ML20046C935

Responds to NRC 930621 GL-93-04, Rod Control Sys Failure & Withdrawal of Rod Control Cluster Assemblies, 10CFR50.54(f). Ltr Requires Licensee to Provide within 45 Days Assessment of Sys Response to Control Rod Failure

ML20046C935
Person / Time
Site:	Beaver Valley
Issue date:	08/06/1993
From:	George Thomas DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-93-04, GL-93-4, NUDOCS 9308130057
Download: ML20046C935 (11)

Text

Beaver Va ey Power Station Shippingport, PA 15077 6 4 (412) 393-5206 (412) f.43-8069 FAX GEORGE S.1HOMA$

Dmsson Vee Pmsident Nuclear Servces Nuclear Poww Division U.

S. Nuclear Regulatory Commission Attn:

Document Control Desk Washington, DC 20555

Subject:

Beaver Valley Power Station, Unit No. I and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Response to Generic Letter 93-04:

Rod Control System 1

Failure and Withdrawal of Rod Control Cluster Assemblics, 10 CFR 50.54(f)

Pursuant to the requirements of 10 CFR 50.54(f), the NRC issued Generic Letter 93-04,

" Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies,"

on

Monday, June 21, 1993 to all licensees with the Westinghouse Rod Control System (except Haddam Neck) for action and to all other licensees for information.

The generic letter requires that, within 45 days from the date of the generic

letter, each addressee provide an assessment of whether or not the licensing basis for each facility is still satisfied with regard to the requirements for system response to a single failure in the Rod Control System (GDC 25 or equivalent).

If the assessment

[ Generic Letter Required Response 1.(a)] indicates that the licensing basis is not satisfied, then the licensee must describe compensatory short-term actions consistent with the guidelines contained in the

' generic

letter, and within 90 days, provide a plan and schedule for long-term resolution

[ Generic Letter Required

Response

1. (b) ].

Subsequent correspondence between the Westinghouse owners Group and the NRC resulted in schedular relief for Required Response.l.(a)-(NRC-Letter-to Mr. Roger Newton dated July 26, 1993).

This portion of the l

required actions will now be included with the 90-day licensee-response.

Duquesne Light Company (DLC) hereby submits in the Attachment to this letter-its-response to -the-Generic Letter as it applies to Beaver Valley. Power Station (DVPS)

Unit No.

1 and No.'2.

This response summarizes the compensatory actions taken by DLC in response to the Salem rod control system failure event.

It also provides a summary of the results of the generic safety analysis program conducted by the Westinghouse owners Group and its applicability to BVPS Unit No.

1 and No.

2.

DLC considers this action to be complete.with respect to the 45 day required response to GL 93-04 (as amended by July 26 NRC letter to Mr. Roger Newton).

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.Foiluro and Withdrawal.of Rod Control Cluster.

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-Assemblies, 10 CFR 50.54(f)

Page.2

. j If you 'have any questions.concerning this response, please contact

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Drexel Williams at (412):393-5226.

Sincerely,.

Geor }e S.b /?Y n 3

~

g Thomas.

Attachment cc:

Mr. L. W. Rossbach, Sr. Resident Inspector Mr. T.

T. Martin, NRC Region I Administrator Mr.

G.

E.

Edison, Project Manager h

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AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA)

) SS:

COUNTY OF BEAVER

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1

Subject:

Beaver Valley Power Station, Unit No. 1 and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Response to Generic Letter 93-04:

Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblics, 10 CFR 50.54(f)

Before me, the undersigned notary public, in and for the County and Commonwealth aforesaid, this day personally appeared George S.

Thomas, to me known, who being duly sworn according to law, deposes and says that he is Division Vice President, Nuclear Services of the Nuclear Power Division, Duquesne Light Company, he is duly authorized to execute and file the foregoing submittal on behalf of said

Company, and the statements set forth in the submittal are true and correct to the best of his knowledge, information and belief.

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MR Geor6e S. Thomas Subscribed and sworn to before me inid, /f(h on this day of

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Notary Public NotarialSeal TraceyA BaczeK NotayPubic i

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DUQUESNE LIGHT COMPANY Nuclear Power Division ATTACHMENT Response to NRC GL 93-04, Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies. 10 CFR 50.54(f)

I.

Compensatory Actions e

Describe any compensatory short-term actions taken or that will be taken to address any actual or potential degraded or nonconforming conditions (see Generic Letter 91-18, Reference 1) such as:

1.

