ML20090D584

Safety Evaluation Supporting Amend 154 to License NPF-4

ML20090D584
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Site:	North Anna
Issue date:	03/03/1992
From:	Office of Nuclear Reactor Regulation
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NUDOCS 9203060296
Download: ML20090D584 (10)

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UNITED STATES 3,., 7

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i NUCLEAR REGULATORY COMMISSION

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i

RELATED TO AMENDMENT NO.154

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TO FACILITY OPERATING LICENSE NO. NPF-4 VIRC,1NIA ELECTRIC Ah0 POWER COMPANY OLD 00 Mill 10N ELECTRIC COOPERATIVE NORTH ANNA POWER STATION, UNIT NO. 1 DOCKET NO. 50-338 i

1.0 INTRODUCTION

By. letter dated January 8, 1992, as supplemented by letters dated January 31, February 10 and February 25, 1992, the Virginia Electric and Power Company (the licensee) proposed changes to the Technical Specifications (TS) for the North Anna Power Station, Unit No. 1 (NA-1). The proposed revisions would reduce the minimum allowable Reactor Coolant System.(RCS) total flow rate specified in Table 3.2-1 of TS 3.2.5 from the current value of 284,000 gpm to a value of 268,500 gpm. This revision is temporary and would remain in effect until the currently scheduled 1993 steam generator replacement.

Additionally, an administrative change has been proposed to Table 2.2-1 of TS 2.2.1 -which would~ permanently revise the footnote to specify that the " Design flow per loop is one-third the minimum allowable Reactor Coolant System Total Flow Rate as specified in Table 3.2-1."

The January-31 and February 25, 1992 letters provided additional information requested by the staff regarding the requested changes. The February 10, 1992 letter revised the minimum allcwable RCS flow rate based on a revised increase in SG tube plugging (SGTP)-projections. These submittals are further discussed under Section 3.0 of this SE. The additional information requested by the staff, as-well as the revised minimum allowable RCS flow, did not alter the proposed action or affect the staff's-initial determination of no significant hazards consideration as noticed in the Federal Register on

-February 5, 1992 (57 FR 4563).

The proposed reduction'in RCS minimum flow rate has been prompted by inspection data obtained during a mid-cycle NA-1 SG inspection outage. These results, based on extensive eddy current inspection of the NA-1 SG ttbes, together with the use of conservative analysis guidelines and plugging criteria, indicate that a substantially increased number of tubes will require' plugging. The attendant. reduction in RCS flow rate through the tubes will increase the likelihood that the current TS 3.2.5 requirement on minimum 9203060296 920303 PDR ADOCK 05000338 p

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g RCS flow rate may be violated during continued opeiation with the existing SGs. The proposed decrease of approximately 31 in specified minimum RCS flow rate is intended to bound ary future measurements of RCS flow (required by the TS cnce per fuel cycle) and any anticipated tube plugging up until the 1993 SG replacement. The 3% reduction in RCS flow rate correlates with an approximate 32% level of tube plugging.

RCS total flow rate is a critical input parameter to the analyses presentec in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). Accordingly, to support operation of NA-1 with extended SUTP, the impact of the proposed reduction in flow rate on Chapter 15 analyses must be evaluated. The licensee's January 8,1992 submittal provides a summary of these reevaluations and, in addition, provides assessments of the following:

(1) whetter the current engineered safety features (ESF) and reactor protection system (RPS) setpoints sat forth in the TS ccntinue to provide adeouate plant protection under the reduced flow conditicns due to extended SGTP, (2) whether the current core thermal limits remain bounding under the reduced flow conditions due to extended SGTP, and (3) whether the nuclear steam supply system (NSSS) and balance of plant (B0P) systems and components continue to meet the applicable acceptance criteria under the reduced fbw conditions due to extended SGTP.

The sthf f has complettd the review of the licensee's proposed TS revisions and the technical evaluations submitted to support these revisions. Our evaluation follows.

2.0 EVALUATION The licensee has examined all of the transients addressed in Chapter 15 and has determined which ones require reanalysis and which require only reevaluation.

