ML20205G299

Core Spray Line Cracking Safety Evaluation

ML20205G299
Person / Time
Site:	Peach Bottom
Issue date:	10/31/1985
From:	Bloomstrand R, Stoll C, Tran P GENERAL ELECTRIC CO.
To:
Shared Package
ML20205G295	List: ML20205G289 ML20205G299
References
DRF-E21-00090, DRF-E21-90, MDE-216-1085, NUDOCS 8511130208
Download: ML20205G299 (6)

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MDE-216-1085 f s DRF-E21-00090 l.j October 1985 PEACH BOTTOM ATOMIC POWER STATION UNIT 3 CORE SPRAY LINE CRACKING SAFETY EVALUATION Prepared by: W P. T. Tran, Engineer Application Analysis Services Verified by: s C. H. Stoll, Principal Engineer Application Analysis Services Reviewed by: h R. R. Bloomitrand, Senior Licensing Engineer Licensing Services Approved by: M G. L. Sozhi, M6h/ger Application Analysis Services

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• G EN ER AL h, ELECTRIC NUCLEAR ENERGY BUSINESS OPERATIONS GENERAL ELECTRIC COMPANY e 175 CURTNER AVEtAJE e SAN JOSE, CAUFCUNIA 95125

4 PEACH BOTTOM ATOMIC POWER STATION UNIT 3 CORE SPRAY LINE CRACKING 4: SAFETY EVALUATION

'f September 1985

.Durilngthecurrentrefuelingandmaintenanceoutage,invesselinservice -

inspection located a crack on one side of the piping to junction box weld

. heat affected zone of the 'A' core spray Line (header) at Peach Bottom Unit 3. The crack was identified during inspection in accordance to IE Bulletin 80-13. Subsequent to identifying the crack air bubble tests

were performed to verify that the crack was through-wall. The air bubble test verified that the crack was through-wall and also revealed another through-wall crack on the opposite pipe to junction box weld (see Figure 1).. .Both cracks are in the piping'in the weld heat affected zone (RAZ).

One crack appears to run approximately 180 degrees in length and appears to be approximately 120 degrees through-wall. The other crack appears to be about 120 degrees in length but the through wall air bubble leakage appears to resemble a pin-hole.

General Electric has performed evaluations regarding the safety

. significance of these cracks. A summary of these results is provided as i follows:

i A. Area of discussion

1. Analysis.ha's been directed toward the following areas of significance.

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a. An estimate of the leakage flow through the cracks

~b. Structural integrity

c. Emergency core cooling system performance limits

d. Structural repairs

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e. Loose parts

'B. Results 1.a. Crack Leakage Estimate From a review of both inspections, the visual and air bubble test, the estimated leakage through both cracks appears to be less than half the leakage flow through the 1/4-inch vent hole present in the tee-box. A bounding estimate indicates that the flow through this vent hole would be less than 13 gpm during core spray operation. During normal reactor operation the flow through this line is expected to be negligible. Therefore,

.during the core spray injection phase of a'LOCA the total leakage through both cracks is expected to be less than 7 spa -

4 (less than half that already present through the vent hole).

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The core spray system includes a design margin leakage .-

allowance of approximately 100 gpm to allow for leakage through r the vent holes and thermal sleeve between the tee-bo; and vessel nozzle. No air bubble leakage was observed through the thermal sleeve during the air bubble test, although some small leakage is expected during a LOCA. Therefore, during a LOCA the combined leakage through the cracks, vent hole, and thermal sleeve are expected to be well within the margin inherent.in the core spray system.

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, 1.h. Structural Integrity

The core spray piping' cracks at Peach Botton Unit 3 are

-' - - expected to be caused by the influence of weld sensitization or

. prior sensitization of the core spray line materisi and subsequent cold work forming and installation. Sources of

, ' stress for possible Intergranular Stress Corrosion Cracking

~(ICSCC) are dependent.on residual stresses from welding bending

'and deflection during installation.

O All identified stresses during normal reactor operation were

. , found to be negligible. The loading considered include impingement load, seismic loading, pressure, weight and thermally induced loads. It is concluded that these normal operating loads do not result in stresses which are sufficient to cause the cracks observed in the piping.

In the evaluation of crack arrest, the stresses due to bracket

"- restraint and the fabrication residual stresses were also evaluated. Because the applied normal loading is predominantly displacement controlled, the stresses relax as the cracks grow.

Crack arrest is therefore expected when the crack grows to-7

- about 180 degrees.

4 Stresses incurred during core spray injection are the design stresses for the piping. Design loadings include those listed above plus those resulting from system activation. It is concluded that the structural integrity of the piping will be maintained during core spray injection.

