ML20205S575

From kanterella
Jump to navigation Jump to search
Forwards Topical Rept Evaluations Accepting BAW-1890 & BAW-1893, Justification for Raising Setpoint... & Basis for Raising Arming Threshold..., Respectively,For Ref in License Applications
ML20205S575
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 05/27/1986
From: De Agazio A
Office of Nuclear Reactor Regulation
To: Williams J
TOLEDO EDISON CO.
References
NUDOCS 8606120661
Download: ML20205S575 (2)
v • d • e


Text

{{#Wiki_filter:r - - . .

  • c OMdO/ 6 May 27, 1986 Docket No. 50-346 Mr. Joe Williams, Jr.

Senior Vice Presidant, Nuclear Toledo Edison Company Edison Plaza Stop 712 300 Madison Avenue Toledo, Ohio 43652

Dear Mr. Williams:

SUBJECT:

ACCEPTANCE OF BABC0CK & WILC0X TOPICAL REPORTS BAW-1890 AND BAW-1893 FOR REFERENCE IN LICENSE APPLICATIONS By letters dated April 22 and April 25, 1986, the staff informed Babcock & Wilcox Company of the acceptability of the following Topical Reports for referencing in 'icer.se applications:

1. BAW-1890, " Justification for Raising Setpoint for Reactor Trip on High Pressure", and
2. BAW-1893, " Basis for Raising Arming Threshold for Anticipatory Reactor Trip on Turbine Trip".

Copies of the staff letters are enclosed for your information, tagsenetthea8 W Al De Agazio, Project Manager PWR Project. Directorate #6 Division of PWR Licensing-B

Enclosure:

As Stated cc w/ enclosure: See Next Page DISTRIBUTION ACRS-10 WPaulson Docket File BGrimes NRC PDR JPartlow L PDR ADe Agazio PBD-6 Pdg RIngram FMiraglia Gray File OELD EBrach EJordan H0rnstein PBU-6 P - ADe Agazio;jak WPaulson 5 86 5/p]/86 8606120661 860527 PDR P ADOCK 05000346 PDR

Mr. J. Williams Davis-Besse Nuclear Power Station Toledo Edison Company Unit No. I cc: Donald H. Hauser, Esq. Ohio Department of Health The Cleveland Electric ATTN: Radiological Health Illuminating Company Program Director P. O. Box 5000 P. O. Box 118 Cleveland, Ohio 44101 Columbus, Ohio 43216 Mr. Robert F. Peters Attorney General Manager, Nuclear Licensing Department of Attorney Toledo Edison Company General Edison Plaza 30 East Broad Street 300 Madison Avenue Columbus, Ohio 43215 Toledo, Ohio 43652 Mr. James W. Harris, Director Gerald Charnofi, Esq. (Addressee Only) Shaw, Pittman, Potts Division of Power Generation and Trowbridge Ohio Department of Industrial Relations 1800 M Street, N.W. 2323 West 5th Avenue Washington, D.C. 20036 P. O. Box 825 Columbus, Ohio 43216 Mr. Paul M. Smart, President Mr. Harold Kohn, Staff Scientist Toledo Edison Company Power Siting Commission 

 -    300 Madison Avenue                   361 East Broad Street Toledo, Ohio 43652                   Columbus, Ohio 43216 Mr. Robert B. Borsum                 President, Board of Babcock & Wilcox                       County Commissioners of Nuclear Power Generation               Ottawa County Division                           Port Clinton, Ohio 43452 Suite 200, 7910 Woodmont Avenue Bethesda, Maryland 20814 Resident Inspector U.S. Nuclear Regulatory Commission 5503 N. State Route 2 Oak Harbor, Ohio 43449 Regional Administrator, Region III l      U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137

   ,i,1q.fi $,
                          .,.l
                           ^'

SG]y*. ffSyg.'? j Q l *); ' 

                                                                                                            ,f-
      ' j Jr s O
                                                          /NUC' LEAR REGULATORY COMMISSION             <-    '
              !6 ~ -{                            5                   WASHINGTON, D. C. 20656 April 22, 1986 Mr. J. H. Taylor, Manager, Licensing Babcock & Wilcox Company 3315 Old Forest Road Post Office Box 1260 Lynchburg,. Virginia 24505-1260                            ,

Dear Mr. Taylor:

SUBJECT:

ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REPORT BAW-1890,- 

                                               " JUSTIFICATION FOR RAISING SETPOINT FOR REACTOR TRIP ON HIGH PRESSURE" The Nuclear Regulatory Commission (NRC) staff has completed its review of the Babcock & Wilcox Licensing Topical Report BAW-1890 entitled, " Justification For Raising Setpoint For Reactor Trip On High Pressure," that was prepared for the B&W Owners Group. The report discusses the effect of the high pressure reactor trip setpoint on overpressure transients in B&W reactors. The report describes the impact of the setpoint on reactor trip frequency, the plant transient data, the analysis methodology, the NRC requirements that must be met, and the results that were obtained.

