ML20235G244

Steam Generator Downcomer Mods Safety Evaluation for North Anna Units 1 & 2

ML20235G244
Person / Time
Site:	North Anna
Issue date:	09/25/1987
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
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Shared Package
ML20235F948	List: ML20235F944 ML20235G244
References
NUDOCS 8709290508
Download: ML20235G244 (16)

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ATTACHMENT STEAM GENERATOR 00WNCOMER MODIFICATIONS SAFETY EVALUATION FOR NORTH ANNA UNIT NOS. 1 AND 2 l

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SAFETY EVALUATION NORTH ANNA UNITS 1 AND 2 STEAM GENERATOR MODIFICATIONS Following the recent determination that a likely cause of the July 15, 1987 tube rupture event un North Anna Unit I was vibration-induced metal fatigue, Virginia Power began investigating possible design changes which would reduce the likelihood of this phenomenon. Based on this investi-gation, it was determined that the installation of flow restricting plates near the top of the upper-to-lower downcomer transition cone provided the best option for reducing fluid selocities in the tube bundle region.

The plates are perforated and will increase resistance to flow through the annulus between the steam generator shell and the tube bundle wrapper.

This increased resistance has the effect of reducing the design recircu-16 tion ratio, reducing the mass flow through the tube bundle and in-creasing the void fraction in the bundle. Since no change in the steady state operating steam generator water level is proposed, this change also results in a reduction in the steam generator secondary side mass inven-tory at normal operating conditions due to higher average void fraction in the tube bundle region.

This evaluation assesses the impact of these modifications on the results of the analyses of postulated accidents presented in Chaptee 15 of the North Anna UFSAR. All of the accident analyses presented in Chapter 15 1

i were reviewed. For purposes of this discussion, the accidents can be divided into three categories:

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1) Events for which there is no impact because the limiting conditions or assumptions used in the analyses are unchanged by the modifications.

This category encompasses a large majority of the events. Included in this category are events which are limiting at hot zero power conditions (e.g. steam line rupture, uncontrolled rod withdrawal from a subcritical condition) where the modifications will have no effect on thermal-hydraulic conditions; and events where the limiting conditions occur so quickly (i.e. less than a reactor coolant loop transport time) that steam generator conditions have no effect on the results. Events of this latter type include control rod ejection and complete or partial loss of fortad reactor coolant flow. Other events that are not impacted by "at power" steam generator operation (e.g. waste gas decay tank rupture) also fall into this category.

There are other events (e.g. , loss of external electrical load) for which the modifications result in a minor effect in the transient reponse, but those results can be shown to remain well within the appropriate accept-ance criteria. The conclusions of the UFSAR remain unchanged for these events. All of the UFSAR Chapter 15 events except the three events dis-cussed below were found to be in this first category.

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2) Accidents which were reanalyzed due to the potential for adverse impact on the existing results, with the reanalysis results showing no increase

in the consequences of the events. There were two events in this cate-gory: loss of normal feedwater (including the case of concurrent loss of j offsite electrical power) and main feedwater line rupture. These events  ;

i were reanalyzed because the lower initial steam generator masses result-ing from the modification could affect the steam generator heat removal capability during these events.

The reanalysis of the loss of normal feedwater event was performed with Virginia Electric and Power Company's RETRAN code (Reference 1). Two i cases were examined: with and without a loss of offsite power at the time of reactor trip. The analysis assumptions were consistent with or con-servative with respect to those in the UFSAR. In addition, the assumed steam generator inventories at the initiation of the event and at the time of reactor trip were modified to conservatively reflect the impact of the higher operating void fractions discussed previously.

The analyses demonstrated that the available auxiliary feedwater flow remains adequate to prevent the water level in the steam generators from receding below the level at which adequate heat transfer is available to dissipate the stored and residual heat from the core. Specifically, the results of each case demonstrate that there is no overpressurization of the reactor coolant system or liquid relief through the pressurizer safety 1 valves. Thus there is no increase in accident consequences. The ac- l l

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ceptance criteria for this event continue to be met with existing pro-tection system setpoints and the modified steam generator downcomers.

The main feedline break event was reanalyzed by Westinghouse using the LOFTRAN (Reference 2) computer code. The limiting cases presented in the UFSAR were examined: break area of 0.717 square feet (the total area of all feedring 'J' tubes), with and without offsite power available.

The feedline breah event is assessed against the following criteria:

pressures in the reactor coolant and main steam systems shall remain below 110 percent of their respective design pressures, the core shall remain in place and geometrically intact with no loss of core cooling capability and any activity release shall be such that the calculated doses at the site boundary are well within the guidelines of 10 CFR 100.

The results of the cases examined meet these criteria. The auxiliary feedwater system capacity is adequate to remove decay heat, to prevent overpressurization of the reactor coolant system and to prevent uncover-ing of the reactor core. Further details of the analysis assumptions and response will be provided in conjunction with a forthcoming UFSAR update.

