Performance ( chrome plated) control rods after Cycle 10 at each unit. The original North Anna 1 control rods were replaced with Westinghouse Enhanced Performance control rods during the 1996 refueling outage, and the North Anna 2 control rods are similarly scheduled to be replaced during the Unit 2 refueling outage in the Fall of 1996.

7.0 ADDIDONAL PIANS In addition to the testing and reporting commitments per Bulletin 96-01, Virginia Power has further plans to support the Westinghouse root cause evaluation by permitting Westinghouse to conduct additional tests on North Anna fuel assemblies in early Sprmg 1996. Data ootained from these tests will directly benefit the root cause evaluation project.

Virginia Power will continue to follow Westinghouse efforts to determine the root cause of the problem. Upon completion of the Westinghouse root cause evaluation, a determination will be made as to the applicability to continued operation of North Anna Units 1 and 2 and Surry Units 1 and 2.

- ** Virgipia Power has responded to the request from the Westinghouse Owners Group to provide additional trip and RCCA test data from the last five years to the WOO. The WOO plans to incorporate these data into their data base on control rod operability.

Page9

ATIACHMENT 1 CON1ROL ROD OPERABILflY REPORf NOR'Ill ANNA UNilS 1 AND 2

1.0 IN'IRODUCilON On February 21, 1996, during the insert shuffle in preparation for loading North Anna 1 Cycle 12, two new Rod Cluster-Control Assemblies (RCCAs) couldnoLbe removed with normal operation of the RCCA handling tool from the fuel assemblies in which they were temporarily stored. The fuel assemblies were OA2 and OA8. Both of these assemblies are of the Vantage-SH

__ design with average assembly bumups of 47,782 and 49,613, respectively. These two assemblies were irradiated for two cycles and discharged after Cycle 10 in September 1994.

A special proced~e was written to !emove the a.tiected. ~CCAs from the two assemblies using

the RCCA handhng tool and the bndge crane hmst. Initial tests were then performed to detennine if the resistance was the result of a defective RCCA handling tool, an undetected RCCA manufacturing or design flaw (the two RCCAs were new and delivered with the N1C12 feed batch (Batch 14 assembhes)), or the result of excessive drag from the fuel assembly guide tubes. A succession of tests were developed and implemented which isolated the problem to excessive resistance in the fuel assembly guide tubes.

There have been reported problems at two other US utilities with RCCA insertion interference in Vantage-SH fuel assemblies following a reactor trip. Another insertion problem was reported at a European reactor using fuel manufactured by Fragema. While North Anna has not experienced any problems with RCCA insertion in Vantage-SH fuel assemblies during operation, Virginia Power conservatively approached the problem as if it were linked to the RCCA interference problems experienced at other sites.

-This report documents the experiences with RCCA insertion problems into Vantage 5H assemblies at other utilities, the event at North Anna, and act10ns taken by Virginia Power to support the startup and operation ofN1C12 and the continued operation ofN2Cl 1. This report will also address the root cause investigation of the problem which is currently in' progress at Wolf Creek. A safety evaluation is included to demonstrate that there would be no unreviewed safety question as a result of certain RCCAs failing to fully insert during.operation ofN1Cl2.and contmued operation of N2Cl 1.

2.0 -

EXCESSIVE RCCA DRAG FORCFS AT OIHER PI.ANTS On Jan~ 30, 1996 following a reactor trip at the Wolf Creek Power Station, five RCCAs failed to fully insert. The indicated position immediately following the reactor trip for these RCCAs ranged from 6 to 18 steps withdrawn. During subsequent cold drop tests, the same five RCCAs plus an additional 3 RCCAs failed to fully insert. All of the affected RCCAs were located in thrice-burned fuel assemblies with bumups greater than 47,600 MWD/MfU. Subsequent to this event Wolf Creek entered a refueling outage.

Similar events have occurred during reactor trips at the South Texas and Ringhals plants. On December 12, 1995 during a reactor trip at the South Texas plant, 3 or 4 RCCAs failed to fully insert at an indicated position of approximately 6 steps withdrawn. South Texas has a 14 foot core, and the affected RCCAs were located in assemblies with bumups between 42,000 and 43,000 MWD/MfU. On August 22, 1994, one RCCA failed to fully insert into the Ring!_lals Unit 4 core during a reactor trip. Subsequent drop tests at both Ringhals Units 3 and 4 identified additional RCCAs which either failed to fully insert or exhibitcil abnormal drop times. The Ringhals units use fuel assemblies manufactured by Fragema On February 6, 1996, the NRC contacted Westinghouse and the Westinghouse Owners Group (WOO) and expressed a concern regarding the safety significance of this issue. The NRC

- :forwarded*** 9 questions for consideration and requested quick response. An initial response was transmitted to the NRC on February 15, 1996 and a meeting was held at NRC headquarters on February 20, 1996 to discuss the responses. Additionally, extensive examinations are currently being performed at Wolf Creek to detennine the root cause of the Wolf Creek event.

Page 1

3.0 ASSEMBLIFS OA2 AND OAS Section 3.1 will describe the problem encountered with assemblies OA2 and OA8 and the necessary recovery efforts. -Section 3.2 will describe additional details.with regard to the operation of control rods in these two assemblies during North Anna 1 Cycle I 0.

3.1 HANDUNG PROBTEM WITII NEW RCCAS IN FUEL ASSEMBLIFS OA2 AND OAS

.. In January 1996, new RCCAs were staged into the North Anna spent fuel pool in pr~aration to replace the North Anna Unit 1 control rods during the Cycle 11 to Cycle 12 refueling outage. These RCCAs were placed in discharged assemblies, many of them VANTAGE-SH assemblies. Two oftlie new RCCAs, RCCAs R120 and Rl21, were inserted into assemblies OA2 and OA8, respectively. Both OA2 and OA8 are Vantage-SH assemblies. An RCCA handling tool problem was encountered early during the process of staging the new RCCAs into the spent fuel pool, however, the tool problem was resolved and all new RCCAs were fully inserted into their host temporary storage fuel assemblies.

On February 21, 1996, during the insert shuffle which was performed in preparation for the onload of the Cycle 12 core, RCCAs R120 and R121 could not be removed from their host assemblies because the RCCA tool overload setpoint was exceeded. The hoist on the RCCA handling tool has the overload trip setpoint *nominally set at 280 lbs. to meet applicable Westinghouse F-Spec. requirements pertaining to handling loads on RCCAs.

At that point it was not clear whether a problem was bemg experienced with the new RCCAs or with the fuel assemblies, although there was some concern that it was related to recent problems e~erienced at other Westinghouse plants. Due to the concern over the safety implication of the inability of RCCAs to fully insert during a trip, a program was set up to remove the two affected RCCAs and to determine whether the RCCAs or the

assemblies were the cause of the handling problem.

A procedure was then written to remove R120 and R121 from OA2 and OA8. using the...

RCCA handling tool with extra lifting force provided by the bridge crane hoist. The method used to remove the two RCCAs did not provide an accurate reading of the sliding drag load of the RCCAs as they were pulled from the assemblies. However, the breakaway forces could be approximated by the digital load cell on the bridge crane hoist.

The breakaway force encountered in removing RCCA R120 from Assembly OA2 was * *

~proximately 110 pounds. The breakaway force encountered in removing RCCA R121 from Assembly OA8 was approximately 140 to 170 pounds. The breakaway forces are

related to static friction and-will generally be greater than the sliding drag load.

Once these RCCAs were removed they were placed into a series of other assemblies to determine if the handling problem was related to these particular RCCAs. Recall that these RCCAs are new and it is possible that they contamed an undetected design or manufacturing flaw. Neither of the two RCCAs e~erienced any excessive resistance during the subsequent insertion and removal exercises, and the test plan called for insertmg these RCCAs into their respective Cycle 12 host fuel assemblies. The potential of an existing RCCA design or manufacturing flaw was ruled out based on the fact that no additional handling problems were experienced with these two RCCAs and difficulty was experienced when trying to insert an old RCCA (from the Cycle 11 core) into assemb1ies OA2 and OA8. This determined that the cause of the high drag was related to the assemblies and not the RCCAs.

3.2 RCCA OPERATION IN OA2 AND OAS

Both OA2 and OA8 were rodded assemblies during Nl ClO and were discharged at the end of Nl ClO after two cycles of irradiation. The rod drop time tests performed at the

. beginning of Cycle 10 indicate that the drop times for these rodded assemblies were very close to the average of the other RCCA drop times. Also, the rod drop time traces for these two assemblies indicate that the RCCAs recoiled when the RCCAs impacted the top Page2

nozzles of the fuel assemblies. Therefore, these assemblies did not exhibit any tendencies to restrain the travel ofRCCAs at the beginning ofNlClO.

At the end of NI ClO the rods were stepped in to their fully inserted position in lieu of tripping the rods in at shutdown. Note that gravity provides the only motive force to insert the rods following either reactor trip or manual rod insertion. Stepping the rods in is the normal practice for shutting a unit aown at North Anna. There was no abnormal control rod-behavior noted during the shutdown process ofNlClO nor was there any reported difficulty in removing either of the RCCAs in.OA2 or OA8 during the Cycle 10 to Cycle 11 refueling outage.

High breakaway forces were experienced in these two assemblies during removal of the new RCCAs. However, there were no problems encountered during operation ofNlClO.

The submerged weight of the RCCA dnve shaft is approximately 115 pounds. The submerged wei~t of an RCCA is approximately 134 pounds.. Therefore the static load.of an RCCA coupled to its drive shaft is approximately 249 pounds. As indicated in Section 3.1, the maximum breakaway force experienced by removmg these control rods from OA2 and OA8 ranged from 140 to 170 pounds. Although this breakaway_ force is greater than the F-Spec limit of 100 pounds, tlie static load of the RCCA and drive shaft 1s greater than the maximum experienced breakaway force ( and thus greater than the expected sliding drag force). These forces alone would not be sufficient to prevent the RCCA from fully msertmg.

4.0 NORIII ANNA RCCA DRAG TESTING 4.1 PROGRAM DFSCRIPilON

--An RCCA drag testing program was set up to measure the sliding drag forces in a series of Vantage-SH fuel assemblies having burnups ranging from approximately 20,000..

~ :.:.

MWD/MIU to approximately 50,000 MWD/MIU. The purpose of the program was to determine RCCA clrag forces with varying assembly burnup.

A special latching tool was developed-by Westinghouse and sent to North Anna specifically for RCCA*drag tests. The drag tests were conducted by hanging a digital load

- cell off the bridge crane hook and attaching the RCCA latching tool riggmg to the load.

cell. After attaching the tool to an RCCA, the load cell was zeroed out and the* crane operator began to lift the RCCA while noting the maximum sliding drag force that was experienced in the dashpot region. The operator continued the lift for approximately 12 more inches above the clashpot into the guide tube region and noted the maximum sliding drag force experienced in tlie guide tube region.

Before any drag tests were performed, the RCCA handling tool was used to exercise the RCCA drag test candidates by lifting the RCCA approximately four feet and then reinserting the RCCA back into the fuel assembly. The intent of this evolution was to identify any drag test candidates that may exhibit excessive drag forces which may have precluded the RCCA from being reinserted upon completion of the test. That situation would have resulted in a potentially difficult recovery operation with an unrestrained RCCA extending up beyond its normal elevation. For this reason, assemblies OA2 and OA8 were specifically omitted from the test. Data had already been collected to approximate the breakaway forces in OA2 and OA8. If additional candidate assemblies haa resulted in excessive RCCA drag forces using the RCCA handling tool during the

screening test, additional consideration would have been given to perform further tests on those assemblies. However, no handling problems were encountered during the screening tests.

