ML20046C391
| ML20046C391 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 08/05/1993 |
| From: | Drawbridge B NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-93-04, GL-93-4, NYN-93109, TAC-M86865, NUDOCS 9308100244 | |
| Download: ML20046C391 (10) | |
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J NOdn,h P.O. Box 300 seebrook. s s 03874 Telephone (603)474 9521 AtIantic Energy Service Corporation N YN-93109 August 5,1993 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention:
Document Control Desk
References:
(a)
Facility Operating License No. NPF-86, Docket No. 50-443 (b)
NRC Generic Letter 93-04, dated June 21,1993, " Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies"
Subject:
Initial (45-Day) Response to Generic Letter 93-04, (TAC No. M86865)
Gentlemen:
The NRC issued Generic Letter 93-04 in response to the rod control system event at the Salem Nuclear Generating Station, Unit 2.
The generic letter requires that, by August 5,1993, licensees of Westinghouse-designed nuclear power reactors, including Seabrook Station, provide plant-specific rod control system information. The generic letter requires an assessment of whether or not the licensing basis for the facility is still satisfied with regard to the requirements for system response to a single failure in the rod control system [ Required Response 1.(a)]. If this assessment indicates that the licensing basis is not satisfied, the generic letter requires a further assessment of the impact of potential rod control system single failures on the facility licensing basis and a 1
description of compensatory short-term actions taken or planned [ Required Response 1 (b)].
A follow-up response is required by September 20,1993 providing a plan and schedule for long-term resolution of the applicable issues [ Required Response 2.],
Subsequent correspondence between the Westinghouse Owners Group (WOG) and the NRC
[
(NRC Letter, A. C. Thadani to Roger Newton, dated July 26,1993) resulted in schedular relief from i
Required Response 1.(a) and the first part of Required Response 1.(b) of the generic letter. This information is now required to be included with the 90-day, follow-up response.
Pursuant to the above, this letter provides the North Atlantic Energy Service Corporation j
_(North Atlantic) response to the second Part of Required Response 1.(b) of Generic Letter 93-04 for Seabrook Station. Enclosure (1) to this letter summarizes the actions taken or planned to date i
by North Atlantic in response to the Salem rod control system event. Enclosure (2) provides a summary of the results of the generic safety analysis program conducted by the WOG and its applicability to Seabrook Station. North Atlantic has reviewed the results of the WOG generic safety analysis program and, based upon discussion with Westinghouse, concurs with the conclusion
- that the results are applicable to Seabrook Station.
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U.S. Nuclear Regulatory Commission ~
August 5,.1993 Attention:
Document Control Desk Page two Should you have any questions on this matter, please contact Mr. James M; Peschel, Regulatory Compliance Manager at (603) 474 9521 extension 3772, Very trt ly yours,
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Bruce L. Dra(vbridge Executive Director Nuclear Production BLD:GK/ tad i
STATE OF NEW llAMPSHIRE Rockingham, ss.
August 5,1993 Then personally appeared before me, the above-named Bruce L. Drawbridge, being duly' sworn, did state that he is Executive Director - Nuclear Production of the North - Atlantic Energy Service' Corporation that he is duly authorized to execute and file the foregoing information in the name and on the behalf of North Atlantic Energy Service Corporation and that the statements therein are true to the best of his knowledge and belief.
AAl 2/flhk Tracy A/ Decredico, Notary Public My Commission Expires: October 3,1995 i
cc:
Mr. Thomas T. Martin i
Regional Administrator U.S. Nuclear Regulatory Commission l
Region 1 475 Allendale Road
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King of Prussia, PA 19406 M'. Albert W. De Agazio, Sr. Project Manager l
r Project Directorate 1-4 Division of Reactor Projects
- U.S. Nuclear Regulatory Comraission Washington,-DC 20555 Mr. Noel Diidley NRC Senior Resident inspector P.O. Box 1149 Seabrook, Nil 03874
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August 5,1993 -
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ENCLOSURE 1 TO NYN-93109
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NORTil ATLANTIC RESPONSE TO GENERIC LETTER (GL) 93-04, REQUIRED RESPONSE 1.(b), SECOND PART FOR SEABROOK STATION
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NRC Reauired Response 1.(bt Second Part:
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), such as additional cautions or modifications to surveillance and preventive maintenance i
procedures additional administrative controls for plant,tartup and power operation 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 North Atlantic Response:
Based on early sources ofinformation, North Atlantic took immediate action to inform Seabrook Station personnel of the rod control systern events at Salem Nuclear Generating Station, Unit 2.
