ML20056B497

Clarifies Info Submitted in 871207 & s Re Steam Generator Tube Rupture Analysis Demonstration Runs. Demonstration Runs Met plant-specific Requirements in Section D to NRC SER on WCAP-10698

ML20056B497
Person / Time
Site:	Catawba
Issue date:	08/20/1990
From:	Tucker H DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9008280363
Download: ML20056B497 (16)

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L Duke Ikurr Compa:y Hu B kher 70 thu33198 Uce President Charlotte. N C 28242 ' %uclearProduction l (104)373-4531 DUKEPOWER August 20, 1990 U. S. Nuclear Regulatory Commission (

A'ITN: Document Control Desk  !

Washington, D.~ C. 20555

Subject:

Catawba Nucicar Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Steam Generator Tube Rupture Gentlemen:

Your letter dated April 20, 1990 transmitted a request for additional information regarding operator response times for mitigating a Steam Generator Tube Rupture (SGTR) at Catawba Nuclear Station. Specifically, the NRC Staff requested operator response times for the recovery actions identified in Tabic 2.3-1 of the Westinghouse Owner's Group WCAP-10698, "SGTR Analysis Methodology tc Determine the Margin to Steam Generator Over fill . The Staff also requested assurance that the operator response times were representative of most, if not all, the operators currently at the plant. As. discussed with members of 1.s.e NRC Staff, Duke Power Company does not have operator response times for the recovery actions licted in Tabic 2.3-1 of WCAP-10698. Additionally, Duke Power did not evaluate each individual licensed operator when performing the simulator demonstration runs. In June 1988, when the demonstration runs were performed, Duke Power did not'believe that this information was necessary to comply with the intent of the NRC's Safety Evaluation Report (SER).

The purpose of this letter is to clarify the information submitted by Duke Power to the NRC on December 7, 1987 and August 8, 1988 regardinB Steam Generator Tube Rupture analysis demonstration runs.

Background

Lit.ense Conditions 16 and 10 of Fac'lity Operating Licenses NPF-35 and 52, respectively, required that Duke Pousr Company submit for NRC staff review and approval a SGTR analysis for the Catawba Nuclear Station. This analysis was to demonstrate that the LGTR analysis presented in the FSAR r was the most severe case with respect to the release of fission products and calculated doses. This issue was b31ng pursued generically by the Westinghouse Owner's Group SGTR Subgroul, of which Duke Power was an active participant. On March 30, 1987, the NRC issued a Safety Evrluation l Report accepting the Subgroup's analysis methodology documented in l

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U. S. Nucir r R;gulctory Commiccion

. August 20, 1990 Page 2 WCAP-10698, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill." The NRC Staff required plant specific input, however, from each utility referencing the WCAP-10698 analysis. My letter dated December 7, 1987 addressed the plant specific submittal. requirements for Catawba Nuclear Station as described in Section D of Enclosure 1 of the SER.

One of the plant specific items required by the SER stated that Each utility in the SGTR subgroup must confirm that they have in place simulators and training programs which provide the required assurance that the necess.ory actions and times can be taken consistent with those assumed for the WCAp-10698 design basis analysis. Demonstration runs should be w -med to show that the '

accident can be mitigated within a periot 6 ,smo compatible with overfill prevention, using design basis ass q mions regarding available equipment, and to demonstrate that the operator action times assumed in the analysis are realistic.

Duke Power has confirmed to the NRC that Catawba has in place a simulator and a training program that provides assurance that the operators are trained to perform the necessary actions to mitigate a SGTR accident. It is Duke's understanding that the demonstration runs required by the SER (as' stated above) were to show that the accident could be mitigated while preventing steam generator overfill, and to demonstrate that the operator response times were realistic.