We find the report to be acceptable for referencing in license applications to the extent specified and under the limitations delineated in the report and the associated NRC evaluation, which is enclosed. The evaluation defines 

         '                        the basis for acceptance of the report.

We do not intend to repeat our review of the matters described in the report and found acceptable when the report appears as a reference in license applications, except to assure that the material presented is applicable to the specific plant involved. Our acceptance applies only to the matters described in the report.

In accordance with procedures established in NUREG-0390, it is requested that
       !                        B&W publish an accepted version of this report within three monthes of receipt of this letter. The accepted version shall incorporate this letter and the enclosed evaluation after the title page. The accepted version shall include an -A (designating accepted) following the report identification symbol.

CONTACT: Daniel Fieno, RSB/DPL-B x27742 

           , y N/JCoff3 Q

, -: . .. c. . 

          ..   ..                                                                            l 1
 .            J. H. Taylor                                            April 22, 1986 Should our criteria or regulations change such that our conclusions as to the acceptability of the report are invalidated, B&W and/or the applicants referencing the topical report will be expected to revise and resubmit their respective documentation, or submit justification for the continued effective applicability of the topical report without revision of their respective documentation.

Sincerely, 

                                                             $hk8      -

enn s M. Crutch le' ssistant Director for Technical Supp t 

                                          .        Division of PWR Licensing-B

Enclosure:

Topical Report Evaluation cc: C. Rossi G. Lainas l l i

r ENCLOSURE SAFETY EVALUATION OF TOPICAL REPORT BAW-1890, 

                     " JUSTIFICATION FOR RAISING SETPOINT FOR REACTOR TRIP ON HIGH PRESSURE" TOPICAL REPORT EVALUATION
1. INTRODUCTION ,

This Babcock & Wilcox (B&W) report was submitted on behalf of the B&W Owners Group to justify increasing the high pressure trip setpoint .from its current value of 2300 psig to 2355 psig. The current value of the 2300 psig high pressure trip setpoint was based on changes required by the staff (Ref.1)r subsequent to the TMI-2 accident, to reduce challenge to and opening of the power operated relief valve *(PORV). Two other changes that are pertinent to this report were required: (1) raising the PORV setpoint from 2255 psig to 2450 psig and (2) implementation of a safety-grade automatic anticipatory reactor trip for, among other things, a turbine trip for power levels of 20 

       .       percent and higher. These modifications have met the NRC requirements that' (1) the PORV will open less than 5% of the time for all anticipated over-pressure transients (Ref. 2, Item II.K.3.7) and (2) the probability of a small-break LOCA (SBLOCA), caused by a stuck-open PORV, is not a significant contributor based     on theto the probability)of WASH-1400             a small-break (Ref. 3 probability        LOCA(Sequence of a SBLOCA  (Ref. 2, Item S2 )- II.K.3.2)

Although these TMI required modifications have met the objectives of re-ducing challenges to and opening of the PORV during anticipated high pressure transients, they have increased the frequency of reactor trips. Each reactor trip results in a challenge to plant safety systems. Appro-priate reductions in reactor trip frequency will contribute to overall plant safety as well as plant availability. The report states that a number of high pressure transients would not have resulted in a reactor trip if more margin had been available to the high pressure trip setpoint. The report further states that the present analysis demonstrates that the NRC requirements would be met with the high pressure trip setpoint at 2355 psig rather than at 2300 psig. Moreover, if the anticipatory reactor trip (ART) on turbine trip setpoint is raised from 20% to 45% power, an additional reduction in reactor trip frequency would occur. The total reduction in reactor trip frequency is estimated to be about 10%. The B&W report (Ref. 4) on raising the ART setpoint power is the subject of a separate staff evaluation. The report discusses the post-TM1 high pressure reactor trip data base and the impact on the reactor trip frequency. A discussion is provided of the analysis methodology. The results of the present study are compared to previous results and are demonstrated to meet NRC requirements. The staff evaluation of this licensing topical report follows. WlZ TG2 W VA ts/