In addition to the Vantage-SH assemblies, the drag force was measured in three once-Page3

burned Performance Plus assemblies with average assembly bumup ranging from 20,000 to 23,000 MWD/MTU. The p_urpose of measunng the drag force m the Performance Plus assemblies was to determine 1f tliere was any difference in the drag force between the two different fuel designs at approximately the same bumup. One of the differences between Performance Plus assemblies and Vantage-SH assemblies is that Performance Plus assemblies have ZIRLO guide tubes instead of Zircaloy-4 guide tubes. ZIRLO has been proven to be more dimensionally stable with irradiation than Zircaloy-4. All rodded

** *assemblies in NlC12 will be Performance Plus assemblies.with ZIRLO guide tubes.

4.2 DATA EVAUJATION Table 4.1 is the tabular drag test data. The column for the calculated wet weight represents the sum of the submerged weight of the drag test latching tool with conduit and the submerged weight of an RCCA The submerged weight of the clrag test tool was..

measured to be 21 pounds.

The mass of the spider hub divided by the density of stainless steel determined the displacement volume of the hub. The weight of a single RCCA which was calculated to be 150founds. The displacement volume of the RCCA equated to a displaced water mass o 16.3 pounds. The net submerged weight of the RCCA is the dry weight of 150 pounds less tlie buoyant force of 16.3 pounds which is 133.7 pounds. lhe total

submerged static weight of the RCCA coupled with the drag test tool is 154.7 pounds.

The hanging weight was recorded following the dashpot drag force measurement and

before tlie upper guide tube region drag force measurement. There are three outliers with respect to tlie measured hanging weight of the RCCA and tool. These outliers are 5Bl (170 pounds), 2A5 (164 pounds), ana lAO (174 pounds)*and appear to have some residual drag tension in the rigging when the measurement was obtaineo. The average of the ;",,,

hanging wei@t (omittmg the outliers from the average) is 154.7 pounds whichis in agreement with the calculated weight. ** Therefore, calculated weight is verified by measurement.

. The**drag forces were determined by subtracting the calculated hanging weight.from the-

.

pulling forces. Westinghouse F-Spec limits were used as acceptance criteria for the drag tests which are 100 pounds or less of drag force in the dashpot region and 40 pounds or less of drag force in the guide tube region. Figure 4.1 is a plot of the dashpot drag force data. Figure 4.2 is a plot of the guide tube region drag force data.

All drag test results were within the acceptance limits of 100 pounds in the dashpot region and 40 pounds in the guide tube region and the limits on RCCA drag are not exceeded

~t!1 average assembly burnup up !o ~pprox4mttely s9,ooo MWDlfyITU. The data d?

md1cate, however, that there 1s a s1gmlicant mcrease m drag force m the dashpot region and the guide tube region with average assembly bumup greater than 45,000 MWD/MTU.

There appears to be very little increase in drag forces with average assembly bumups less than 45,000 MWD/MfO. Also, based on limited data, the drag forces in the once.:bumed ZIRLO Performance Plus assemblies appear to be approximately 20% less than the drag forces in the VANTAGE-SH assemblies with Zircafoy-4 guide tubes at low bumups.

Performance Plus assemblies that were drag tested have the letter "C" as the center character of the assembly identification.

Page4

e Table 4.1 North Anna RCCA Drag Test Data Assembly Burnup Dashpot Pulling Guide Tube Pulling Hanging Calculated Wet Dashpot Drag Guide Tube Drag Drag Test Shutdown Cooling ID (MWD/MTU)

Force (lbf)

Weight (lbf)

Force (lbf)

Date, Date Days Y47 19581 177 164 154 154.7 22 9

2/24/96 2/26/92 1459 4C2 20192 173 156 156 154.7 18 1

2/24/96_

2/11/96 13 Y40 23152 178 159 154 154.7 23 4

2/24/96 2/26/92 1459 3C4 23304 165 155 154 154.7 10 0

2/24/96 2/11/96 13 6C1 24272 163 155 155 154.7 8

0 2/24/96 2/11/96 13 483 37838 178 161 153 154.7 23 6

2/24/96 2/11/96 13 288 38126 168 155 155 154.7 13 0

2/24/96 2/11/96 13 283 41410 174 158 158 154.7 19 3

2/24/96 2/11/96 13 185 41564 179 159 159 154.7 24 4

2/24/96 2/11/96 13 484 43953 181 156 155 154.7 26 1

2/24/96 2/11/96 13 387 43967 170 158 152 154.7 15 3

2/24/96 2/11/96 13 588 44064 183 158 154 154.7 28 3

2/24/96 2/11/96 13 287 45342 181 162 153 154.7 26 7

2/24/96 2/11/96 13 584 45638 192 163 154 154.7 37 8

2/24/96 2/11/96 13 482 47018 198 157 154 154.7 43 2

2/24/96 2/11/96 13 581 47284 193 181 170 154.7 38 26 2/24/96 2/11/96 13 188 48239 207 165 154 154.7 52 10 2/24/96 2/11/96 13 083 48430 221 166 157 154.7 66 11 2/24/96 2/11/96 13 282 49135 191 162 153 154.7 36 7

2/24/96 2/11/96 13 2A5 49274 207 179 164 154.7 52 24 2/24/96 9/9/94 533 1AO 49970 239 187 174 154.7 84 32 2/24/96 9/9/94 533

elf C.l * -

0

~

~ -

~

u u 1Z 110 100 90 80 70 60 50 40 30 20 10 0

15000

)I(

it:

20000 Figure 4.1 North Anna RCCA Drag Test Data Dashpot Region

)I(

X

)I(

X 25000 30000 35000 40000 45000 Assembly Burnup, MWd/MtU

)

)I(

)I( )I(

)I(

50000 55000

9

~

CJ

a..

0

~

bJ)

~

a..

~

u u

p:::

50 40 30 20 10 0

15000

)I(

~

20000 Figure 4.2 North Anna RCCA Drag Test Data Guide Tube Region

)I(

X

0:

)I(

'J{ 'J{

'J{

25000 30000 35000 40000 45000 Assembly Burnup, MWd/MtU

)I(

X

)I(

50000 55000

) 5.0 NORIH ANNA EXPERIENCE WITH RODDED V ANTAGE-5H FUEL ASSEMBI.JFB The failure of some RCCAs to fully insert, as experienced at other plants, has occurred following a reactor trip. At Wolf Creek, the five rods that ilid not fully insert "drifted" to the bottom within an hour. At South Texas Unit 1, one of the three RCCAs that did not fully insert "drifted" to the bottom within one hour. The other two were stepped in. During subsequent

... testing, the same three RCCAs plus an additional one did not fully insert. Of tliese four, two "drifted" to the bottom ~d the other ~o w~e stepped in. It has not been ~etermin~d why some RCCAs would not.fully msert after,bemg tripped mto the core only to passiyely "drift" or be stepped to the bottom later. The act of steppmg the rods into the core ooes not forcefully "drive" the rods downward. Gravity (i.e., the weight of the RCCA coupled to its drive shaft) is the only motive downward force when rods are stepped in. There is no evidence indicating that this phenomenon would not occur if the rods were being stepped into the fully inserted position. To elate however, the phenomenon has only been observed following reactor trips.

The historical use of Vantage-SH fuel assemblies in rodded locations at both North Anna Units 1 and 2 was reviewed. Table 5.1 lists the average fuel assembly bumup of Vantage-SH fuel assemblies in rodded locations at the end of each cycle they were used. Some cycles did not have Vantage-SH assemblies in all rodded locations, hence the reason for not necessarily displaying 48 rodded assemblies for all the cycles listed on Table 5.1.

North Anna Unit 1 first operated with Vanta_ge-5H fuel in Cycle 9 which began in March 1991.

None of the Vantage-SH fuel assemblies in cycle 9 operated in a rodded location. Cycle 10 was the first operating cycle to have Vantage-SH assemblies in rodded locations. There were no reactor trips during Cycle 10 and there were no reported problems associated with the operation of the RCCAs durmg this cycle. Cycle 11 experienced a reactor trip on January 27, 1995.

However, the maximum assembly average bumup in a rodded locat10n was apr,roximately 31,500 MWD/MIU which is well below the bumup where drag forces appear to sigmficantly increase.

There were no other reactor trips during Cycle 11, and there were no reported problems associated with the operation of the RCCAs during this cycle.

North Anna Unit 2 first operated with Vantage-SH fuel in Cycle 8 which began in November 1990. None of the Vantage-SH fuel assemblies in Cycle 8 o~ted in a rodcled location. N2C9 was the first OJ?erating cycle to have Vantage-SH assemblies m rodded locations. There were three reactor tnps durmg Cycle 9. The first reactor trip occurred on August 6, 1992. The maximum assembly average burnup in a rodded location during this trip was approximately 29,500 MWD/MIU. The next two reactor trips occurred on April 16, 1993 and April 24, 1993.

The maximum assembly average burnup in a rodded location at the time of these trips was approximately 42,000 MWD/MfU.

There were no reported RCCA problems subsequent to these reactor trips nor were there any reported operational problems associated with the RCCAs throughout the remainder of tlie cycle.

Also, during the insert changeout process during refueling outages, all RCCAs have been removed from their host assemblies with normal operation of tlie RCCA handling tool.

Page 8

Table 5.1 END-OF-CYCLE RODDED ASSEMBLY BURNUP NlClO NlCll N2C9 N2C10 e

I ID I Bumup I I ID I Bumup I I ID I Bumup I I ID I Bumup I OA4 50047 2B2 49135 Yl4 47374 4L5 47946

!AO 49970 OB3 48430 Y24 46841 3L2 47846 IA5 49928 OB7 48119 Y62 46826 4L6 41n1 OAS 49613 2BO 48076 Y53 46630 6L5 47692 2A5 49274 4B9 47970 Y22 46582 4L7 47686 2A9 48873 5B3 47903 Y03 46540

!LS 47668 OAS 48719 3B9 47767 Y06 46429 2L5 47587 OA9 48416 3B6 47699 Y31 46400 6L4 47439 2Al 48353 5B7 47653 Y28 46275 5L3 47385 4Al 48075 3B3 47639 Y23 46209 5L7 4n43 IA?

48031 2B9 47550 Yl6 45977 6L3 46915 2A3 47948 5B6 47515 Y21 45931 IL6 46881 6A4 47873 2B5 47325 Y45 45775 2L4 46837 OA2 47782 5Bl 47284 Y05 45181 3L7 46674 6A8 47101 4B2 47018 Yl3 44441 IL2 46466 4A8 47024 4B6 46937 Y26 44171 OL9 45957 5A6 46998 5B4 45638 YIS 43960 5L2 45371 6A2 46944 2B7 45342 YOS 43858 5L5 45204 5A5 46871 4B8 45229 Y07 43303 4L4 44879 SAO 46866 3B4 45097 Yll 43160 5LO 44583 3A9 46637 5B9 44894 Y02 43112 ILi 44233 3A3 46467 5B5 44733 Y25 43091 5L8 44146 5A4 46143 5BO 44515 YlO 42917 2L6 44034 4AO 46031 3B2 44474 Yl2 42542 6L7 43950 4A6 46029 3BI 44105 Y55 41226 6L6 43846 6A3 45736 5B8 44064 Y46 41055 3LI 43816 6Al 45654 3B7 43967 Y33 41020 5L6 43484 6AO 45114 4B4 43953 Y56 41004 3L5 43154 4A4 45100 6BO 43047 Y54 40992 6L8 42943 4A7 44930 3B8 42855 Y49 40860 6L2 42605 3A8 44736 4B5 42830 Y30 40689 4L3 42560 3A5 44635 OBS 42478 Y29 40485 2L2 42029 OA6 39154

!BO 42172 Y37 37606 4L9 41993 OA3 39033 3BO 42034 Y60 37345

!LS 41307 2A7 39008

!BS 41564 Y52 37273 3L6 41066 IA2 38985 2B3 41410 Y44 37214 3L9 40693

!Al 38920 4BO 39217 Y58 36820 IL4 37657 IA3 38790 3B5 38578 Y34 36278 3L8 37448 5A9 38577 2B8 38126 2LO 37193 4A2 38430 4B3 37838 2L7 36821

!AS 38208 IA6 38163 5A3 38153 3A6 37855 6A7 37824 5A2 37490 4A9 37485 4A5 37354

6.0 SAFEIY EVALUATION 6.1 N1C12 AND N2Cll SHUTDOWN MARGIN Given the similarities between the Wolf Creek and North Anna fuel designs, there exists the potential that North Anna may experience the same problem as Wolf Creek.