Basis Behind SGTR Demonstration Huns WCAP-10698 Table 2.3-1, lists the following operator actions:

1) Identify and isolate the ruptured steam generator within the latter of: 10 minutes after reactor trip or prior to ruptured steam generator narrow range level exceeding 33% increasing

2) Initiate plant cooldown within five minutes of isolating the ruptured steam generator

3) Cool the RCS

4) Commence RCS depressurization within two minutes of completion of cooldown

5) Depressurize the RCS to ruptured SG pressure to stop break flow ,

6) Initiate SI termination within 1 minute of completion of depressurization

7) Terminate SI

U. S. Nuclear Regulatory Commission

. August 20,'1990 Page 31 i

.i Actions 1, 2, tnd 4 were included as Items 2, 3, and 4 in my August 8, 1988 letter.

Actions 3 and 5 are physical plant processes. As discussed in the s footnote to Table 2.3-1 of WCAP-10698, they therefore depend on plant design and parameters and on the equipment used to perform the operations.

The times for the three loop reference plant would therefore not be directly applicable to a four loop plant like Catawba. Further, the Catawba emergency procedures use a simultaneous cooldown and depressurization, rather than the saquential method assumed in Table 2.3-1. Therefore these actions were not included in the demonstration runs.

Actions 6 and 7 deal with termination of safety injection. While this is an important event with respect to stabilizing the plant in the context of1 recovery from a tube rupture, as long as pressure equilibrium is maintained across the break, as a result of Action 5, no overfill consequences are associated with SI termination timing. Therefore these actions were not included in the demonstration runs.

The preceding paragraphs explain the reasons for selecting Items 2, 3, and 4 in my August 8, 1988 letter. Item 1, while nat relevant for overfill margin, is an assumption of the " site specific SGTR radiation offsite consequence analysis" required in the WCAP-10698 SER, Enclosure 1, Part D.

Item 2. Since this assumption has not previously been a part of the Catauba SGTR licensing basis, it was felt prudent to inc_ude it in the demonstration runs. Similarly, Item 5 is an assumption of the dose analysis which was included in the demonstration runs. Item 5 also was discuss d in the generic overfill evaluation. The five minute operator action time requirement for isolation is based on WCAP-10698, p. 4-65.

Finally, Items 2,.3, and 4 were also assumed in the dose analysis.

In summary, Actions 1, 2, and 4 of WCAP-10698, Table 2.3-1 were included as Items 2, 3, and 4 of my August 8, 1988 letter, and were also

-assumptions of the dose analysis, as shown in Table 15.6.3-1 of my August i 24, 1988 letter. Actions 3, 5, 6, and 7 of Table 2.3-1 are not appropriate, as explained above, to the demonstration runs. Items 1 and 5 i

of the August 8, 1988 letter were included to verify assumptions in the site specific dose analysic. Item 5 also confirms an assumption in the generic single failure evaluation for overfill.

Description of Demonstration Runs The following paragraphs provide additional information regarding the demonstration runs conducted by Catawba Nuclear Station and corrects an error in my August 8, 1988 letter regarding the number of operators involved.

U.-S. Nuclist R:gulctory Commie: ion August- 20,-1990 Page 4 The demonstration runs were conducted on June 9, June 13, and June 23, 1988. The June 9th demonstration involved five licensed operators. The June 13th demonstration run involved four licensed operators.. The June 23rd demonstration run involved five licensed operators. A total of 14 ,

oporators participated in the simulator runs. My August 8, 1988 letter incorrectly stated that only three operators per group participated in the demonstration runs.

The 14 operators which' participated represent 20 per cent of all operators at the plant in June 1988. The demonstration runs were administered as ,

part of the operator rnqualification training program. The simulator teams consisted of operators who normally worked together when on duty in the control room. Two separate crews were involved, however, so that all of the participants were not from the same operating shift. There were no efforts made to administer the demonstration runs to operators with any particular level of experience. Additionally, the scheduling of the simulator runs were not made with any underlying motive. The three dates chosen were-merely the first convenient opportunities to perform the

. demonstration runs. The dates depended not only on simulator availability and current training schedules, but on the development of the specific simulator scenario and demonstration run documentation.

