ML20141M344
| ML20141M344 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 07/31/1992 |
| From: | Opeka J CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20141M345 | List: |
| References | |
| B14087, NUDOCS 9208110370 | |
| Download: ML20141M344 (30) | |
Text
.
l NORTHEAST UTILITIES o.nor.i ori.c... seioen sir i. Bernn. connecticut I
fOaNIn#
LL==:===
<20ai ees-sooo
- n. comteTicuT osmm70 t
t July 31,_1992 Docket No. 50-213 B14087.
Re:
U.S. Nuclear Regulatory Commission Attention: Document Control Desk l
Washington, DC 20555 Gentlemen:
Haddam Neck Plant i
Proposed Revision to Technical Specifications S,tfam Generator Rooair Criteria Pursuant to 10CFR50.90, Connecticut Yankee Atomic Power Company. (CYA%',/ hereby -
proposes to at.;end Facility Operating License DPR-61 by incorporating tW changes identified in Attachment 2 into the Technical Specifications of the Haddam Neck 4
Plant.
4 INTRODUCTION t
The reason for the proposed change is to reduce unnecessary -plugging of the Haddam Neck Plant steam generator tubes in the tubesheet expansion zone (EZ) roll transition area.
This. change will have the beneficial -effect of minimizing' further reductions in thermal hydraulic operating margins'and plant efficiency.
In - addition, radiation exposure,. cost, andc outage _ time. associated with the.
plugging of these tubes will be reduced.- The proposed. change would also allow -
the unplugging and returning to _ service of tubes which'were:previously plugged due to cracks in the expansion region, with its associated-improvement in the thermal hydraulic operating margins and plant; efficiency.
' BACKGROUND
_ Primary water stress corrosion cracking (PWSCC)-'of alloy 600 steam generator tubes has been diagnosed in the ' tubesheet EZ roll-transition area of steam generator tubes in many pressurized water reactor plants.
When existing tube-repair (plugging / sleeving) limits based -on crack depth are applied to axial cracks, many tubes may require repair which is unnecessary from either a safety-or reliability standpoint. Allowing tunes with-PWSCC to remain.-in service can.
be justified based on a combination of enhanced in-service. inspection,:a repair -
limit based on crack length rather than crack depth, and a limit on the number of tubes with characterized cracks retained in service.
100rs 0v 9208110370 920731
/
PDR ADOCK 05000213 g
osam nev.ne
U.S. Nuclear Regulatory Commission 4
B14087/Page 2 July 31, 1992 The combination of augmented insper. ions and an acceptable crack length-based repair limit provides an alternativ, to the current 50 percent depth-based repair criterion while maintaining safey and acceptable reliability for the Haddam Neck steam generators which have experienced PWSCC. The criteria used in the proposed technical specification changes are in addition to the degradation specific evaluation criteria in the existing technical specifications which have provided significant operational benefit while maintaining adequate safety margins.
The length-based repair limit for PWSCC induced axial cracks described herein was adopted from the generic criteria that was developed for steam generator tubes with full or partial depth for roll EZs. For tubes with partial capth roll EZs, as in the Haddam Neck Plant steam generators, confinement of the cracked section of the tube within the tubesheet precludes tube burst at the flaw location.
In this case, the repair limit may be based on maintaining acceptable leakage limits at faulted load for the distribution of cracked tubes which are retained in service.
This proposed change applies to axial or inclined axial cracks where the axial length of the inclined crack is greater than the circumferential length. Tubes with identified distinct circumferential cracks will be repaired. In addition, multiple axial cracks will be limited such that the length of possible undetected circumferential cracks remains acceptable.
The elements of the proposed approach include:
At-each scheduled inspection outage, performance of eddy-current inspections of 100 percent of the tube roll is in regions of the steam generator where the tube roll-expansion zones are susceptible to PWSCC.
l Repair of tubes with axial cracks longer than a -conservatively l
established repair limit which includes the following elements:
Use of a tube rupture curve bounded by tube burst test correlations for cracks located above the top of the tubesheet, Application of Regulatory Guide 1.121' safety factors,-
Use of lower bound tube material properties, Correction for tubesheet constraint,.
Allowance for crack growth between ' inspections, Allowance for cddy-current crack lcngth measurement uncertainty, and Restriction on spacing between adjacent ax ul cracks.
b
U.S. Nuclear Regulatory Commission B14087/Page 3 July 31, 1992 Calculation of the potential theoretical leak rate expected during postulated accident loads from the cracked tubes that remain in service.
i Calculation of maximum allowable site-specific le k rate at postulated faulted load to ensure that a small fraction of 10CFR100' dose limits are not exceeded.
Implementation of these elements constitutes a defense-in-depth approach that was developed to ensure adequate levels of safety.
The inspection scope and procedures, crack length and distribution limits, and the leak rate limits developed for tubes with PWSCC ensure adequate margins against failure and ensure calculated 10CFR100 doses within acceptable levels.
4 The benefits of this change are reduction of occupational radiation exposure due to tube plugging; minimizing further reductions in thermal margins due to plugging of steam generator' tubes; minimizing the reduction in plant efficiency due to tube plugging; and reduced ' cost and outage durations.
The Advisory Committee on Reactor Safeguards (ACRS) has previously reviewed the conceptual industry apg)oach that justifies keeping tubes with certain throughwall cracks in service.
They concluded that:
"The-continued use of the 40 percent depth limit as a repair limit results in a large effort by the licensees and a significant exposure to workers,'gnd leads to the repair of many tubes that have
)
a negligible risk of failure."
This industry approach was subsequently documented in detail,g) and is the basis for the approach provided in this request.
DISCUSSIOJ A.
Existin, Technical Soecifications The _ existing technical specifications prescribe plugging limits.
One-portion of the plugging limits permit imperfections t>elow the tppermost one inch of sound roll.
Imperfections above the uppermost one inch of sound roll within the tubesheet have a plugging limit criterion equal to 50 percent of the thickness -nominal tube wall.
The-technical specifications also set -forth tube sample selection - criteria for inspections.
(1)
Meeting Among ACRS, Nn :,, EPRI, M, Laborela, and Alabama Power Co.,
Bethesda, November 6, 1991.