2-II. EVALUATION A. Impact of Previous and Proposed Post-TMI Changes B&W compared the average high pressure trip frequency for its plants in the pre-1979 and post-1979 periods. B&W found that the average trip frequency for its plants remained about the same although individual plant data varied. However, B&W notes that ARTS in the post-1979 period are, in effect, anticipatory high pressure trips

and should be included in the post-1979 data base. When these trips are included in the data base, the post-1979 high pressure reactor trip frequency is about double the pre-1979 frequency. None of the --

plant data presented in the report reached the PORY pressure setpoint thereby demonstrating the efficacy of the post-TMI modifications to the PORY and high pressure setpoints and the ART in preventing the PORV from opening. This analysis of plant data on high pressure trip frequency is acceptable and demonstrates the increased reactor trip frequency caused by the TMI modifications. . B&W evaluated the potential for reactor trip frequency reduction for (1) an increase in the high pressure reactor trip setpoint by 55 psi back to the original FSAR value of 2355 psig and (2) an increase in the power level threshold for the turbine trip ART from 20% to 45% (Ref. 4). The first change would provide more margin to the high pressure reactor trip setpoint and would allow some minor plant upsets to either avoid reactor trip or provide the operator sufficient time to perform an action which would not result in a reactor trip. The second change in conjunction with an increased high pressure reactor trip setpoint, would not require a reactor trip for additional low power turbine trips. This second change will be the subject of a separate staff evaluation. The analysis of potential reactor trip frequency reduction demonstrates, from the data, that a number of high pressure and anticipatory reactor trips could be avoided. That is, a potential 10% reduction in reactor trip frequency may be possible. B. Staff Reviews of NUREG-0737 Requirements on the PORV 

      .                      Raising the high pressure reactor trip setpoint may reduce the frequency of reactor trips but NRC imposed post-TMI requirements on the PORV must still be met. The report contains a new analysis which is the main subject of this review, to demonstrate that these requirements are met. A report (Ref. 5) had previously been provided by B&W in response to Item II.K.3.2 of Reference 2 that demonstrated that a stuck-open PORV, with a high pressure trip setpoint of 2300 psig and a PORV setpoint of 2450 psig, would not be a significant contributor to a SBLOCA (Sequency S ). This report was reviewed by a staff coasul-2 tant, Franklin Research                                     Center (A Division of the Franklin Institute),

who submitted an evaluation (Ref. 6) which concluded that the B&W

licensees met the requirements of Item II.K.3.2. The staff issued its own safety evaluation report (Ref. 7) concluding that
"We have l

determined that the requirements of NUREG-0737, Item II.K.3.2 are met with the existing PORV, SV, and high pressure reactor trip setpoints..." This staff safety evaluation report trip implies, in addition, that the requirement of NUREG-0737, Item II.K.3.7, with regard to the frequency of PORV opening per high pressure transient, is met. C. Method of Analysis of Effect of Proposed High Pressure Reactor Trip Setpoint on PORV Openings The report presents analyses to demonstrate that the proposed high pressure reactor trip setpoint will meet the NRC requirements on PORV openings during high pressure transients. Those transients with excessive HPI or total loss of main and auxiliary feedwater are not - considered since they qould result in the PORV opening regardless of the high pressure reactor trip setpoint. The report reviews the actual high pressure reactor trip setpoints and the allowance made for instrument drifts and uncertainties. This error was assumed to vary 

      ~

from 0 to +5 psi in the Monte Carlo simulation to be discussed later. . The determination of the error to be applied to the analysis of PORV openings is, therefore, acceptable since the error increases the high pressure reactor trip setpoint (i.e., less reactor pressure overshoot would be required to open the PORV). The amount of pressure overshoot (i.e., the maximum reactor pressure minus the high pressure reactor trip setpoint) that occurs during a high pressure transient is a function of the heat transfer rates between the primary and secondary systems. The maximum reactor pressure is dependent on the pressurization rate prior to reactor trip and the time after trip when the reactor power is decreasing suf-ficiently. Some 47 plant transients were examined to detennine the actual pressure overshoots that occurred. Although instrument string errors downstream of the Reactor Protection System (RPS), uncertainties due to print out device readability, and uncertainties due to,daja recording frequencies are included in the data, the indicated maximum pressure minus the high pressure reactor trip setpoint was conservatively assumed to be entirely due to pressure overshoot. The various errors will, however, be included in the Monte Carlo simulation to be discussed later. These errors are, therefore, counted twice in the analysis. The 47 transients indicated that the three most important categories of high pressure trip events are: (1) total or partial loss of feedwater, (2) feedwater/ power mismatches during turbine runbacks, and (3) load rejections /MSIV closures. The pressurization rates for these transients varied from about 2 to 40 psi /sec with a corrsponding time to maximum reactor pressure varying from about 2 minutes to about 5 seconds. Our review of the information and data presented indicates that the overshoot distribution that was obtained is acceptable since (1) pressure a sufficient number and range of applicable transients were evaluated. (2) a conservative detennination of the overpressure was made, and (3) the capabilities of the recording devices were taken into account. I