Therefore, as a precaution, shutdown margin calculations were performed to verify that adequate shutdown margin exists assuming that the control rods in assemblies greater than

, 45,000 MWD/MIU are 18 steps withdrawn. This is judged to be a conservative assumption' and basis for this calculation. Tue 18 steps withdrawn was based on the experience at Wolf Creek and the assembly bumup greater than 45,000 was based on the North Anna drag force measurements. The results of the shutdown margin calculations are summariz.ed in Tables 6.1 and 6.2 for Nl C12 and N2Cl 1, respectively.,

The shutdown margin calculations indicate that, using conventional bounding conservatism, adequate shutdown margin does not exist using the above assumptions.

However, adequate margin can be gained by limiting the rod insertion during the end-of-cycle moderator temperature coefficient (MfC) measurement to something less than the rod insertion limit. This will reduce the rod insertion allowance (RIA) such that the shutdown margin requirement of 1770 pcm is met. Therefore, an administrative rod insertion requirement must be developed for each cycle that allows the RIA to be reduced to a sufficient level such that the shutdown margin requirement is met during the MTC test. The administrative limit would only apply when the MTC test was being performed.

The RIA will be reduced to 60 pcm during the MTC tests for N1C12 and N2Cll. The corresponding D-bank insertion limit to meet the 60 pcm RIA is 216 steps. The MTC test procedure will be modified by May 6, 1996 which is in advance of the date that the procedure will be required.

Page 10

TABIE6.1 N1C12 SHUTDO\\ffi MARGIN CALCUIATION Comeivative MIC Test Non-Test Parameter RSAC Gllculation Calculation Calculation (pcm)

(pcm)

(ocm)

N-1 Rod Worth 6800 6727 6727 6727 N-1 w/ Uncertainty 6120 6054 6054 6054 1PD 3704 3750 3750 3750 Void Effects 50 50 50 50 RIA 300 300 60 300 MfC Test 260 260 260 0

Total Requirements 4314

4360 4120 4100 SDM 1806 1694 1934 1954 SDMLimit 1770 1770 1770 1770 Excess SDM 36

-76 164 184 TABLE6.2 N2Cll SHUTDO\\ffi MARGIN CALCUIATION Comeivative MIC Test Non-Test Parameter RSAC Gllculation Calculation Calculation (ocm)

(pcm)

N-1 Rod Worth 6832 6744 6744 6744 N-1 w/ Uncertainty 6149 6070 6070 6070 1PD 3743 3704 3704 3704 Void Effects 50 50 50 50 RIA 290 290 60 290 MfC Test 260 260 260 0

. Total Requirements 4343 4304 4074 4044 SDM 1806 1766 1996 2026 SDMLimit 1770 1770 1770 1770 Excess SDM 36

-4 226 256 Page 11

6.2 ADDfilONAL SAFEIY ANALYSIS ffiNSIDERA1I0NS The Westinghouse Owners Group (WOO) has performed additional evaluation of existing

safety analyses with-respect to tlie potential for having some RCCAs not fully inserted.

The results of these analyses were presented to the NRC at a meeting at NRC headquarters on February 20, 1996. The WOO concluded that there 1s sufficient margin in the existing analyses to cover the effects of having certain RCCAs not fully inserted.

Virginia Power technical staff personnel have reviewed existing analyses in addition to the WOO material and concurred with the WOO conclusion.

6.3 10 CFR 50.59 EV AUJA1ION The following section addresses the questions asked in I0CFR50.59 to determine if an unreviewed safety question exists for the operation ofN1C12 and the continued operation ofN2Cll upon evaluation of the RCCA drag tests performed at the plant during the Cycle 11 to Cycle 12. refueling outage.

6.3.1 Does the char)2e incre~e the probability of occurrence or the comequences of an accident or malfmction of equipment important to safety previomly evaluated in the safety analY.sis report?

This change will not mcrease the probability of occurrence for any Chapter 15 accident. Operation of the reactor is not changed based on the evaluation of the RCCA drag tests perfonned in the spent fuel pool. In addition, there are no physical alterations to the unit. :* Therefore, the probability of occurrence of any accident evaluated in the safety analysis report will not be increased.

This change will not increase the consequences of the accidents evaluated in the safety analysis report. The result of the RCCA drag tests show only minimal drag forces in the upper guide tube region, even for fuel assemblies with high bumup.

RCCA drop times (measured to the. beginning of the dashpot) will be.unaffecteo, Based on the forces measured in the dashpot, it is also expected that the RCCAs will fully insert following a reactor trip at any time during operation ofN1C12 or

N2Cll. However, calculations have.oeen performed to ensure-that adequate

~.,

shutdown margin exists in the unlikely event that a number of RCCAs m high --

bumup assemblies fail to fully insert. Therefore, the consequences of any of th,e accidents analyzed in Chapter 15 of the UFSAR are not increased.

This change will not increase the probability or consequence of malfunction of equipment important to safety. Drag forces between the fuel assemblies and the RCCAs, in the dashpot region, were shown to increase with burnup. This drag force did, however, remain below the combined weight of the RCCA and drive shaft combination for the entire range of fuel assemoly burnups tested. Based on this, it is expected that the RCCAs will fully insert following a reactor trip at any time during operation ofN1C12 or N2Cll. Recent industry e~erience has shown that there 1s the possibility that some number of RCCAs may fail to fully insert during a reactor trip. The operators have been made aware of these industry events. Per existing Emergency Operating Procedures, the operators are instructed to borate should some of the RCCAs fail to fully insert. Also, calculations have been performed which demonstrate that adequate shutdown margin exists should some of the RCCAs fail to fully insert. Since the control rods will continue to meet the required safety function, the probability or consequence of a malfunction has not been increased.

6.3.2 Does the change create 1he possibility.for an accident or malfmction of equipment of a different type than any previomly evaluated in the safety analysis report?

Design and operation of the reactor protection and control system has not been Page 12

changed. Therefore, operation ofN1C12 or N2Cll following evaluation of RCCA drag test results does not create the possibility of an accident or malfimction of eqwpment of a different type than previously evaluated in the safety analysis report.

' 6.3.3 Has the mmgin of safety of any part of the Technical Specificatiom as described in the Bases section been reduceil?

The margin of safety for any Technical Specification as described in the Bases section has not been reduced. Results of the drag test data show very little RCCA drag in the upper portion of the fuel assembly guide tube, and, therefore, maximum RCCA drop time (measured to the beginning of the dashpot) will be unaffected.

Calculations have demonstrated that adequate shutdown margin exists even if the

-RCCAs in the high burnup fuel assemblies fail to fully insert into the core. No other Bases for Section 3.0 and 4.0 in the Technical Specifications are affected by the evaluation of the RCCA drag tests.

7.0 REmMMENDED ACTIONS The following actions were recommended in a Westinghouse Owners Group (WOO) letter to Westinghouse Owners Group Primary Representatives, dated February 23, 1996:

Plant operators should be trained and aware of the potential for the incomplete rod insertion event and be prepared to act in accordance with plant procedures. This will be accomplished oy operator awareness training and Licensed Operator Requalification Training.

Manually trip the control rods into the core at the end ofN2Cl 1 and N1C12 in -

lieu -of steppmg the control rods to the fully inserted position. (Reactor Engineering will initiate changes to the operating procedures to -implement this "

recommendatiori.)

Provide reactor trip data with regard to the behavior of the RCCAs immediately following the trip to the WOO. The behavior of the RCCAs is particularly important with liigh assembly burnup in rodded positions. Virgmia Power will prepare this information and transtn1t to the WOO as available.

Provide additional trip and RCCA test data from the last five years to the WOO by March 8, 1996. Virgmia Power transmitted this information to the WOO by letter dated March 7, 1996.

In addition to the above WOO recommendations, Virginia Power will perform the following tasks:

Continue to follow the efforts of Westinghouse and Wolf Creek to determine the root cause of the problem. The Westinghouse effort will include evaluation of the North Anna drag test data. Following evaluation of these results, a determination will be made as to the applicability to continued operation of North Anna Unit 1 and Unit 2. Supplemental information on the cause, applicability to North Anna, and any required corrective actions will be provided prior to exceeding a burnup of 45,000 MWD/MfU in a rodded assembly.

Reactor Engineering will modify the end-of-cycle MIC test procedure to limit D-Bank rod insertion to no less than 216 steps withdrawn during the test.

Page 13

8.0 CONCLUSION

The highest average assembly bumup in a rodded location in N1C12 at the time when the cycle reaches end of reactivity is approximately 45,500 MWD/MIU. The highest average assembly bumup in a rodded location m Nl C12 is approximately 49,000 MWD/MfU at the end of the cycle's design bumup limit. The highest average assembly bumup in a rodded location in N2Cl 1

... *. at the time when the cycle reaches end ofreactivity is approximately 45,800 MWD/MIU. The

highest average assembly bumup in a rodded location in N2Cl 1 is approximately 49,400 MWD/MIU at the end of the cycle's design bumup limit. Based on data and evaluation presented in this report, it is concluded that the*operation ofN1C12 and N2Cll to the end of reactivity and to the design bumup limit is acceptable based on the following:

RCCA drag test data indicate that the 17Xl 7 V ANTAGE-5H fuel assemblies previously irradiated in North Anna 1 do not exceed the criteria established for

~RCCA drag for average assembly bumups up to 50,000 MWD/MIU. North Anna drag test data do exhibit a significant increase in RCCA drag force at assembly bumups greater than 45,000 MWD/MTU, but the magnitude of the RCCA drag force cloes not exceed limits established by Westinghouse.

i Two RCCAs could not be removed with normal operation of the RCCA handling tool in the spent fuel pool from two discharged fuel assemblies ( discharged in September 1994 from NlClO) on February 21, 1996. The*two RCCAs were temporarily placed into these assemblies in Jan1:1fil)' 1996 for staging P!llPOSes to be removed during the insert changeout phase durmg the upcommg refueling outage. The RCCAs were fully inserted into these two fuel assemblies. Also, these two assemblies were rodded assemblies during operation of Cycle 10. There were no reported problems with the operation of RCCAs during Cycle 10, and there were no reported difficulties in removing the RCCAs from tliese assemblies during the Cycle 10 to Cycle 11

refueling outage.

