ML20236F497

Safety Evaluation Re Util 870925 Request for Approval of Steam Generator Downcomer Flow Resistance Plate Mods & Increase in Calculated Offsite Dose Per 10CFR50.59.Mods Acceptable

ML20236F497
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Site:	North Anna
Issue date:	10/23/1987
From:	Office of Nuclear Reactor Regulation
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ML20236F484	List: ML20236F480 ML20236F497
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NUDOCS 8711020165
Download: ML20236F497 (5)

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1. '% UNITED STATES j' g: NUCLEAR REGULATORY COMMISSION

% :j . WASHINGTON, D. C. 20555

\...../ f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO OPERATION AT GREATER THAN 50% POWER FACILITY OPERATING LICENSE NO. NPF-4 AND NO. NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATION, UNITS NO. 1 AND 2 DOCKET NOS. 50-338 AND 50-339 INTRODUCTION By letter dated September 15, 1987, the Virginia Electric and, Power Company (the licensee):provided the results of its evaluation of the North Anna Power Station, Unit No. 1 (NA-1) Steam Generator Tube Rupture (SGTR) event which occurred on July 15, 1987.- To preclude a similar event from recurring, the

' licensee has taken corrective actions, which include the installation.of downcomer flow resistance plates (DFRP) to the SGs for NA-1&2.

By letter dated September 25, 1987, the licensee provided a Safety Evaluation (SE) to. assess the impact of the DFRP modification on the NA-182 Updated Final Safety Analysis Report (UFSAR) Chapter 15 accident analyses. The result of the licensee's SE concluded that the DFRP modifications for the SGTR event resulted in a calculated offsite dose greater than originally reported in the NA-182 UFSAR.

Therefore, the licensee's September 25, 1987 letter with the enclosed SE requested NRC review and approval of the SG DFRP modifications with respect to the SGTR accident analysis and the increase in calculated offsite dose as defined in 10 CFR Part 50.59.

I Our discussion and evaluation of these matters are presented below. l DISCUSSION l' Following the licensee's recent determination that the most likely cause of the July 15, 1987 SGTR event at NA-1 was vibration-induced metal fatigue, the ]

' licensee began investigating possible design changes which would reduce the l likelihood of this phenomenon. Based on this investigation, it was determined J that the installation of DFRPs near the top of the upper-to-lower downcomer I transition cone of the SGs would provide the best option for reducing fluid velocities in the tube bundle region.

These plates are perforated and will increase resistance to flow through the annulus between the SG shell and the tube bundle wrapper. This increased resistance has the effect of reducing the design recirculation ratio, reducing 8711020165 DR 871023 ADOCK 05000338 PDR

l the mass flow through the tube bundle and increasing the void fraction in the bundle. Since no change in the steady-state operating SG water 1gyel is proposed, this change also results in a reduction in the SG secon@ry side mass inventory at normal operating conditions due to higher average void fraction in the tube bundle region.

The licensee's evaluation assessed the impact of these modifications on the analyses of the postulated accidents presented in Chapter 15 of the NA-1&2 UFSAR. With the exception of the SGTR event, all accidents or events were no' impacted because the limiting conditions or assumptions used in the analyses were unchanged by the modifications or the conclusions as stated in the NA-1&2 UFSAR and were within the guidelines of 10 CFR Part 100.

The one accident reanalyzed with a resulting increase in the consequences from those presented in the UFSAR was the SGTP accident for power levels in excess of 59 percent of rated thermal power. While the SG modifications are expected to have an insignificant impact on the overall thermal / hydraulic response of the plant, the offsite dose calculations were reperformed. This was done for i two reasons: (1) calculations had demonstrated that, with the revised SG j inventory, 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 (2) the licensee's experience  !

with the July 1987 NA-1 SGTR event had shown that the tube break location can  !

be as high as the seventh support plate in the tube bundle. .