3) Accidents which were reanalyzed with a resulting increase in the con-sequences from those presented in the UFSAR. Only one event is in this 4

category, namely, the steam generator tube rupture. 9hile the steam generator modifications are expected to have an insignificant impact on the overall thermal / hydraulic response of the the plant, the offsite dose calculations were reperformed. This was done for two reasons: a) cal-culations had demonstrated that, with the revised steam generator inven-tory, the secondary side water level could drop below the top of the tubes for the first few minutes following a reactor trip for power levels in excess of about 59% of rated thermal power and b) Virginia Power experi-ence with the July 1987 event on Unit I showed that the tube break lo-cation can be fairly high in the tube bundle.

The implication of the post-trip tube uncovery for the steam generator tube rupture is that the assumed effective iodine partition factor (PF) could increase significantly from the value used in the UFSAR analysis for power levels in excess of 59%.

The existing UFSAR analysis uses a PF of 0.01 throughout the entire 30 minute interval of assumed releases from the faulted SG. This effectively assumes that the tubes remain covered with fluid. Virginia Power per-formed a conservative calculation to quantify the duration of post-trip tube uncovery associated with the reduced initial mass in the modified steam generators, and concluded that this period could last up to 9 min-utes. The offsite dose reanalysis assumed that the tubes were uncovered for the first 10 minutes of the event, with an associated PF of 1.0.

Thereafter, the tubes were assumed to be covered with a resulting PF of 5

0.01. A summary of key thermal hydraulic input to the analysis is given in. Table 1.

Three' cases were analyzed: Case 1 included the revised assumptions con-cerning SG tube uncovery, initial mass and RCS break flow, but assumed RCS activity equivalent to 1% failed fuel. This is the same as the ex-isting UFSAR analysis, with the addition of including the steam generator modification effects; Case 2 assumed that RCS and steam generator sec-ondary side activities equal the values set forth in the Technical Spec-ifications, and includes the effects of a pre-accident iodine spike. Case 3 is the same as Case 2, except with an iodine spike concurrent with the accident. The analysis was performed by Stone and Webster Engineering j Corporation.

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In addition to those assumptions governed by steam generator initial conditions, the following additional major assumptions were made in per-forming the dose calculations:

1. The guidance of NUREG-0800 (Section 15.6.3) is used.

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2. There is no failed fuel due to the tube rupture.

3. The radioactive releases from the three SGs are released )

directly to the environment. The faulted steam generator is j isolated 30 minutes after initiation of the accident. )

4. Steam dump to the main condenser is not available, i.e., offsite j

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i l 5. Concurrent iodine spike appearance rates and duration are

l. assumed which are bounding for North Anna uprated core conditions.

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Models were used which separately track the release to the environment from each source of radioactive material. This includes: the initial RCS coolant activity transferred to the faulted generator via the break, and to the unfaulted generators via primary-to-secondary leakage, the initial generator secondary coolant activity (liquid and vapor) and the iodine spike activity.

Each source is followed along its path leading to ultimate release. Separate thyroid, gamma and beta doses are calculated from these sources for the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ). The results are summarized in Table 2.

Section 15.6.3 of NUREG-0800 (NRC Standard Review Plan) provides the review criteria for tube rupture dose consequences. The doses in Table 2 from the steam generator tube rupture meet the guidelines of NUREG-0800 for the EAB and LPZ, and thus the 10 CFR 100 limits are met with considerable margin (See Table 3.) Therefore, the results of this analysis meet the acceptance criteria for the tube rupture event.

Further details of the analysis results for steam generator tube rupture will be provided as part of an upcoming regularly scheduled UFSAR update.

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10 CFR 50.59 EVALUATION This safety evaluation has described the effects of the proposed steam generator modification upon the postulated events of- UFSAR Chapter 15.

For all events investigated, it is concluded that the North Anna units will continue to meet the applicable acceptance criteria, assuming the current control and protection system setpoint values and the installa-tion of the flow resistance plates. Each event except the steam generator tube rupture has been demonstrated to meet acceptance criteria without any increase in . calculated consequences. For the tube rupture, operation with the modification above 59% rated thermal power causes the predicted offsite dose consequences to increase with respect to those reported in the - UFSAR.- On this basis, it is therefore concluded that these proposed steam generator changes create an unreviewed safety question as defined

-in 10 CFR 50.59, and the results are being submitted for NRC review and approval. The results of this evaluation can be stated as follows.

1. No increase in the probability of occurrence of an accident will result from the steam generator downcomer modification.

The plate design and installation is being performed in such a manner that there is no increased probability of accidents created. The effects of this modification upon existing accident analyses has been investigated. Each accident continues to meet its applicable acceptance criteria, but a reanalysis of the steam generator tube rupture event has 8

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resulted in a calculated offsite dose greater than currently

' reported in the UFSAR. It'has been determined that for core power greater than 59% of RATED THERMAL POWER, the consequences of the tube rupture event will exceed those of the existing analysis.