The two assemblies in the spent fuel pool which exhibited excessive resistance to removing RCCAs stored in them were specifically not included in the RCCA drag test measurement program due.to concerns with RCCA handling. However, the RCCA breakaway force was measured as the RCCAs were pulled free from the

assemblies. _The breakaway forces in these two ~semblies were approximately

110 pounds m one and from 140 to 170 rounds m the other. While these. forces exceed the F-Spec limit for drag in a fue assembly, these forces alone would not be sufficient to prevent the control rod from fully mserting in the core with the additional weiglit of the drive shaft attached.

Shutdown margin calculations Eerfonned for N1C12 and N2Cll indicate that excess shutdown margin exits If all rods in fuel assemblies with bumup greater than 45,000 MWD/MTU were to not fully insert (assumed to be 18 steps withdrawn). An administrative change will be irry)lemented to limit D-Bank insertion to no less than 216 steps withdrawn durmg the end-of-cycle MTC test to ensure adequate shutdown margm exists for the duration of the test.

The safety evaluation applicable to Nl C12 and N2Cl 1 in Section 6.3 indicates that an unreviewed safety question does not exist even if all rods in fuel assemblies with bumup greater than 45,000 MWD/MIU were to not fully insert (assumed to be 18 steps withdrawn).

..The conclusion that.Nl C12 and N2Cl 1 can operate to their respective design bumup limits is based upon the data obtained from the RCCA drag tests performed at Nortli Anna in February 1996 and current industry experience. However, Virginia Power technical staff personnel will

evaluate additional data as it becomes available prior to exceeding 45,000 MWD/MIU in a rodded fuel assembly to ensure this conclusion is still valid. If tlie Westinghouse root cause Page 14

evaluation links the phenomenon to fuel assembly growth, then the drag forces between fuel assemblies and RCCAs in N1C12 may be reduced at higher burnup relative to Zircaloy-4. The guide tubes in the rodded fuel assemblies in Nl C12 are manufactured of ZIRLO and not Zircaloy-4. ZIRLO has been proven to be more dimensionally stable with irradiation than Zircaloy-4. It should be noted however that ~ide tubes in the rodded fuel assemblies in N2Cl 1

.are manufactured of Zircaloy-4 and will transition to ZIRLO in the following cycle.

Page 15

AITACHMENT 2 CON1ROL ROD OPERABIU1Y REPORf SURRY UNITS 1 AND 2

Introduction

The NRC issued Bulletin 96-01, "Control Rod Insertion Problems," on March 8, 1996. The Bulletin was issued to alert-.licensees of control rod insertion problems at South Texas Unit 1 and Wolf Creek. North Anna was included in the Bulletin because of difficulty experienced while attempting to remove new control rods from discharged fuel assemblies in which they were

_.:temporarily stored. Even though North Anna was included in the Bulletin, there have been no reported control rod insertion problems identified during operation of either Unit 1 or Unit 2.

The Bulletin requires licensees to take specific actions. The purpose of this report is to specifically address Action 2 in the Bulletin which is:

Promptly determine the continued operability of control rods based on current information.

As new information becomes available from plant rod drop tests and trips, licensees should consider this new information together with data already available from Wolf

Creek, South Texas, and North Anna, and other industry experience, and make a prompt determination of control rod operability.

Continued Qpernbili1Y of Control Ro~ In Suny 1 and Sim 2 A review of the trip oata at South Texas Unit 1 and Wolfreek indicates a dependence of the anomaly on fuel assembly burnup. The phenomenon is not linear with burnup but appears to have a threshold between fuel assembly burnups of 40,000 to 45,000 MWD/MTU. All control

rods -at Wolf Creek and South Texas in assemblies with burnups below this apparent threshold have inserted properly. Drag force measurements performed at Wolf Creek indicate a similar dependence on burnup. That is, fuel assemblies iliat experience insertion problems also. exhibited higher than normal drag force. When comparing the measured drag values to Westinghouse.

acceptance criteria of less than 100 pounds in the dashpot and less than 40 pounds out -of the -

dashpot (Westinghouse F-Spec 5-1), it was noted that all assemblies that experienced insertion problems had drag forces above both criteria. However, the control rods that experienced no, :.

msertion problems had at least one drag force value below the acceptance criteria.

Drag force measurements were performed in the North Anna spent fuel pool in February 1996-(dunng the NlCl 1 to N1C12 refueling outage) when it was believed that the difficulty in

. removmg two new control rods from discharged fuel assemblies may be linked to the

phenomenon at Wolf Creek and South Texas. The reason why there was difficulty removing the control rods from the fuel assemblies that had been discharged for approximately 18 months has not yet been determined. However, both assemblies were rodded assemblies during NlClO, and there were no problems with control rod insertion during operation of Cycle 10 nor were there any reported problems with removing the control rods from these two assemblies during the NlCIO to NlCl 1 refueling outage.

The North Anna drag force measurements were obtained over a wide range of fuel assembly burnup, with emphasis on assemblies between 40,000 and 50,000 MWD/MTU. The data indicate an increase in drag forces, particularly with assembly burnups greater than 45,000 MWD/MTU.

Even though drag forces increased with high burnup, all assemblies that were tested met both of the Westinghouse F-Spec acceptance critena for control rod drag force.

Several Westinghouse 15X15 plants have provided recent trip data to Westin~ouse. Turkey Point Unit 4, Z10n Unit 2, and D. C. Cool( Unit 1 have all recently tripped with no anomalous control rod behavior. Westinghouse is continuing to develop a data base of information on Westinghouse plants that have experienced triQS within the past 5 years. This database indicates that, for the 24 plants evaluated thus far, 501 fuel assemblies had burnup greater than 40,000 MWD/MTU at the time of the trip. --Of these 501 rodded fuel assemblies, 106 had burnups greater than 45,000 MWD/MfU. None of the plants operating with these fuel assemblies reported problems with control rod insertion.

Page 1 J

e

. There have been no reported problems with control rod insertion at either Surry 1 or Surry 2 similar to the events tliat occurred at Wolf Creek and South Texas 1. Both Surry 1 and Surry 2 have operated with control rods in high bumup assemblies. Table 1 provides the rodded assemoly end-of-cycle (EOC) bumup history for Surry 1 and Surry 2 beginning with Cycle 10 of each Unit.

Table 1 No. of Rodded No. of Rodded

. lVJaximum Rodded Assemblies "With EOC Assemblies "With EOC Assembly EOC Cycle Bumgg Greater Than B~o8 Greater Than Bumup 40,00 ~/MIU 45,00 MWD/MIU (MWD/Mru)

SlClO 4

4 46,370 SlCll 20 8*

48,991 S1C12 25 0

42,703 S1C13 32 11 56,138 S1C14 0

0 35,803 (Current Cycle)

(Not EOC, but as of 2/29/96)

S2C10 4

0 42,190 S2Cll 29 0

42,359 S2C12 26 12 46,160 S2C13 4

0 42,111 (Current Cycle)

(Not EOC, but as of 2/29/96)

Table 1 indicates that there is considerable operating experience with rodded fuel assembly bumups exceeding 40,000 MWD/MfU with additional experience with rodded fuel assemblies exceecling 45,000 MWD/MfU up to 56,000 MWD/MIU. There have been no problems with control rod insertion in any of tliese assemblies.

The maximum rodded assembly bumup in the current Surry 2 cycle (Cycle 13) is expected to be less than 44,000 MWD/MfU.

This cycle is expected to end in May 1996. The maximum rodded assembly bumup in the subsequent Surry 2 cycle will be approximately 49,000 MWD/MfU at the maximum cycle design bumup. Similarly, the maximum rodded assembly bumup in the current Surry 1 cycle is approximately 46,500 MWD/MfU at the maximum cycle design bumup.

All these rodded assembly bumups are well within the experience base for operating with rodded assemblies at Surry. Therefore, operating with rodded assembly bumups of these magnitudes is not expected to. result in a control rod insertion problem similar to that experienced at Wolf Creek or South Texas 1.

While latching the RCCA drive shafts and performing the RCCA drag tests prior to operation of S1CI3, it was noted that at least six control rods exceeded the guidelines in l>rocedure OP-4.4 of 20 pounds of drag force. The greatest drag force was measured in location H-10 and was recorded as 51 pounds during withdrawal and 50 pounds during insertion. During a subsequent retest, the greatest measured drag force was 38 pollllds. The fuel assembly in location H-10 was a twice-burned assembly with a begipning of cycle bumup of 38,575.MWD/MfU. The drag forces are expected to be greater in fuel assemblies with higher bumup, and all measured drag

-.,forces were within their respective Westinghouse F-Spec limit. Subsequent rod drop time tests demonstrated that all rod drop times were well within their respective Technical Specifications limits. Also, there were at least four occurrences when the control rods were tripped into the Page2

\\

, reactor during the operation of Cycle 13. No anomalous control rod behavior was reported during any of these trips.

Control rod drop times for recent cycles were reviewed for any anomalous behavior. Control rod drop time tests are performed lUlder hot full flow conditions prior to initial startup of a cycle.

. Most control rods are resident in once-burned fuel assemblies, however, depending on the core

design, some twice burned fuel assemblies may be the host assemblies for control rods. Because of lower assembly bumup at the beginning of a cycle, rod drop times should not be affected significantly by control rod drag. The review of control rod drop* times confirmed that there were no anomalous trends in rod drop times recorded during the rod drop time tests.

The available shutdown margin for Surry Units 1 and 2 was evaluated in the unlikely event that control rods in fuel assemblies with burnups greater than 45,000 MWD/MfU fail to fully insert and remain withdrawn at 18 steps. Virginia Power technical staff personnel concluded that there is excess shutdown margin available in S1C14, S2Cl3, and S2Cl4 should the control rods with the above conditions fail to fully insert. This conclusion is based primarily on the fact that there are fewer affected control rods (i.e., control rods in assemblies with bumups greater than 45,000 MWD/MTU) in either of these Surry cycles when compared to the number of affected control rods in the analysis for shutdown margin performed for North Anna 1 Cycle 12 or North Anna 2 Cycle 11. The similarity in core design, total rod worths, stuck rod worths, and other core physics parameters between Surry and North Anna means that the net loss in available rod worth aue to 28 rods failing to fully insert assumed in the North Anna analysis would bolUld the net

loss due to at most 20 control rods failing to fully insert at Surry.

Recommended Actions Conclmion In addition to the operability evaluation documented in this report, Virginia Power will develop a testing plan to measure and evaluate rod drop t1me data and control rod drag forces (NRC Bulletin 96-01 Requested Action (3)). Also, Virginia Power will develop a plan to verify that control rods properly insert following a trip.

(NRC Bulletin 96-01 Requested Action (4)).

Virginia Power will continue to follow the Westinghouse root cause evaluation and lllaKe a determination as to the applicability to Surry Units 1 and 2 based on the root cause evaluation results.

Based on the data evaluation provided in this report, Virginia Power concludes that the control rods in Surry Units 1 and 2 remain operable in light of the information provided in NRC Bulletin 96-01 and that operation of S1Cl4, S2C13, and S2C14 to their respective design bumup limits is acceptable based on the following:

There have been no reported problems with control rod insertion at either Surry 1 or Surry 2 similar to the events that occurred at Wolf Creek and South Texas 1.

Both Surry 1 and Surry 2 have considerable operating experience with rodded fuel assembly bumups exceeding 40,000 MWD/MTU with acfditional experience with rodded fuel assemblies exceeding 45,000 MWD/MfU up to 56,000 MWD/MfU.

There have been no problems with control rod insertion in any of these assemblies.

In addition, there have been no reported problems with control rod insertion at other plants with similar 1 SXl 5 fiiel designs.

