ML20198A553

Safety Evaluation Supporting Amend 76 to License DPR-61

ML20198A553
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	05/14/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20198A523	List: ML20198A520 ML20198A541 ML20198A553 ML20198A576
References
NUDOCS 8605210129
Download: ML20198A553 (11)

Text

.'

p uas e

$ UNITED STATES y

g NUCLEAR REGULATORY COMMISSION

5. j WASHINGTON D. C. 20555

\...../ .

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.16 TO FACILITY OPERATING LICENSE N0. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT DOCKET N0. 50-213

1.0 INTRODUCTION

On May.6, 1986, during preparation of the Haddam Neck Steam Generator Tube Inservice Inspection Final Report, Connecticut Yankee Atomic Power Company (CYAPC0) discovered that tube 37-73 of the No. 2 steam generator should have been plugged during the recently completed refueling outage. A transcription error during the evaluation of the eddy-current test data for this tube resulted in the tube at row 37, column 33 (tube 37-33) being plugged by mistake. Tube 37-73 has a small diameter pit that is 55%

of the nominal tube wall thickness. The plugging limit, as defined in Specification 4.10.1, is greater than or equal to 50% of the nominal tube wall thickness.

By letter dated May 7, 1986, CYAPC0 requested a temporary waiver of compliance from Technical Specifications 3.3.0 and 4.10.1 for tube 37-73 in the No. 2 steam generator. NRC approval of the temporary waiver was issued on May 8, 1986. Since the temporary waiver expires on May 14, 1986, CYAPC0 submitted an amendment request on an emergency basis, in accordance with recent discussions with the staff, and pursuant to 10 CFR 50.91(a)(5). The proposed license amendment would authorize operation for Cycle 14 without removing tube 37-73 in the No. 2 steam generator from service by plugging.

The staff's evaluation of the licensee's request for an emergency license amendment is provided below.

8605210129 860514 PDR ADOCK 05000213 P PDR

}

2.0 EVALUATION This evaluation is based on material provided by the licensee in its application dated May 9, 1986. The defect located on tube 37-73 was identified by eddy current testing (ECT) as being 8.5 inches above the third support plate in the hot leg. The defect depth is 55 per-cent through-wall. The amplitude of the defect signal is approximately 1.3 volts. Historical ECT data for tube 37-73 was also provided by the licensee.

Two other indications of 45 percent and 24 percent through-wall were also ,

identified by ECT at. 7.5 indhes and 10.5 inches, respectively, above the third support plate (hot leg). All three defects have been determined to be pits, based upon their signal characteristics, location and prior destructive examination results from two steam generator No. 2 hot leg tubes removed during the 1984 refueling outage.

2.1 Eddy Current Testing Accuracy i t

As part of a probe qualification program, the vendor produced and tested flawed tube standards. Flaws representative of Haddam Neck steam generator pits, in both diameter and depth (based on prior destructive examination),

were machined into representative Haddam Neck steam generator tubing. ECT of the tube standarcs was performed in the laboratory. A comparison of the actual to the ECT-estimated flaw depth showed a good correlation.

For 30 flaws of various sizes (35 percent through 100 percent through-wall, the average error in the estimated flaw depth was -3.1 percent with a standard deviation of 5.8 percent. The ECT error band was demonstrated by this test to be from an 8.9 percent through-wall undercall to a 2.7, percent through-wall overcall of the actual flaw depths, with a confidence limit of one standard deviation of 8.9 percent. The greatest difference between an actual ard ECT-estimated flaw depth was a 15 percent through-wall under-call (80 percent 0.040 inch diameter actual versus 65 percent through-wall call) on one of the 30 tube flaw depths.

One tube in row 31, column 64, of two which were removed from the steam generator No. 2 hot leg during the 1984 refueling outage, was tested twice by ECT before removal. The pits in the removed tube were located above the second tube support platc. These pits showed similar ECT characteris-tics to the flaws located above the third support plate hot leg of tube 37-73. Two pitted regions were ident#fied in the pulled tube. The maximum pit depths in each region, as measured by destructive examination (40 per-cent and 23 percent through-wall respectively). This demonstrates that the maximum estimated depths of pitted field tubes were conservatively over-estimated by ECT in this case.