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

The present NA-1&2 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. The licensee perfonned a L conservative calculation to quantify the duration of post-trip tube uncovery j associated with the reduced initial mass in the modified SGs and concluded  !

that this period could last up to 9 minutes. 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 0.01. l Three cases were analyzed by the licensee: Case 1 included the revised I assumptions concerning SG tube uncovery, initial mass and Reactor Coolant System (RCS) break flow, but assumed RCS activity equivalent to 1% failed fuel. This is the same as the existing NA-1&2 UFSAR analysis, with the l addition of including the SG DFRP modifications. Case 2 (the most limiting case) assumed that RCS and SG secondary side activities equal the values set forth in the NA-182 Technical Specifications (TS), 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.

In addition to those assumptions governed by SG initial conditions, the following additional major assumptions were made in performing thydose calcu-lations: (1) The guidance of NUREG-0800 (Section 15.6.3) is used, (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 SG is isolated 30 minutes after initiation of the accident, (4) Steam dump to the main condenser is not available, i.e., offsite power is unavailable, and (5) Con-current iodine spike appearance rates and duration are assumed which are

• l bounding for the NA-1&2 uprated core conditions.

l Models were used which separately track the release to the environment from l each source of radioactive material. This includes: the initial RCS coolant I activity transferred to the faulted SG by way of the break and to the unfaulted SGs by way of primary-to-secondary leakage, the initial SG secondary coolant activity (liquid and vapor) and the iodine spike activity. Each source was followed along its path leading to ultimate release. Separate thyroid, gama and beta doses were calculated from these sources for the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ). The results of the licensee's calculations are summarized in the tables below.

NORTyANNASTEAMGENERATORTUBERUPTURE (BASED ON OFRP MODIFICATIONS)

COMPARISON OF CALCULATED DOSE TO LIMITS INTEGRATED 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 iodinespike(most limiting) 26.7 300 300 Case 3 - SGTP with concurrent iodine spike 1.52 30 300

u INTEGRATED'WHOLE BODY DOSE'AT EAP (REM) f Calculated .SRP Section 15.6.3 10 CFR 100 Result- Acceptance Guidelines- Limit UFSAR Case 0.36 Not Addressed ~ 25 Case 1 - SGTR with 1% failed fuel /no iodine spike 'O.297 Not Addressed 25 Case 2 - SGTR with pre-accident Iodinespike(most i limiting)

O.119 25 25

. Case 3 - SGTR with concurrent i iodine spike 0.081 2.5 25  :

EVALUATION- l

.The staff has analyzed the SGTR event with DFRP modifications in accordance with Standard Review Plan Section 15.6.3. The pre-accident iodine spike case is most limiting; but even in this case the thyroid dose at the EAB is less than 30 rems which is.well below the acceptance criterion of 300 rems. In addition, the whole-body dose and the LPZ doses are less than 1 rem. The staff concludes that the acceptance criteria of NUREG-0800 and the requirements of 10 CFR Part 100 are met. Therefore, the SG DFRP design changes are acceptable.

CONCLUSIONS-The SG DFRP design and installation have been performed in such_a manner th'at there is no increase in the probability of accidents. The effects of this modification upon existing accident analyses have been investigated. Each accident continues to meet its applicable acceptance criteria. A reanalysis of..the SGTR accident analysis has resulted in a calculated offsite dose greater than currently reported in the NA-182. 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 Part 100 limits and meets the'guidelinesofNUREG-0800(Section15.6.3).

No new accident types or equipment malfunction scenarios will be introduced as -

. 'a result of the DFRP modification. The original design of the plant included

-such plates in the SGs. Therefore, operation with this modification does.not create the possibility of an accident of a different type from any evaluated l previously in the NA-182 UFSAR.

- There is no significant reduction in the margin of safety. An evaluation of NA-1&2 UFSAR accidents has concluded that the existing analyses continue to j meet their acceptance criteria for operation with the SG modification. The

present mar TS' 3/4.4.8)has gin also of safety been fo, the SGTR maintained, (asthe since defined in dose revised the basis resulfs forremain the NA-1&P a small fraction of the appropriate 10 CFR Part 100 limits. T Therefore, based en all of the above, we find the NA-1&P SG DFRP modifications to be acceptable for operation at all power levels.

Finally, the staff's evaluation and conclusions, as stated above, will be

. included on the staff's SE of the SGTR event for operation of NA-1 at power levels greater than 50% of rated thermal power.

Dated: October 23, 1987 Principal Contributors:  !

C. Willis "

L. Engle

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