" additional cautions or modifications to surveillance and preventive maintenance procedures" -

Westinghouse did not make any initial recommendations regarding surveillance or preventative maintenance procedures.

Based on the response provided in Westinghouse Owners Group (WOG) letter OG-93-42, there was no perceived need to increase the frequency of testing on a

permanent or generic basis.

PSE&G had committed to a temporary increase in testing, but only until it was demonstrated that the rod control system was operating properly.

A recommendation was made for utilities to ensure that their surveillance testing will demonstrate rod control system operability and address maintenance trouble-shooting.

Increased surveillance testing

can, in and of itself, result in higher rates of system and component failures.

Therefore, the WOG and Westinghouse have concluded that increased frequencies in surveillance testing is not required or appropriate in response to the Salem rod control system failure event.

Surveillance testing at BVPS Unit No. 1 and No. 2 meets the requirements of the individual plant -technical specifications.

Every 31 days, when in Modes 1 or 2, each unit verifies that each control rod not_ fully inserted in the core is operable by moving each control rod at least 10 Steps in any one direction.

Comparison' of the individual control rod position-indication with group position indication confirms proper control rod 1

motion.

Existing surveillance procedures are adequate I

to assure rod control system operability, and the existing site / station guidance with respect to maintenance troubleshooting is sufficient to handle-j anticipated problems.

At BVPS Unit No.

1 the i

individual rod position indication and group _ demand position indication are logged every four hours to i

ATTACHMENT j

Response to NRC GL 93-04 Page 2 assure proper alignment of the control rods and proper response of the system to demands for rod motion.

At J

BVPS Unit No. 2 the rod deviation alarms will alert the operators to system malfunctions should the operator i

not detect'the problem when proper control rod response-is observed during/following demands for control rod notion.

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-ATTACHMENT Response to NRC GL 93-04 Page 3 2.

" additional administrative controls for plant startup and power operation" -

PSE&G committed the Salem units to startup by dilution.

Neither Westinghouse nor the WOG has endorsed this requirement.

In actual operation, the operators would be aware of abnormal rod movement and terminate rod demand prior to ever reaching criticality.

The operator would be manually controlling the rod withdrawal such that the detection of rod mis-stepping in under 1

minute would be reasonable.

In fact, as demonstrated during the R.

E.

Ginna

event, abnormal rod motion was terminated after only one step both in automatic and manual rod control.

It is unrealistic to believe that the operators would permit an unchecked rod withdrawal during startup such that criticality would be reached.

Thus, the WOG and Westinghouse have concluded that startup by dilution is not required in response to the Salem rod control system failure event.

i All of the licensed operators for BVPS Unit No. 1 and No.

2 have read the Westinghouse Nuclear Safety Advisory Letter (NSAL-93-007) regarding the rod control system failure which occurred at Salem Unit No.

2.

Thus, the operators are alerted to the potential for improper rod motion.

As a matter of routine operation, the operators at BVPS Unit No.

1 and No. 2 confirm proper response of the control rods to a demand for control rod motion.

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ATTACHMENT Response to NRC GL 93-04 Page 4 3.

" additional instructions and training to heighten operator awareness of potential rod control system failures and to guide operator response in the event of a rod control system malfunction" -

Both Westinghouse and the WOG

have, at various

times, recommended that licensees provide additional discussion,

training, standing orders, etc. to ensure that operators are aware of the Salem event.

The recommendations of the Westinghouse Nuclear Safety Advisory

Letter, which was subsequently endorsed by the WOG via Letter OG-93-42, recognize the benefits of ensuring that plant operators'are i

knowledgeable of Salem rod control system failure event.

As stated previously, the licensed operators'at both BVPS Unit No.

1 and No.

2 have read the Westinghouse Nuclear Safety Advisory Letter (NSAL-93-007) regarding the rod f

control system failure which occurred at Salem Unit No.

2.

This served to heighten the operators' awareness for potential improper response of the rod control system to a 4

demand for control rod motion.

In addition, operators are programmatically trained in normal and abnormal rod control system operation, and drilled on unit-specific simulators in response to rod control system failures.

Following completion of the WOG programs, which are underway i

to investigate and better understand the Salem Unit No. 2

event, recommendations for classroom and simulator training i

and event response procedures may be made by WOG and/or Westinghouse for guidance for all affected utilities.

DLC will utilize such guidance in determining appropriate revisions to training and event response procedures at BVPS Unit No. 1 and No.