If an event is potentially impacted by RCS flow rate and also by other effects of SGTP (e.g., increased hydraulic resistance, reduced heat transfer area, reduced RCS volume), t N n the event is either reanalyzed or evaluated using available sensitivity tato for that spec 1fic event.

If a departure from nucleate boiling ratio (DNBR)-limited event is impacted by RCS flow rate but not by other effects of SGTP, a DNBR penalty is assessed, as describcd helow.

Finally, events which are unaffected by RCS flow rate but are impacted by other effects of SGTP, and those events which are impacted by neither flow nor other effects, h6ve not been addressed further. These events have no bearing on the requested TS revisions.

The DNBR design limit and the statistical DNBR limit (SDL) for NA-1 are 1.46 and 1.26, respectively. The SDL is based on the WRB

• CHF correlation with DNBR parameter and correlation uncertainties combined in a statistical manner. The minimum DNBR value computed as part of a transient analysis is assessed against th, A sign limit. The percentage difference between the computed DNBR value and the design limit is termed the "DNBR analysis margin." The " generic ret 61ned margin" is defined as the percentage difference between the design limit and the SDL (i.e., 13.7%).

As noted above, a DNBR penalty is assessed for certain Chapter 15 events to compensate for adverse effects on DNBR due to the proposed reduction in RCS flow rate. The i

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3 penalty, which is the same for all these events, is assessed against the generic retained margin without taking credit for the analysis margin. The magnitude of the penalty (4.8!) is con @uted as the product of the bouncing DNBR partial derivative with respect to RCS flow rate (1.6), and the proposed percent reduction in ficw (31). The value of this partial derivative was based on the UFB-1 CHF correlation for DNBR and determined by considering a wide ronge of statepoint conditions which bounded both normal operation and accident conditions.

The penalty is directly subtracted from the 13.71 generic retained margin, leaving a margin balance of 8.9% against which other penalties such as, for example, the effects of fuel rod bowing, can be assessed. The Chapter 15 transimts for which the 4.8% penalty has been assessed are the following:

- Accidental depressurization of the RCS

- Accidental depressurization of the Main Steen System

- Excessive load increase

- Excessive heat removal (feedwater malfunction)

- Partial loss of RCS ficw

- Rod withdrawal from subtritical

- Spurious operation of Safety injection System

- Single rod withdrawal at power The disposition of the remaining Chapter 15 transients which were reevaluated but not reanalyzed are as follows:

1)

For the main steamline break (MSLB) event, a different penalty (4.3%) was assessed in an analogous manner against a different retained margin (107) because the ItSLB analysis employs another correlation (W-3 CHF rather than WRB-1 CHF) for DNBR calculations. The correlation has 3 different DNBR sensitivity to change in RCS flow rate and therefore yields a different value (1.4) for the bounding partial derivative.

2)

For the control rod drop / misalignment event, revised DNBR limit lines applicable to reduced RCS flow rates have been developed for this transient for application to reload cores.

As such, no penalty need be assessed against generic retained margin.

Although the negative flux rate trip has been retained at NA, the dropped rod analysis methodology employed does not take credit for this trip.

3)

The Chemical Volume Control System (CVCS) malfunction (boron dilution) event is not affected by a reduction in RCS flow rate but is, however, sensitive to the reduction in RCS volume resulting from SGTP. There fore, this event has no impact on the propostd TS revisions regarding RCS flow rate.

4) - The small break loss of coolant accident (LOCA) transient has been cemonstrated to be insensitive to marginal changes in RCS flow rate. The critical consideration in conservatively treating the reactor coolant pumps (RCP) for this event is the time during the event when these punps

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are tripped rather than their steady-state flow rate.

The analysis of record, therefore, rew ins valid for operation at the proposed reduced flow rate.

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The DNBR impact of reduced RCS flow rate on the minor secondary steam pipe break transient is bounded by the F.SLB event discussed above.

6)

The rupture of a main feedwater pipe (main feedline break) event can result in either RCS cooldown or heatup, depending on break size and operating conoitions at the time the break occurs.

The bounding cooldown scenario for a secondary system pipe rupture is the MSLB transient addressed above.