(2)
Ward, David (Chairman), " Steam Generator Tube Repair Limits,"' Letter to Selin, Ivan, November 15, 1991.
l (3)
EPRI, "PWR SG Tube Rtpair Limits for PWSC, in Roll Transitions," Rev.1, Np6864-L, December, 1991.
U.S. Nuclear Regulatory Commission B14087/Page 4 July 31, 1992 B.
Descript. ion of Prooosed Chanae The proposed change modifies the requirements for repairing or removing from service defective steam generator tubes. The proposed changes would allow tubes with certain well characterized axial cracks to remain in service. The allowable cracks are limited in length and circumferential extent, but are conservatively assumed to be through the tube wall. Tubes with cracks characterized in this fashion have been shown to be structurally sound and exhibit acceptable levels of leakage with the required safety margins. The number of tubes with cracks to be retained in service is limited by the primary-to-secondary leakage limits. During ncrmal operation, leakage from all steam generator tuber, including any tubes with cracks retained in service, must remain below the specified operational limits. The calculated theoretical leakage from these cracks under postulated accident conditions is limited such that the dose contribution from the aggregate leakage will remain a small fraction of the 10CFR100 dose guidelire values.
SAFETY ASSESSMENT CYAPC0 has reviewed the implementation of this revised steam generator tube plugging limit as a proposed design change pursuant to 10CFR50.59, to assess the impact on the accidents evaluated as part of the design basis, the potential for creation of a new unanalyzed event, and the impact on the margin cf safety.
CYAPC0 has determined that the proposed design change constitutes an unreviewed safety question due to an increase in postaccident radiological consequences, and is considered a malfunction of a different type than previously evaluated. For this reason, CYAPC0 is pursuing prior NRC Staff approval of this design change via the license amendment request process. CYAPC0 has determined the proposed-design change to be acceptable and safe.
CYAPC0 has reviewed -the proposed technical specification changes in accordance with 10CFR50.92 and has determined that the changes do involve a significant hazards consideration. Along with this amendment request, supporting documentation is provided as follows: provides the safety evaluation for the proposed design changes.
Attachment 2
forwards the revised pages of the technical specifications.
provides the significant hazards consideration determination.
The safety evaluation concludes that a plant-specific application of generic alternate repair criteria provides the same margin of safety for steam generator tube structural design with-respect to bursts. The plant-specific application of the generic alternate repair criteria provides a calculated theoretical leakage rate for steam generator tubes having cracks that are permitted to remain
'in service.
The accident which is affected by this proposed change is a main
U.S. Nuclear Regulatory Commission B14087/Page 5 July 31, 1992 steam line break (MSLB). When the calculated theoretical leakage rate is used to evaluate the dose consequences of a MSLB, dop)e consequences are higher than those calculated previously by the NRC Staff (
in accordance witu Standard Review Plan 15.1.5, Appendix A methodology.
This n.ethodology is overly conservative and inconsistent with the original dose calculation methodology associated with the initial licensing of the Haddam i:eck Plant. In addition, the results of recent research indicates that the SRP 15.1.5, Appendix A methodology may contain overly conservative assumptions. CYAPC0 has revised the radiological assessment methodology accordingly.
Using this revised dose ' calculation
^
methodology and the calculated theoretical leakage rates, dose contributions from the aggree". cube leakage will be limited to a small fraction of 10CFR100 dose guideline values.
Further details are provided in Attachment 1.
Hence, these proposed technical specification changes are considered safe.
Summary Based upon the information contained in this submittal and the environmental assessment for the Haddam Neck Plant, there are no unacceptable radiological or nonradiological impacts associated with the proposed change and the proposed license amendment will not have a significant effect on the quality of the human environment.
The CYAPC0 Nuclear Review Board and the Plant Operations Review Committee hcVe reviewed and approved the proposed license amendment and have concurred with the above determination.
Regarding our schedule for this proposed amendment, CYAPC0 intends to fully implement the license amendment within 30 days of its. issuance by tne NRC.
CYAPC0 desires to implement these changes during the spring 1993 refueling outage. Hence, NRC action on this proposal is requested by April 1993.
The Haddam Neck Final Safety Analysis Report (FSAR) will.be revised within six months once the Staff issues the license amendment.
This revised-FSAR will reflect the results of the analyses performed for the proposed design basis changes and the technical specification changes.
4 i
(4)
Dennis Crutchfield letter to W.
G.
Counsil, "SEP Topic XV-2," dated November 9, 1982.
.s l-U.S. Nuclear Rcgulatory Commission 4
B14087/Page 6 July 31, 1992 I
j.-
In accordance with 10CFR50.91(b), we are providing the State of Connecticut with a copy of this proposed amendment.
l Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY John ka _ ()
i Execut Vice President-l Attachment i
cc:
T. T. Martin, Region I Administrator A. B. Wang,- NRC Project Manager, Haddam Neck Plant W. J. Raymond, Senior Resident Inspector, Haddam Neck Plant i
Mr. Kevin McCarthy Director,-Radiation Control Unit-Department of Environmental Protection Hartford, Cl 06116 STATE OF CONNECTICUT)
-) ss. Berlin COUNTY OF HARTFORD )
Then personally appeared before me,-J. F. 0peka, who being duly sworn, did state that he is Executive Vice President of Connecticut Yankee Atomic Power Company, j
a Licensee herein, that ;he is -authorized: to execute and: file the foregoing information in the name and on behalf. of the Licensee. herein,. and that the statements contained in said information are true and correct to the best of his
' knowledge and belief.
dust k t;bt.4JV Notary Public lhyletMiss104 $Vpirec:JlBll1la i
i i
i
a l
1 Docket'No. 50-213 B14087 t
e h
E i
4 5
T 4
Haddam Neck Plant Safety Evaluation r
4 5
i 8
i u
i 4
2 July 1992 1
U.S. Nuclear Regulat. Commission /B14087/Page 1 July 31, 1992 Proposed Modifications to Steam Generator Tube Repair Criteria This Safety Evaluation evaluates the impact of allowing cracked steam generator tubes to remain in service. The cracks would be limited in length and the total number of tubes with cracks retained in service are limited such that the dose contributions from the theoretically calculated aggregate tube leakage will be limited to a se?ll fraction of 10CFR100 dose guideline values in the event of a steam line break.