2.

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i ATTACHMENT I

Response to NRC GL 93-04 page 5 II. Summary of the Generic Safety Analysis Procram Introduction As part of the Westinghouse Owners Group initiative, the WOG Analysis subcommittee is working on a

generic approach to demonstrate that for all Westinghouse plants there is no safety significance for an asymmetric RCCA withdrawal.

The purpose of the program is to analyze a series of asymmetric rod withdrawal cases from both subcritical and power conditions to demonstrate that DNB does not occur.

The current Westinghouse analysis methodology for the. hank withdrawal at power and from subcritical uses point-kinetics and one dimensional kinetics transient models, respectively.

These models use conservative constant reactivity feedback assumptions which result in an overly conservative prediction of the core response for these events.

A three-dimensional spatial kinetics / systems transient code (LOFT 5/SPNOVA) is being used to show that the localized power peaking is not as severe as current codes predict.

The 3-D transient analysis approach uses a representative standard 4-Loop Westinghouse plant with conservative reactivity assumptions.

Limiting asymmetric rod withdrawal statepoints (i.e., conditions associated with the limiting time in the transient) are established for the representative plant which can be applied to i

all Westinghouse plants.

Differences in plant designs are addressed by using conservative adjustment factors to make a plant-specific DNB assessment.

Description of Asymmetric Rod Withdrawal The accidental withdrawal of one or more RCCAs from the core is assumed to occur which results in an increase in the core power level and the reactor coolant temperature and pressure.

If the reactivity worth of the withdrawn rods is sufficient, the reactor power and/or temperature may increase to the point that the transient is automatically terminated by a reactor trip on a High Nuclear Flux or Over-Temperature Delta-T (OTDT) protection signal.

If the reactivity rise is small, the reactor power will reach a peak value and then decrease due to the negative feedback effect caused by the moderator temperature rise.

The accidental withdrawal of a bank or banks of RCCAs in the normal overlap mode is a

transient which is specifically considered in safety analysis reports.

The consequences of a bank withdrawal accident meet Condition II criteria (no DNB).

If, however, it is assumed that less than a full group or bank of control rods is withdrawn, and these rods are not symmetrically located around the core, this can cause a

" tilt" in the core radial power distribution.

The

" tilt" could result in a radial power distribution peaking factor which is more severe than is normally considered in the plant safety analysis report, and therefore cause a loss of DNB

ATTACHMENT Response to NRC GL 93-04 Page 6 l

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I margin.

Due to the imperfect mixing of the fluid exiting the core before it enters the hot legs of the reactor coolant loops, j

there can be an imbalance in the loop temperatures, and therefore in the measured values of T-avg and delta-T, which are used in the over-Temperature Delta-T protection system for the core.

The radial power

" tilt" may also affect the ex-core detector signals used for the High Nuclear Flux trip.

The axial offset (AO) in the region of the core where the rods are withdrawn may become more positive than the remainder of the core, which can result in an additional DNB penalty.

Methods The LOFT 5 computer code is used to calculate the plant transient response to an asymmetric rod withdrawal.

The LOFT 5 code is a combination of an advanced version of the LOFT 4 code (Reference 1),

which has been used by Westinghouse in the analysis of the RCS behavior to plant transients and accidents, and the advanced nodal code SPNOVA (Reference 2).

LOFT 5 uses a

full-core model, consisting of 193 fuel assemblies with one node per assembly radially and 20 axial nodes.

Several

" hot" rods are specified with different input multipliers on the hot rod powers to simulate the effect of plants with different initial FAH values.

A

" hot" rod represents the fuel rod with the highest FAH in the

assembly, and is calculated by SPNOVA within LOFT 5.

DNBRs are calculated for each hot rod within LOFT 5 with a

simplified DNB-evaluation model using the WRB-1 correlation.

The DNBRs resulting from the LOFT 5 calculations are used for comparison purposes.

A more detailed DNBR analysis is done at the limiting transient statepoints from LOFT 5 using THINC-IV (Reference 3) and the Revised Thermal Design Procedure (RTDP).

RTDP applies to all Westinghouse

plants, maximizes DNBR margins, is approved by the

NRC, and is licensed for a number of Westinghouse plants.

The LOFT 5-calculated DNBRs are conservatively low when compared to I

the THINC-IV results.

Assumptions The initial power levels chosen for the performance of bank and multiple RCCA withdrawal cases are 100%, 60%, 10% and hot zero power (HZP).