Regarding RCS heatup, the consequences of a feedline break upstream of the feedline check valve are bounded by the consequences of the loss of normal feedwater event. This is one of the events that has been reanalyzed by the licensee and will be discusseo below. A break downstream of the check valve, however, may also result in a loss of SG secondary side inventory and can prevent auxiliary feedwater addition to the affected generator. This scenario is not DNB-limited but, rather, is 11mited by RCS subcooling margin.

For the analysis of recoro, this margin is 36*F.

Existing sensitivity data for changes in RCS flow rate and SGTP level has indicated that a 5% reduction in RCS flow ra6e for this event results in e reduction in subcooling margin of less than 10*F.

The impact of the proposed 3% reduction in RCS flow rate, therefore, should be readily accommodated by the existiog subcooling analysis margin.

7)

For the control rod ejection event, existing sensitivity data indicates that a SY reduction in RCS flow rate results in a 36'F increase in peak cladding temperature (PCT). When this increase is applied to the most limiting value of 2575*F (for the zero power, end-of-life case), an 89'F margin to the clac embrittlement temperature of 2700*F still remains.

Therefore, the impact of the proposed 3% reduction in ficw rate on PCT should be acccmmodated by this margin.

8)

In the acalysis of the SG tube rupture event, operator action is assumed to terminate the primary-to-seconcory side mass transfer withM 30 minutes and bounding values of key parameters are assumed in the calculation of consequences. These assumptions are not impacted by a reduction in RCS flow rate and, therefore, the analysis of record remains valid.

9)

The requirements imposed by 10 CFR Part 50 Appendix K with regard to RCP for purposes of analysis of the large break LOCA event result in u relative insensitivity of the event to marginal changes in RCS flow rate.

Reduced flow rates cause a decrease in core inlet temperatures with an attendant decrease in the RCS Tavg.

Existing sensitivity data shows that, for the proposed decreese in RCS flow rate, the corresponding reduction in inlet temperatures results in an increase in PCT of approximately 2* F.

Considering the conservatisms dictated by Appendix K, this increase is deemed insignificant. Therefore, with regard to the proposed TS revisions, the analy:is of recore remains valid.

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10) Those Chapter 15 ever,ts which are unaffected by reduced RCS flow rate or other effects of SGTP have been identified as:

- Inactive loop startup frorr reduced power

- Misioaded fuel assembly

--Volume control tank rupture

- Paste gas oecay tank ruPure

- Fuel handling accident outside contair, ment

- Fuel handling accident inside containment As noted earlier, those Chapter 15 transients which are potes;tially impacted by both RCS flow rate reductions and other effects of $GTP have either been reanalyzed or evaluated using existinq event-specific sensitivity data.

The latter category of events has already been addressed above. 'ihe disposition of the reanalyzed events follows. All reanslyses were performed using the RETRAN single-and doubie-loop models and all assumptions were consistent with or conservative with respect to the assumption employed in the analyses of record. Modifications were made to the models to reflect the various effects of the extended SGTP.

1)

For the loss of external load event, the BOC "With Pressure Control" case and the BOC "Without Pressure Control" case represent the limiting cases for DNB and overpressurization, respectively. Thesa cases were reanalyzed.

Results indicated that the DNBR increased throughout the transient from an initial value of 2.15 (for the DNB-limited case). In addition, the peak RCS and secondary pressures remained below their respective acceptance criteria values for the overpressure-lihiited case.

2)

For the loss of normal feedwater event, both the case of " Loss of Offsite Power" and the case of "Offsite Power Available" we m reanalyzed (the latter being more limiting). In both cases, results indicateo that extended SGTP does not adversely impact the ability of the auxiliary feedwater system to adequately perform its safety function.

3)-

for the rod withdrawal at power event, a wide range of initial plant conditions were reanalyzed to ider,tify the most linaiting cases.

Results indicated that, in all cases, the n.inimum DNBR remained above the design limit value. Additionally, the reanalysis confirmed that the current TS setpoints for overtemperature and overpower delta-T (OT delta-T and OP delta-T) trip continue to provide bounding core thermal limit protection under extended SGTP conditions.