In summary, the potential sources of stresses in the piping resulting from fabrication, installation, normal operation, and operation during postulated Loss of Coolant Accidents were s

reviewed. Potential causes of cracking and the likelihood of crack propogation were also evaluated. It is concluded that the structural integrity.of the piping will be maintained for-

, all conditions of operation.

1.c. Emergency Core Cooling It has been concluded based on the analyses discussed below that there is no change in the Maximum Average Planar Heat Generation Rate (MAPLHCR) limit of the upcoming Cycle 7 due to the presence of the cracks. ,

, The estimated total leakage through the cracks is expected to be less than 7 gpm, compared to the rated flow, assumed in the

• reload licensing analyses, through one core spray system of

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6250 gpm. Considering the additional 100 gpm leakage allowance

! (conservatively neglected in the licensing analyses) the -

)

effects are negligible. However, to conservatively bound the 4

, leakage CE has performed a sensitivity study assuming that 10%

(or 625 spm) of the core spray flow from the system with the cracks is lost from injecting into the reactor core region.

This is almost 100 times the estimated leakage rate.

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With this flow loss. assumed. the bounding effect of the crack j , , on ECCS performance was evaluated by reanalyzing the limiting i

Design Basis Accident (DBA) LOCA event. The limiting break

- I.

] size or single failure combination does not change with this

. assumption. The original limiting LOCA event analysis for

' Reload 6, Cycle 7 resulted in a maximum PCT of 1954*F. The *

,g . analyses with an assumed 10% flow loss and.all other .

assumptions identical to the reload analysis, results in a PCT of approximately 1960*F'(a negligible change). This results in >

240'F margin to the 10CFR50.46 limit of.2200'F. .Since

reanalysis indicates that sufficient margin remains, no q reduction is required in the-Cycle 7 MAPLHGR limits.

To further bound the effects of the cracks, the above analysis p was repeated but with no credit for core spray heat transfer from the core spray system with the cracks. The resulting PCT ,

increased to 2074*F, again well within the 2200*F FCT limit and again indicating that.no reduction to MAPLHGR is required.

1.d. Structural Repair The above ECCS and structural analyses indicated that: 1) the cracks are not expected to grow significantly beyond 180', 2) there is sufficient remaining structural integrity to accommodate the normal and injection loads, and 3) the cracks' do not result in exceeding ECCS PCT limits. Nevertheless, to assure that the piping can accommodate any of the postulated l loads two brackets will be welded across the piping arms and tee-box =(see Figure 1). These brackets assure that the structural integrity of the piping is restored. The design basis of the bracket welds balanced strength and installation '

ALARA considerations.

1.e Loose Parts Analysis Although it is anticipated that the piping will not break,

.particularly with the bracket design being installed, an evaluation of the possible consequences of a potential loose piece was performed. The evaluation addressed the following safety concerns: 1) Potential for corrosion or other chemical reaction to reactor materials; 2) Potential for flow blockage to a fuel bundle and subsequent fuel damage and; 3) Potential for interference with control rod operation.

The probability for unacceptable corrosion or other chemical action is zero. The piping and bracket ma'terial are selected ~

for the reactor vessel environment.

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. It is not possible to uniquely describe a loose part since none has occurred. Therefore, four different types of loose pieces

I were postulated. They'are: a section of pipe; a small piece of

. the pipe; a piece of the bracket and;'a whole bracket., The

' potential for'these postulated pieces to cause flow blockage or. -

control rod interference was evaluated by considering -

appropriate hydrodynamic principles and available flow paths.

' The possible effects of these concerns on safe reactor operation was also addressed. It is concluded that the potential for unacceptable flow blockage of a fuel assembly or for control rod interference is essentially zero.

C. Safety Evaluation The following is a summary of the evaluations:

1. Leakage through the cracks is negligible compared to the margin inherent in the PECo Unit 3 core spray system.

2. The cracks are most probably due to IGSCC and will most likely self-arrest.

3. - Structural integrity can be restored by welding support brackets to the piping and tee-box.'

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4. Bounding ECCS analyses demonstrate that NRC 10CFR50.46 limits are not exceeded and that no reduction in the upcoming Cycle 7

- MAPLHCR limits are required.

5. There are no safety concerns due to potential loose parts.

Based on the above it is concluded that:

1. There is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to safety,

2. There is no increase in the possibility for an accident or malfunction of a different type than analyzed, and

3. There is no reduction in the margin of safety as defined in the basis for the Technical Specification.

4 DRF E21-00090  :

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