Since the overshoot distribution was obtained from transients with a 2300 psig high pressure reactor trip setpoint, analyses were performed with the POWERTRAIN (Ref.8) program to determine if the distribution would be valid at the 2355 psig setpoint. POWERTRAIN has been reviewed and approved by the staff (approval letter dated November 28,1983). A turbine trip from full power with no anticipatory reactor trip was selected for study since it would cause the largest pressure overshoot. The results indicated that POWERTRAIN was in agreement with plant data obtained at the 2300 psig high pressure reactor trip results. Analyses at the higher setpoint of 2355 psig indicated that pressure overshoot is a weak function of the high pressure reactor trip setpoint. In fact, the overshoot actually decreases as the setpoint is raised because of the complex behavior of the nucleate boiling region in the ~ steam generators. The.over-pressure distribution from plant high pressure reactor trips at the 2300 psig setpoint is conservative and is, therefore, acceptable when used at the higher setpoint in the Monte Carlo simulation to be discussed below. The report describes the Monte Carlo analysis used to stochastically simulate the response of the four channels of the RPS and the control instrumentation for the PORV on the receipt of a pressure signal. The major sources of uncertainty included in the simulation are the uncertainties in the RPS and the NN1 signal processing and the un-certainties in the high pressure trip and PORV setpoints. The NNI channel provides the signal for opening the PORV. The high pressure trip uncertainty is taken to be a uniform distribution from 0 to +5 psi while the PORV setpoint uncertainty is taken to be a uniform distri-bution from 0 to -5 psi. The pressure overshoot results obtained from the plant high pressure reactor trip data is treated as a physical phenomenon having an exponential distribution. This distribution is truncated between 10 psi and 60 psi. Cases in the Monte Carlo analyses that gave overshoots less than 10 psi were set to 10 psi and cases that gave overshoots greater than 60 psi were set to 60 psi. This resulted in a conservative representation of the distribution derived from the 47 plant transients, as the pressure overshoot in these transients was always less than 60 psi. A successful Monte Carlo simulation resulted when 2 out of 4 RPS l channels trip on the assumed high pressure trip setpoint. The l pressure, chcsen as the highest value from the 2 of 4 channels that j caused the trip, is next incremented with the pressure overshoot chosen from the exponential distribution. This pressure is then processed by the Monte Carlo program using the NNI channel to determine if the PORV setpoint has been reached. This Monte Carlo process is repeated until a sufficient number of high pressure trip events have been accumulated to provide adequate statistics for the specified high i pressure trip setpoint. Based on our review, we conclude that the treatment of the uncertainties used and their distribution, the treatment of the pressure overshoot distribution, and the Monte Carlo simulation process itself are conventional and appropriate and are, therefore, acceptable.

D. Comparison of Results for PORY Opening with NRC Requirements The Monte Carlo simulation indicated that there would be one FORV opening per 100,000 high pressure trips at the proposed high pressure reactor trip setpoint of 2355 psig. This frequency of 0.00001 is much less than the NRC requirement of less than 0.05 PORV openings per overpressure transient events that required a L reactor trip. Therefore, Item II.K.3.7 of NUREG-0737 remains satisfied. The report states that there were 65 high pressure trips from 1980 through 1984 for the 7 operating B&W reactors. This yields -- an average of 65/35 or 1.86 events per reactor year. Thus, the probability of a PORV opening per reactor year is given by: 1.86 events

  • 1.0 x 10-5 PORV openings = 1.86 x 10-5 PORV openings reactor-year event reactor-yea r The PORV opening frequency from all other causes is 8.06 x 10-2 (Ref.6). Therefore, the total PORV opening frequency at the proposed i

setpoint of 2355 psig is : L 8.05 x 10-2 + 1.86 x 10-5 = 8.06 x 10-2 total PORV openings reactor year The total PORY openings per reactor year is negligibly changed over the values presented in References 5 and 6 since operator actions under AT0G (abnormal transient operating guidelines) and, to a lesser degree, instrumentation and control faults dominate the togal PORV opening frequency. Using the Reference 7 value of 2 x 10- failures per demand for the PORY failure probability gives:

-0RV failures = 8.06 x 10-2 PORV openings
  • 2 x 10-2 failures reactor-year reactor-year demand
                                                         = 1.6 x 10-3 SincetheprobabilityofaSBLOCA(Sequences)caysedby-gstuck-open PORV is within the WASH-1400 (Ref. 3) range of 10- to 10 per reactor-year, the requirements of Item II.K.3.2 of NUREG-0737 remains satisfied.

