ML20056D266
| ML20056D266 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/05/1993 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-93-087, CON-NRC-93-87 GL-93-04, GL-93-4, VPNPD-93-138, NUDOCS 9308110183 | |
| Download: ML20056D266 (8) | |
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Wisconsin J lih?ctnc POWER COMPAth' 1
234 W Michwyin. PO. Mx ' 046. MJ*aukee V.4 53201 (414)221 2345 2
VPNPD-93-138 i
NRC 087 i
August 5, 1993 l
l Document Control Desk U.S.
NUCLEAR REGULATORY COMMISSION Mail Station P1-137
-l Washington, DC 20555 Gentlemen:
DOCKETS 50-266 AND 50-301 RESPONSE TO GENERIC LETTER 93-04 ROD CONTROL SYSTEM FAILURE AND WITHDRAWAL OF ROD CONTROL CLUSTER ASSEMBLIES POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 On June 21, 1993, the Nuclear Regulatory Commission issued Generic Letter 93-04, " Rod Control System Failure and Withdrawal of Rod Control Cluster Assemblies."
The Generic Letter was issued to all licensees with the Westinghouse rod control system.
The letter discussed a potential single failure concern with the Westinghouse rod control system discovered as a result of the event that occurred on May 27, 1993, at the Salem Nuclear Generating Station, Unit 2.
Licensees were requested to respond to Generic Letter 93-04 in accordance with the requirements of 10 CFR 50.54(f).
An assessment of whether or not the licensing basis for their facility was satisfied with regards to system response to a single failure in the rod control system was requested within 45-days of the date of the Generic Letter.
If the licensing basis for the facility is not satisfied, licensees were requested to describe any short-term compensatory actions taken consistent with the guidelines of the Generic Letter and to provide, within 90-days, a plan and schedule for long-term resolution.
Subsequent to the Generic Letter, the Westinghouse Owner's Group (WOG) requested by letter dated July 14, 1993, that licensees be allowed additional time to assess the impact of potential single failures in the rod control system on the plant specific licensing basis.
The NRC granted the WOG's request in a letter dated July 26, 1993, to Mr. Roger Newton, Chairman of the Westinghouse l
Owner's Group.
Accordingly, our assessment of the plant-specific f
licensing basis will be included with our 90-day response.
l 9308110183 930805 N3 I
DR ADOCK 0500 6-yp g g
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Document Control Desk August 5, 1993 Page 2 Included as an attachment to this letter is our response to the Generic Letter as it applies to the Point Beach Nuclear Plant.
The response summarizes the compensatory actions taken in response to the Salem rod control system failure event.
As requested by the NRC's July 26, 1993, letter to the WOG, we have also included the results of the generic safety analysis performed by Westinghouse as they relate to the Point Beach Nuclear Plant.
We believe this letter satisfies the 45-day response requirement of Generic Letter 93-04 as modified by the NRC's July 26, 1993, letter.
If you have any additional questions, please contact us.
Sincerely, Bob Link Vice President Nuclear Power TGM/jg cc:
NRC Regional Administrator NRC Resident Inspector Subscribed and sworn before me 4
this 5 day of (lad
, 1993.
0 x
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L Qanab,a L ekh (JHotany Public, State of Wisconsin My commission expires 10 96 x
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ATTACHMENT RESPONSE TO GENERIC LETTER 93-04 ROD CONTROL SYSTEM FAILURE AND WITHDRAWAL OF ROD CLUSTER CONTROL ASSEMBLIES Generic Letter 93-04 as modified by NRC letter dated July 26, 1993, to Mr. Roger Newton, Chairman of the Westinghouse Owner's Group, requested the following information related to potential single failure concerns with the Westinghouse rod control system:
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) such as additional cautions or modification to surveillance and preventive maintenance procedures additional administrative controls for plant startup and power operation additional instructions and training to heighten operator awareness of potential rod control system failures and to guided operator response in the event of a rod control system malfunction a
provide results from the generic safety analysis program and its applicability to individual licensees The first item listed above is the second part of requested action 1(b) of the Generic Letter.
A response to the second item was requested in the July 26, 1993, NRC letter.
Our response to each of these items follows.
1.
" additional cautions or modifications to surveillance and preventive maintenance procedures."
Response
Westinghouse did not make any initial recommendations regarding surveillance or preventative maintenance procedures in their safety advisory letter (NSAL-93-007) dated June 11, 1993, Based on the Westinghouse Owner's Group evaluation, there was no perceived need to increase the frequency of testing on a permanent or generic basis.
A recommendation was made to ensure surveillance testing will demonstrate rod control system operability and address maintenance troubleshooting.
Increased surveillance testing is contrary to the normal testing philosophy.
That is, increased testing, in and of itself, can result in higher rates of system and component failure.
Therefore, the WOG and Westinghouse have concluded that and increased frequency of rod control system surveillance testing in response to the Salem event is not
~
a'ppropriate.
We agree with this assessment and have not instituted increased testing due to this event.
However, we performed our Technical Specification required bimonthly rod motion test on a twice weekly frequency for three weeks in June of this year for PBNP Unit 2 and-in July for PBNP Unit 1 in order to determine the.cause of intermittent electrical failures in our rod control system.
During this period of increased surveillance testing, we did not experience any events similar to the Salem event.
We have reviewed maintenance.and test procedures related to the rod control-system at PBNP and have determined that, with one exception, modifications to these procedures are not required.
Rod drop test procedures are being reviewed and will be modified-as necessary to ensure that only'one shutdown bank or control bank may be withdrawn at.any given 3
time.
Procedure revisions, if necessary, will be completed j
prior to their next use.
2.
" additional administrative controls for plant startup and power operation."
