ML20204H016

Resubmits Util 990302 Response to Violations Noted in Insp Repts 50-327/98-11 & 50-328/98-11.Corrective Actions:Lessons Learned from Event Have Been Provided to Operating Crews

ML20204H016
Person / Time
Site:	Sequoyah
Issue date:	03/19/1999
From:	Bajestani M TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-REGGD-01.033, RTR-REGGD-1.033 50-327-98-11, 50-328-98-11, NUDOCS 9903260364
Download: ML20204H016 (8)

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. Tennessee Volley Authority. Post Offce Box 2000, Soddy Daisy, Tennessee 37370-2000 Masoud Bajestani Site Vce President

%qw anNucbuPWM j March 19, 1999 U. S. Nuclear Regulatory Commission 10 CFR 2.201 ATTN: Document Control Desk Washington, D.C. 20555 1

Gentlemen:

In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 SEQUOYAH NL EAR PLANT (SQN) - NRC INSPECTION REPORT 50-327, 50 i3/98 REPLY TO NOTICE OF VIOLATION (NOV)

The purpose of this letter is to resubmit our March 2, 1999, reply to the NOV as discussed in a conference call with NRC on March 10, 1999. The NOV contains two violations as documented in the subject inspection report dated February 1, 1999. One violation addresses the failure to properly implement an emergency procedure following a reactor trip on ' ivember 9, 1998.

The other violation was for a failure to enter Technical Specification 3.0.3 when the limiting condition for reactor coolant system flow instruments was not met.

The enclosure contains TVA's clarified response to the NOV in its entirety. Revisions associated with the clarification have been annotated with a revision bar in the right hand margin. This submittal does not contain additional commitments.

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PDR G ADOCK 05000327)f PDR Jl $ llf ee-L_

U.S. Nuclear Regulatory Commission

'Page*2 March 19, 1999 If you have any questions regarding this response, please contact me at extension (423) 843-7001 or Pedro Salas at extension (423) 843-7170.

Sincerely,

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/Wbkhs Os Bafestani Enclosure cc (Enclosure) 1 I

Mr. R. W. Hernan, Project Manager Nuclear Regulatory Commission One White Flint, North i 11555 Rockville Pike Rockville, Maryland 20852-2739 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission Region II Atlanta Federal Center ,

61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 4

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ENCLOSURE j I

TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 INSPECTION REPORT NUMBER 50-327, 50-328/98-11 REPLY TO NOTICE OF VIOLATTON (NOV) l l

I. RESTATEMENT OF VIOLATION A (50-327/98-11-02) J

, " Technical Specification 6.8.1.a requires, in part, that procedures shall be established, implemented, and maintained-covering activities recommended in Appendix A of Regulatory Guide-1.33, Revision 2, February 1978, ' Quality Assurance Program Requirements (Operations).' Regulatory Guide 1.33 Appendix A, Paragraph 6: recommends procedures for combating emergencies and other significant events. l 1

Emergency Subprocedure (EOP) ES-0.1, Reactor Trip Response, Revision 22, Step 3.c., contains the continuous action requirement 'IF temperature greater than 552 F and rising, THEN... DUMP steam USING atmospheric reliefs.' Step 3 of the subprocedure foldout page, which presents actions which apply  ;

at all times during procedure performance, also contains the requirement to 'MONTTOR reactor coolant system (RCS) temperatures stable at or trending to between 547 F and 552 F.'

EOP ES-0.1, Reactor Trip Response, Revision 22, step 8.b.,

contains the continuous action requirement 'IF pressure greater than 2235 psig and rising, THEN... CONTROL pressure USING one pressurizer power operated relief valve (PORV).'

Step 8.b.2 of ES-0.1-foldout page also contains the ~

requirement '(if pressurizer pressure greater than 2235 psig and rising)- CONTROL pressure.'

Contrary to the above, following a Unit 1 reactor trip on November 9, 1998, the procedural requirements of EOP ES-0.1 were not properly implemented, in that:

1. RCS temperatures were not monitored, stable and trending to between 547 F and 552 F and subsequently steam was not dumped using the atmospheric reliefs when RCS temperature was greater than 552 F and rising. RCS temperatures exceeded 552 F for a period of approximately 23 minutes.

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2. RCS pressure was not controlled using one pressurizer l PORV when RCS pressure was greater than 2235 psig and rising. RCS pressure exceeded 2235 psig for approximately 25 minutes.