This is as expected since the PORV opening frequency due to over pressure reactor events that cause a high pressure trip is neglibibly affected by the proposed high pressure reactor trip setpoint of 2355 psig. E. _ Comparison of Present Analysis to Previous Analysis The report states that the main difference between the present analysis and the previous analysis was in the treatment of the pressure over-shoot. The analysis methodology and other statistical components are similar. In the previous analysis the overpressure had to be based on plant data where the PORV opened. This led to a large uncertainty in

the actual pressure overshoot determination. This was reflected in the use of a nonnal distribution with a large standard deviation (27.5 psi) to accommodate the wide scatter in the data. The present analysis uses plant data for transients for which the PORV did not open. It is believed that the present analysis has a more realistic assessment of the actual overpressure that occurs for the high pressure transients considered in this report. The staff concurs with this assessment of the differences between the present and previous analyses. III. CONCLUSION The staff has reviewed the Babcock & Wilcox licensing topical report on the high pressure reactor trip setpoint and concludes that it is -. acceptable to increase the high pressure reactor trip setpoint for B&W plants from 2300 ptig to 2355 psig while the PORV setpoint remains at 2450 psig. The staff concludes that this setpoint change meets the NRC requirements of NUREG-0737, Items II.K.3.2 and II.K.3.7 regarding PORV openings and PORV caused SBLOCA. Similarly, the requirements on , this matter embodied in IE Bulletin 79-05B are also met. Accordingly, the staff concludes that the licensing topical report may be referenced in licensing submittals by the B&W Owners Group members. Since this report, of necessity, must use analyses based on a statis-tical approach, uncertainties are inherent in the results obtained. Additional uncertainty in the results are caused by the modeling, the assumptions made, and the data that are used. Therefore, as plant experience is accumulated with the proposed high pressure reactor trip setpoint, the staff should be kept infonned of any sig-nificant deviation from the assumptions and results presented in the report. IV. REFERENCES

1. " Nuclear Incident at Three Mile Island - Supplement," IE Bulletin 79-05B, April 21, 1979.
2. " Clarification of TMI Action Plan Requirements," NUREG-0737, November 1980,
3. " Reactor Safety Study - An Assessment of Accident Risks in U. S.

Commercial Nuclear Power Plants," WASH-1400, 1975.

4. " Basis for Raising Arming Threshold For Anticipatory Reactor Trip i

on Turbine Trip," BAW-1893, October 1985.

5. " Report on PORY Opening Probability and Justification for Present Systems and Setpoints," 12-1122779 Rev. 1, Babcowk & Wilcox report, January 1981.

I l

l 7-

6. " Operating Reactor PORV Reports (F-37), Generic Report - Babcock
                                                              & Wilcox Designed Units," Franklin Research Center, July 20, 1983,                                                                                                                                                i
7. NRC Memorandum from F. H. Rowsome to G. C. Lainas dated August 24, 1983; entitled " Safety Evaluation of the B&W Licensees' Responses '

to THI Action Item II.K.3.2."

8. "POWERTRAIN: Hybrid Computer Simulation of a Babcock & Wilcox Nuclear Power Plant," N.S. Yee and J. A. Weimer, BAW-10149, Rev.1, November 1981.

1 l l l 

                   ,}.    [ .he .                                                                       f.-            .              UNITEhSTAT'ES
                    [                                                      g                                         NUCLEAR REGULATORY COMMISSION -                                            ,_

f. 5 - WASHINGTON, D.,C.20655 .- 

                           *g April 25, 1986 Mr. J. N. Taylor, Manager, Licensing Babcock & Wilcox Company 3315 Old Forest Road Post Office Box 1260 Lynchburg, Virginia 24505-1260

Dear Mr. Taylor:

SUBJECT:

ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REPORT BAW-1893, 

                                                                                                           " BASIS FOR RAISIfjG ARMING THRESHOLD FOR ANTICIPATORY REACTOR TRIP
       ,                                                                                                   ON TURBINE TRIP" The Nuclear Regulatory C mnission (NRC) staff has completed its review of the Babcock & Wilcox Licensing Topical Report'BAW-1893 entitled, " Basis For Raising Arming Threshold for Anticipatory Reactor Trip On Turbine Trip," .

that was prepared for the B&W Owners Group. The report discusses the effect of the power threshold for the anticipatory reactor trip (ART) on turbine trips and power runbacks in B&W reactors. The report describes the impact of the turbine trip ART power level threshold on reactor trip frequency, the plant transient data, the analysis methodology, and the results that were obtained. The staff finds the report to be acceptable for referencing in license applications to the extent specified and under the limitations delineated .a in the report and the associated NRC evaluation, which is enclosed. The evaluation defines the basis for acceptance of the report. The staff does not intend to repeat its review of the matters described in the report and found acceptable when the report appears as a reference in license applications, except to assure that the material presented is applicable to the specific plant involved. The staff's acceptance applies only to the matters described in the report. In accordance with procedures established in NUREG-0390, it is requested that B&W publish an accepted version of this report within three months of receipt of this letter. The accepted version shall incorporate this letter and the enclosed evaluation after the title page. The accepted version shall include an -A (designating accepted) following the report identification symbol. Should the staff's criteria or regulations change such that its conclusions as to the acceptability of the report are invalidated, B&W and/or the applicants CONTACT: Daniel Fieno RSB/0PL-B x27742 (_,_ .ld n A w c - 

                           * ( U M ' " l-                                                                         ... . & , ,1 D'                                                              ,

                                                                                                     \
 .           J. H. Taylor                                               April 26, 1986 referencing the topical report will be expected to revise and restbmit their respective documentation, or submit justification for the continued effective applicability of the topical report without revision of their respective documentation.