The maximum rodded assembly burnup in the current Surry 2 cycle (Cycle 13) is expected to be less than 44,000 MWD7MfU. This cycle is expected to end in Page 3

May 1996. The maximrun rodded assembly bumup in the subsequent Surry 2 cycle will be approximately 49,000 MWD/MTU at the maximrun cycle design bumup. Similarly, the maximrun rodded assembly bumup in the current Surry 1 cycle (Cycle 14) is approximately 46,500 MWD/MTU at the maximum cycle desigr}. bumup. All these rodded assembly bumups are well within the experience base for operating with rodded assemblies at Surry.

, A review of beginning-of-cycle control rod drop time results for recent cycles confirmed that there were no anomalous trends in control rod drop times.

Toe available shutdown margin for Surry Units 1 and 2 was evaluated even in the unlikely event that control rods in fuel assemblies with bumups greater than 45,000 MWD/MfU fail to fully insert and remain withdrawn at 18 steps. This evaluation concluded that there is excess shutdown margin available in S1C14, S2Cl3, and S2Cl 4 should these specific control rods fail to fully insert.

Page4

ATIACHMENT 3 FUELASSEMBLY AND CYCLE DATA IN ACCORDANCE Wl1H REQUIRED RFBPONSE (2)

e

, Cycle and fuel data for Surry and North Anna fuel cycles operating in 1996 are provided as to this report m accordance with Rt:x1urred Response (2) of Bulletin 96-01. A description of the attached.cycle and fuel data follows:

1) For each of the four cycles currently in operation (Surry 1 Cycle 14, Surry 2 Cycle 13, North Anna 1 Cycle 12 and North Anna 2 Cycle 11 ), the actual core map is provided which gives the identification numbers of both the fuel assemblies and any insert components for each core location. The insert component identifiers are listed above the fuel assembly numbers in these core maps, and the RCCA identification numbers begin with an 'R'. ( e.g., R97 or Rl 13).

A fmal core loading plan is provided for Surry 2 Cycle 14, which will begin operation in June. This figure differs from the actual ma_ps for the operating cycles in that 1t identifies the type of insert to be-used, but does not give identification numbers for insert components.

2) For each of the five cycles noted in Item 1, a page is provided which identifies all the*

batches of fuel (to beJ used in the cycle, the assembly ID notation used for each batch, and notable information about the design of each fuel batch. It should be noted that the Surry fuel is a l 5xl 5 design, and the North Anna fuel is a 17xl 7 design with standard (versus optimized) diameter rods. The grid and the fuel rod cladding materials are listed for each fuel batch. The guide thimble material is typically the same as the fuel rod cladding material; although during the implementation of low tin Zircaloy-4, the tin content of the cladding and guide thimbles could differ slightly. The gmde thimble inner diameters are S!Ven for both the upper portion of the tube and the daslipot region. In conjunction with the core map (Item 1 ), this information will allow characterization of the type of fuel assembly under each RCCA. For convenience, fuel batches which are used in control rod locations are also identified in these lists with an asterisk. These sheets also note the startup and shutdown dates for each cycle, and the maximum* cycle burnup.

3) In response to the request for current and projected EOC burnup for each rodded assembly, we have J>rovided lists of the assembly burnups in the rodded locations as of the end of February for the three cycles which were in operation at that time (Surry 1.

Cycle 14, Surry 2 Cycle 13, and North Anna 2 Cycle 11 J. For North Anna 1 Cycle 12, which just began operation during March, a quarter core map is provided which shows* the burnup distribution at the beginmng of the cycle. For these four operating cycles~ quarter core maps at the maximum cycle burnup are also provided, which: identity the maximum EOC burnups under the RCCAs. The RCCA locations are highlighted on all these quarter core maps.

For Surry 2 Cycle 14, which will not begin operation until June, a quarter core map is provided which shows limiting values for the beginning of cycle and end of cycle fuel assembly burnups for each rodded location in the core. The actual values for Surry 2 Cycle 14 may slightly differ from the maximum values shown, and will depend on the actual end of cycfe burnup for Surry 2 Cycle 13.

Page 1

e

Similar information will be provided for North Anna 2 Cycle 12 when it becomes available later this year.

It should be noted that both Suny and North Anna use Ag-In-Cd control rods. The original Surry control rods were replaced with Westinghouse Enhanced Performance ( chrome plated)

....control rods after Cycle 10 at each unit. The original North Anna 1 control rods were replaced with Westinghouse Enhanced Performance control rods during the recent refueling outage, and

, the North Anna 2 control rods are similarly scheduled to be replaced.during the Unit 2 refueling outage this fall.

Page2

Surry 1 Cycle 14 Criticality Date: 10/19/95 Projected Shutdown Date: 2/18/97 Current Cycle Burnup (as of2/29/96): 4522 MWD/MTU (128 EFPD)

Maximum Cycle Burnup: 19,700 MWD/MTU (559 EFPD)

Batch No. ofF/As F/AIDs Description Sl/10 17 nEn Inconel grids, std Zr-4 clad Sl/12B*

1 nGn Zr-4 grids, std Zr-4 clad Sl/14A*

7 nJn Zr-4 grids, low Sn Zr-4 clad Sl/14B 12 nJn Zr-4 grids, low Sn Zr-4 clad Sl/15A*

28 nKn Zr-4 grids (rotated), low Sn Zr-4 clad Sl/15B*

28 nKn Zr-4 grids (rotated), low Sn Zr-4 clad Sl/16A 36 nnA ZIRLO grids, ZIRLO clad Sl/16B 28 nnA ZIRLO grids, ZIRLO clad Notes:

1. Guide thimble composition is same as fuel cladding.

Thimble ID, _in.

(Upper/Lower) 0.512 I 0.455 0.499 I 0.455 0.499 I 0.455 0.499 I 0.455 0.499 I 0.455 0.499 I 0.455 0.499 I 0.455 0.499 I 0.455

2. Assemblies from batches marked with an asterisk are in rodded locations.

£ e

!"'surry - Unit 1 Core Map created on 03/04/1996 Actual Map 15 14 13 12 11 10 9

8 7

6 5

4 3

2 A

SJO FS003 4JO SS1 3J8 B

C FS011 1E8 SS7 BP437 4E9 03A R145 BP421 1J4 SSA BP406 R148 41A 5K9 R103 BP389 3K6 45A BP405 R110 62A 4KB R113 BP423 OJS SOA BP439 2E3 31A FS009 3EO D

E F

R131 1E9 2J4 FS013 BP435 BP419 SEO 33A 63A R146 BP427 R102 3K3 43A 5KB BP429 R105 BP411 48A 4K5 16A R142 BP413 R118 4KO 35A 1K1 BP385 BP386 11A 2KO 23A BP390 R130 OK7 13A 3KO BP393 BP394 36A 2K1 29A R126 BP415 R129 6KO 06A 3K1 BP431 R121 BP417 39A SKS 25A R140 BP433 R147 5K1 52A 5K4 FS007 BP441 BP425 OE3 09A 46A R108 OE9 2J7 e

!'!*"cycle Nl.lllber 14 ~

G H

J K

L M

N p

R FS004 3J9 6J1 5J8 BP407 R134 BP408 R119 57A 4K9 37A 2J6 4E7 R136 BP379 R141 BP420 BP436 FS012 3K4 54A 4K3 47A 22A 3E4 BP380 BP381 R114 BP428 R137 FS010 27A OK3 14A 5K7 60A 5K6 2E4 BP382 BP412 R127 BP430 BP438 2K5 20A 3K2 10A 4K1 40A 15A OE8 BP383 R139 BP384 R122 BP414 R144 BP422 R128 08A 1K4 OSA 2K8 32A 4K7 51A 2J8 R109 R117 BP387 BP388 R123 BP409 SS2 OK2 1K3 1K6 18A 2K6 24A 4K2 58A 3J6 BP378 R111 BP391 BP392 R135 FS002 OK4 3E8 OKS OK6 04A 1KO 64A 3K5 5J7 R120 R112 BP395 BP396 R138 BP410 1KB 2K3 1K9 02A 2K2 12A 5K2 44A 6JO BP397 R104 BP398 R106 BP416 R115 BP424 R125 19A OK1 28A 2K7 30A 3K7 61A OJ2 BP399 BP418 R124 BP432 BP440.

2K4 34A 2K9 07A 5K3 53A 26A 2E5 BP400 BP401 R107 BP434 R149 FS008 01A 1K2 21A 3K8 42A 4K4 1E3 R132 BP402 R143 BP426 BP442 FS006 3K9 38A 4K6 56A 17A 1E5 BP404 R116 BP403 R133 59A SKO 49A 4G7 4E8 FS001 4J9 4J7 5J6

e THE SAS SYSTEM e

CYCLE=S1Cl4 ------------------------------------------------------------

CORE ASSEMBLY TYPE BURNUP BANK LOCATION 4G7 SIF 35803 A BANK K 2 2.J6 SIF 34733 A BANK K14 2.J7 SIF 34675 A BANK F 2 2.J8 SIF 34622 A BANK PIO 2.J4 SIF 34511 A BANK Fl4 O.J5 SIF 34139 A BANK B 6 O.J2 SIF 34099 A BANK p 6 1.J4 SIF 33812 A BANK BIO 2K7 SIF 28100 D BANK K 6 3Kl SIF 28089 D BANK F 6 lKl SIF 28072 D BANK no 2K8 SIF 28042 D BANK KIO OK2 SIF 27824 SB BANK G 9 1K6 SIF 27806 SB BANK

.J 9 1K9 SIF 27776 SB BANK

.J 7 3KO SIF 27729 C BANK F 8 OKI SIF 27669 C BANK H 6 1K8 SIF 27590 SB BANK G 7 1K4 SIF 27377 C BANK HlO OK6 SIF 27170 C BANK K 8 5K2 SIF 26086 SA BANK N 7 5K9 SIF 26073 SA BANK C 9 3K4 SIF 26021 SA BANK G13 5K4 SIF 25990 B BANK F 4 3K9 SIF 25901 SA BANK G 3 4KO SIF 25897 B BANK D10 4K3 SIF 25853 SA BANK

.J13 3K7 SIF 25845 B BANK M 6 4K2 SIF 25829 SA BANK N 9 3K6 SIF 25794 D BANK B 8 4K8 SIF 25733 SA BANK C 7 4K7 SIF 25657 B BANK MIO 5K8 SIF 25647 B BANK F12 3K8 SIF 25590 B BANK K 4 5K7 SIF 25500 B BANK Kl2 4K6 SIF 25498 SA BANK

.J 3 6KO SIF 25445 B BANK D 6 4K9 SIF 25327 D BANK H14 5KO SIF 25325 D BANK H 2 3K5 SIF 25164 D BANK p 8 4K5 SIF 23256 SB BANK Ell 5K3 SIF 23061 SB BANK L 5 5K5 SIF 22992 SB BANK E 5 4Kl SIF 22949 SB BANK Lll A5ScM8J.'I GIJA.NIJP.5 5K6 SIF 20947 C BANK M12 J..OCAT10IJ S 4K4 SIF 20899 C BANK M 4 JI,./ RO~AE'b 3K3 SIF 20892 C BANK D12 AS OF 2/Z9/9t.-

SKI SIF 20667 C BANK D 4

j

~ *

";I

i f.'

[i i

i' '=

1-

i.