A maximum theoretically calculated post steam line break primary-to-secondary leakage of 100 gpm is used in this safety evaluation.
I.
Description of the Chance The,. posed design change modifies section 3/4.4.5 of the Technical Specifications.
Details of the-change are contained in Attachment 2.
Basically, the change would allow some steam generator tubes that may leak following a postulated steam line break (SLB) to remain in service.
The number of tubes allowed to remain in service would be limited such that the dose contribution from the aggregate tube leakage will be limited to a small fraction of 10CFR100 dose guideline values in the event of a steam line break.
II.
Discussion A.
Backaround The proposed alternate repair criteria would allow tubes with postulated throughwall cracks 11the Haddam Neck steam generator to remain in service without repair.
The cracks.are located in the roll expansion region, which is defined as the_ bottom 5 inches of the tube, approximately 18 inches from the secondary side of the tubesheet.
The c,teria defines an allowable axial crack-length such that the structural integrity of the tubs is assured during normal operation and postulated accident conditions as required by Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes" August 1976. The basis for the proposed allowable length is_the Haddam Neck plant-specific application of a generic alternate repair _ criteria developed by a committee of industry experts under the cognizance of EPRI (Reference [1]).
The generic alternate repair criteria was-initially developed for steam generator tubes which are fully expanded through the tubesheet with cracks located at or above the top of the tubesheet in the expansion transition region. The length limitations of the generic criteria are to preclude burst of the cracked tubes with the required safety margins. -The length limitation includes allowances for inspection measurement uncertainty and potential crack growth based on a 95% confidence level of the composite industry data base.
U.S. Nuclear Regulatory Commission /B14087/Page 2 July 31,1992 As indicated in Reference (1), app'ication of the same length limitations to the Haddam Neck steam generator tubes, which are cnly partially expanded through the tubesheet, is extremely conservative.
The tubesheet completely surrounds the tube in the region where the cracks are located. Depending on the length of axial cracks assumed to be present, the tube can either withstand a pressure load of three times the normal operating pressure on its own or will expand and come in contact with the tubesheet-prior to reaching the pressure which would cause burst if the crack were in the free span of the tube.
Thus, burst of the tube within the tube-sheet is precluded regardless of the length of the axial crack assumed to be present in the region where the alternate repair criteria is applicable.
B.
Structural Consideration Primary water, tress corrosion cracks (PWSC) have been identiried in steam generator tubes at the Haddam Neck Plant.
The cracks are located in the roll expansion region of the tube.
- The roll expansion region consists of the expandea portion of the tube plus the transition zone between the expanded and unexpanded portions of the tube, Figure 1.
Since the tubes in. the Haddam Neck steam generators are expaaded only a small porticn of the length within the tubesheet, any cracks which develop in the roll expansion region are totally contained within the tubesheet. For the purpose of this analysis, the bottom 5 inches of the tube will be considered the roll expansion region and cracks outside of this arca will not be addressed.
Steam generator tubes with cracks which are retained in service must -
remain structurally sound over the next operating cycle under normal operating and postulated accident conditions.
Structurally sound means that the tube will not burst or separate and pull out of the tubesheet with the required safety margins.
Use of the generic-repair criteria ensures that crack tubes retained in-service will remain structurally sound.
(1)
Structural limits Based on Generic Criteria A generic alternate repair criteria has been developed to address cracks in the expansion region of steam generator tubes, Reference [1].
The structural basis of the generic criteria is an experimentally determined burst curve for axial cracks of various lengths. The generic criteria assumes that the tubes are fully expanded through the tubesheet and the cracks are located in the expansion transition region at or above the top of the tubesheet.
Cracks are limited in axial length and circumferential extent under these conditions in order to preclude burst of the tube with the required safety
U.S. Nuclear Regulatory Commission /B14087/Page 3 July 31,1992 margins. Tube plugging limits developed using this criteria are extremely conservative for tubes which are only partially expanded through the tubesheet as in the case of the Haddam Neck Plant, since tube burst is prevented by the presence of the tubesheet. The axial limiting crack length is determined for the Haddam Neck Plant using the following equation from -
Reference [1):
A : = a +a,, - a, - aa where: A - allowable axial crack length a - tube rupture equation reference crack length e
a,, - correction for tube sheet constraint a, - allowance for crack growth c
amu - allowance for NDE uncertainty For the Haddam Neck Plant, the relevant input parameters are listed ir. Table 1 along with the tabulation of the results.
For the Haddam Neck Plant, the limiting crack length is 0.45 inches.
Hence, tubes with axial cracks less than or equal to 0.45 inches in length in the expansion region meet the applicable regulatory requirements and are safe to be 'matatained in operation.
(2)
Other S'.ructural Considerations j
Additional plant specific testing and anaiysis was pc-formed to ensure the structural acceptability'of retaining cracked-tubes in service in the Haddam Nack Plant. Table 2 shows-the tube loads used in this testing and ~ analysis. Burst of a tube with axially oriented cracks of any length in the defined roll-expansion region cannot occur due to the-physical. presence of the tubesheet which surrounds the tube.. ;An increaseE in internal pressure would cause the tube to bulge outward and would be restrained by the tubesheet. The maximum expansion of the tube is limited to the clearance between the tubesheet hole and the tubes or 0.014 inch nominal clearance on the diameter. A worst case stack-up of tolerances would add 0.010 inch to the nominal clearance for a maximum of 0.024 inch clearance on the diameter or 0.012 inch on the radius. Since the thickness of the tube wall is 0.055
.0055 inch, a flap caused by cracks in the tubes could be-postulated to expand radially to the tubesheet-contact point while still overlapping the thickness of the tube by 0.043 inch. Leakage through-this opening would be greatly-inhibited and would be orders of magnitude less than the classic fishmouth type rupture which could occur if the tubesheet was not present.
)
U.S. Nuclear Regulatory Comission /814087/Page 4 July 31,1992 Since tube burst is precluded by the presence or the tubesheet, structural integrity of the tube is ensured provided the cracks cannot result in physical separation of the tube and pull out from the tubesheet hole.