Attachment I is a copy of the simulator exercise guide used to conduct the demonstration runs in June 1988. On page 9 of Attachment I are the five operator actions that were evaluated relative to the SGTR overfill ,

analysis as discussed in my August 8, 1988 letter.

Conclusions The demonstration runs performed by Duke Power met the plant specific requirements in Section D of Enclosure 1 of the NRC's Safety Evaluation Report on WCAP-10698. Duke Power views the purpose of the demonstration runs as 1) providing assurance that a sample of the operators at.a given site were tested, and 2) showing that the procedures used by the operators and the training received by the operators enable them to respond as

'quickly as the generic operator response times assumed in WCAP-10698. The two actions above give. assurance that the SGTR accident can be mitigated within a period of time compatible with overfill prevention and that the operator action times assumed-in the analysis are realistic.

In response to the NRC's specific request that Duke provide the operator response times demonstrating that the SGTR event can be' mitigated within a period of time compatible with overfill prevention, response times are-not available. .Only the fact that the critical operator actions were performed within the bounding times as specified by the WCAP analysis was documented. In response to the request that Duke provide assurance that the response times are representative of most, if not all, the operators

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L U. S. Nuclocr R:gulctory Commiccion August 20, 1990 Page 5 currently at the plant, this assurance has been demonstrated by three operator teams, selected without regard to expected team response to a tube rupture, successfully completing the demonstration runs.

To help facilitate closure of this long standing issue, Duke Power will commit to perform additional demonstration runs to show that tha applicable WCAP response times are representative of most operators ,

currently at the plant. These demonstration runs will be performed during the third segment of licensed operator reenalification training in 1991,  ;

currently scheduled to be completed by Au - 30, 1991. The simulator scenario and operator response times eval .d will be the same as those used during the June 1988 demonstration runs. Where appropriate, the actual duration of the response times will be recorded. To assure the times are representative of most of the plant operators, a minimum of 80%

of the licensed operators will be evaluated. This percentage includes the i fourteen operators evaluated in 1988.

Very truly yours, N. h rle mr

11. B. Tucker RGM/06279001 Attachment cc Mr. S. D.~ Ebneter Regional Administrator, Region II U. S. Nuclear Regulatory Commission 101 Marietta St., NW, Suite 2900 Atlanta Georgia 30323 Dr. K. N. Jabbour U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 9114 Washington, D. C. 20555 l Mr. W. T. Orders l Senior Resident Inspector Catawba Nuclear Station l.

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ATTACHMENT I

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SIMULATOR EXERCISE GUIDE

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PROGRAM: . Operations.Trdining MODULE:- License Preparatory Training TOPIC: ' Transients and' Accidents .

EXERCISE: Steam Generator Tube Rupture - CNS FSAR Study (Unannounced Casualty) i OVERVIEW: a s

This exercise will demonstrate the operator's ability to recognize a Steam

. Generator Tube Rupture based on the analysis made in the Catawba Nuclear Station FSAR..

PREREQUISITE KNOWLEDGE LEVEL:

1 As a minimum, the trainee should be in the RO License Preparatory Module, and have completed the.EP lessons 1.through 4. .

REFERENCES:

1. Catawba Nuclear Station FSAR Section 15.6.3

2. Engineering Calculation CNC-1552.08-00-dO36

3. EP/1/A/5000/01 Reactor Trip or Safety Injection

4. 'EP/1/A/5000/1E Steam Generator Tube Rupture

5. EP/1/A/5000/1El Post SGTR Cooldown and Depressurization

6. AP/1/A/5500/10 Reactor Coolant Leak

7. RP/0/A/5000/01 Classification of Emergency

8. RP/0/B/5000/13 -NRC Notification Requirements AIDS:

1. EP/1/A/5000/01 Reactor Trip or. Safety Injection

2. EP/1/A/5000/1E Steam Generator Tube Rupture

3. EP/1/A/5000/1El Post SGTR Cooldown and Depressurization

4. AP/1/A/5500/10 Reactor Coolant Leak

5. RP/0/A/5000/01 Classification of Emergency

6. .RP/0/B/5000/13 NRC Notification Requirements TRAINING USE ONLY s.