Response
During reactor startup operations, PBNP operators' 3
manually control rod motion.
During this time, operators are closely monitoring rod motion and would be immediately aware of any abnormal rod motion.
During normal operation, an auto rod motion alarm alerts the operators any time that auto rod motion occurs.
This prompts the' operators to verify that the attendant rod motion is correct for the plant conditions.
It is normal practice for PBNP Operators to verify proper rod motion for the demand.
As suggested by Westinghouse in NSAL-93-007, we have verified the functionality of the rod deviation alarms.
i This provides assurance that operators will be alerted to abnormal rod operation.
We also believe that additional controls are not necessary since the PBNP units are operated as' base load units.
Due to the stable operating conditions in this mode of operation, the low frequency of rod motion demand reduces i
the possible frequency of an event similar to.the occurrence documented in the Generic Letter.
Also, at this time in the present PBNP operating cycles, the units are operating with all rods at or near the top of the core.
In this condition, little rod worth is left, reducing the potential consequences of a similar event.
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."
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4..
m
Response
Both Westinghouse and the WOG have recommended that licensees provide additional discussion, training, etc.
y to ensure that operators are aware of the event documented in the Generic Letter.
An Operations Night Order book entry was written to discuss the Salem event.
This entry emphasized the normal PBNP practice of verifying that rod motion is proper for the demanded movement.
A package of information concerning the event was assembled for review by each operating crew.
The information included an event summary, NRC Information Notice 93-46, Westinghouse NSAL-93-007 and other information pertinent to the event and necessary to heighten operator awareness to the potential consequences.
Necessary operator response to the event was emphasized.
Each crew reviewed the information.
4.
" provide results from the generic safety analysis program and its applicability to individual licensees"
Response
Summary of Generic Safety Analysis Program 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 i
safety significance for an asymmetric rod withdrawal.
The purpose of the generic analysis program is to analyze a series of asymmetric rod withdrawal cases from both subcritical and power conditions to demonstrate that departure from nucleate boiling (DNB) does not occur.
The current Westinghouse analysis methodology for the bank 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 l
prediction of the core response for these events.
A three-dimensional spatial kinetics / systems transient code (LOFTS /SPNOVA) is being used to show that the lucalized 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 designs are H
addressed by using conservative adjustment factors to make a plant-specific DNB assessment.
i
Description
~
The accidental withdrawal of one or more rod control cluster assemblies (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 Overtemperature 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 II criteria (no DNB).
If, however, it is assumed that less than a full group or bank of rods is withdrawn, and these rods are not symmetrically located around the core, a " tilt" in the core radial power distribution will occur.
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 in the measured values of T-avg and delta-T, which are used in the Overtemperature 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
(. O) in the region of the core where the rods are Offset A
withdrawn may become more positive than the axial shape in the remainder of the core, which can result in an additional 1
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 (Ref.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 (Ref. 2).
LOFT 5 uses a full-core model, consisting of 193 fuel j
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 LOFTS.
Departure from nucleate boiling ratios (DNBRs) are calculated for each hot rod within LOFT 5 using the WRB-1
{
correlation in a simplified DNB-evaluation model.
The DNBRs
. resulting.from the LOFT 5 calculations are used for comparison purposes.
1 O-
. --.-L.
e a
w A more detailed DNBR analysis is done at the limiting transient statepoints from LOFT 5 using THINK-IV 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 the THINK-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.
Therefore, uncertainties will not 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 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.
Results A review of the results presented in WCAP-13803, " Generic Assessment of Asymmetric Rod Cluster Control Assembly Withdrawal," 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 demonstrate that the cases analyzed with the LOFT 5 computer code are sufficiently conservative for a wide range of plant configurations.
When the design initial FAH is taken into account on the representative plant, the DNBR criterion is met for the at-power cases.
4 At HZP, a worst-case scenario (3-rods withdrawn from three-i different banks, which is not possible) demonstrates a non-limiting DNBR.
This result is applicable to all other Westinghouse plants.
Plant Applicability The 3-D transient analysis approach uses a representative j
standard 4-Loop Westinghouse plant with bounding reactivity-assumptions with respect to the core design. This results in limiting 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 the transient was terminated by a high neutron flux
(' TDT) reactor trip.
For the Overtemperature Delta-T O
6 reactor trip, no credit is assumed for the FAI reset.
function.
The FAI penalty reduces the OTDT setpoint for; highly skewed positive or negative axial power shapes.
compared to the plant-specific OTDT setpoints including credit.for the FAI penalty, the setpoint used in the LOFT 5 analyses is conservative.
For those cases that tripped on OTDT, a plant-specific OTDT setpoint with the FAI penalty will result in an earlier reactor trip than the LOFT 5 setpoint.
This ensures that the statepoints generated for those cases that trip on OTDT are. conservative for all Westinghouse plants.
Differences in core designs are accounted for by an adjustment factor ("A-factor") which was calculated for a wide range of plant types and rod control configurations.
i The A-factor is 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.
Differences in thermal hydraulic conditions, including power level, RCS temperature, pressure, and flow are addressed by sensitivities performed using THINK-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 adjustments, plant specific DNBR calculations were generated for all Westinghouse plants.
Conclusion Using this approach, the generic analyses and their plant specific application demonstrate that for PBNP, DNB does not occur for the worst-case asymmetric rod withdrawal.
References 1.
- Burnett, T.W.T, et. al.,
"LOFTRAN Code Description,"
WCAP-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.
3.
Friedland, A.J.
and S.
Ray, " Improved THINC IV Modeling for PWR Core Design," WCAP-12330-P, August 1989.
4.
- Huegel, D.,
et, al.,
" Generic Assessment of Asymmetric Rod Cluster Control Assembly Withdrawal," WCAP-13803, August 1993.
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