CY-85-062, Rev 2 to Steam Generator Tube Sleeving & Plugging

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Rev 2 to Steam Generator Tube Sleeving & Plugging
ML20205T142
Person / Time
Site: Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date: 01/22/1986
From: Blaisdell J, Gharakhani A, Kai M
CONNECTICUT YANKEE ATOMIC POWER CO.
To:
Shared Package
ML20205T117 List:
References
ISE-CY-85-062, ISE-CY-85-062-R02, ISE-CY-85-62, ISE-CY-85-62-R2, NUDOCS 8606130063
Download: ML20205T142 (10)
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Text

r Appendix B INTEGRATED SAFETY EVAL!!ATION NO.:

ISE/CY-85-062 (REV. 2)

PLANT: CONNECTICUT YANKEE PDCR NO.:

798 REV. 1 TITLE:

STEAM GENERATOR TUBE SLEEVING AND PLUGGING REFERENCES 1.

Connecticut Yankee Steam Generator Sleeving 'and Plugging PDCR #798, Rev. O.

2.

J. A. Blaisdell, Integrated Safety Evaluation No. ISE/CY-85-062 (Rev.

1) 11/21/85 " Steam Generator Tube Sleeving and Plugging".

3.

Connecticut Yankee Steam Generator Sleeving and Plugging PDCR #798, Rev. 1.

4 M. L. Van Haltern, L. W. Ward, "Small Break LOCA evaluation of Steam Generator Tube Plugging", Safety Evaluation SE/HN-85-085.

5.

M. R. Galler, J. F. Ely " Component Engineering Safety Evaluation",

Dated Januars 13, 1986 6.

L. J. Laskowski, J. 4. Klisiewicz, " Nuclear Materials and Chemistry Safety Evaluation", Dated October 18, 1985 7.

Calc. File C2-517-585-RE, " Connecticut Yankee Plant Parameters vs.

Steam Generator Tube Plugging", Rev. O, Dated April 8,1985 8.

Calc. File C2-517-550-RE " Connecticut Yankee Minimum R. C. Flow Rate with 2000 Steam Generator Tubes Plugged," Rev.1, April 18,1985 9.

" Core Bypass Flow Summary report for Haddam Neck Plant" D. R. Forsyth, September,1985, Westinghouse Electric Corporation.

10. Memo, " Impact of tube plugging on Neutronic Safety Parameters", W. M.

Herwig to A. Gharakhani, NE-85-R-545, Nov.18,1985 l

11. W. M. Herwig, M. P. Hills, " Connecticut Yankee Cycle 14 Reload PDCR No.85-795" SE/CY-85-096, Dated December 31, 1985 DESCRIPTION OF CHANGE In the repair of the steam generator tubes, tube sleeves are installed where possible, in lieu of plugs, because they minimize the reduction in primary coolant flow rate.

In regions where it is not practical, steam generator tubes are repaired by plugging the tube at the location of the tubesheet.

Reference 1, Revision 0 of the PDCR, proposed to install sleeves or plugs

,-in steam generator tubes which exhibit defects in excess of technical

. specification limits.

Inconel 690 sleeves 33 inches long, were to be installed in accessible tubes to span defects within six inches of the top

.cf the tube sheet. Defects above this height, or in tubes not accessible h

-to sleeves, were to be plugged.

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9 Prior to this repair, 35fr tubes have been plugged. An Integrated Safety Evaluation was performed (Reference 2) which indicated that the existing i

Accident Analyses (LOCA and Non-LOCA) remain valid if no more than 400 effective tubes were plugged.

Reference 3, Revision 1 of this PDCR, increased the number of effective tubes that can be plugged from 400 total to 500 per steam generator.

This Integrated Safety Evaluation is being performed to evaluate the impact of the additional plugged tubes.

DISCUSSION The effect of tube plugging / sleeving was assessed by evaluating the thermal performance of the reactor during hypothetical incidents to ensure that it is not degraded.

Therefore, an evaluation of each transient was performed to determine if it should be reanalyzed.