4)

For the complete loss of flow event, two cases were reanalyzed:

complete more limiting 5.0 Hz/sec decoy in supply frequency (ge case"), ard the loss of voltage at the RCP brehkers (the "undervolta the "underfrequency case").

Results indicate that, in both cases, the transient DNBR remained above the SDL DNBR at all times.

5)

For-the locked rotor / sheared shaft event, only the locked rotor case has been reanalyzed, since previous analyses have shown this case to be bounding. The analysis ;onsists of two parts:

(1) calculation of peak 1

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6 RCS pressure (assuming no fuel rods experience DNB), and (2) a conservative determination of the fraction of the core experiencirg DNB.

For the former case, results indicate that peak RCS pressure ren.ains well within the acceptarce limit of 2750 psia.

For the latter case, the criterion of less than 13% of fuel rods experiencing DNB at the limiting time in core life (for the current operating cycle) continues to be met.

The core therral limit lines of TS Figure 2.1-1 define the range of acceptable operating conditions which sutisfy two important limits:

DNBR and vessel exit boiling. The DNBR-limited segment of each line bounds the design DNBR limit of 1.46 and is based on a minimum neasured flow of 289,200 gpm. As discussed earlier, the generic retained margin is large enough to absorb the DNBR penalty assessed against it and to compensate for the proposed reduction in RCS flow without taking credit for the analysis margin.

Therefore, the existing DHBR portions of the limit lines continue to remain bounding. The vessel exit boiling-lin.iteo segments of the limit lines shown in Figure 2.1-1 are based on an RCS flow rate of 278,400 gpm. The proposed reduction in flow rete results in less than a 1 F reduction in subcooling margin. The licensee states that the existing vessel exit boiling limits contain sufficient margin to offset this impact and, therefore, the existing limit lines continue to remain bounding. These conclusions are based on the current TS values for OT (celta-T) and OP (delta-T).

The licensee has performed evaluations of key NSSS and BOP systems and components to confirm whether their operation under the proposed reduced flow and extended SGTP conditions remain in compliance with the applicable codes and standards.

Results of these evaluations indicate continued acceptable performance for all NSSS and 80P systems / components considered.

3.0 SUPPLEt1 ENTAL EVALUATION As SG tube inspections progressed durino the current outage, it was found that the original tube plugging projections (on which the above TS change to reduce minimum total RCS flow to 275,300 gpm was based), needed to be modified.

In particular, updated projections indicetea that the plugging level fcr SG "C" may exceed 30%. To compensate for tLe effects of this anticipated increase in tube plugging, the licensee proposed to limit the maximum reactor power level to 95% of rated thermal power until the planned 1993 SG replacement. This revision, requested by letter cated January 28, 1992, was necessary because the analysis of record for the large break LOCA event would not support 100% power operation with more than 30! SGTP.

(The proposed change and.evised large break LOCA analysis are being reviewed under separate cover). To provide the required margin, the event was reanalyzed assuming a reduced power level of 95%, a 35% level of SGTP, and an RCS flow rate of 264,400 gpm. Note, however, that no adoitional reduction in RCS flow rate (from 275,300 gpm) was requested in the January 28, 1992 submittal. Following that request, the licensee proposed an adoitional 2 1/2% reduction in minimum total RCS flow rate to 268,500 gpm to account for the uncertainties inherent in RCS flow measurement and in plugging estimates. This further revision to TS Table 3.2-1 was requested by letter cated February 10, 1992.

Following a meeting with the licensee on February 10, 1992 and several subsequent telephone conferences,

.:acitional information of a clarifying nature was submitted by letter dated February 25, 1992.

To support the additional reduction in RCS flow rate, an accompanying revision in the reactor core safety limits was requested. The revision would add a footnote to TS 2.1.1Lreferencing new Figure 2.1-la in lieu of existing Figure 2,1-1.

The r.ewly generated figure presents revised thermal limit lines which are-based or, the proposed flow rate of 26P,500 gpm and which bound the existing design DNBP limit of 1.46.