A series of mechanical _ pull tests specifically for the Haddam Neck Plant (Reference [7]) were performed to determine the structural strength of SG tubes with cracks present.
Test specimens were manufactured using prototypical 0.75 00, 0.055 inch wall, Inconel Alloy 600 tubing. rolled into carbon steel collars. Slots were electric discharge machined (EDM) through the tube wall at various distances below the roll expansion transition to simulate throughwall stress corrodon cracks.
Specimens were manufactured with either 7 or 15 slots inclined at an angle of 15, 30, or 45 degrees to the longitudinal axis of the tube. In addition, specimens were manufacturd with 36 slots inclined at a 30* angle to the longitudinal axis.
Thirty-six was the maximum number of slots which could be machined using the available equipmen;..
The length of the EDM slots was 0.5 inch, which was considered larger than the average length of eddy current test (ECT) indications present in the SG tube roll expansions. The width of the machined slots was 7 mils.
This width is much wider than the width of stress corrosion cracks and while structural-tests would produce a prototypical to conservative result, leakage tests would be expected to be overly conservative.
Complete specimens were heated to 1100*F for-four hours, to simulate the effects of the highest possible temperature the tubes could have experienced during SG assembly or subsequent operation. The mechanical pull tests were then performed at room temperature.
The results were adjusted to account for-the decrease in material strength at operating temperature and the minimum material properties of actual Haddam Neck Plant tubing based on a 90/95 lower tolerance limit of the tubing certification values.
Some specimens were tested without collars in order to determine the effect of cracks present in the free span of the tube.
The test fixture which held the specimens was specifically constructed to allow the tube to rotate when an axial load was applied to the tube. This decision was made after an analysis was performed which determined that allowing the tube to rotate would result in a lower, and hence more conservative, failure load. The tube has-a. tendency to rotate when a load is applied due to the moment generated by the. slots machined at an angle to the longitudinal axis.
This moment acts to 1
=
U.S. Nuclear Regulatory Commission /B14087/Page 5 July 31, 1992
" straighten" the array of slots and in the process causes plastic deformation at the tip of the slot.
If rotation is prevented, the portion of the axial load causing the rotation r> ' be absorbed by the test fixture. Since the test fixture w..i now be absorbing a portion of the load placed on the specimen, the observed specimen failure load would be higher and nonconservative. In the actual steam generator, friction between the tube and the tubesheet would absorb a portion of the axial and ratational loads placed on the tube.
- However, this effect was neglected for conservatism.
The results of the mechanical pull tests indicated that the load bearing capability of degraded tubes was strongly dependent on the angle of the cracks from the longitudinal i
axis.
The number of cracks present and the location of the cracks within the roll expansion had little effect on the load bearing capability of the tube.
As expected, nonco' tared specimens failed at a slightly lower load than coliared specimens, since the collar would tend to inhibit deformation of the tube.
In all cases, the test failure load was significantly abcVe the limiting case load of three times normal operation.
Following the machap.al testing, an analytical model was constructed to diow tne formulation of design curves for various crack lengths and orientations.
The model bounds crack lengths up to 1.5 inches and crack angles up to 60' from the longitudinal axis. The model conservatively predicts the results of the mechanical testing.
The design curves constructed from the analytical model are shown in Figure 2.
The conclusions of this anal.ms ano teding are that tubes with arrays of cracks less than i.5 inches in length and 60' in angle from the longitudinal axis ut all the requirements of Regulatory Guide 1.121, and art safe to be maintained in cperation.
Additional experimental tests were performad with tubes with circumferential cracks.
Experimental data (Reference (2])
indicatos that the extent of circumferenMal cracking needed to produce axial separation of the tube is very'large (i.e.,
approximately a through wall crack extending 250' around the circumference, or a part-through wall crack 80% of the wall thickness and 360' around the circumference). ' These crack sizes are readily detectable by standard 'lDE methods for tube inspection. By limiting the maximum allowable circumferential crack length, which theoretically could be present between axial cracks and remain undetected, to less than 1 inch, structural integrity of the tube is ensured.
U.S. Nuclear Regulatory Commission /B14087/Page 6 July 31, 1992 l
C.
Leakaae Considerations The cracks in the ex ansion region of the tube to be retained in service are assurrN to be through the tube wall and as such could potentially leak primary water to the secondary side of the SG. The level of lettage under nt,rmal operating conditions is limited to the current Technical Specification requirement of 150 gallons per day (gpd).
Under postulated accident conditions, cracks could potentially open up and leak at an increased rate.
The number of tubes with cracks retained in service will be limited such that the calculater' theoretical leak rates will remain acceptable.
A main steam lint reak is the limiting accident considered in this case since it results in the highest possible pressure differential between the primary and secondary sides and provides a leak path outsr.de of containment.
(1)
Haddam Neck Plant Specific Aeolication of Generic Methodolocy for Determinina (33Jgtqg Theoretical leakage under accident conditions through cracks of a given size are calculated using the methodology prr tided in Reference [1)-adcpted to plant specific conditions.
Cold len cor.ditians are assui;.ed since these result in slightly higher leakage rates.
Table 3 provides a list of the parameters used.
As per Refarence [1], the leakage rate from an axial crack in a table with internal pressure is given by the following formula.
O=KbfEF[ji where 3
Q is leak rate (in /sec)
K isthedischargecpefficient(dimensionless) 6 is leakage area (in )
-is effective differentia)) pressure (psi)
P is liquid density (lb/in p
The leakage area 's obtained by the summation of an elastic and plastic co@onent-of the crack opening area during steam line break conditions.
The discharge coefficient will increase with crack size.
Experimental data is used to-determine the discharge coefficient as a function of crack size in Reference.[1].
For the Haddam Neck Plant, the l
_ ~., -
g
..-y-
.y
U.S. Nuclear Regulatory Corrnission /B14087/Page 7 July 31, 1992 discharge coefficient has been benchmarked against test data as described below.
For the Haddam Neck Plant, Table 4 p ovides a summary of the elastic and plastic component; of
.ae cracx opening area and the resulting leak rate for crack lengths from 0.05 to 0.5 inches.