\\*

~11

I I

I l e r'f ti' 8

9 10 11 12 13 14 15 FIGURE 3-13 SURRY UNIT 1, CYCLE 14 FUEL BURNUP DISTRIBUTION MAP AT 19700 MWD/T BURNUP H

G F

E D

C B

A J..-LJ..J..J..J..J..J..J..J..J..J..J..J..J..-LJ..J..J..J..J..J..J..J..J..J-1-L-l-J.-1-.L,J-J.-J-J.J.J-1.J...I t l!J t tJ..J..J..J..J..J-LJ-L.J-1.J-J....1.J..J..J-L nnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnn--nnnnnnnnnnnnnnnnnnnnn-nnnnnn

38042

43493

44872

26129

46187

24839

39587 45240 *

1080

1235

1274

742

1311

705

1124 1284 *

.J...J.. J..J..J....,...J.. J..J..J.... L-LJ...J...J.. J....J..;

J..J...l...LJ.-1-LJ....J..-!-t.....t..LJ.J...J..J..~J..J..J..J..J.,J.. f I

I P I I

t O.L-L-1 nnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnnn-nnnn

43496

44389

25490

46666

26115 43461 20363

39630 *

1235

1260

723

1325

741 1234 578

1125 *

- - - - "fr'*'*~'#""...... 0""...... 7.J..++J..J..J..J.,*********

44869

25497

26405

44687

24290

42625

1274

724 749

1268

689

1210

- - - - -f**** n *-f-*******~*****~.. *********

n n n n n n

n '-k

26140

46682

26463 42518

24290

19708

42285

742

1325

751

1207 n

689

559

1200 *

- - - - +

-!: 46194

26130 -

24287

34169

42574 *

1311

742 1269 689

970 -

1208 *

- - - - -;':**J.,**-';-**i-*********************************

24844

43462

24288

19661

42863

705

1234

689

558

1217 *

-1****************..1:******************* J.. :.. ;,, J..

-/ 39588 20362

42621 42284 *

1124 578 1210 1200 *

45246

39631 *

1284

1125

J..-LJ..J..J...LJ.J..J..-LJ..J..J..J..J..J..J..

MWD/T EFPD S1C14 Design Report - NE-1045, Rev. 0 Page.. 61 of 154

Surry 2 Cycle 13 Criticality Date: 3/19/95 Projected Shutdown Date: 5/1/96 Current Cycle Bumup (as of2/29/96): I0,857'MWD/MTU (315 EFPD)

Maximum Cycle Bumup: 14,000 MWD/MTU (404 EFPD)

Batch No. ofF/As F/AIDs Description Sl/9A 1

nDn Inconel grids, std Zr-4 clad Sl/13B*

4 nHn Zr-4 grids, mixed std and low Sn Zr-4 clad S2/9A 16 nRn Inconel grids, std Zr-4 clad S2/13A 12 nVn Zr-4 grids, low Sn Zr-4 clad?

S2/14A*

32 nWn Zr-4 grids, low Sn Zr-4 clad S2/14B*

36 nWn Zr-4 grids, low Sn Zr-4 clad S2/15A 28 nXn Zr-4 grids (rotated), low Sn Zr-4 clad S2/15B 28 nXn Zr-4 grids (rotated), low Sn Zr-4 clad Thimble ID, in.

(Upper/Lower) 0.512 I 0.455 0.499 I 0.455 0.512 I 0.455 0.499 I 0.455 0.499 I 0.455 0.499 I 0.455 0.499 I 0.455 0.499 I 0.455 Notes:

1. Guide thimble composition is typically same as fuel cladding. (Tin contents of Sl/13 guide thimbles and S2/13 cladding and guide thimbles are unverified.)

2. Assemblies from batches marked with an asterisk are in rodded locations.

. ~--Surry -

2 Core

Map created on 03/04/1996 Actual Map 2

3 4

5 6

7 8

9 10 11 12 13 14 15 R

6RO 1V2 SS3 4R6 p

N 1V6 SS4 3R8 1W1 R183 BP373 4W7 sxo BP361 R177 4X3 SW1 R157 BP337 4W2 3X8 BP357 R176 4X1 6W2 R162 BP374 6W6 sxs SR7 1W9 1V7 M

3V2 R175 3H6 BP372 SX3 R159 3W5 BP348 2X3 2W4 BP349 2X7 R164 SW4 BP375 4X6 R192 3H9 2V3 L

K J

1R8 R187 BP354 4R4 6W3 3X2 BP370 R150 2W3 3XS SW2 BP371 R172

. BP346 SX2 3W9 oxs R178 BP353 OW6 1X3 2WS BP347 R158 BP329 1X2 SW7 1X6 BP330 R184 1W4 OX4 4W1 BP331 R193 1X9 1WS 1W3 BP332 R196 1W8 OX3 3W6 BP350 R153 BP333 2X1 sws 2X5 R189 BP351 ows 2X4 OW4 BP376 R197 BP352 4X4 4W4 2XO BP377 R169 1W6 4X5 6W4 R194 BP360 0R1 4W8 4X7 2R9

.. e

,,., Cycle Nllli:Jer 13

.e H

G F

E D

C B

A 1VB OR2 R155 BP358 R174 3W7 3X9 6W1 4R2 BP336 R160 BP369 3X1 6WS 3X6 2WO 3V1 BP345 R173 BP368 R168 OW3 1X8 4W6 3XO 4H1 OVB BP328 BP344 R182 BP367 OXB 3WO 1X7 OW2 3X4 2W7 SRO R191 BP327 R161 BP343 R180 BP366 R170 2W2 OX7 6W7 1X4 3W3 SX6 SW6 R195 BP326 BP342 R186 BP355 2W1 4W9 1X1 2W8 OX9 6WO SX4 3R4 R190 BP325 BP335 R154 1DS OW9 2W9 OX1 1W7 2X9 3W4 zvo R166 BP324 BP341 R163 BP359 SS5 OWB 4W3 2X8 3W1 2X6 6W8 3X7 5R3 R181 BP323 R151 BP340 R152 BP365 R188 3W2 1XO SW9 OX6 3W8 4XO 4WO BP322 BP339 R167 BP364 OX2 2W6 1XS 1WO 3X3 OW1 5RS BP338 R165 BP363 R185 OW7 2X2 SW3 4X8 4H4 1V3 BP334 R179 BP362 5X1 SWB 4X2 1W2 2V4 R156 BP356 R171 4WS 4X9 swo 3R1 OV3 1R9

THE SAS SYSTEM -

CYCLE=S2Cl3 ------------------------------------------------------------

CORE ASSEMBLY TYPE BURNUP BANK LOCATION 3H6 SIF 42111 C BANK M 4 4Hl SIF 41885 C BANK D 4 4H4 SIF 41800 C BANK D12 3H9 SIF 41775 C BANK Ml2 6W7 SIF 38596 D BANK F 6 5W7 SIF 38434 D BANK K 6 5W9 SIF 38006 D BANK FlO 5W5 SIF 37954 D BANK KlO 2W9 SIF 37811 C BANK F 8 3W2 SIF 37755 C BANK HlO 1W5 SIF 37632 C BANK K 8 2W2 SIF 37603 C BANK H 6 3W8 SIF 37521 B BANK D10 4W9 SIF 37434 SB BANK G 7 4W6 SIF 37263 B BANK F 4 4W3 SIF 37252 SB BANK G 9 3W6 SIF 36923 SB BANK J 9 4Wl SIF 36917 SB BANK J 7 4W4 SIF 36895 B BANK Kl2 5W3 SIF 36815 B BANK Fl2 5W4 SIF 36692 B BANK HlO 3W5 SIF 36459 B BANK H 6 3W3 SIF 36308 B BANK D 6 3W9 SIF 36011 B BANK K 4 6W8 SIF 35594 SA BANK C 9 6W5 SIF 35573 SA BANK G 3 6W4 SIF 35410 SA BANK Jl3 5Wl SIF 35279 SA BANK N 7 6WO SIF 35013 SA BANK C 7 5W2 SIF 34740 SA BANK J 3 5W8 SIF 34646 SA BANK Gl3 6W2 SIF 34369 SA BANK N 9 3W4 SIF 33436 D BANK B 8 lWO SIF 33414 SB BANK Ell 4W2 SIF 33307 D BANK p 8 3W7 SIF 33276 D BANK H 2 4W5 SIF 33122 D BANK Hl4 OW2 SIF 32871 SB BANK E 5 OW6 SIF 32796 SB BANK L 5 OW5 SIF 32614 SB BANK Lll 6Wl SIF 32385 A BANK F 2 4WO SIF 32050 A BANK BlO 5W6 SIF 31939 A BANK B 6 6W6 SIF 31861 A BANK PlO ASSSH8LY 81JRNIJP5 6W3 SIF 31794 A BANK K 2 IN AObDetl J...OCAnc/JS 4W8 SIF 31765 A BANK Kl4 4W7 SIF 31527 A BANK p 6 212.q/q, 5WO SIF 31427 A BANK Fl' AS OF

I I **

I I

I II I

11

I I

I I

I I*

I 1*

I F'IGURE 3-10 SURRY UNIT 2, CYCLE 13.

FUEL BURNUP DISTRIBUTION MAP AT 14000 MWD/T BURNUP H

G F

E D

C B

A J..J....r-,..J,.**tl l!J..J..J....t-1.J.J-!..J..J..J..J..J..J..J..J..J..l!II I 11111 I 11***-I.J-LJ.J.II !I I I I I 11 IJ..J..J..J..J..

- ~~~~;-:**;~~;~ -:r~~~~;-~*- ~~~~~**; -~~;~;-; -~~;~~ --*;~;~ ~. =** ;; ~~~**:

8*

9 10 11 12 13 14 15 J..J..J-L.-'-'-'-',.J..J-L.J..I 11!1 IJ-J..J-L.J-t.~1111I111111111111 II I I

I Ill P !IJ..J..J..J.-l.-L,J-1-l..

34891 40994

18072

40973

18444 - 39038.. 15227 *

J..J..-t-J..J..J-L*~*""'"-t-.t.J..

40447

18063

18720

40731 16846

- - - - ~*~*1\\ A J.. n "" n ********** J...,\\ J,. !. l"*"'" n

18404

40965

18706

36533

17403

29603

37332 *

-:~~;;*1

~;~~T~~~~~*~*-

- - - - AAAAAAAAAAAAAA**AAAA

18319 ~ 39047 ~ 16851: 29542: 40576:

38345 J..J..J..al-L.J..*J..J.J..J..J.J..J..J.-t-1.J..J..J..J..J..J..J..J..J.-t..l II! I I..L..L..L-LJ..i-1..

D. * *

~~~;~;**:r ;;;~;r;;~~;-~*--******

- - - - AA:...,\\!.. l.. :.. :..

MWD/T

- "':************\\ :.. !.. l.. :.. J.. ;,. :.. J.. :.. **** :

.. J:.. A A A J.. :.. :.. **AA J,. J,. I'****

37556

38279

S2C13 Design Report - NE-1012, Rev. 0 Page 53 of 136

Surry 2 Cycle 14 Projected Startyp Date: -6/10/96 Maximum Cycle Burnup: 20,800 to 22,000 MWD/MTU (Depends on actual EOC shutdown of S2C13)

Batch Sl/15A S2/13B S2/14A*

S2/14B*

S2/15A*

S2/15B*

S2/16A S2/16B Notes:

Thimble ID, in.

No. ofF/As FIA IDs Description (Upper/Lower) 1 nKn Zr-4 grids (rotated), low Sn 0.499 I 0.455 Zr-4 clad 8

nVn Zr-4 grids, low Sn Zr-4 clad?