On June 10,1993, NRC Information Notice 93-46 and other industry information on this event were widely distributed within North Atlantic. On June 11,1993, a " Station Information Report" (SIR) was initiated. Normally, an SIR is initiated in response to a potentially reportable event or condition at Seabrook Station. As a conservative measure, an SIR was initiated in response to the Salem event to effect an expeditious operability and reportability determination for Seabrook Station. Based upon the information available at the time, it was determined that there was neither an operability concern with the Seabrook Station rod control system.nor the need to report the potential for a similar problem at Seabrook Station to the NRC. This initial determination will be reviewed as part of the SIR closure process upon resolution of the applicable issues, l
On June i1,1993, Westinghouse issued Nuclear Safety Advisory Letter (NSAL) No.93-007.
NS AL 93-007 contained the following recommendations; j
(1)
Licensed operators should co itinue the normal process of verifying that rod motion is proper for the required movement L
(2)
Confirm the functionality of rod deviation alarms.
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(3)
Complete the WOG survey of historical data regarding rod mispositioning events;
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(4)
Licensed operators should review NSAL 93-007 to ensure their understanding of the.
Salem event.
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In response to the above NSAL recommendations, the following actions were taken:
On June 11, 1993, the Assistant Operations Manager issued a night order to shift personnel emphasizing the need to verify proper rod motion for any rod movement. The night order also required licensed operators to review NSAL 93 007.
On June 11, 1993, a test was conducted which verified'the functionality of the rod deviation alarms at Seabrook Station.
On June 12,1993, NSAL 93-007 was widely distributed for review by North Atlantic personnel.
The Seabrook-specific historical data of rod control system performance and reliability was forwarded to Westinghouse as requested by letter WOG-93-11I dated June 25,1993.
The following discussion addresses the three specific compensatory short-term actions identified in the second part of GL 93-04, Required Response 1.(b). It also describes North Atlantic actions taken in addition to those described above.
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- additional cautions or modifications to surveillance and preventive maintenance procedures" Westinghouse did not make any initial recommendations regarding surveillance or preventive maintenance procedures. Ilased on the response provided in 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 and with confidence. 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 is contrary to the general trend and philosophy of surveillance relaxation in that increased testing can, in and of itself, result in higher rates of system and component failures. Therefore, the WOG and Westinghouse have concluded that increased surveillance testing frequencies are not an appropriate response to the Salem rod control system failure event.
North Atlantic has reviewed applicable surveillance pracedures and concluded that they include the steps necessary to ensure rod control system operability. Administrative controls governing.
corrective maintenance address trouble-shooting. These administrative controls are judged to be effective; and no revisions are planned.
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 rod withdrawal such that the detection of rod mis-stepping would be reasonably expected within one minute. In fact, as demonstrated during the R.E. Ginna event, abnormal rod motion was terminated after only one step in both 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 event.
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2 Nor.th Atlantic has reviewed the applicable procedures for normal operation, including reactor startup, and has concluded that no procedure revisions or additional administrative controls are needed at this time as a result of the Salem rod control system event.