(2)
Qualification of Methodoloav for Determinina leakaae As en additional check of the Reference [1] methodology for determining leakage,1eakage levels were determined for ten tubes removed from the McGuire 1 steam generators and compared eith leakage determined in laboratory tests (Ref erence (8]).
The test leakage levels for the ten tubes were a factor of 40 greater than the leakage measured in the entire SG prior to tube removal, indicating some crack opening occurred during the tube removal process.
Using average tube material procerties, a good correlation is obtWd between the test values and the predicted results.
The leakage values calculated for the Haddam Neck Plant specific case employed the same methodology with the exception that 95 percent confidence lower 1:mit material properties were used.
The Haddam Neck Plant ca'culated values are expected to be te Jte conservttive when compared to actual steam generator conditions.
(3)
Leakaae- -hrouah Tube to Tubcsheet Agaglgi The maximum theoretical leakage under steam line break conditions is calculated assuming that the tube is severed circumferential1y 18 inches below the top of the tubesheet and the only restriction to leakage is that tube to tubesheet annulus.
The maximum tolerance stack-up and cold leg conditions are used.
A value of 6,4 gal /miu was calculated for a single tube.
)
D.
Jack Growth Rate A smparison of the crack length determined by ECT is made for tubes with cracks present during consecutive inspections.
The data consists of 72 cracks retained in service between 1987 and 1989 as part of the F* criteria.
The 1989 length is subtracted from the 1987 length to obtain a delta length. On average the delta length is zero, indicating no growth of the cracks.
Although zero crack growth rate has been found,- as a conservatism a crack growth rate was calculated using the Reference
[1]
methodology.
For the Haddam Neck Plant the value used is 0.029 inches /EFPY or 0.037 inches per operating cycle.
J
U.S. Nuclear Regulatory Commission /B14087/Page 8 July 31, 1992 E.
lee Mate-fal Pronerties The material test certifications for the tubes used in the construction of the Haddam Neck Plant steam generators, were reviewed and the mean, standard deviation and 95 percent cenfidence lower limit values determined.
The 95 percent lower limit represents the value for which 95 percent of the tubes have an equal or gretter value. Material property data was not available for 500 J tL3 approximately 15,000 tubes in the Haddam' Neck steam gen rators.
Lack of data for these small number cf tubes is not expected to significantly affect these values.
F.
Other Considerations Tubes with identified cracks whose circumferential extent exceeds the axial extent do not meet the proposed criteria and wculd be repaired or removed from service.
Circumferential cracks i
theoretically could be present between identified axial cracks and remain undetected due to nondestructive test limitations.
The currrnt state of the art rotating pancake coil (RPC) eddy current testing can reliably identify circumferential cracks in the presence of axial cracks spaced greater than approximately 0.4 inches.
A throughwall circumferential crack conservatively will be assumed to be present between axial crecks spaced closer than 0.4 inches.
The extent of axial cracking retained in service is limited such that the theoretically undetected circumferential crack extent does not exceed 1 inch of the tube circumference.
The 1-inch limit on the circumferential extent assumed to be present is a conservative number based on burst tests of circumferential fl aws, Refer-ence [2].
Tubes with throughwall circumferential flaws up to 1.4 inches in circumferential extent were tested and found to meet Regulatory Guide 1.121 requirements for margin to bursts.
The 1-inch limit assures that axial separation of a tube with characterized cracks retained in service will not occur under normal operation or postulated accident conditions.
In addition to remaining structurally sound, tubes with cracks retained in service must not exhibit unacceptable levels of leakage under normal operation or postulated accident conditions.
Leakage under normal operation is limited to tha current Technical Specification requirement of 150 gallons per day (gpd) per steam generator.
Because of the location of the cracks approximately 18 inches into the tubesheet, the likelihood that the tube to the tubesheet annulus is packed with sludge, and the fact that not all cracks would be ex)ected to be through the tube wall, the contribution to the 3aseline operating leakage by these tubes is expected to be small.
Examination of the steam generator leakage history of the Haddam Neck Plant steam generators shows no change in the baseline leakage levels following the plugging of 200 tubes with
.,,,..,c.
-.w---..---
-0
U.". 'uclear Regulatory Commission A:tachment 1/Bl4087/Page 9 Ju's." 31, 1992 cracks in the expansion region.
A postulated increase in the operating baseline leakage does not decrease margins of safety since the required shutdown limits remain the same and any developing condition which increases the rate of leakage would require plant shutdown sooner than the condition where the baseline leakage was Zero.
Leakage from undetected circumferential cracks which theoretically could be present is assumed to be negligible, since the actual through wall length would be very small, the increase in load in the axial direction due to an increase in pressure is one half of the increase in load in the hoop direction, and the presence of the tubesheet restricts leakage.
The calculated maximum leakage under -
accident conditions from the approximately 200 tubes which were previously plugged, but meet the proposed criteria, was determined to be 2.6 gpm.
As an additional check, the maximum leakage through the tube to tubesheet annulus was determined.
The annulus was assumed to be
/
clean and the tube was assumed to be severed circumferential1y 18 inches from the secondary face of the tubesheet.
Even given these highly unlikely and very conservative assumptions, the maximum leakage would not exceed 6.4 gpm during a postulated steam line break accident.
Cracks which are through the tube wall ir the roll expansion region could potentially allow primary coolant in contact with the tubesheet.
The offect of primary coolant on the tubesheet was considered. Corrosion of the tubesheet is not a concern due to the icw level of oxygen present in the primary coolant and the tendency of any corrosion products which do form to expand and block access of the coolant to the tubesheet base metal. The limited amount of tubesheet corrosion possible would tend to restrict the tube tr 4
tubesheet annulus, further limiting leakage.
Ill.
Impact of Chance on Consecuences of an Accident A.
Imnact on Previousiv Evaluated Accidents The only accident potentially impacted by-the change is the steam line break accident. However, since the change is related to steam generator tubes, the impact on a postulated steam generator tube rupture (SGTR) is also evaluated.
