ML20059F223

Rev 1 to Justification for Continued Operation (Jco) Jco 93-0001, Solid State Rod Control Sys

ML20059F223
Person / Time
Site:	South Texas
Issue date:	11/01/1993
From:	Myers L, Pacy M, Sheppard J HOUSTON LIGHTING & POWER CO.
To:
References
JCO-93-0001, JCO-93-0001-R01, JCO-93-1, JCO-93-1-R1, NUDOCS 9401130207
Download: ML20059F223 (8)

Text

-.. - -. -. -. = -- -. - -.

i

.ll '8 ) 'O.0001 R es.

i

,ll STIFl('ATION FOR CONTINUEl) Ol'EltATION (,1CO)

Al'I'ItOVA1 COVEllSilEET Inaiation Date:

Julv 9,1993 INeiration Dates: (Unit 1) 12/31/95 (Unit a 12/31/95 1

Revisien: __ Date 10/28/93 s onic a sol >lD STATE ItOD CONTItOL SYSTEM IC() Nu:

93 0001 Applicable Units:

I&2 i

sanmuu r This Justification For Continued Operation (JCO) is applicable for all operational inodes. This JCO addresses the issue of non compliance with the Licensing 11ases and allows for Rod Control system operation in i

accordance with approved plant procedures

~~7 T

D2 Picpared by.

I

-ll'1 /-Q' JJ G(\\

M Pac)

Date Mwapr. Deugri lM enng e

//

. + e'

/C!? \\ fr '(

Cout nrictl;c; j f

J Sheppard [

/[

Date

/

1 Ger.nal Manager. Nuclear 1.icensing Recoluniensied by PORC Mtp No'

'O

[O bb Date Approsed by:

v >, /u/ [ O N -f_.E td

/

' // /

N Dale '

/

L %. Myers Mant M.inap Unit I

/!I t\\pproved by

/

g g

G L Parkey I

Date o

na usarc, unir :

1100i4 9401130207 931101 6(

PDR ADOCK 05000498 i

P PDR w ou n o Page 1 ot 8 4mn

y v s nous w\\

1.

Idenlification Un Slay ?7,1993, operators at the Salem Nuclear Generating Station, Unit 2, everienced problems with the rod control system. While attempting to withdraw Shutdown Bank A, it was noticed that the rods were not moving based on the analog rod position indication (ARPI). The group demand indicator counter indicated that the rods should have stepped out approximately 20 steps. The operator attempted to insert the rods; however one rod (ISA3) withdrew to 8 steps as indicated by the ARPI. This f

occurred while the group demand indicator counted down frorn 20 steps to 6 steps.

Further attempts to insert the rods withdrew rod ISA3 to 15 steps causing a rod deviation alarm to annunciate. The power for rod ISA3's Control Rod Drive Niechanism (CRDN1) was removed resulting in rod insertion. The ARPI indicated 0 steps.

The U.S. Nuclear Regulatory Commission (NRC) issued a generic letter (GL 93-04) to notify utilities about a single failure vulnerability within the Westinghouse solid state rod control system that could cause an inadvertent withdrawal of control rods resulting in a power distribution not considered in the design basis analysis.

The Salem failure, which can result in i single rod withdrawal, is believed to have been an IC chip on one of the slave cycler decoder cards. The slave cycler decoder card decodes the current orders that tell the stationary, movable and lift coils when to energize. The chip that failed caused the corrupted current order that resembled the withdrawal command to be present continuously at each slave cycler decoder card.

When the IN-IlOLD-OUT switch was moved to the IN position, the OUT command was also present. The current for the lift coil is present during a full cycle creating a constant upward pull on the rod. If the stationary gripper drops out before the lift coil is de-energized, the rod may withdraw from the core. Since the failure maintains the lift coil energized during the cycle, the rods would either move out or not move at all.

STP's initial response to the Generic Letter involved the issuance of JCO 93-0001 Rev 0 and training of all licensed Unit 1 o;x rators on the Salem rod control event. The Salem rod control event was included in the Lessons Learned portion of licensed operator requalification training completed September 15, 1993.

11.

Operation The South Texas Project (STP) units use Westinghouse model L-106RL CRDh1's to position the Rod Control Cluster Assemblies (RCCAs) in the reactor core. The CRDh1 periouns two major functions. These functions consist of (1) positioning the rod clusters in the core thereby providing a means of controlling the core reactivity, and m e w aa a o rye.' or s

~

~

c

. _ _ - - ~ _ _ _ _ _ _ _

A P

Jm io.oom ei G ei.ist a, part of the Reactor Coolant Pressure lloundary. The RCCAs are used foi shuidov.n and conuol purposes to provide negative reactivity for required shutdown margin.