The consequences of steam generator tube plegging/ sleeving are as follows:

0 Reactor coolant flow is reduced due to increased flow resistance through the steam generator (SG).

O The SG heat transfer area is reduceu.

Thus, to maintain the 100%

steam flow, Tavg must be increased or steam pressure reduced.

O Primary rea: tor coolant mass inventory is reduced.

For the purpose of this evaluation, the following assumptions are appropriate:

a. The plant will maintain 100% thermal power and design steam flow by lowering the SG pressure by about 34 psi, as indicated by the results of the calculations outlined in Reference 1.
b. A reduction of 34 psi in SG pressure corresponds to a T reduction and a T increase of approximately 0.5 F (REi1Yence 7).

hot

c. The Sh secondary inventory (indicated level) is maintained as a function of power level and is not anticipated to change significantly as a result of tube plugging. Also, the density variations are considered to be insignificant.
d. The reduction in RCS flow rate as a result of 500 effective tubes plugged per SG has been calculated based on the original measured flow rate of 278,000 GPM and the change in the SG resistance due to

~

tube plugging (Reference 8). This results in a new estimated flow of 271,300 GPM.

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e. There will be no significant change in the key neutronic parameters used in the safety analysis as a result of the tube plugging / sleeving (i.e. moderator and fuel temperature coef ficients, scram worth, ejected rod worth, axial shapes and radial peaking factors.)

(Reference 10)

INTEGRATED SAFETY EVALUATION EFFECT ON DESIGN BASIS ACCIDENTS

. In order to facilitate this evaluation, CY design analyses were grouped into five categories as outlined below:

A - Departure from Nucleate Boiling (DNB) event B - Overcooling events C - Overheating events D - Various reactivity events E - Loss of coolant events A - DNB DESIGN BASIS EVENTS Standard Review Plan (SRP) Section 4.2 specifies the acceptance criteria for various fuel design limits.

One of these states that there will be at least a 95%/95% probability / confidence level that DNB will not occur on the limiting fuel rods during norr 31 operation, operational transients, or any transient conditions arising f rom Condition 1 and 2 events.

In order to meet these bases for CY, the minimum transient DNB ratios were calculated for various transients in the original and/or subsequently revised reference analyses.

Table 1 sunmarizes the current CY design parameters.

The important parameters for analyzing DNB along with their limiting d.irections are also summarized in Table 2.

Tube plugging could affect the DNB design basis events as a result of:

1 - Reduced initial core flow 2 - More rapid primary flow coastdown for the loss of flow event due to increased flow resistance through the steam generator.

'In order to accommodate the first effect, a stu'dy has been performed by Westinghouse to determine the core bypass flow fraction.

The Westinghouse study (Reference 9), confirmed that the bypass flow fraction can be reduced

from 9% to 4.5%.

Therefore, CY Tech. Spec. 3.20 is being revised to reflect this change.

The propgsed Tech. Spec. is as follows:

Original Design Proposed Vessel flow rate, GPM 268,000 257,000 Bypass flow 9%

4.5%

Note:

257,000 GPM was derived at by placing 5% uncertainty on the calculated flow rate of 271,300 GPM.

With a vessel flow rate of 257,000 GPM and a 4.5% bypass flow, the net core flow is 245,400 GPM which exceeds the original design value of 244,600 GPM.

Since the net core flow rate, which is the key input in the safety analyses, remains above the design value, it can be concluded that the initial starting point for*the DNB design basis events remains valid.

The slightly nonconservative effect of more rapid primary flow coastdown caused by increased flow resistance for the loss of flow event is more than compensated for by the proposed change in the Hot Full Power (HFP) moderatur coefficient of the plant. The proposed change imposes a more restrictive limit on the HFP moderator coefficient (CY Tech. Spec. 3.16),

as outlined below:

HFP Moderator Coefficient Original Design Proposed Tech. Spec 10.

PCM/F 0.0 PCM/F Based on the above discussion, it has been concluded that plugging 500 effective tubes per SG at CY will not impact the DNB design bases of the plant.

f B - OVERC00 LING DESIGN BASIS t

j Overcooling t/ansients are performed to address the following:

To assure that worst containment pressure and temperature remain below the containment design conditions and also to provide the most adverse containment response to be used for Equipment Qualification.