The change cannot impact the probability of occurrence of a steaat line break (SLB). As discussed above, the impact of the change on structural integrity of the SG tubes is such that the tubes that s
7 l
l I
U.S. Nuclear Regulatory Commission
' /B14087/Page 10 July 31, 1992 would be allowed to remain in service would be structurally sound.
Therefore, there is no increase in the probability of occurrence of an SGTR.
Since the tubes will recain structurally sound during normal operation, the change does not increase challenges to safety systems. There may be the potential for increasing the likelihood of plant shutdown if the leak rate increases during normal o,seration.
However, as discussed above, operating experienco at the Haddam Neck Plant indicates no additional leakage is expected.
Thus, this potential is expected to be small.
The change does not affect any malfunctions evaluated for the SLB or SGTR accidents.
The change does not increase the postulated primary to secondary leakage following a steam generator tube rupture since no significant increase in DP across the unaffected tubes could occur.
Therefore, it does not affect the consequence 2 of a steam generator tube rupture.
The change does not modify any failure modes. Therefore, the change cannot impact the consequences of a
previously evaluated malfunction.
The postulated primary to secondary leakage following an SLB could increase as a result of the change.
This would increase the calculated dose consequences of a steam-line break.
B.
Potential far a New Unanalyzed Accident The change could increase the primary to secondary leakage following a postulated SLB. This leakage would be limited such that the dose contribution from the aggregate tube leakage will be limited to a fraction of 10CFP.100 dose guidelines values in the event of a steam line break. Based on current analysis, this value will be limited to 100 gpm. This leak rate is larger than currently assumed but is not large enough to compromise the mitigation of an SLB. Therefore, it is concluded that the poJsibility of an accident of a different type than previously evaluated is not created by the change.
The change would allow steam generator tubes that could potential;y leak following an SLB to remain in service.
In-the design basis analyses, it is cssumed that there is no increase (over the existing level specified in the Technical 9ecification) in the primary to secondary leakage following a steam line break.
The existing technical specification does not allow any tubes to remain in service that are known to have the potential for increased leakage following an SLB.
Since the change would allow a postulated postaccident leak rate, this is considered a malfunction of a i
t U.S. Nuclear Regulatory Ccmmission /814087/Page 11 j
July 31, 1997 different type than previously evaluated, i.e., increased primary to secondary leakage post-SLB.
The proposed change will maint a the r? quired margin of safety of the steam generator tubes.
M so, the change does not compromise accident mitigation following an SLB.
The increased leakage is judged to negligibly impact core or containment predicted postaccident parameters.
Therefore, the margin of safety with respect to mechanical issues is not negatively impacted.
Radiological consequences are discussed in Section III.C below.
4 The proposed change does not introduce any new malfunctions.
The only malfunction evaluated is failure of steam generat.or tubes following an SLB.
1 C.
Eldioloaical Consecuences (1)
Licensing Basis Background The Haddam Neck Plant's initial Facility Description and Safety Analysis (FDSA) contains a radiological assessment for steam line break incidents. The methodology assumes 1% failed fuel and a 3 gpm primary to secondary leak rate resulting in a steam generator iodine activity of 9.7 curies dose equivalent I-131.
For purposes of the dose assessment, the i
entire inventory of I-131 was assumed to be released.
The thyroid dose to an individual at the site boundary was calculated to be 3 Rem. However, with an ' average steam flow rate of 300% fer the 10 seconds required for tne steam line isolation valves to close and 5% moisture carryover, the dose was calculated to be only 0.06 Rem. Since the analysis shows no failed fuel, no other activity releases were assumed, i.e.,
no post break primary to secondary leakage was assumed. These assumptions are consistent with those found in similar radiological assessments of steam line breaks for PWRs licensed in the-1967-1972 time frame (cee Table 5).
It should be noted that these assumptions are significantly different than those made in the Standard Review Plan (SRP).
Since the FDSA and its successor, the FSAR represent the licensing basis for the Haddam Neck Plant, the assumptions set forth in the FDSA can be used to assess this proposed Technical Specification change.
During the Systematic Evaluation Program (SEP)'the NRC 3erformed a radiological dose assessment for the steam line-areak using SRP assumptions.
The following assessment uses this NRC assessment, discusses -the conservatism in the SRP assumptions and adjusts the assumptions based on the FDSA
- s
(
U.S. Nuclear Regulatory Commission /014037/Page 12 July 31, 1992 4
licensing bases assumptions or based on operations data from the Haddam Neck Plant and other operating plants.
2 (2)
Assessment i
for this proposed change the only accident which needs to be considered from a radiological viewpoint is the main steam line break. This is the only accident which would result in a larger Ap across the tubes than that which exists during normal operations and hence, may increase the leakage rate i
through the cracks.
For the steam generator tube rupture i
accident and the control rod ejection accident, the primary to secondary leakage is not expected to increase as a result of the accident and hence there would be no change in accident i
assumptions.
The radiological consequences for a Haddam Neck Plant main steam line break were calculated by the NRC Staff in 1982 4
during SEP and are presented in Reference [3].
The calculation was done in accordance with Standard Review Plan 4
15.1.5, Reference [4].
The assumption used for primary to secondary leakage in the Standard Review Plan and by the NRC Staff is that the postbreak primary to secondary leakage is equal to the Technical Specification primary to secondary leak rate limit.
This is a nonmechanistic simplifying assumption that may, under certain conditions, be nonconservative.
Prior to the accident (steam line break), operation with through wall i
defects in the tubes 1.4 permitted. Depending on the type of defect and location, leakage could increase as a result of the break and depressurization of the secondary side.
The theoretical calculated leakage values discussed are calculated assuming the cracks are above the top of the tubesheet and takes no credit for the fact-that the cracks are located in tubes in an area entirely within the tubesheet.
This tube /tubesheet crevice would further restrict primary /
secondary flow even for the case where the annulus area was assumed to be clean. The annulus arca may contain corrosion 4
products that would additionally restrict flow.
Hence, the source of leakane for the radiological assessment is itself a conservative number. In aMition to source term conservatism, the fc11owing dose assestnent assumption conservatisms exist.
1.
95% worst-case meteorology -approximately 10-100 times higher than typical meteorology, especially for a'
release such as this where thermal plume rise can be expected.