0.499 I 0.455 12 nWn Zr-4 grids, low Sn Zr-4 clad 0.499 I 0.455 20 nWn Zr-4 grids, low Sn Zr-4 clad 0.499 I 0.455 28 nXn Zr-4 grids (rotated), low Sn 0.499 I 0.455 Zr-4 clad 28 nXn Zr-4 grids (rotated), low Sn 0.499 I 0.455 Zr-4 clad 32 nYn ZIRLO grids, ZIRLO clad 0.499 I 0.455 28 nYn ZIRLO grids, ZIRLO clad 0.499 I 0.455

1. Guide thimble composition is typically same as fuel cladding. (Tin content of S2/13 cladding and guide thimbles are unverified.)

2. Assemblies from batches marked with an asterisk are in rodded locations.

VP-NES-N.

0 90 INCORE DEVICE DESCRIPTIONS:

RCC-FULL LENGTH C SSX-SECONDARY SOU 4P-4 BURNABLE P SP-8 BURNABLE P 16P-16 BURNABLE P 20P-20 BURNAB_LE P R

p N

I I

I OW1 I

I SS4 4P I

6W8 1Y8 I.

I RCC 16P I

4WO 5YO 4P RCC 3VS 5Y7 3X7 RCC 20P 3W4 2X9 4Y5 SS3 4P RCC 4V5 5Y5 5X4 RCC 16P SW6.

SY9 4P 6WO OYS 2W7 ONTROL ROD RCE OISON ROD CLUSTER OISON ROD CLUSTER OISON ROD *CLUSTER OISON ROD CLUSTER M

L I

I I

I swa

. 4P 1W2 OY1 RCC 16P 2W1 4Y1

16P RCC 4Y4 OX4 RCC 20P 5X3 2YO 20P 1Y1 2X3 20P OX1 5X5 20P 2YS 2X7 RCC 20P 4X6 OY2 16P RCC SY2 OX3 RCC 16P.

1W3 4Y3 4P 2WO 2Y9 6WS e

SURRY ~N~2 -- CYCLE 14 PAGE 1 OF 2 FINAL CORE LOADING PLAN K

I I

I RCC swo 16P 3Y5 RCC 5X2 20P 1Y9 RCC 1X6 20P 2Y6 RCC*

1X7 20P 1Y6 RCC 2XS

20P 1Y7 RCC 4X4 16P 3Y9 RCC 6W1 REVISION NO, 0 J

sva 4P 3Y3 RCC 4X9 20P OY4 oxs 20P 2Y4 RCC 1X2 BP 5XO RCC 2X1 20P OY9 2XO 20P 2Y8 RCC 3X9 4P 3Y7 4VO H

4W5 RCC 5X1 20P

.5Y1 OX2 20P 3X6 RCC 1X5 BP 3X5 OKS SP 4X2 RCC 1X3 20P 4XS oxa 20P 5Y3 RCC 3X1 3W7 I 0 0

G 6V2 4P 3Y8 RCC 4X7 20P 1Y2 1X8 20P 1YO RCC 1X4 BP 4XO RCC OX6 20P 2Y7 2X2 20P 2Y1 RCC 3X2 4P 4Y7 SV2 F

E D

C B

A I

I I

RCC I

I I

I 4W8 6W4 I

I I

16P 4P I

I I

4YO OYB 1W6 I

I RCC 16P RCC I

I 3XO 5Y6 OW9 1W9 I

20P RCC 16P 4P I

3Y1 1X1 4Y2 1Y4 6W2 I

RCC 20P RCC 16P RCC I

OX7 3Y2 3X4 3Y6 6W6 20P 20P RCC 4P OY3 OX9 OY7 4X1 3Y4 4V9 RCC 20P 20P RCC 2X4 5X6 1X9 6YO 3X8 4W2 20P 20P RCC 4P sss 2Y3 2X6 3YO 4X3 4Y6 svs RCC 20P RCC 16P RCC 1XO 1Y3 3X3 4Y9 4W7 20P RCC 16P 4P OY6 2X8 5Y4

1Y5 5W1 RCC 16P RCC 4X8 4Y8 owe 1W1 * -----------

16P 4P SYS 2Y2 2W3 RCC 6W3 5W2 CONCURRENCE BY: e,t.f_ ~

APPROVED BY: ~~~

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 DATE: t/11/tre, DATE: \\ / 11 \\q.(_

S2C14 Assembly Average Burnups For Assemblies In Control Rod Locations (Limiting Burnups From The Low And High Burnup Windows - Preliminary)

C D

~.. "* **-

.,._.. -.~*-. **-*

18687~

18325 -

43323, 37905 -

SB SA 18725,.

15235-43272--

40430, C

D B

A 18681 18062-17411 *

33886, 43317-44512' 43844,.

45706

SB

18070, 43737 -

B C

17400-34878,.

43822

49101,.

SA

15253, 40434,.

D A

18328 -

33923.-

37899,.

45727, Bank BOC Burnup EOC Burnup Notes: All Burn ups are in units of MWD/MTU.

These burnups are best estimate and contain no uncertainty.

Low-Window: EOC13=12,000 MWD/MTU, Cycle Length=22,000 MWD/MTU.,

High-Window: EOC13=14,000 MWD/MTU, Cycle Length=20,800 MWD/MTU.

North Anna 1 Cycle 12 Criticality Date: 3/10/96 Projected Shutdown Date: 5/16/97 Maximum Cycle Bumup: 20,500 MWD/MTU (515 EFPD)

Batch No. ofF/As FIA IDs Description N2/11B 1

nLn Zr-4 grids, low Sn Zr-4 clad Nl/1 lB 8

nAn Zr-4 grids, low Sn Zr-4 clad Nl/12A 8

nBn Zr-4 grids, low Sn Zr-4 clad Nl/12B 12 nBn Zr-4 grids, low Sn Zr-4 clad Nl/13A*

28 nCn ZIRLO grids (rotated),

ZIRLO clad Nl/13B*

36 nCn ZIRLO grids (rotated),

ZIRLO clad Nl/14A 28 nDn ZIRLO grids (rotated),

ZIRLO clad Nl/14B 36 nDn ZIRLO grids (rotated),

ZIRLO clad Thimble ID, in.

(Upper/Lower) 0.442 I 0.397 0.442 I 0.397 0.442 I 0.397 0.442 I 0.397 0.442 I 0.397 0.442 I 0.397 0.442 I 0.397 0.442 I 0.397 Notes:

1. Guide thimble composition is typically same as fuel cladding.

2. Assemblies from batches marked with an asterisk are in rodded locations.

North Anna - Unit 1 Core Map created on 01/22/1996 Conflict Check Map 2

3 4

5 6

7 8

9 10 11 12 13 14 15 R

VS016 4B5 VS024 OB1 VS028 3B8 p

N VS027 1BO VS042 BP502 5A9 506 R139 BP492 2C4 303 BP475 R117 400 3C9 R157 BP458 4C7.

4D8 BP476 R118 3D4 4C9 R150 BP491 1CO 4D3 VS048 BP507 5A3 601 VS022 1BS tJAP.S 1 Cyr:./e /2. Core.;£1"/J ~v d l

After NAPS 1 CYCLE 12 CORE ONLOAO

~

M L

K J

H G

F E

0 C

B A

VS023 VS007 VS025 5B2 0B4 2B6 VS043 R125 BP470 R146 BP469 R128 VS044 4A5 2C2 209 4C8 307 ocs 4A9 VS026 BP501 BP485 R154 BP457 R156 BP486 BP504 VS003 4B3 404 501 5C8 401 3C4 405 308 4BO R114 BP493 R152 BP450 SS19 BP449 R149 BP496 R138 VS015 2C6 602 5C6 1D4 2CO 2D7 3CS SD4 OC9 OB8 BP494 R119 BP481 BP448 BP480 R135 BP495 BP503 VS045 407 6C3 1D6 2C1 OD2 1C1 1D3 2C9 3DO 502 4A2 R121 BP482 R134 BP462 R155 BP461 R123 BP479 R124 BP487 R140 6C1 OD7 2C3 1D9 4C1 009 1C3 2DS scs 600 OC7 BP455 BP467 R136 BP444 R147 BP464 BP456 R133 BP472 VS017 1D2 1C7 1DO 3C3 3C6 4C3 1D5 OC6 1D7 5C9 309 6BO BP451 R112 BP443 BP442 R127 BP447 BP460 R142 VS018 OC8 202

3C1 sco 4L8 5C7 6C4 1D8 OC3 SD3 SC4 1B7 BP454 BP468 R158 BP441 R111 BP463 BP453 R129 BP471 VS020 ODS 1C4 2DO 6C2 5C1 4C6 203 2C8 0D6 4C5 305 3BO R131 BP483 R132 BP466 R144 BP465 R122 BP478 R126 BP488 R151 5C3 2D1 1C8 2D8 SC2 003 OC4 OD8 4CO 4D6 OC2 BP499 R120 BP484 BP445 BP477 R130 BP498 BP506 VS041 306 4C2 2D4 1CS 0D1 OC1 OD4 3C8 4D9 5D7 3A6 R115 BPSOO R141 BP452 SS20 BP446 R137 BP497 R113 VS030 2CS 3D1 3C2 2D6 2C7 1D1 3C7 4D2 1C9 2B3 VS032 BP508 BP490 R143 BP459 R145 BP489 BPSOS VS029 3B5 sos 603 3CO 604 6CO 3D2 SD9 2B8 VS046 R148 BP474 R153 BP473 R116 VS047 6A7 1C6 508 4C4 5DO 1C2 5A2 VS031 VS019 VS021 4B1 0B9 4B7 PREPARED B'l fl* -

DATE ~-:.f.;)i REVIEWED BY

_ DATE

. (

,.,1.,.

FIGURE 3-1 FUEL BURNUP DISTRIBUTION MAP AT O MWD/MTU NORTH ANNA UNIT 1, CYCLE 12 H

G F

E D

C B

A

24425

19314

24163 0

23968

0 33252

8

615

486

609 0

604

516

837

I I

I I

I I I.I I I

I I

I

19326 9

19314 486 0

24235

0

23276

0

42433

10 11 12 13 14 15 487 0

0

24027 *

605 0

0

516

33258

838

0 0

41696 *

1050

AAAAAAAAAAAA*AAAA 0

610

0 *

0

0

0

0

0

37537 *

945

N1Cl2 Design Report - NE-1063, Rev. 0 0

0

0

24820 625 38284 *

964

586 0

1069

0

38159

0

41577

1047

961 *

.: *.: *.,\\...,..... l *.... :. :. :. :.,\\,\\

MWD/T EFPD i\\. J. !.

I, !. ~*.,\\ A!. l. ***

Page 46 of 152 I

I I

11 I

I I

8 9

10 11 12 13 14 15 H

FIGURE 3-13 FUEL BURNUP DISTRIBUTION HAP AT 20500 MWD/MTU NORTH ANNA UNIT 1, CYCLE 12 G

F E

D C

B A

- l.. J.. :.. :.. :.. !, J.. !.. ! J., !. :.. :..,'. !.. '" !.. J..,'... !. J.. :.. J.. ! l* :.. !.. !.. :.. A A A J, A :.. A *'* k I; !.. :.. ;. *\\ J.. J; a\\ :.. !,. h :.. *1rn*irn* l, a',,!.. :.. !.. :..

47232

42817

48033

26128

47841

26167

39301

40542 *

1190

1078

1210

658

1205

659

990

1021

J.J...,J.J.

J-1.. I I

I I

I

,..,, ** n ** ** ** """'"....., ** n ""..,,,. ""'"""...,..,,, ** " ** *** ** h ** n **.,..,,.,..,...,,,,..,, '""",,,.. **** **** '""

42911

43598

26145

48288

25982

46277

21163

I*

48326 *

1081

1098

658

1216

654

1165 I*

533

1217 *

-/:~** !.. :.. }, J. J..,'. :.. :.. :., :.. :.. -!.!rl.*-lrl: !.. !.. !.. :.. !.. !.. !.. :..,'.. !, !.. A 1,',., A A:.. !.. !.. -1rn*** :.. A a\\ :.. J.,, ;\\ 1.. l, ***a\\!.. !.. a\\ a\\ :. !.. !.. !