These limits would be in effect for the period

-until the SGs are replaced.-

Based on the revised core thermal limits, new setpoint values.for the OT (delta-Y) and OP (delta-T) trips and the power range neutron flux (PRNF) high trip, as specified in TS Table 2.2-1, were generated. These values would also apply until SG replacement. Specifically, the PPNF high trip st. point is decreased from 109% to 103% thermal power, while the allowable value is reduced from 110% to 104%. For the OT (delta-T) setpoint, the value of K is reduced from 1.264 10 1.132 ana f or the OP (delta-T) setpoint, the value bf K, is

. reduced from 1.079 to 1.016. The following describes the staff's evaluation of the additional TS revisions requested by the licensee's February 10, 1992 letter and the technical bases submitted in support of these revisions.

As noted in Section 2.0 above, for the group of Chapter 15 transients affected by RCS flow only, the adverse DNER impact of a reduction in RCS flow to 275,300 gpm was accounted for by assessing a 4.8% DNBR penalty against the generic retained margin of 13.7%. With the proposed further reduction in RCS flow to 268,500 gpm, it was determined that for certain events, this additional reduction could not be. compensated fo. by simply increasing the DNBR pena'ty.

Because other factors (e.g., rod howing) require additional penalties to be assessed against generic retained margin, the margin balance was insufficient.to: accommodate a greater DNBR penalty. However, the licensee's proposed TS revision to limit maximum powe to 95% of rated thermal power has a favorable impact on DNBR. Sensitivity tcudies were performed to determine the overall DNBR impact of the 5% reduction in maximum power level,-

the additional reduction in FCS flow rate to 268,500 gpm, and the 1.5%

increase in F (delta-H) associated with the 5% power reduction. To accomplish this, selected thermal hydraulic-statepoints representing normal operating conditions and_ limiting accident conditions were perturbed on these three

-variables. The. studies indicated that the favorable DNBR impact of a 5%

decrease in power more than compensates for the adverse impacts of the additional RCS flow reduction-and increase in F (delta-H). The net result was a DNBR benefit of 1.2% to 2.5%.

Therefore, the proposed reduction in flow to 268,500 gpm, in combination with the 5% reduction in power, is acceptable with respect to this group of transients.

Concerning the group of events for which reanalyses were performed and

- documented in the January 8,1992 submittal, these events were reevaluated to determine the overall impact of the additional reduction in flow from 275,300 gpm to 268,500 gpm combined with the power reduction to 95% thermal rated power. The following summarizes the results of these reevaluations.

1)

For the loss of external load event, the effect of the proposed reductions is-bounded by the DNBR and overpressurization reanalyses documented in the January 8, 1992 submittal.

8 2)

For the loss of normal feedwater event, the additional reduction in RCS flow has an insignificant adverst impact on heat removal capability.

This is more than compensated for by the proposed power reduction which reduces the initial stored energy and post trip decay heat.

Therefore.. the reanalysis of this event, as documented in the January 8, 1992 submittal, remains bounding.

3)

The DNBR sensitivity studies described above indicated a net DNBR benefit resulting from the proposed reductions in power and RCS flow for a wide range of thermal-hydraulic statepoints covering normal operation and limiting accident conditions.

For the complete loss of flow event, the impact of the proposed reductions was evaluated by examining their effect on RCS flow coastdown vs. time behavior.

The thermal hydraulic statepoint for the revised coastdown was shown to be bounded by these sensitivities. Therefore, the reanalysis of this event, as documented in the January 81992 subniittal, remains bour. ding.

4) for the locked rotor event, the proposed reductions result in a smaller rumber of rods experiencing DNBR than was the case for the reanalysis docurented in the January 8, 1992 submittal. This is attributed to the net DNBR benefit shown to exist through the above described sensitivity analyses. Therefnre, the acceptance criterion of less than 13% of fuel rods experiencing DNBR (bastd on the current offsite dose calculation) continues to be met.

For the peak RCS pressure portion of the analysis, the locked rotor still represents the dominant source of flow resistance in the loop. Therefore, the increased resistance resulting from additional SGTP has only a minor effect on the flow coastdown vs. time behavior. This is more than offset by the favorable impact of the 5%

power reduction on peak RCS pressure during the event.