There are fifty seven CRDM's mounted on the Reactor Vessel Head, each consisting of a magnetic jack mechanism that moves a single RCCA in the reactor core The magnetic jack mechanism consists of four coils. Three of the electromagnetic coil assemblies lift, hold, or lower a 168 inch long drive rod, which is attached to a rod cluster assembly. Each of the three coils actuate an armature within the pressure housing. The movable and stationary gripper armatures operate latches that engage the grooved drive rod. The stationary gripper latches are used to hold the drive rod in position, and the movable gripper latches, which are raised and lowered by the lift armature. are used to raise and lower the drive rod. The fourth (holdout) coil provides the ability to lock the drive rod / rod control cluster assemblies in the fully withdrawn position.

The CRDM's are divided into four control and five shutdown banks. Each of the control banks and come of the shutdown banks are further subdivided into two groups.

The groups in these banks move sequentially so that the two groups are always within one step of each other. The CRDM's within each group are electrically connected to step simultaneously.

Rod movement is in discrete increments (steps). Each step of the mechanism raises or lowers the drive rod 5/8 of an inch and requires 780 ms for completion. Complete withdrawal or insertion of the rod cluster requires 268 steps of the L-106RL mechanism for complete withdrawal or insertion of the rod cluster in a 14-foot core.

The mechanism sequence for one OUT or IN step is as follows:

OLTI' SEQUENCE (1)

The rod cluster is held with the stationary gripper coil energized and the stationary latches closed.

(2)

The movable gripper coil is energized, closing the movable latches.

(3)

The stationary gripper coil is de-energized releasing the stationary latches.

H)

The lift armature is energized, raising tb movable gripper armature and the drive rod 5/S" j

\\

(5)

The stationary gripper coil is energized, closing the stationary latches.

i l

m)

The inovable cripper coil is de-energi/ed, releasirig the movable latches.

an.m:r,nunoao l'ap 3 ot x

m_..___

1

.lFO 9.L0001 Rev l 47)

'l lR-h!! coil is de-energi/cd, allowing the hit almalute to drop.

IN SEQUENCE (l}

'Ihe rod cluster is held with the stationary grippec coil energized and the stationa*y latches closed.

(2)

The lift coil is energized, raising the lift armature.

(3)

The movable gripper coil is energized, closing the movable latches.

l i4)

The stationary gripper coil is de-energized, releasing the stationary latches.

(5)

The lift armature is de-energized, lowering the movable gripper armature and the dove rod 5/8" dd The stationary gripper coil is energized, closing the stationary latches.

(7)

The monble l: ripper coil is de-energized, releasing the movable gripper latches, i

1 III.

Safety Function

'l The Control Rod Drive System (CRDS) performs its intended safety function, (ie., a reactor trip), by placing the reactor in a suberitical condition when a safety system setting is approached; single failure criteria is applicable. Despite the remote probability of a common mode failure impairing the ability of the reactor trip system to perform its safety function, Anticipated Transient Without Scram (ATWS) analyses have been performed in accordance with the requirements of WASH 1270. These analyses, documented in WCAP-7706-L and WCAP-8330, have demonstrated that acceptable safety criteria would not be exceeded even if the control rod drive system were rendered incapable of functioning (due to multiple SSPS and RTS failures) during a reactor transient for which their function would normally be expected. An ATWS is the most severe event associated with potential rod failures. The condition covered by this JCO does not involve failure mechanisms which result in an ATWS.

i hwipi a o P.gr 4 of 8 t.n o a s

s

-, + -,

.e er

. w

-~

<-n

+ +, - +

-r e

y o u nnoi un i IV.

Safely Atialysis A review was perfoimed to assess the impact of operating with the cunent single j

failure polemial in the rod control system. The Westinghouse Rod Control System livaluation Program (WCAP-13864) determined that of all single Rod Control System failmes identified, only the failures which occurred at the Salem plant can result in uncontrolled asymmetric rod motion following a demand for the rods to move. All other identified failures result in rod movement in the demanded direction and are hence limited to a finite number of steps. These failures may result in some asynunetric rod movement following a rod motion demand signal; however, the movement is " limited" by the rod demand signal. These events have been evaluated i

and determined to result in less severe consequences than the limiting single Rod Control System malfunction presented in the accident analysis chapter in the Updated Final Safety Analysis Report (UFSAR). Therefore, all the single Rod Control System failures, other than those identified to have occurred at Salem, are bounded by the cunent licensing basis safety analysis.