Sincerely, 

                                                           *)11.

Dennis M. Crute fiel , ssistant Director for Technical S;ppo Division of PWR Licensing-B ^

Enclosure:

Safety Evaluation cc: C. Rossi G. Lainas ,

ENCLOSURE SAFETY EVALUATION OF TOPICAL REPORT BAW-1893, 

                                " BASIS FOR RAISING ARMING THRESHOLD FOR ANTICIPATORY TRIP ON TURBINE TRIP" I.       INTRODUCTION
            . This Babcock & Wilcox (B&W) report was submitted on behalf of the B&W Owners Group to justify increasing the anticipatory reactor trip (ART) setpoint on turbine trip from its current value of 20% power to 45% power. The current value of the 20% power ART setpoint on turbine trip was based on changes required by the staff (Ref.1) subsequent to the TMI accident to reduce challenges to and opening of the power operated relief valve (PORV). Two other changes that are pertinent to this report were required: (1) raising the PORV setpoint from 2255 psig to 2450 psig and (2) lowering the high pressure reactor trip setpoint from 2355 psig to 2300 psig. These modi-fications have met the NRC requirements that (1) the PORV will open less than 5% of the time for all anticipated overpressure transients (Ref. 2, Item II.K.3.7) and (2) the probability of a small-break LOCA (SBLOCA),

caused by a stuck-open PORV, will be less than 0.001 per reactor-year (Ref. 2, Item II.K.3.2) which is based on the WASH-1400 (Ref. 3) median probability of a SBLOCA (Sequence S ). Although these TMI required 2 modifications have met the objectives of reducing challenges to and opening of the PORV during anticipated high pressure transients, they have increased the frequency of reactor trips. Each reactor trip results in a challenge to plant safety systems and any reduction in reactor trip frequency will contribute to overall plant safety as well as plant availability. J gy 502 T1 9 liff'

s .. The report states that a number of turbine trips would not have resulted in a reactor trip if more margin had been available in the ART power level setpoint. The report further states that the present analysis demonstrates that the NRC requirements would be met with the ART power level setpoint at 45% power rather than at 20% power. In fact, the report states that these requirements on PORV openings would be met regardless of whether or not ART is implemented. Moreover, if the high pressure reactor trip setpoint is increased from 2300 psig to 2355 psig, an additional reduction in reactor trip frequency would be posssible. The total reduction in reactor trip frequency is estimated to be about 10%. The B&W report (Ref. 4) on raising the high pressure reactor trip setpoint has been evaluated by the staff (Ref. 5). This staff safety evaluation report concluded that it was acceptable to raise t'.s high pressure reactor trip setpoint from 2300 psig to 2355 psig. This increased high pressure reactor trip setpoint is assumed in the analyses performed in support of raising the turbine trip ART power level threshold to 45%. The report discusses the post-TMI turbine trip / reactor trip data base and the impact on the reactor trip frequency. A discussion is presented of the analysis methodology. The results of the present study are used to justify the turbine trip ART proposed threshold power level of 45%. The staff evaluation of this licensing topical report follows. } __ _ __ - _ . - - - _ - 

            .                                                                      ~
 -m                                                                                       ,
                                                                                              .c,,.s     'l[FMEM              ;

II. EVALUATION A. Impact of Previous and Proposed Turbine Trip ART Power Threshold The report discusses the response of B&W plants to turbine trips. Prior c to the TMI accident, a turbine trip caused a reactor power runback. For some plants successful runbacks were demonstrated for power levels as high as 100%. However, these runbacks were dependent, to some

  • 3 degree, on the PORV opening. Since the TMI accident, the turbine trip ART, among other changes, was instituted to reduce challenges to the 4

PORV. This turbine trip ART now results in a reactor trip whenever the turbine trips and the reactor power level is 20% or higher. Although the NRC requirements on PORY challenges are met by the various post-TMI changes, an undesired side-effect of increased frequency of reactor j trip and consequent challenges to the plant safety system has occurred. ' The data presented in the report show that 52 turbine trips occurred in the period from January 1,1980 to January 1,1985. Twelve of these trips occurred between power levels of 20% to 40%. Raising the turbine trip ART power level threshold has the potential for reducing the reactor trip frequency without affecting PORY opening frequency. Based on our i review, we concur with the applicant that the analysis of plant data ) on reactor trips caused by turbine trips demonstrates that reactor trip k frequency increased as a result of TMI modifications. i __ , u - .. . _ . . _._ _.. _ _ _ _ _ _ _ _ _ __ l _. . _ . _ _ _ _