U.S. Nuclear Regulatory Commission /B14087/Page 13 July 31, 1992 2.
No iodine DF in the steam generators--although the affected generator will be dry, there would still be significant platcout on generator surfaces such as the steam dryers and in our specific case here.in the roll region.
A minimum DF of 10 has been used on other dockets (See Table 5.)
3.
Iodine spiking factor of 500--this is a factor of 2-100 conservative depending on the plant.
For the Haddam Neck Plant a spiking factor of approximately 25 is typical.
A factor of 3
has been reported (Drerence [5]).
4.
The plant is operating right at its Tech. Spec. primary coolant limit of 1 yCi/gm DEQ 1-131. Typical operation is 10-1000 times less than this.
5.
The plant is operating right at itt secondary side Tech.
Spec. limit of 0.1 yC/gm 1-131, This is over a 1000 factor conservative ano should not even be considered in the dose calculation, as the contribution will be small upared to the primary activity released.
In s
Lference [3), initial secondary side activity accounts for 60% of the dose.
6.
The plant is operating right at its primary to secondary Tech. Spec. leak rate limit.
Typically, the plant operates at 10-1000 times less than this limit.
Hence, the conservatisms in the steam line break dose calculation more than offset an increased leak rate.
Since mechanistically leakage will increase through the cracks following a steam line break and with the above conservatism
.in mind, the SRP guidelines can be modified and an assumed ir creased leakage used to determine if the proposed change is safe.
The proposed Technical Specification correlates theoretical crack leakage determined in accordance with the Haddam Neck Plant specific application of Reference [1] with the dose calculation methodology described below.
A main steam line break event theoretical leakage is calculated.
This leakage rate is then used in the dose calculation methodology.- A postaccident leak rate of 100 gpm is used.
The potential leak rate is 1000 times the current Technical Specification operations 1 limit. Even at this leak rate the proposed change is safe. To show this, the NRC Staff calculation in Reference [3] is modified using realistic assumptions.
U.S. Nuclear Regulatory Commission /B14087/Page 14 July 31, 1992 Fer Reference (3), it can be determined that the limiting
- dose, i.e.,
the one closest to its limit, is the Exclusion Area Boundary (EAB) thyroid dose for the accident initiated spike. Whole body doses are relatively insignificant without any fuel failures.
The NRC calculated EAB thyroid dose was 10.8 REM.
Of this, only 4 REM was due to the primary to secondary leakage.
The 7 REM due to preexisting secondary side activity can realistically be neglected.
The NRC assumed a primary to secondary leakage of 0.4 gpm in the affect generator. Thus,-
using an 0.4 gpm leakage rate assumption, the resulting dose is approximately 10 REM per gpm leakage.
Assume that leakage increases by a factor of 1000 in the affected generator.
That is,- the leakage goes from 0.1 gpm f
(the normal operating Tech. Spec. limit for any one generator) to 100 gpm in the affected generator.
Also, assume a realistic iodine spiking factor of 100 in lieu of 500 Reference [5] and 3 realistic iodine DF of 30 Reference (6),
in the steam generator. -The resulting dose would be:
10 Dm=10 x100ppmx
=20 rem 0
0 This is within the 30 REM acceptance criteria of SRP as being a small fraction of 10CFR100.
A 20 REM thyroid dose is equivalent to a whole-body dose of approximately 0.6 REM and is certainly safe.
This dose still assumes worst-case meteorology and operation at the Technical Specification coolant activity limit.
An even more realistic estimate of the dose would reduce tne X/Q by c factor of 30 and the coolant activity by an activity factor of 100. This would yield a dose of:
Dm=20remx x
=7(-3) rem 3
This is certainly well within 10CFR100 Timits and not a safety concern given the low dose and low probability of the initiating event.
U.S. Nuclear Regulatory Commission /B14087/Page 15 July 31, 1972 IV.
SAFETY DETERMINATI(,N The evaluation above concluded that the proposed change would increase the dose consequences of an SLS and introduce a malfunction of a different type than previourly evaluated. This means that the proposed change is an unreviewed safety question under 10CFR50.59.
Therefore, the sections below evaluate the safety of the proposed change.
The proposed change could- -increase the primary-to-secondary leakage following an SLB. This postulated increased leakage would be limited such that the dose contribution from the aggregate tube leakage will be-limited' to a small fraction of 10CFR100 dose guideline values in the event of a steam line break using the more realistic assumptions in the dose _
2alculation.
The Emergency Operating Procedures-contain appropriate instructions to mitigate the consequences of a postulated-SLB concurrent with an SGTR.
Therefore, the procedures provide appropriate operator guidance for mitigating this postulated event.
Based on this, the probabi'ity that the operators _ will fail to mitigate an SLB is not increased.
The proposed change would allow higher primary-to-secondary leskage following a postulated SLB. This increased leakage would.Le limited such that it would not compromise the ability of the high-pressure _ safety injection (HPSI) system to mitigate an SLB. The HPSI capccity exceeds the anticipated leak rate.
Sufficient water exists in the refueling water tank to compensate for such leakage until the primary side can bc cooled down and put on RHR. Therefore, the proposed change does not affect the i
probability of failure-of mitigation equipment.
The postulated consequences following an SLB would increase as a result of the change.
However, as discussed _ above in Section III.2 Radiological Consequences using more realistic assumptions than the design basis, but still conservative assumptions, the consequences would be a small fraction of 10CFR100 limits. Usir J even more realistic assumptiors results in the increased public risk being negligible.
The change does not increase the probability of any cccident..Also, the postulated increased consequences following an SLB will negligibly. impact public risk. Therefore, the proposed change is judged to be safe.
u
5 U.S. Nuclear Regulatory Commission B14087/ References /Page 1 July 31, 1992 EEFERENCES (1)
J. A. Maccoun, Electric Power Research In.stitute, letter to NRC Attention Emmett L. Murphey dated March 26, 1992 transmitting EPRI NP-6864-L, PWR Steam Generator Tebe Repair Limits:
Technical Support Document-ror Expansion Zone PWSCC in Roll Transitions (Revision 1), December 1991.
(2)
WCAP 12916, " Northeast Utilities Tube specimen Burst Tests," dated May 1991.