This is a Severity Level IV Violation (Supplement 1)."

TVA'S REPLY TO THE VIOLATION

1. Reason For Violation A (50-327/98-11-02)

Based on an evaluation of the condition, the root cause for the violation was the crew'.s failure to properly monitor plant parameters (i.e., reactor coolant system

[RCS) temperature and pressure).

Relative to RCS temperature control, contributing causes were a combination of understanding procedural flow path and hierarchy following a reactor trip due to equipment / component failures that effect post trip  !

response; the role, priorities, and focus of the unit f supervisor and shift technical advisor; desirability of having controlling bands in Procedure ES-0.1; " Reactor Trip Response," and lack of training crews on reactor trips concurrent with the loss of inverter (abnormal operating procedure (AOP) type events). As a result of these causes, the operating crew did not take manual control of the steam generator (SG) atmospheric relief valves and maintain reactor coolant (RCS) temperature.

Relative to RCS pressure control, the crew failed to properly monitor RCS pressure as implied by the

" Response Not Obtained" column of Procedure ES-0.1, Step 8.b and the crew should have discussed the method  ;

of pressure control. Their actions were influenced by {

past operating and industry practices, and were i consistent with previous operator training. As a i result, the crew chose to control pressure using the i pressurizer power-operated relief valves (PORV) in j automatic, i i

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2. Corrective Steps Taken And Results Achieved Lessons learned from the event have been provided to the operating crews. Each crew is familiar with the event and the deficiencies that have been identified.

The operating crew has been counseled on their individual and collective performance deficiencies.

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Their understanding and correction of these deficiencies have been demonstrated through successful simulator performance. Additionally, the otner licensed operators have had simulator tiaining on the reactor trip with loss of the 1-IV vital inverter transient. Based on this training, licensed operators fully understand the importance of closely monitoring and controlling the RCS temperature during these types of transients.

Procedure ES-0.1 has been revised to ';1early designate Step 8.b as a continuous action step (" Monitor"). Also, the " Response Not Obtained" column t.) Step 8.b has been clarified to recognize that control 1mng RCS pressure by use of a pressurizer PORV in automatic is acceptable provided that the PORV is performin properly. This was done by use of the action verb "ENSLM ," requiring the operator to observe that an expected ce dition exists and, if necessary, take action to make the condition occur. An excerpt of the revision is illu trated below:

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lBTEP l ACTION /EXPECTE D . l RE5PONSE NOT l

3. M ON?TO R pressuruer pressure j con trol: - 1

s. Pressuriser pressure . a. ENSU R E SI ACTU ATED. 1 greater than 157 pal 8 G O T O E-0, Reactor Trip or Safety Inieetion b Pressurher pressere b, IF pressure less than 223$ pais and dropping, stable at or trending to 2215 psis - T II E N (beiween 2210 psig and 2260 psig)- PER F0 MM the following:

1) EN#1J R E pressuriser FORV s CLOSE D f IF pressuruar PO RV taa NO T be closed.

THEN CLOSE sts block valve.

2) E N S U R E spray valt es C LO S E D.

?T spray valve (s) can N O T be slesed.

THEN STO P RC P(s) supplying failed spray valvef t)

2) E N EU R E ell pressurizer healers ON.

IF pressure greater than 223$ psig een rising, T 11 E N P E R FO R M the following:

1) E N S U R E all pressurizer besters O F F.

2) CO N T R O L pressure 0 5 3N G norm al spray, IF n,ormal spray NO T available A N D leiden a en sati sc e.

T il E N CO N TRO L pressure U5tMC E A42 4, Estabisshtag A vailiary S pes) -

le aormal sprey N OT swailable A N D letdown N O T in service, l THEN P E NSU RE pressure sostrelled with pressuruer PORY 7

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l i As an interim. action to the evaluation of operator  ;

performance for the November 9, 1998 Unit 1 trip, an

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i independent review was performed. The review team consisted of a Westinghouse Nuclear Services representative, the Comanche Peak Operations Support I l Manager, the McGuire Operations Training Manager, the l Watts Bar Operations Training Mnnager, a senior operator from the Sequoyah Operations Training Department, and a Sequoyah Operations unit supervisor.

i The team reviewed the reactor trip report, observed the ,

performance of titree operating crews responding to a j reactor trip with a loss of Vital Instrument Power '

Board 1-IV on the simulator, and reviewed plant j procedures used to respond to this type of event. 1

( Enhancements tc the emergency operating procedures  !