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. LPRO OBJECTIVES-The student'will be able to: ,

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1. Correctly diagnose the symptoms of a SGTR.

- 2. Determine when to enter procedure EP/1/A/5000/1E, Steam Generator Tube Rupture, from' procedure EP/1/A/5000/01,_ Reactor Trip or-Safety Injection.

3. Determine when to enter procedure EP/1/A/5000/1E1, SGTR Cooldown and Depressurization, from EP/1/A/5000/1E.

4. Respond to the procedure reader during coverage of the procedure immediate and subsequent actions.

5. Perform all tasks with.a minimum of supervision.

6. . Correctly interpret all information obtained from control room indications and alarms, and take the proper action based on these indications.

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LPSO OBJECTIVES

.The student will be able*to:

1. ~ Ensure procedural steps are performed correctly.

2. Determine'the class of emergency per Rp/0/A/5000/01 and make proper notifications per RP/0/B/5000/13/

- 3. Ensure critical safety functions are maintained.  !

4. Correctly diagnose the symptoms of a SGTR.

5. -Initiate corrective actions per approved procedures to safely shutdown the plant and protect the health and safety of plant personnel and the general public.

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. Page 4 of 10 1.0' INITIAL' CONDITIONS 1.1 Unit 2 Steam Generator Model at-100% FP (EOL) i 1.2 "D" S/G PORV is inoperable and will not open. This is a preexisting condition that should be made known to the-students. .,

'1. 3 "A" S/G PORV is-failed such that it will not reseat itself once it-operates.(Student will not know this)..

1.4 When the Generator is Tripped off following a Reactor trip, a malfunction will cause the entire switchyard to be lost for the

' duration of the transient. (Student will not know this).

2.0 SIMULATOR BRIEFING,

, 2.1 Description of Exercise - The purpose of this exercise is to ensure that the operators can respond to the SGTR transient analyzed in.our FSAR. In addition to responding to the SGTR with approved Emergency.

Procedures, the operators will be expected to perform certain actions, as the conditions arise, within the time frame assumed in the FSAR analysis. This exercise should continue until break flow ceases due NC system depressurization.

2.2 Plant Status - Unit 2 S/G model operating at 100% FP (EOL), NOT/NOP All systems are stable except as noted in.section 1.0 above.

3.0 EXERCISE PRESENTATION

. 3 .1 Insert the.following malfunctions for the duration of this exercise:

A. EPX2 (Loss of Switchyard); Select Choice 3 (for both buses)

Conditional on: PSM:5210 .LE. 100 (Turbine Impulse Pressure drops off due trip on Reactor Trip)

B. SGX1A (S/G "A" Tube Leak); Select Leak Rate = 566 GPM, Select Ramp Time = 0 seconds. Set Delay for 5 minutes.

1 C. SMX2A ("A" S/G PORV); Select position 100% open Conditional on:

PSM:5520.ge.1125.("A" S/G PORV Pressure)

D. SMX2D ("D" S/G PORV); Select position 0% open 3.2 Insert the following malfunction for your own use to delay the expected indications on the EMF racks caused by a SGTR A. EMF 2 Select 26 and fail low B. EMF 1 Select 34L; Set to 10% of value For these two (2) EMF malfunctions, an appropriate delay in returning l them to operation, will be useful in making this exercise more I representative of actual plant conditions with a full rupture of one (1) tube sized break. Remember, these EMF's are part of the SGTR I determination indications.

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Page 5 of 10 3.3 Select any 100% FP (EOL) IC Set and File RUNU2 3.4 As the SGTR malfunction begins, the instructor will have to judge when to' clear the EMF malfunctions and allow the operators these.

indications.

3.5 When asked about the Lost Switchyard, inform operators that the Transmission Department will be unable to return those buses to service for several hours.