48137 : 26209 : 48874 ~ 26761 : 48383

25375 =137459 ~

1212 660

1231

674

1219 639

943 *

- '\\. J,. J..,'..***}.,,!.. !..,\\.'t.,\\ J. }., n **,11

"" "" ** '",,,. **** * '",,,, n,,.,, '""

, J.,.1\\ J. I* J,. ;\\ }.*

26169

48215

26756 659

1214

674

47905

25994 n 48413 45334 1142 25768

21812

45372

649

549

1143

25775

41621

49309

1242 *

- - - - /I: },.!/,.~)*,.:.:.:.,,*********\\~/..::;;:'J,.~:.....

i.::,.~

.... ~

... :~

... \\~,',...Jll.. ~1.-!,,',.+.J,.****

1206

655 n

1219 649 1048

26195

46308 25376

21874

46328

660

1166 639

551

1167 *

!.. :.. *\\ *\\ J...... !.. :.. J.. !..,'.. ***

21198

37443 534

943 44787

1128 *

- -lffi-!:**** :

.. A.\\ ! *.'..\\ l.. :..'* :.,*. :. !. *..............................

40582

47667 *

1022

1201 *

- - - .*. :~.... ! * * *. ! * * * ** *.. :.. !.. J.. }. *'..

MWD/T EFPD J..l,.....l.t t t

I I

I I**

N1C12 Design Report - NE-1063, Rev. 0 Page 58 of 152 I

I I

I

North Anna 2 Cycle 11 Criticality Date: 5/31/95 Projected Shutdown Date: 9/14/96 Current Cycle Bumup (as of2/29/96): 10,684 MWD/MTU (269 EFPD)

Maximum Cycle Bumup: 20,200 MWD/MTU (508 EFPD)

Batch No. ofF/As F/AIDs Description N2/6 8

Tnn Inconel grids, std Zr-4 clad N2/9A 1

Xnn Inconel grids, std Zr-4 clad N2/11A 4

nLn Zr-4 grids, low Sn Zr-4 clad N2/11B 16 nLn Zr-4 grids, low Sn Zr-4 clad N2/12A*

28 nMn Zr-4 grids (rotated), low Sn Zr-4 clad N2/12B*

36 nMn Zr-4 grids (rotated), low Sn Zr-4 clad N2/13A 28 nNn ZIRLO grids (rotated),

ZIRLO clad N2/13B 36 nNn ZIRLO grids (rotated),

ZIRLO clad Thimble ID, in.

(Upper/Lower) 0.450 I 0.397 0.450 I 0.397 0.442 I 0.397 0.442 I 0.397 0.442 I 0.397 0.442 I 0.397 0.442 I 0.397 0.442 I 0.397 Notes:

1. Guide thimble composition is typically same as fuel cladding.

2. Assemblies from batches marked with an asterisk are in rodded locations.

North Anne - Unit 2 Core Hep created on 01/26/1995 Conflict Check Hep 15 14 13 12 11 10 9

8 7

6 5

4 3

2 A

VSOOS 4L9 VSOOS 1L4 VS014 6LS B

C VS034 4L4 BP43S T09 6NO RS7 BP422 1H7 4N7 BP407 R6S 4N2 6H3 R103 BP390 4H9 5N3 BP40S R76 5N7 3H6 R97 BP421 2H7 3NS BP439 T29 5NO VS036 6L6 D

VS010 SLO R67 2H3 BP432 3N7 R62 5HS BP3SO 2N4 OH4 BP384 1N1 R107 3H9 BP431 3N2 R65 1H4 VS001 SL6 E

F G

VS035 6L1 R88 BP403 T16 OH7 5N5 BP437 BP419 R73 4N6 3N6 SH2 BP427 lis2 BP377 3N1 5H7 1N6 RS5 BP409 3H2 ON3 1HO BP411 R64 BP393 1NS 2HS 2N6 BP395 R101 1H9 1N4 SH1 BP3S2 R7S BP374 2N1 3HO 4H1 BP397 RS1 OH2 ON4 SH9 BP413 R93 BP399 2N7 1H8 2N3 R100 BP415 4M7 1NO 0M6 BP42S R75 BP3S7 6N2 SH6 1NS BP440 BP41S R102 4N4 4NS 4H2 RS9 BP401 T39 2H6 2N9 VS033 SL1 e

H J

K VS004 VS040 2L7 4L1 R104 BP404 R69 4H3 3NO ZHO BP3S9 R74 BP420 5N1 4H4 4NS SS17 BP37S R77 1H2 ON9 4H6 BP379 BP410 ONS OHS ZNS Rn BP394 R66 5HO 1N7 2H4 BP373 R96 BP396 3HS SH4 1N3 BP375 R61 X19 3H5 4H5 BP376 RBO BP398 2H9 SHS 2N2 R105 BP400 R10S 3H7 ON7 1H1 BP386 BP416 ON6 1M6 ONZ SS1S BP388 R106 OH9 ZNB 3H1 BP392 R71 BP417 3NS 3H3 4N9 RS4 BP402 R91 6H4 5N4 OH3 VS038 VS037 ZLO SL9 After NAPS 2 CYCLE 11 CORE ONLOAD, REV. 0 L

M N

p R

T53 BP435 VS039 4N1 SL2 BP426 R63 VS009 6N4 OH1 SLS R9S BP429 BP436 SH3 4NO SNS T12 BP412 R94 BP423 R86 ONS 4HS SN2 OHS BP3S1 RS3 BP405 VS011 2H2 2ND 6HO 3N4 6L2 BP383 BP391 R99 VS002 ON1 2H1 4N3 6H2 3LS BP385 R79 BP406 VS006 1HS 1N2 6H1 3N9 4L3 BP414 R92 BP424 R9S 1N9 3H4 SN9 ZHS R90 BP430 BP434 4HO 6N3 3N3 T13 BP425 R70 VS013 6N1 1H3 SLS BP433 VS012 5N6 6L7 TSO

\\

PREPAREDBv.{f*k~ DAlE £-:J.6-'/S REVIEWED BY

,._ ~i: DATE t -z. w- '"f 5"

THE SAS SYSTEM e

CYCLE-N2Cll ----------------------* -------------------------------------

CORE ASSEMBLY TYPE BURNUP BANK LOCATION 1M8 NAIF 39003 D BANK F 6 2M4 NAIF 38980 D BANK KIO 2M5 NAIF 38561 D BANK FlO 3M9 NAIF 38455 B BANK D 6 lMl NAIF 38308 D BANK K 6 3M4 NAIF 38178 B BANK M 6 3Ml NAIF 38115 B BANK K 4 5M7 NAIF 38092 B BANK Fl2 4M6 NAIF 37832 B BANK Kl2 4M8 NAIF 37800 B BANK MlO 3MO NAIF 37718 C BANK F 8 5M8 NAIF 37606 B BANK D10 5H6 NAIF 37450 B BANK F 4 5HO NAIF 37430 C BANK HlO 4H5 NAIF 37134 C BANK K 8 3H7 NAIF 37101 C BANK H 6 6H3 NAIF 36099 SA BANK C 9 4H2 NAIF 36013 SA BANK G 3 6Hl NAIF 35716 SA BANK N 7 3H3 NAIF 35443 SA BANK J 3 3H6 NAIF 35396 SA BANK C 7 OMl NAIF 35365 C BANK Ml2 2H3 NAIF 35348 C BANK D12 3H2 NAIF 35254 SB BANK Ell 5H2 NAIF 35104 SA BANK Gl3 4H4 NAIF 35011 SA BANK Jl3 6HO NAIF 34988 SA BANK N 9 5H3 NAIF 34537 SB BANK Lll 1H4 NAIF 34509 C BANK D 4 4MO NAIF 34279 SB BANK L 5 4H7 NAIF 34205 SB BANK E 5 1H3 NAIF 34106 C BANK H 4 OM7 NAIF 33001 A BANK Fl4 2HO NAIF 32991 A BANK Kl4 5H5 NAIF 32954 SB BANK J 7 5Hl NAIF 32905 SB BANK G 9 1M7 NAIF 32863 A BANK BlO 5H4 NAIF 32855 SB BANK J 9 5H9 NAIF 32514 SB BANK G 7 2H7 NAIF 32388 A BANK B 6 2H8 NAIF 32360 A BANK p 6 OMS NAIF 32015 A BANK PIO OM3 NAIF 31987 A BANK K 2 SUANUP.S 2M6 NAIF 31733 A BANK F 2 ASSEMBLY 6M2 NAIF 31224 D BANK p 8 AObbl!b J.C)CAT10N.S 4M9 NAIF 31046 D BANK B 8 IN 4M3 NAIF 30673 D BANK Hl4 OF 2./Z'I / 9~

6M4 NAIF 30603 D BANK H 2 AS

8 9

10 11 12 13 14 15 H

45033 1133 FIGURE 3-13 FUEL BURNUP DISTRIBUTION MAP AT 20200 MWD/MTU NORTH ANNA UNIT 2, CYCLE 11 G

F E

D C

B

': 42714 48548

25733

48664

25872

38991 1075 647

1224

651

981 A

43687 *

1099 *

-!:*;':**-!:*':-/:-/~.. ':-_'*;':**--':*.'-*,n.... n n o ii n **'******"****'****;':.'. /. :,. ****m*.., n n n n

n * !.. /. J.. J. *':***

42819 43633

26111

48862

25702

45882 20888

48318 *

'i 1077 'i 1098..

657

1229

647

526

1216 -.':

-!:-.':*-!:-.':*;':;':7,':;':;':;':-!:-.':-;':;,':*':-.':***-f:**********-':***-;,'rlrlrn***... n **.:... ~**"f-****'****"':

'i, 48676 k 26194.. 49356

26536 24908

38425 *

-A

'i, 1225.':

1242 *':

668 627

967 *

': -.:o..;:;':"!:i:-::-::-::-.':-::-;:;':-1:;':*"Jo-.':;.. 0.::-.. **; * *;. *J.-'-'-J....t.J.J;* n n n ** *"' n n * ]., }., }.., "/:********* ** *

25792 -!: 48792

26535 ;': 45421

25823

21445

44064

649 -.':

1228 ;':

668 650

540

1109 *

-;,': 48744 "i,': 25737

48632 ': 25808..

.. 52057

1226 ;':

648

1224 649 1057..

1310 -.':

"'/:;':;':-;':*;':"'.':-/:"';':"",':-;,':;,':*-;':-;':'";':-.*:-.':*;':--,'i"'.':-/:-,':;':".':-.':-_'rirl:*irn'* **,.., ** n n, ** ******""**-/:

25931 24931

21449

52065

1:

652 627

540

1310 *

-!:.J:.. ':**-;':-'***--* ** ** **.: ** n -;,f..J.-';J."':**-f:.J:****;':*-!:****** /,. /. /.,

- - -.':******-k.':**

': 39071

20951

': 44018

MWD/T 527 1107

EFPD

-;': 43759

47556

1101

1196 *

-!:.. ':-!:*1:*7:"i':.. ':*.. ':-!:**.. ':-!:-.':

N2Cll Design Report - NE-1021, Rev. 0 Page 58 of 152 I

I I

I

.1 I

I I

I

ML18151A620

Text

8.0 CONCLUSION

Navigation menu

Search