(3)
Dennis Crutchfield letter to W. G. Counsil, dated November 9,1982, SEP Topic XV-2.
t (4)
NRC Standard Review Plan 15.1.5, Appendix A, " Radiological Consequences of Main Steam Line Failures Outside Containment of a Pl?R."
(5)
EGG-NERD-8648, " Probability of the Iodine Spike Release Rate During an SGTR," September 1989.
(6)
A. T. Body, Jr. letter to K. H. Evers, dated May 3,1991, Amendment No. 93 to Facility Operating Licensing No. DPR-43.
(7)
WCAP 11280, " Development of Tubesheet Roll Region Plugging Criteria for Steam Generator Tubes in the Connecticut Yankee Power Plant," Revision 1, dated January 1987 (8)
C. R. Faye and D. T. Martin, " Leak and Burst Tests of the Expansion Transmitted and Support Areas of Tubes ' Exam McGuire Unit 1 Steam Generators, Volume I,
Steam Generator Tube Test Data," B&W Report RDD:
90-5160-01:01, dated October 1990.
?
4
_-_____.____.--.s
_a.-___.,x--
b e
Figure 1 l
Expanded Roll Area Diagram i
)
l
.7500 + $0 c!A
=
i
.0550 d.0055
.640DIA(NOK.)
j.
WALL THK.
l 7
~Q
/
%/N 1
TUSE
/
-TUBE 5HEET l
/
0.25 N TOP OF ROLL-4 y
BOTTDM OF ROLL 4
4 i
~/
J i
/
/
'/
/
4.25 LENGTH OF ROLL
/
l I
n p/
I 25 i
cuDona i
N ITLEE END N
L s
J 4
p it
.22 4.00 g.
1.031'_+.005 (P]TCH)
.6595 DIA (NOM.)
.03'
=
.7640 4'.0025 DIA 2,
..~.u.._.___.._.._,
Figure 2 Jesign Curves ULTlW ATE PULL LOAD (LBS)
~ 12000 10000-8000-6000-c.
4000-1 i.r' 2000-s 0000 1
0-5 10 15'.20 25 30 35:40I45-50 55 00-MULTIPLE CRACK SLANT ANGLE (DEGREES)-
0.5 inch Cracks-1.5 inch Cracks-amen SAFETY FACTOR.*' 3 i:
I
r
_m
.m.._
m i
=
=.
1 U.S. Nuclear Regulatory Commission j
Table 1/814087/Page 1 July 31,1992 i
Table 1 Haddam Neck Parameter for Determining Allowing Crack Size 1
Inout Parameters 2
0.055 inches Tube wall thickness 0.348 inches Average tube radius Normal Primary Pressure - Psig 2050 Psig Normal Secondary Pressure - Psig 675 Psig Tube. Yield Strength (psi) 9 600 F 36,960
=
- 95% lowe. confidence limit Tube Ultimate Strangth (psi) 9 600 F 91,675
=
- 95% lower confiilence limit t
Results Reference crack length a
0.528 inches Corr 9ction for tubesheet constraint a,,
0.060 inches
=
Allowances for Crack Growth 0.037 inches a,
=
c NDE uncertainty allowance a,
0.102 inches
=
y Allowable crack length A
.528 +.060
.037 - 0.102 0.45 i
l
l l
U.S. Nuclear Regulatory Cornission Table 2/B14087/Page 1 July 31, 1992 Table 2 l
TUBE LOADS USED IN ANALYSIS I
DIFFERENTIAL AXIAL.
ANALYSIS 1
PRESSURE FORLE SAFETY VALUE
)
fosi) lbs)
FACTOR (lbs)
Normal Operation 1375 633 3
1899-I Steam Line Break 2650 1220 1.43 1744 i
?
5
?.
r i
i l
I s
I s
,s a
en
+ -
r 2---m we v w"wn=
~r m
n
-em- - - &
m'w e
+ ~
L
.D U.S. Nuclear Regulatory Commission i
Table 3/B14087/Page 1 l
July 31, 1992 Table 3 l
Haddam Neck Plant Parameters for Determining Leakage for Crack Lenaths0.05 to 0.5 inches 0.055 inches Tube wall thickness - inches 0.348 inches Average tube radius
=
l 0.33 l
Poisson's Ratia
=
6 29.5 x 10 psi Elastic modulus 0 600 F
=
i 2650 psi Maximum Accident Pressure
=
3 0.027663 lb/ inches Density 0 539 F
=
954.9'F Saturation Pressure O psi l
Secondary Accident Pressure 91,675 psi Tube Ultimate Strength 9 600 F 95% lower confidance limit l
Tube Yield Strength 9 600 F 36,960 psi
- 95% lower confidcnce limit l
l l
0 U.S. Nuclear Regulatory Commission Table 4/814087/Page 1 July 31, 1992 Table 4 Haddam Neck Plant Elastic and Plastic Components of Leakage Area Openj)ng Openjng
_ leakage Elastic Plastic Rate fin (in i foal / min)
Crack S. irs 0.05 2.11 x 10
O 0.0001 0.10 9.076 x 105 0-0.0004 0.15 2.262 x 10*5 0
0.0009 0.20 4.536 x 10' 0
0.0019 0.25 8.06 x 10'5 1.1 x 10'5 0.0040 1.327 x 10 4.1 x 10'5 0.0081 2.06 x 10',
0.30 1.04 x 10
O.0155 0.35 0.40 3.068 x 10' 2.13 x 10
O.0285 0.45 4.405 x 10
3.87. x 10
O.0503 0.50 6.14 x 10"'
6.31 x 10" O.0858 s
i mm_'m___-m_
___m_
_--.m-
U.S. Nuclear Regulatory Comission Table 5/B14087/Page 1 July 31, 1992 Table 5 Radiological Dose Calculation Assumptions Iodine
. lodine Spiking Plateout Plant
% Fuel Failure Duration Factor Oconee 1% defective fuel rods None NA Surrey 1% defects,e fuel rods None 10 Fort Calhoun 1%
None 1
Robinson 1%
None NA Kewaunee 1%
None 10 e
p
--_------u
s---
.--n
--