UFSAR Section 15.4.3.2.3 (Single Rod Withdrawal Method of Analysis) discusses the analysis of record in the licensing basis. The analysis uses NRC accepted computer codes (LOFFRAN and TIIINC). The analysis is based on the assumption that no single failure will resuh in a single rod withdrawal. By making this assumption, the General Design Criteria (GDC ?.5) criteria that no single malfunction result in the speemed fuel design limits (DNB in this case) being exceeded, is satisfied. Since a multiple failure is assumed, a small fraction of roos in DNB is acceptable. The results of the analysis show that less than 5% of the fuelLexperiences DNB.

+

As a result of the Salem event, the assumption that no ' single failure will result in a single rod withdrawal is no longer valid. To address this issue, Westinghouse perfonued an analysis using state-of-the-art methods and computer codes (LOFT 5 and SPNOVA). The analysis used STP specific values and considered the effects of a 14 foot core. The results of the. analysis are documented in WCAP-13803, Revision I and show that STP Units I and 2 will not experience DNB for a single rod withdrawal i

event like the Salem event with the current fuel design.- Therefore, the GDC 25 criteria i, satisfied and no significant safety hazard exits. Iloweve,' the analysis has not-been reviewed and approved for use by the NRC and cannot be incorporated into the licensing basis at this time.

i m.no u o P.y $ of 8 iun

.m o u nooi

m. i l

V.

Required Compensatory Actions

.I 1.

Perform Control Rod Operability test per Technical Specification surveillance requirement 4.1.3.1.2 to provide assurance that the failure scenarios of an uncontrolled asymmetric rod withdrawal will be detected.

i Operations OPSP03-RS-0001 Y

Per T.S. 4.1.3.1.2 p.n..He De,-i itn;Wcwrtmg th ument Complete (Y/N)

Lhedule Date

.12.

Install a CAlfflON TAG on the IN-IIOLD-OUT switch tr remind the operator that during any rod movement the actual direction of rod motion should be closely monitored.

Prior to plant Operations Caution Tag N

startup

-,mue nm nr ru enny n, omm c.mse:eovs sacaoie o re VI.

Corrective Action Based on the successful demonstration of the timing adjustments at an operating plant and receipt of the official technical bulletin from Westinghouse, STP will modify-the Rod Control System current order timing to prevent any uncontrolled asymmetric rod withdrawal in die event of the failure identified at Salem. If corrupted current orders are present, none of the rods will move (with a high degree of' certainty) once the cutTent order timing adjustments are made. This action will restore compliance with the Licensing liasis.

VII. Reporting This condition presently affects STP Units 1 and 2 rod control systems. The ieportability evaluation was not performed by STP since the NRC was notified by Salem. and a Part 21 was initiated by Westinghouse.

l s.w a co;w n %oilRI a p l'g e 6 of 5

~..,

9 l

w o i.w ooi un i Ylli.

lieferences I

.\\.

Technical Specincations 3.9.1 ("Horon Concentration")

13.

SPR 931987 (" Salem Rod Withdrawal Event")

C NRC Generie Letter (GL 93-04 " Rod Control System Failme And Withdrawal Ol' Rod Control Cluster Assemblies")

l D.

Westinghouse Letter (NSAL-93-007 " Rod control System Failures")

E.

Design Basis Document (9Z529ZB0100 " Additional Instrumentation and Control i

Systems")

l F

Updated Final Safety Analysis Report (sections 4.4, 4.3, 4.5, 15.4.3, 15.4.3.2.3)

G.

Salem Rod Control Event Lessons Learned training closure (ST-HS-IIS-26185) l 11.

Transmittal of 45 Day Response To Generic Letter 93 04 (ST-IIL-AE-4533) 1.

Transmittal of 90 Day Response To Generic Letter 93-04 (ST-HI AE-4589)

J.

Generie Assessment Of Asymmetric Rod Cluster Control Assembly Withdrawal i

(WCAP-13803); Transmittal WOG-93-154R1 K.

Rod Control System Evaluation Progra'm (WCAP-13864): Transmittal WOG l 172 a

i i

i h

i l

i i

i i

)

i i

i l

d i

i i

1 i

a j

.un

..u s; h o Pa;c 7 of h d

s r o o.onni nn i lCO Ol'ERATIONAI, ISIPACT STATEAIENT SOI lD STATE ROD CONTROL SYSTESI 1,un,co A.

To maintain the validity of tinis ICO, Operations must do the following:

l.

Perform Control Rod Operability test per Technical Specification surveillance requirement 4.1.3.1.2 to provide assurance that the failure scenarios of an uncontrolled asymrnetric rod withdrawal will be detected.

2 Maintain a CAUTION TAG on the IN-IIOLD-OUT switch to remind the operator that during any manual rod movement the actual direction of rod motion should be closely n1Dnitored.

I e

'J $' I N II k f ( k)

I4'(

kI I tM ll e,

w n