3.6 Allow transient to continue into the Cooldown and Depressurization i phase of.the recovery.

3.7 Supply sufficient Simulator instructors to ensure the ' specific actions the operators need to perform are adequately documented.-

3.8 Terminate the exercise when the break flow has ceased due to NCS depressurization.

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' Instructor Date Catawba FSAR Steam Generator Tube Rupture 4.0 Evaluation A. Evaluation Criteria

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, 1. Satisfactory (S) Operator performance in this area is fully acceptable, e.g., takes appropriate actions in a reasonable time; understanding of' actions being performed, etc.

2. Marginal (M) Operator performance in this area is acceptable, but some weaknesses are evident, s.g. improper action, but recovers before losing control; delayed response, etc.

3. Unsatisfactory (U) Operator performance in this area is not acceptable, e.g., improper actions, no recovery, loss of control, or no action when-

'needed.

4. Not Observed (N/0)

Note: 1. SRO

2. RO

3. BOP

4. SS

5. SE B. Evaluations of Objectives SRO RO BOP SS SE A. Maintain control board awareness B. Maintain logs per OMP 2-17 C. Proper Use of Procedures per OMP 1-4 D. Teamwork and effective communications E. 'Make notification within required times frames (SRO/SE only)

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. Page 7lof 10 SRO R0 _- BOP SS SE

6. Performs the' command function of the SRO by efficiently directing and coordinating the actions of other.

trainees (SRO only) >

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7. Monitors critical safety functions and makes recommendations to SRO/0ATC (SE only) 3

8. Evaluates plant conditions from all' available indications and diagnoses' event

9. Manipulates'the control board (RO/ BOP only)

10. EP/1/A/5000/01 Reactor Trip or Safety Injection

a. Recognizes Symptoms / Diagnoses Event Correctly.

b. Directs / Performs Immediate Actions [

from Memory

c. Directs / Performs SWosequent Actions

d. Monitors Enclosures

11. EP/1/A/5000/1E Steam Generator Tube Rupture

a. Recognizes Symptoms / Diagnoses Event Correctly

'b. Directs / Performs Immediate Actions from Memory-

c. Directs / Performs Subsequent Actions

d. Monitors Enclosures i

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12. ' EP/1/A/5000/1El- Post SGTR Cooldown= and Depressurization

a. Recognizes Symptoms / Diagnoses Event Correctly

b. Directs / Performs Immediate Actions from Memory

c. Directs /Perforn., Subsequent Actjons

d. Monitors Enclosures-

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13. AP/1/A/5500/10 Reactor Coolant' Leak j i

a. Recognizes Symptoms / Diagnoses' Event '

Correctly

b. Directs / Performs Immediate Actions ,

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c. Monitors Enclosures j i

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14. RP/0/A/5000/01 Classification of Emergency ,

a. Recognizes Symptoms / Diagnoses ~ Event Correctly

b. Directs / Performs Immediate Actions from Memory l

c. Directs / Performs Subsequent Actions

d. Monitors Enclosures

15. RP/0/B/5000/13 NRC Notification Requirements a.. Recognizes Symptoms / Diagnoses Event ,

Correctly

b. Directs / Performs Immediate Actions from Memory

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16. performance of! actions within stated times

a. Manu'al SI actuated within 30 min.

of break initiation

b. Identify and isolate Ruptured S/G within the later of; 10 min after  !

jRx Trip or Ruptured NR S/G Level

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reaching 33% ince.

c. Initiate plant cooldown within 5 min. of isolating Ruptured S/G

d. Depressurize plant 2 min, after completion of cooldown- 1

. e. Isolate Failed S/G pORV within I 5 min.lof the time it should ressat per its setpoint j i

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Page 10 of 10-C., Instructor Comnents (Required for M,' U) l l

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l 5.0 ' CRITIQUE

. Note:- Critique trainee performance at the;end of each scenario.

• A. Review exercise (preferably done by trainees).

iB. Critique performance (Including Positive and Negative Feedback) o C. Allow trainees to ask questions /make comments.

6.0 DOCUMENTATION A. Complete simulator hours and identified weakness documentation for requal records.

B. Complete annual requi:%.nents record for each student for the following:

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