O To demonstrate that the limits of 10CFR20 are met for Condition 2 events.

To demonstrate that radiological doses do not exceed 10CFR100 limits for Condition 4 events..

l 1

L

The worst overcooling transient is the double ended rupture of the main steam line.

Even though this transient is a Condition 4 event, it has been analyzed using more limiting acceptance criteria of "no return to criticality".

Table 3 summarizes the key safety parameters along with their limiting directions for this class of transient.

Tube plugging / sleeving could affect the c'onsequence of this class of transients as follows:

1.

Reduced heat transfer surface area would lower the rate and extent of the primary cooldown.

2.

Increased SG resistance would lower the RCS flow rate, and therefore degrade the heat transfer between the primary and secondary coolant.

As indicated in Table 3, both of these effects are in the conservative direction.

Therefore, tube plugging / sleeving will reduce the consequences of the overcooling events.

C - OVERHEATING DESIGN BASES The purpose of analyzing the overheating transients are as follows:

To demonstrate that the RCS pressure will not exceed 110% of design pressure during Condition 1 and 2 events.

(This criteria is assured by applying the more stringent requirement that the pressurizer "must not be filled with water.)

To detennine the design bases of the auxiliary feedwater system.

The most limiting transients analyzed in this category are the loss of load and loss of nonnal feedwater events.

Table 4 sumarizes the key safety parameters along with their limiting directions for this class of events.

The effects of tube plugging / sleeving on these events are as follows:

1. Reduced RCS flow rate
2. Reduced SG heat transfer area
3. Decreased RCS mass inventory However, the most important parameter for this class of events is the moderatortemperaturecoefficient(MTC). As indicated in Section A, CY Tech. Spec. 3.16 is being revised to impose a more restrictive limit on the

. hot full power MTC. The reduction in the maximum allowable MTC more than compensates for the effects of SG tube repair.

D - VARIOUS REACTIVITY EVENTS The following transients can be grouped under this category:

0 Rod withdrawal Boron dilution 0

Dropped rod Control rod ejection This class of transients is affected by changes in core kinetics characteristics, control rod worths, and core power distributions.

As indicated previously, SG tube repair will not cause a significant change in any of the key neutronic parameters of the plant.

E - LOSS OF COOLANT ACCIDENT (LOCA)

References 4 and 11 address the impact of plugging / sleeving 500 tubes per steam generator on the small and large break LOCA analysis for Connecticut Yankee.

These evaluations concluded that the current LOCA design basis analyses will not be affected.

MECHANICAL ASPECTS OF TUBE PLUGGING / SLEEVING Repairing steam generator tubes by sleeving, although not formerly done at Connecticut Yankee, has been done at several other PWRs.

As stated in Reference 5, sleeving is a repair process that requires sleeves to be inserted inside the defective tube in a manner to completely span the defective tube region.

Plastic expansion of the sleeve against the tube, at the top and bottom of the sleeve, provides a leak limiting joint which exhibits sufficient mechanical strength to withstand normal operating, test and postulated accident loading conditions.

The tube repair program will be conducted in accordance with the requirements 6f the project design specifications.

These specifications (SP-ME-516, SP-ME-518 and SP-ME-538) require that the sleeve and plug design, materials of construction, and installation method are qualified by testing and analyses to establish that:

1.

The upper and lower mechanical joints have been qualified by design, analysis and design verification testing to ensure:

- leak tightness

- pressure retaining / load carrying capability for nonnal and faulted design loading conditions

, j

- adequate resistance to potential stress-corrosion cracking and general corrosion concerns, and

- the hard rolling of the bottom tube sleeve joint will not degrade the tube-to-tubesheet weld.

2.

The tube design minimizes the increase to primary flow resistance.

3.

The tube sleeves are purchased according to appropriate specifications.

4.

The tube sleeves are inspected and tested prior to installation to ensure that they are manufactured according to specifications and have the required structural integrity.