2. No new accident types or equipment malfunction scenarios will be introduced as a result of this modification. The original design of the plant included such plates in the steam-generators. Therefore, operation with the modification does not create the possibility of an accident of a different type than any evaluated previously in the UFSAR.

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3. The margin of safety is not reduced. An evaluation of UFSAR accidents has concluded that the existing analyses continue to meet their acceptance criteria for operation with the SG modification. The results of a reanalysis of the loss of feedwater and feedline break accidents have shown that the l consequences are unchanged, thus maintaining the existing margin of safety. The present mergin of safety for the steam )

generator tube rupture (as defined in the basis for Technical Specification 3/4.4.8) has also been maintained, since the revised dose results continue to meet the appropriate 10 CFR 100 limits.

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BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION It has also been determined that the proposed changes described above do not involve a significant hazards consideration as described in 10 CFR 50.92. This results of this determination can be stated as follows.

1. The changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The plate design and installation is being performed in such a manner that there is no increased probability of accidents created. The effects of this modification upon existing accident analyses has been investigated. Each accident continues to meet its applicable acceptance criteria. A reanalysis of the steam generator tube rupture event has resulted in a calculated offsite dose greater than currently reported in the UFSAR for core power greater than 59% of RATED THERMAL POWER. This increase is not significant because the revised dose remains a small fraction of the 10 CFR 100 limits and meets the guidelines of j NUREG-0800 (Section 15.6.3) )

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2. No new accident types or equipment malfunction scenarios will j be introduced as a result of this modification. The original design of the plant included such plates in the steam generators. Therefore, cperation with the modification does  !

not create the possibility of an accident of a different type than any evaluated previously in the UFSAR.  ;

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3. There is no significant reduction in th'e margin of safety.

l. An evaluation of UFSAR accidents has concluded that the

-existing analyses continue to meet their acceptance criteria for operation with the SG modification. The results of a reanalysis of the loss of feedwater and feedline break accidents have shown that the consequences are unchanged, thus maintaining the existing margin of safety. The present margin of safety for the steam generator tub'e rupture (as defined in the basis for Technical Specification 3/4.4.8) has also been maintained, since the revised dose results continue to meet the appropriate 10 CFR 100 limits.

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Table 1 North Anna Steam Generator. Tube Rupture Thermal Hydraulic Input to Dose Calculation Time. Interval Tube Status Integrated RCS Break Flow 0 - 10 min- uncovered 44000 lbm i 10 - 30 min covered 88000 lbm -l Time' Interval Steaming Mass (1bm)

(hours) Faulted SG Intact SGs 0-2 81,640 341,000 l 2-8 0 605,500 Other Input:

Initial SG mass = 99100 lbm (91600 liquid /7500 vapor) .

1 RCS coolant activity = 1% failed fuel (approx. 4.0 microcuries/ gram) f RCS coolant activity = 1.0 microcuries/ gram (Tech Spec limit)

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1; 's. t-l Table 2 North Anna Steam Generator Tube Rupture Site Boundary Doses INTEGRATED DOSE-(REM) thyroid gamma beta Case 1 - SGTR with 1% failed EAB 1.77 1.33-1* 1.64-1 fuel /no iodine spike LPZ. 6.38-2 4.77-3 5.89-3 Case 2 - SGTR with pre-accident EAB 26.7 6.67-2 5.27-2 iodine spike LPZ 0.96 2.38-3 1.89-3 1

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. Case 3 - SGTR with concurrent EAB 1.52 3.85-2 4.21-2 iodine spike LPZ 7.10-2 1.41 1.50-3 l

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• Notation: 1.33-1 = 0.133 g I

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r. t TABLE 3 NORTH ANNA STEAM GENERATOR TUBE RUPTURE COMPARISON OF CALCULATED DOSES TO LIMITS IRIEGRATED THYROID DOSE AT EAB (REM)

Calculated SRP Section 15.6.3 10 CFR 100 Result Acceptance Guideline Limit UFSAR Case 0.38 Not Addressed 300 Case 1 - SGTR with 1% failed fuel /no iodine spike 1.77 Not Addressed 300 Case 2 - SGTR with pre-accident iodine spike '/6.7 300 300 Case 3 - SGTR with concurrent iodine spike 1.52 30 300 i

INTEGRATED WHOLE BODY DOSE AT EAB (RDf)

Calculated SRP Section 15.6.3 10 CFR 100 l Result Acceptance Guideline Limit  !

UFSAR Case 0.36 Not Addressed 25 Case 1 - SGTR with 1% failed fuel /no iodine spike 0.297 Not Addressed 25 l l

Case 2 - SGTR with pre-accident I iodine spike 0.119 25 25 Case 3 - SGTR with concurrent i iodine spike 0.081 2.5 25 l

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'f REFERENCES

1. N. A. Smith, "Vepco Reactor System Transient Analysis Using the

. RETRAN Computer. code," VEP-FRD-41A, May, 1985.

' 2. T. W. T.'- Burnett, et. al,' "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary) and WCAP-7907-A (Non-Proprietary),l April 1984.

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