Many variables affect the accuracy of measuring flaw locations by ECT (e.g., probe speed, accuracy of steam generator design measurements, etc).

Some estimated accuracy values for ECT flaw location measurements are:

1. The accuracy in measuring the height of a pit above a tube support plate, using the MIZ-18 data acquisition system, is approximately 1/2 inch. If, for example, the location calibration is not consis-tent from one analyst to another, a 1 inch accuracy may be expected.

In comparing heights from earlier inspections (prior to MIZ-18), a much greater inaccuracy can be attributed to variations in probe pull speed. These variations can result in height inaccuracies in excess of one inch.

2. The accuracy in measuring the distance between pits, which are within a 2-inch band, is approximately 1/8 inch.

3. Using the A560SF probe, signals from pits spaced at least 0.10 inch apart can be resolved.

4. Background noise was minimal as shown in the ECT trace for tube 37-73.

2.2 Estimation of Flaw Growth' Rates Progression rates 'for those flaws located above the tube support plate (ATSP) and above the tube sheet have been determined. Only those flaws greater than 20 percent through-wall, during each of the 1984 and 1986 ECT inspectionst were utilized because of the higher degree of confidence associated with these data.

Only steam generator Nos. 1 and 2 had greater than 20 percent through-wall above tube support plate flaws during both 1984 and 1986 inspections.

The calculated average progression rate for these flaws is -2 percent standard deviation of 9 percent.

Similarly, an average progression rate of -2 percent with a standard devia-tion of 8 percent was calculated for the above tube sheet flaws. In this situation, data were available from all four steam generators.

Little or no increase in the indicated progression rates for the above tube support plate and above tube sheet flaws are anticipated during the up-coming cycle of operation. Overall secondary system chemistry control is expected to be comparable to that experienced during the November 1984 through January 1986 cycle 13 operation. In fact, an apparent reduction in the rate of pitting corrosion from that predicted prior to the end of cycle 13 may be a direct result of earlier plant improvements, such as

~

partial condenser retubing and significant removal of copper alloyed - .

componants from the feedtrain. Imolementation of additional secondary water chemistry improverents could further reduce the rate of corrosion.

There are four tubes surrourding tube 37-73 that also have indications of ATSP flaws. In each case, the flaws identified were smfiler then, those present in tube 37-73.

e e

-. _ - - - - . - . - - _ _ -_- - _- - _ . . - - . _ - - ~ . _ -

5-2.3 Structural Integrity of Flawed Tubes An analysis was performed using Millstone Unit No. 2 data to derive 1

an acceptable limiting pit-like defect depth for a Haddam Neck steam l generator tube. The pit in tube 37-73 was conservatively extrapolated  !

to an end of cycle maximum depth of 83%. A limit depth for large single pits was computed based on burst test results conservatively corrected for differences between the lab test and field samples. Using the guidance of Regulatory Guide 1.121, the derived limit depth is 89%

through-wall. Since the maximun estimated end-of-cycle depth of 83%

i is less than this derived limit depth of 89%, it is acceptable.

Tests and analyses have been performed for pit-like defects in steam

generator tubing at Millstone Unit No. 2. For Millstone Unit No. 2 a single 0.187 inch diameter by 0.044 inch deep hole (92% through wall i thickness) was tested and began leaking at 6500 psi (Reference 4). The remaining material (i.e., ligament) in front of this hole was 0.004 inches thick. The necessary ligament for Haddem Neck to provide a burst pressure of 6500 psi was determined to be 0.006 inches or a pit depth of 89%

nominal tube wall thickness.

I l Regulatory Guide 1.121 provides guidance for plugging limits for steam 4

generator tubes. It states that a normal operations safety factor of

! 3 against burst pressure should be assured at the end of an inspection j interval (i.e. to allow for corrosion progression). It also states that

leak rate increases should be gradual to allow time for corrective action; and accident safety margins should be consistent with margins determined via ASME Code Section III, Subsection NB-3225.

l The Millstone Unit No. 2 tubes were evaluated against all of these criteria, 1

and the licensee found that a 92% through-wall pit is acceptable. Since the pit in question is not near a tube support, pressure constitutes the l only major load imposed upon the tube. The conservative equivalent depth f for Haddam Neck is 89%. Thus, an end-of-cycle pit with a diameter less i

9

y ._ .. _ ~ __. _ .. _ . __ _ -

i l i than 0.187 inches is acceptable per Regulatory Guide 1.121 provided it has less than a 89% through-wall depth. This limit was compared to the six criteria of Regulatory Guide 1.121, Section C.2 and found to be

acceptable.