5.

The installation process is monitored to ensure proper installation.

The purchase specifications further require that the tube sleeves and plugs be inspected and tested prior to installation to ensure that they are manufactured according to specifications and have the required structural integrity.

Proper installation is assured by process control monitors and by inspection.

Since all installed tube sleeves will be subject to the qualification program requirements, purchase specification and installation / inspection procedures repairing defective steam generator tubes by sleeving will not increase the probability of any previously analyzed accident (e.g. steam generator tube rupture.)

NUCLEAR MATERIALS AND CHEMISTRY ASPECTS OF TUBt PLUGGING / SLEEVING l.

As indicated in Reference 6, the sleeve repair leads to an upgrade in corrosion resistance as compared to the original tubing.

No unacceptable accelerated corrosion effects as a result of crevices, or tensile stresses associated with the sleeve have been identified based upon available data.

Thus, the probability of occurrence of a steam generator tube rupture event l

is not increased by the sleeving repair process.

Because of the above assessments, this PDCR will not increase the probability of occurrence or the consequences of an accident or malfunction of equipnent important to safety previously evaluated in the safety l

analysis report.

POTENTIAL FOR CREATION OF AN UNANALYZED ACCIDENT Reference 5 considered the possibility of a number of failure modes associated with repairing steam generator tubes by sleeving or plugging.

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. These include:

1.

failures caused by inadequate design or improper installation, and 2.

failures caused by installation of a defective sleeve.

Because of the extensive qualification process, sleeve specification, inspection and testing, and controls on sleeve installation, these f ailure modes are considered to be improbable.

As a result, this PDCR does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report.

EFFECT ON THE MARGIN OF SAFETY The design specification for the sleeves / plugs, the sleeve / plug inspection, sleeve / plug testing and installation procedures will result in a repaired steam generator tube which is structurally sound. The repaired tube will provide a pressure boundary barrier which meets all necessary requirements.

Therefore, repairing sesam generator tubes by sleeving and/or plugging does not reduce the margin of safety as specified in the basis of any technical specification.

SUMMARY

AND CONCLUSION Because of the above assessments, the changes implemented by this PDCR are considered to be safe and do not constitute an unreviewed safety question, as defined in 10CFR50.59, since it does not:

1.

increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report, 2.

create the possibility for an accident or malfunction of a different type than previously analyzed in the safety analysis report, or 3.

reduce the margin of safety as defined in the basis of any techpicals cification, k9h7 I OMA fl//2/fh

'S. Rai, Supervisor

/

A'. Gharikhani, Engineer

/M.

. Transient Analysis Section Transient Analysis Section

$YWErb

/J'A.Blaisdell, Manager C

L5afety Analysis Branch

_9_

TABLE 1 CY DESIGN PARAMETERS Core power 1825 Mwth.

System pressure 2000.0 psig.

Inlet temperature 540.6 F Initial total flow rate 268,000 GPM.

Core bypass fraction 9%

Enthalpy rise hot channel factor 1.78 Heat flux hot channel factor 3.09 TABLE 2 IMPORTANT PARAMETERS FOR ANALYZING DNB EVENTS PARAMETER LIMITING. DIRECTION Core power Maximum System pressure Minimum Inlet temperature Maximum Inlet core flow rate and pump coastdown characteristics Miniraum Core bypass flow Maximum Radial and axial power distribution Maximum o

10-1 TABLE 3 IMPORTANT PARAMETERS FOR ANALYZING OVERC00 LING EVENTS PARAMETER LIMITING DIRECTION Shutdown margin Minimum Moderator feedback Maximum 4

Power peaking factors Maximum RCS flow rate Maximum SG. heat transfer area and inventory Maximum TABLE 4 IMPORTANT PARAMETERS FOR ANALYZING OVERHEATING EVENTS PARAMETER LIMITING DIRECTION Doppler coefficient Maximum Moderator coefficient Maximum Trip reactivity Minimum Decay heat Maximum SG. heat transfer area and inventory Minimum RCS flow rate and inventory Minimum 1

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