_4 B&W evaluated the potential for reactor trip reduction for (1) in-creasing the high pressure reactor trip setpoint by 55 psi back to the original FSAR value of 2355 psig (Ref. 4) and (2) increasing the power level threshold for the turbine trip ART from 20% to 45%. The first change would provide more margin to the reactor trip setpoint and would allow son:e minor plant upsets to either avoid reactor trip or provide the operator sufficient time to perform an action which 

     .                           would not result in a reactor trip. The second change, in conjunction-with an increased high pressure reactor trip setpoint, would not re-quire a reactor trip for some additional low power turbine trips. We find that the analysis of potential reactor trip frequency reduction is reasonable and demonstrates from the data in the report and Reference 4 that a number of high pressure and anticipatory reactor trips could be avoided. That is, a potential 10% reduction in reactor trip frequency may be posssible.

B. Results of Analysis of Turbine Trip ART l The POWERTRAIN (Ref. 6) program was used by B&W to evaluate the factors which are important in power runback on turbine trips without a reactor trip. These factors lead to the determination of the highest initial power level or threshold for the turbine trip ART. Factors evaluated { r l L _ . _ _ _ _ - - -_ - 

                                                                                                                                                                              \

j i 5-included (1) the total bypass steam flow, (2) the moderator temperature coefficient, (3) the initial power level, (4) the power runback rate, and (5) the pressurizer spray flow rate. The cases evaluated were turbine trips with runbacks modeled with a reactor closely resembling Rancho Seco. A successful runback case was defined by B&W to have the 

  ,                                      following desirable performance characteristics:                                                   (1)noreactortrip on high reactor system
  • pressure, (2) no auxiliary feedwater actuation on low steam generator level, (3) no steam generator overfill affecting steam quality, and (4) no loss of subcooled margin as affected by reactor system pressure and temperature. Since the modeling, assumptions, and criteria used in the analysis considers the principal factors in a turbine trip with runback, the staff concludes that the methodology used is, therefore, acceptable. In addition, since the POWERTRAIN program has been reviewed and approved by the staff (approval letter dated November 28,1983) the staff concludes that its use is, therefore, acceptable.

From the POWERTRAIN analyses it was determined that the total steam bypass flow was one of the most important factors in determining whether or not a reactor power runback on turbine trip was successful. The total steam bypass flow included turbine bypass flow, atmospheric vent flow and flow through at least one bank of Main Steam Safety ., Valves (MSSV). At least one bank of MSSVs will open at the high pressure reactor trip setpoint of 2355 psig (Ref. 5). In the analyses, 

           . - -   - - , . , - - - - -          - _ -  , , , , - , . . , ---r _,_.. _ . _    -n.,. .,. _,-. , , - - - - .___ , , , _ , , - - _ , , - , , _ _ , _ , , _ _

s .. 

  .                                                                                                         l e

if the core power decreases because of control rod insertions and negative moderator temperature coefficient, to the total steam bypass flow before the high pressure reactor trip setpoint is reached, suffi-cient primary to secondary heat transfer exists to stop the reactor system pressure from increasing. These results presented show that the larger the total steam bypass flow the higher the power threshold that can be tolerated by the turbine trip ART. The reactor coolant temperature and pressure increases during the early stages of a turbine trip. Themoderator(andDoppler) reactivity coefficient are negative throughout a reactor cycle. These negative coefficients, therefore, help to reduce the reactor power and thus help the reactor power runback process caused by control rod insertion. POWERTRAIN results were obtained for near beginning-of-cycle (BOC) and end-of-cycle (EOC)caseswhichdemonstratesthiseffect. Therefore, for the same total steam bypass flow and control rod insertion rate, successful reactor power runbacks are more probable the more negative the moderator temperature coefficient becomes. The initial power level is a factor in determining a successful power runback along with the total steam bypass flow and moderator temper-i ature coefficient. POWERTRAIN results established, as expected, that successful reactor power runbacks from higher initial reactor power would require higher total steam bypass flow. POWERTRAIN results were also obtained for two other factors. These were the Integrated Control 

     ,                                                                                      1 l

System (ICS) runback rate on control rod insertion and the pressurizer spray flow rates. The ICS runback rate was changed from 20% per minute to 50% per minute but this did not change the overall control rod in-sertion rate during the important early stages of a turbine trip transient where the moderator temperature coefficient is also important. . Therefore, the indicated ICS runback rate had negligible influence on the reactor power runbeck on a turbine trip event. Similarly, the pressurizer spray rate was found to have very little effect in turning around the reactor coolant pressure in the time period of interest. The conclusions of this POWERTRAIN analysis were that, for a given control rod reactivity insertion rate and high pressure reactor trip setpoint, the most important factors, in determining whether or not a reactor power runback, on turbine trip is successful, are the initial power level and the total steam bypass flow. The study also concluded that the negative moderator temperature coefficient helped the reactor power runback especially at E0C when it is more negative than say, for example, near BOC. The report concluded that other factors had negligible impact on reactor power runbacks. Since the results in the report were obtained with the approved POWERTRAIN program and since the principal effects were evaluated, the staff concludes that the POWERTRAIN results are, therefore, acceptable.