Two other pits are in the vicinity of this one; a 45% deep pit is 1 inch I

away and a 24% pit is 2 inches away. ASME Code Section III, Subsection NB-3224-1 provides a test for interaction of local stress discontinuities

! (NB-3213.10). No interaction occurs if the discontinuities are separated

, by 2.5 8 where R is the nominal outside diameter tube radius (0.375 inches) and t is the nominal tube wall thickness (0.055 inches). The j interaction limit in this case is determined to be 0.36 inches. Therefore, the staff concludes that no interaction is expected to occur.

The actual maximum existing pit depth for tube 37-73 is 65% (55% measured depth plus 10% depth due to error in measurement). The best estimate growth rate over the last cycle was -2% 9%. Conservatively ignoring the mean reduction of 2% and allowing for twice the best estimate growth yields a maximum end-of-cycle defect depth of 83%, which allows for corrosion and inspection error.

Since the maximum estimated end-of-cycle defect depth is 83% and the allowable defect depth limit is 89%, this defect meets the provisions of Regulatory Guide 1.121 through the end of this fuel cycle.

i 2.4 Summary The staff has reviewed the test data and analyses provided by the licensee to develop a limiting pit depth for the Haddam Neck Steam generator tube and find adequate justification has been provided by the licensee to continue normal plant operation through the end of cycle 14 without plugging tube 37-73. The licensee has committed to plug this tube during the next refueling outage.

v 3.0 EMERGENCY CIRCUMSTANCES 1

On May 6,1986, during preparation of the Haddam Neck Steam Generator Tube Inservice Final Report, the licensee discovered that tube 37-73 of the No. 2 steam generator should have been plugged during the recently 1 completed refueling outage. As discussed earlier, a transcription error during the evaluation of the eddy-current test data for this tube resulted in the tube at row 37, column 33 (tube 37-33) being plugged by mistake.

A temporary waiver of compliance was issued by NRC letter dated May 8, 1986. If NRC authorization is not receiv.ed for this proposed license amendment by May 14, 1986, then the temporary waiver of compliance i will expire. CYAPC0 will isolate Loop 2 of the reactor coolant system to maintain technical specification compliance for an inoperable steam generator. The net effect of this action is to derate the plant to 65% power in three-loop operation. Long-term operation with three loops will adversely affect cycle length and very likely result in a forced shutdown to repair tube 37-73 in the No. 2 steam generator.

The staff has determined that the above circumstances constitute an emergency situation since, if no action were taken, plant operation would be derated and limited to only 65% of rated power.

l 3.1 No Significant Hazards Consideration Determination

[ In accordance with 10 CFR 50.92, the Commission may make a final

! determination that a license amendment involves no significant hazards considerations if operation of the facility in accordance with the amendment would not:

! (1) Involve a significant increase in the probability or consequences I

of an accident previously evaluated; or

[

(2) Create the possibility of a new or different kind of accident from any accident previously eyaluated; or t

L ______-______-_-__ -____- -_ - _ -________________-_ _ - --___ _ _ _ _ ___ - _

1 i  !

I l

(3) Involve a significant reduction in a margin of safety.

f i

The information in this section provides the staff's evaluation of this l license amendment against these criteria:

I A. Involve a significant increase in the probability or consequences i of an accident previously evaluated.

The accident analysis potentially affected by this proposed change is the steam generator tube rupture. The progression rate of pit defects is bounded by 18% and structural integrity is ensured if the most limiting type defects do not exceed 89% through-wall during the cycle. Since the 55% tube defect (plus a 10% allowance for

measurement uncertainty) is not expected to exceed 83% through-wall

f by the next refueling outage and pit-type defects are not expected to result in structural failure until well beyond 89% through-wall, the probability or consequence of an accident previously evaluated

. is not increased by the proposed change. i

? I I

B. Create the possibility of a new or different kind of accident  ;

from any accident previously evaluated.