5)

The roo withdrawal at power event was reanalyzed at the reduced flow rate of 268,500 gpm and the revised 01 (delta-T) and PRNF high trip setpoints me'itioned above.

In the reanalysis described in the January 8,1992 submittal, it was noted that the limiting case corresponded to initiation of the transient from hot full power. Thus, for the revised reanalysis of this event, a range of reactivity insertion rates from an initial condition of 95% rated thermal power was examined. A comparison of the

-results of the revised and original (January 8,1992) reanalysis indicates the latter is bounding.

Furthermore, the results obtained for the revised case confirm that the updated thermal limits (i.e., TS Figure 2.1-la) are not exceeded for the complete range of possible system conditions with the revised OT (delta-T) and PRHF setpoint values employed. Concerning the OP (delta-T) reactor trip setpoint, the revised value of the constant K is consistentwiththeproposec5%reductioninthePRNFhightripsetpbint.

The OP (delta-T) trip serves as a backup to the PRNF; no credit was taken for it in the Chapter 15 analyses.

For the MSLB event, as noted in Section 2.0, the adverse DNBR impact of the original proposed reduction in RCS flow rate was compensated for by assessing a DNBR penalty of 4.3% against the 10% generic retained margin. With the

9 further reduction in RCS flow to 268,500 gpm. It was necessa y to impose an additional penalty of 3.5%.

It should be noted that the RCS flow rate reduction associated with SGTP and the reduction in heat transfer area resulting from SGTP can be treated as separate effects with opposite impacts.

The latter has a favorable impact on the MSLB transient because it results in reduced SG heat removal capability which, in turn, leads to a decrease in RCS cooldown rate and associated power excursion.

This benefit was not taken credit for in the licensee's evaluation.

The licensee has reevaluated the operation of key NSSS and BOP systems and components under the further reduced RCS flow rate and reduced thermal power limit. Using, as a baseline, the Westinghouse and Stone & Webster evaluations performed fur tFe originally proposeo flow rete reduction, as dccumented in thi. January 8,1992 submittal, the licensee has concluded thtt these evaluations bobnd operation under the new conditions. Therefore, the sys. ems and components reexamined rernaio in compliance with the applicable coces and standards.

4.0

SUMMARY

On the basis of the above evaluation (Sectici. 2.0) and supplementary evaluation (Section 3.0) the stoff finds that, with regard to the proposed TS revisions, the licensee has providec adequate supporting analyses and evaluations to demonstrate the following:

1)

For all UFSAR Chapter 15 events, the proposed reductions in RCS flow rate (to 268,500 gpm) and thermal power limit (to 95% of rated value) are accommodated by current thermal margins or by the assessment of a DNBR penalty against generic retained margin. All acceptance criteria continue to be met.

2)

The revised values of the OT (delta-T), OP (delta-T), and PRNF high trip setpoints, as well as the current values of the remaining ESF and RPS setpoints, provide adequate plant protection at the reduced RCS flow rate and power.

3)

The revised core thermal limits (Figure 2.1-la) bound NA-1 Cycle 9 operation at the reduced R"5 flow rate and power.

4)

NSSS and BOP systeme and components will continue to remain in compliance with the applicable codes and standards for operation at the reduced RCS flow rate and power.

Therefore, the staff finds the proposed TS revisions to be acceptable. With the exception of the revision to TS Table 2.2-1, which is a permanent administrative change, all other proposed TS will remain in effect for the balance of the operating period until the NA-1 SGs are replaced in 1993.

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5.0 STATE CONSULTATION

1 In accordance with thv. Comission's regulations, the Virginia State official was notified of the proposed issuance of the amendment. The State official had no comment.

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6.0 ENVIRONMENTAL CONSIDERATION

This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined thot the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative cccupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding (57 FR 2291). Accordingly, this amendment meets the eligibility criteria for categorical Oxclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safet not be endangered by operation in the proposed manner, (2) y of the public will such activities will be conducted in compliance with the Corrmission's regulations, and (3) the issuance of the amendtrent will not be inimical to the common defense and security or to the health ant safety of the public.

principal Contributor:

H. Abelson Date: March 3, 1992 l

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