The report states that the results are applicable to all the B&W 177 fuel assembly (FA) plants. Based on the review of the analyses presented, we concur on the applicability of these results to the 177 FA B&W plants. The report concludes that, for the total steam bypass flow credited in the analysis, the reactor trip on turbine trip power level threshold could be increased from 20% to 45% with a high pressure reactor trip setpoint of 2355 psig. Based on the review of the plant data presented in the report and the POWERTRAIN results, the staff concludes that the B&W assessment regarding the raising of the turbine trip ART power level threshold to 45% is, therefore, acceptable. C. Effect of Turbine Trip ART Proposed Power Level Threshold l On PORV Openings and NRC Requirements Although the results presented in the report are applicable to all B&W 177 FA plants, differences in a number of plant ' parameters may not lead to successful reactor power runbacks on turbine trips with a turbine trip ART power level threshold of 45% and a high pressure reactor trip setpoint of 2355 psig. An unsuccessful power runback will lead to a high pressure trip. Therefore, it is essential to evaluate the effect of these potential additional high pressure trips on the frequency of PORV openings and to determine whether or not NRC requirements on PORV openings are met. l \

The report assumes that 30% of the reactor power runbacks will be unsuccessful. Assuming the same turbine trip frequency at power levels equal to or beiow 45% as occurred in the post-THI period, the report finds the following: 12 (turbine trips) * .30 (reactor trip / turbine trip) 5(years)

  • 7 (reactors)
                       = 0.10 high pressure trips reactor year Then the high pressure trip frequency would increase from 1.86 per          -

reactor-year (Ref. 4) to (1.86 + .10) or 1.96 per reactor-year. The number of PORV openings from high pressure trip events would now be: 1.96 events

  • 1.0 x 10-5 PORV opens = 1.96 x 10-5 PORV openings year event year The total number of PORV openings per reactor-year for all events, as given in Reference 4, is 8.06x10
  • and is negligibly affected by this change. The results of Reference 4 on PORV openings and the probability of a SBLOCA (Sequence2S ) remain applicable. Therefore, the staff concludes that the requirements of Item II.K.3.2 and Item II.K.3.7 of NUREG-0737 (Ref. 2) are met even if a number of reactor power runbacks are unsuccessful at the proposed turbine trip ART power threshold of 45%.

III. CONCLUSION The staff has reviewed the Babcock & Wilcox licensing topical report on the turbine trip ART power level threshold and concludes that it is 4 4

acceptable to increase the turbine trip ART power level threshold for B&W plants from 20% to 45%. The staff concludes that this power level threshold change meets the NRC requirements of NUREG-0737, Items II.K.3.2 and II.K.3.7 regarding PORV openings and PORV caused SBLOCA while benefitting plants by potentially reducing the reactor trip frequency. Similarly, the requirements on this matter embodied in IE Bulletin 79-05B are also met. Accordingly, the staff concludes that the licensing topical report may be 

     ,         referenced in licensing submittals by the B&W Owners Group members.        .

Due to the modeling, assumptions made, and data used, the results presented in the report, as is the case for any analysis, may contain uncertainties. Therefore, as plant experience is accumulated with the proposed turbine trip ART power threshold, the staff should be kept informed of any signi-ficant deviations from the results presented in the report. IV. REFERENCES

1. " Nuclear Incident at Three Mile Island - Supplement," IE Bulletin 79-05B, April 21, 1979.
2. " Clarification of TMI Action Plan Requirements," NUREG-0737, November 1980.
3. " Reactor Safety Study - An Assessment of Accident Risks in U. S.

Commercial Nuclear Power Plant," WASH-1400, 1975. 

 ~~                                                             _

a . l

4. " Justification For Raising Setpoint For Reactor Trips on High Pressure," BAW-1890 September 1985.
5. Letter from D. M. Crutchfield (NRC) to J. H. Taylor (B&W) on
                 " Acceptance For Referencing of Licensing Topical Report BAW-1890, ' Justification For Raising Setpoint For Reactor Trip on High Pressure,1" April 1986.
6. "POWERTRAIN: Hybrid Computer Simulation of a Babcock & Wilcox
  • Nuclear Power Plant," N. S. Yee and J. A. Weimar, BAW-10149, Rev.1, November 1981.

l i i 

                  .                                      -                  _ _ _ _ _ _ _ _}}
Retrieved from "https://kanterella.com/w/index.php?title=ML20205S575&oldid=3325923"