These power levels are the same_ powers considered in the RCCA Bank Withdrawal at Power and Bank Withdrawal from Subcritical events presented in the plant Safety Analysis Reports.

The

plant, in accordance with RTDP, is assumed to be operating at nominal conditions for each power level examined.

i Therefore, uncertainties will not affect the results of the LOFT 5

)

I transient analyses.

For the at-power cases, all reactor coolant pumps are assumed to be in operation.

For the hot zero power case (subcritical event),

only 2/4 reactor coolant pumps are assumed to be in operation.

A " poor mixing" assumption is used for the reactor vessel inlet and outlet mixing model.

ATTACHMENT Response to NRC GL 93-04 Page 7 Results A

review of the results presented in Reference 4 indicates that for the asymmetric rod withdrawal cases analyzed with the LOFT 5

code, the DNB design basis is met.

As demonstrated by the A-Factor approach (described below) for addressing various combinations of asymmetric rod withdrawals, the single most-limiting case is plant-specific and is a function of rod insertion

limits, rod control

pattern, and core design.

The results of the A-Factor approach also demonstrates that the cases.

analyzed with the LOFT 5 computer code are sufficiently conservative for a wide range of plant configurations for various asymmetric rod withdrawals.

In addition, when the design FAH is taken into account on the representative plant, the DNBR criterion is met for the at-power cases.

At

HZP, a

worst-case scenario (3-rods withdrawn from three different banks which is not possible) shows a non-limiting DNBR.

This result is applicable to all Westinghouse plants.

Plant Applicability I

The 3-D transient analysis approach uses a

representative standard 4-Loop Westinghouse plant with bounding reactivity assumptions with respect to the core design.

This results in conservative asymmetric rod (s) withdrawal statepoints for the various asymmetric rod withdrawals analyzed.

The majority of.the cases analyzed either did not generate a reactor trip or were 1

terminated by a

High Neutron Flux reactor trip.

For the Overtemperature Delta-T reactor

trip, no credit is assumed for

~

the f(AI) penalty function.

The f(AI) penalty function i

reduces the OTDT setpoint for highly skewed positive or negative I

axial power shapes.

Compared to the plant-specific OTDT setpoints including credit for the f(AI) penalty

function, the setpoint used in the LOFT 5 analyses is conservative; i.e.,

for those cases that tripped on OTDT, a plant-specific OTDT l

setpoint with the f(AI) penalty function wil'1 result in an i

earlier reactor trip than the' LOFT 5 setpoint.

This ensures that i

the statepoints generated for those cases that trip on OTDT are conservative for all Westinghouse plants.

With respect to the neutronic

analyses, an adjustment factor

("A-factor")

was calcu3ated for a wide range of plant types and rod control configurations.

The A-factor is defined as the ratio between the design FAH and the change in the maximum transient FAH from the symmetric and asymmetric RCCA withdrawal cases.

An appropriate and conservative plant-specific A-factor was calculated and used to determine the corresponding DNBR penalty or benefit.

With respect to the_ thermal-hydraulic

analyses, differences in plant conditions (including power level, RCS temperature,

pressure, and flow) are addressed by sensitivities performed using THINC-IV.

These sensitivities are used to_

determine additional DNBR penalties or benefits.

ATTACHMENT

' Response to NRC GL 93-04 Page 8 Uncertainties in the initial conditions are accounted for in the i

DNB design limit.

Once the differences in plant design were accounted for-by the adjustment approach, plant-specific DNBR calculations can be generated for all Westinghouse plants.

i conclusion Using this approach, the preliminary results of the generic analyses' and their plant-specific application demonstrate that for Beaver Valley Power Station Unit No.

1 and No. 2 DNB does not occur for their worst-case asymmetric rod withdrawal.

References 1)

Burnett, T.W.T.,

et al.,

"LOFTRAN Code Description," WCAP-7907-A, j

April 1984.

l 2)

Chao, Y.A.,

et al.,

"SPNOVA A Multi-Dimensional Static and i

Trnasient Computer Program for PWR Core Analysis," WCAP-12394,-

September 1989.

3)

Friedland, A.

J.

and S.

Ray, " Improved THINC IV Modeling for PWR Core Design," WCAP-12330-P, August 1989, j

4)

Huegel, D.,

et al.,

" Generic Assessment of Asymmetric Rod Cluster Control Assembly Withdrawal," WCAP-13803, August 1993.

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