Since the proposed change to the technical specification does not I affect plant operating conditions, does not affect the way in which any equipment is operated, and does not create any new failure modes, no accidents would be created of a new or different kind from any previously analyzed.

C. Involve a significant reduction in a margin of safety.

Steam generator tubes form a part of the RCS pressure boundary and are manufactured with a wall thickness much greater than the minimum wall thickness necessary to withstand the stresses due to normal operation and design basis accidents. The 50% technical specification plugging limit is chosen to ensure that, assuming a conservative

i .  ;

l i

l r

l progression rate, defects do not degrade to the minimum thickness

• before the next tube inspection. Using the guidance of Regulatory Guide 1.121, CYAPC0 calculated the allowable through-wall pit defect to be 89% through-wall degradation.

, Based on burst testing and structural analysis, a single large pit f

i up to 92% through-wall has a greater margin to burst than 50%

l through-wall uniformily thinned tubes.

4 In addition, the progression rate of defects in the Haddam Neck steam generator tubes has been shown to be conservatively bounded by 18% per cycle. As a result, the 55% through-wal.1 defect (plus a 10%

allowance for measurement uncertainty) is not expected to increase beyond a maximum of 83% through-wall before the next refueling outage.

Because of the conservative progression rate, and the structural integrity of the tube material, the margin of safety that is the

, basis of this technical specification will not be significantly reduced during one cycle of operation.

3.2 State Consultation Mr. K. McCarthy, Director, Radiation Control Unit, Department of Environmental Protection, State of Connecticut, was contacted concerning .

the waiver of compliance issued on May 8,1986. The issue of plant i

! operation with an unplugged defective steam generator tube was discussed at that time. Mr. McCarthy expressed no concern about the waiver of compliance from the standpoint of plant safety at that time. On May 14, 1986, an attempt was made to notify Mr. McCarthy of our intent to issue an emergency license amendment following the licensee's formal application.

However, he was unavailable due to official state business.

~

4.0 ENVIRONMENTALCbNSIDERATION This amendment involves a change to a requirement with respect to the installation or use of facility components located within the restricted

area as defined in 10 CFR Part 20 and changes to the surveillance requirements. The staff has determined that 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 occupational radiation exposure. The Commission has determined that this amendment involves no significant hazards consideration. Accordingly, this amendment meets the eligibility criteria for categorical exclusion 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.

5.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that: (1) the amendment does not (a) significantly increase the probability or consequences of an accident previously evaluated, (b) increase the possibility of a new or different kind of accident from any previously evaluated or (c) significantly reduce a safety margin and, therefore, the amendment does not involve significant hazards consideration; (2) there is reasonable assurarice that the health and safety of the public will not be endangered by operation in the proposed manner; and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and the security or to the health and safety of the public.

6.0 REFERENCES

1. J. F. Opeka (CYAPC0) letter to C. I Grimes (NRC), dated May 9,1986

Subject:

Proposed Revision to Technical Specifications.

2. J. F. Opeka (CYAPCO) letter to C. I. Grimes (NRC), dated May 7,1986,

Subject:

Temporary Waiver of Compliance From Technical Specification Limiting Condition for Operation (LCO).

l .

i .

l . .

i I

4

! I j 3. C. I. Grimes (NRC), letter to J. F. Opeka, (CYAPCO), dated May 8,1986,  !

e i

Subject:

Emergency Waiver of Compliance from Technical Specifications  ;

1 j 4.1.0.1.D.2 and 3.3.D. ,

i I

4. Detailed information regarding this test and related information was i provided in Northeast Nuclear Energy Company's letter of June 3, 1983,  !

l as supplemented on August 18, 1983 and November 17, 1983, on Docket No. l f 50-336. This information formed part of the basis of License Amend- f I ment No. 89, dated December 30, 1983, and the supporting Safety i 1

i Evaluation authorizing sleeving at Millstone Unit No. 2. l 7.0 ACKNOWLEDGEMENT i

! i

l. This Safety Evaluation has been prepared by J. Rajan and F. Akstulewicz.

h l Dated: May 14, 1986 l

f i

l I

- t

_ _ _ _ _ _ _ . _ _ _ _ . _ . . _ . _ . . _ . _ _ _ . _ _ _ _ . . _ . _ _ . _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ ________ -_ _ _ .