" additional instructions and training to heighten operator awareness of potential rod control 3.
system failures and to guide operator response in the event of a rod control system malfunction" 130th Westinghouse and the WOG have, at various times, recommended that licensees provide additional discussion, training, standing orders, etc. to ensure that their operators are aware of what transpired at Salem. The recommendations of the Westinghouse NSAL, which was subsequently endorsed by the WOG via Letter 0G-93-42, recognize the benefits of ensuring that plant operators are knowledgeable of the Salem rod control system event.
b in response to the Salem rod control system event, North Atlantic management has briefed and issued written orders to Seabrook Station shift personnel to inform and update them on developments regarding this event. In the briefings and written orders to shift personnel, management has stressed the need for control room operators to closely monitor any control rod motion (manual or automatic) and to ensure that it is proper for the circumstances. The Training Department plans to include a review of the Salem rod control system event and North Atlantic's responsive actions in a future cycle oflicensed operator requalification training. This training will be delivered upon completion of the WOG activities on this issue, or sooner, as judged appropriate.
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SUMMARY
OF Tile GENERIC SAFETY ANALYSIS PROGRAM i
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 j
asymmetric rod withdrawal cases from both subcritical and power conditicas to demonstrate that DNB does not occur.
The current Westinghouse analysis methodology for the bank withdrawal at power and from suberitical 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.
The 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 all Westinghouse plants. Differences in plant design 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 plant safety analysis reports. The consequences of a bank withdrawal accident meet Condition 11 criteria (no DNB). I f, 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 margin. Due to the imperfect mixing of the fluid exiting the core before it enters the hot legs of the reactor coolant loops, 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 in 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 for many years by Westinghouse in the analysis of the RCS behavior to plant transients and accidents, and the advanced nodal code SPNOVA (Reference 2).
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LOFTS, 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 Fall values. A " hot" rod represents the fuel rod with the highest Fall in the assembly, and is calculated by SPNOVA within LOFTS. 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 TillNK-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 conservat!vely low when compared to the TillNC-IV results.
Assumptions The initial power levels chosen for the performance of bank and multiple RCCA withdrawal cases w ere 100%, 60%,10%, and hot zero power (llZP). These power levels are the same powers considered in the RCCA Bank Withdrawal at Power and Bank Withdrawal from Subcritical events 7
presented in the plant Safety Analysis Reports. The plant, in accordance with RTDP, is assumed to be operating at the nominal conditions at each power level examined Therefore, uncertainties will n( t affect the results of the LOFT 5 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 two of the four reactor coolant pumps are assumed to be in operation. A " poor mixing" assumption is used for the reactor vessel inlet and outlet mixing model, Results
-j 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-l I
Factor approach (described below) for addressing various combinations of asymmetric rod withdrawals, the single most-limiting case is plant srcific and is a function of rod insertion limits,
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rod control pattern, and core design. The results u the A-Factor approach also demonstrate that i
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 Foll 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 other Westinghouse plants, Plant Applicability
.j 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 terminated by a high neutron flux rector trip, For the overtemperature delta-T reactor trip, no credit is assumed for the f(AI) penalty function. The f(Al) penalty function reduces the OTDT setpoint for highly skewed positive.or negative 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 setpoint with the f(AI) penalty function will result in an earlier reactor trip than the LOFT 5 setpoint. This ensures that the 2
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statepojnts 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 calculated for a wide range of plant types and rod control configurations. The A-factor is defined as the ratio between the design Fall and the change in the maximum transient Fall from the symmetric and asymmetric RCCA withdrawal cases. An appropriate and conservative plant-specific A-factor was calculated and used to determine the corresponding DNilR 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 sensitivies performed using TlilNC-IV. These sensitivities are used to determine additional DNBR penalties or benefits. Uncertainties in the initial conditions-are accounted for in the 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.
Conclusion Using this approach, the generic analyses and their plant-specific application demonstrated that for i
Seabrook Station DNB does not occur for their worst-case asymmetric rod withdrawal.
References
- 1) Burnett, T.W.T., et al., "LOFTR AN" Code Description," WC AP-7907-A, April 1984.
- 2) Chao, Y. A., et al., "SPNOVA - A Multi-Dimensional Static and Transient Computer Program for PWR Core Analysis," WCAP-12394, September 1989.
12330-P, August 1989
- 4) lluegel, D.,
et al., " Generic Assessment of Asymmetric Rod Cluster Control Assembly Withdrawal." WC Ap-13803, August 1993.
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