(EOPs) were identified, which will eliminate the i confusion added by having a parenthetical operating band l in the " Action / Expected Response" column.

l l Recommendations made by the independent review have been l incorporated into the Corrective Action Program as

" Actions" and are being tracked to completion.

l Following the independent review, we ran simulator j sessions with each crew for a loss of the Vital

! Instrument Power Board 1-IV with a reactor trip scenario. We ran them for two reasons. First, they sxposed the crews to the lessons learned-(monitoring and controlling RCS temperature) and reinforced expectations j regarding command and control, communications, and proper procedure usage. Second, the scenarios would >

provide the information necessary to determine the best combination of plant parameter control and procedural l pace that would provide the most effective mitigative l strategy (for this scenario). One crew was specifica1.!y directed to take manual control of the PORV for RCS-pressure control. Observation of that crew determined that precise pressure control was not achievable by taking manual control of the PORV- and mar,ual control could induce a safety injection actuation.

Following each scenario, crew specific critiques ware held:to discuss command and control crew perform'nce.

These critiques did not discuss crew differences to preserve the independence of each crew. The training

analysis data from each crew is being combined with

benchmarking data from training observations recently made at three Westinghouse sites to further enhance EOPs i and AOPs. Additional scenarios are being conducted with El-4 l

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other AOPs to identify if similar conflicts exist between AOPs and EOPs. This will result in further procedure enhancements. These procedural enhancements, along with expectations concerning the role of crew members, procedural hierarchy, and monitoring plant parameters, will be provided to operators in future ,

training cycles.  !

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3. Corrective Steps That [Have Been Or] Will Be Taken To Prevent Recurrence No additional actions are necessary ralative to the cause'of the violation.

4. Date When Full Compliance Will Be Achieved j i

With respect to the violation, TVA is in full compliance.  !

II. RESTATEMENT OF VIOIATION B (50-327/98-11-01)  !

" Technical Specification 3.3.1.1 requires that the reactor trip system instrumentation channels and interlocks of i Table 3.3.-1 shall be operable. Table 3.3-1, Functional Unit 12, Loss of Flow-Single Loop, requires a minimum of 2 RCS flow channels per loop to be operable in Mode 1.

Technical Specification 3.0.3 requires, in part, that when a Limiting Condition for Operation is not met, except as provided in the associated action requirements, within one hour action shall be initiated to place the unit in a MODE in which the Specification does not apply.

Contrary to the above, Technical Specification 3.0.3 was not met, in that on November 20, 1998, all three channels of the Unit 1 RCS Loop 1 flow instruments (1-F1-68-6A, 6B & 6D) were inoperable, indicating off-scale high, and TS 3.0.3 was not entered to initiate action to place the unit in a MODE in which the Specification does not apply (Mode 2 or less).

This is a Severity Level IV " .on (Supplement 1)."

a TVA'S REPLY TO THE VIOLATION

1. Reason For Violation B (50-327/98-11-01)

The reason for the violation was our interpretation of Technical Specification (TS) 4.0.3 applicability. The individuals involved incorrectly believed that entry into El-5

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i TS 4.0.3 was acceptable based on an evaluation that the computer data for Unit 1, Loop 1, RCS flow was valid and that the surveillance could be performed successfully if the procedure was revised to allow use of the computer data. As a result, we failed to understand that TS 4.0.3 I does not apply to a surveillance, which could not be I performed. l 1

2. Corrective Steps Taken And Results Achieved '

l The Operations Superintendent issued an Operations Department Standing Order to the shift crews. The i standing order reinforces the correct use of TSs 4.0.3 and 3.0.3. The standing order states that TS 4.0.3 entry for the RCS flow indication condition was inappropriate.

The operating crews have reviewed the standing order and understand correct application of TS 4.0.3.

The SQN site management team discussed TSs 4.0.3 and 3.0.3 entries and applicability. The management team and the Operations department now understands that the TS 4.0.3 entry was an inappropriate decision and that a TS 3.0.3 entry was required.

3. Corrective Steps That [Have Been Or] Will Be Taken To Prevent Recurrence No additional actions are needed to prevent recurrence.

4. Date When Full Compliance Will Be Achieved With respect to the violation, TVA is in full compliance.

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