Text

Proposed Tech Specs Revising once-through Steam Generator Tech Specs by Incorporating once-through Requirements of Proposed Amend 63,Rev 2
	ML20128G361
Person / Time
Site:	Rancho Seco
Issue date:	06/17/1985
From:	; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	;
Shared Package
ML20128G353	List: ML20128G350; ML20128G361;
References
	NUDOCS 8507090165
	Download: ML20128G361 (17)
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS LIST OF TABLES Table Page

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2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 2-9 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS 3-27 3.6-1 SAFETY FEATURES CONTAINMENT ISOLATION VALVES 3-40 3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS 3-55 3.15-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3-61 3.16-1 RADI0 ACTIVE GASES EFFLUENT MONITORING INSTRUMENTATION 3-64 3.22-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 3-83 3.22-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS 3-86 IN ENVIRONMENTAL SAMPLES 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS 4-3 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY 4-8 4.1-3 MINIMUM SAMPLING FREQUENCY 4-9 4.2-1 CAPSULE ASSEMBLY WITHDRAWAL SCHEDULE AT DAVIS-BESSE 1 4-12a 4.2-2 INSERVICE INSPECTION SCHEDULE 4-13 4.10-1 ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-42 4.10-2 OPERATIONAL ENVIRONMENTAL RADIATION MONITORING PROGRAM 4-22a 4.14-1 SNUBBERS ACCESSIBLE DURING POWER OPERATIONS 4-47c 4.17-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE 4-51 INSPECTED DURING INSERVICE INSPECTION 4.17-2A STEAM GENERATOR TUBE INSPECTION - 4-57 4.17-28 STEAM GENERATOR TUBE INSPECTION 4-57a 4.17-3 SPECIAL PERIPHERAL GROUP TUBES 4-57b 4.17-4 SPECIAL LANE REGION GROUP TUBES 4-57f 4.19-1 RADI0 ACTIVE LIQUIU EFFLUENT MONITORING INSTRUMENTATION 4-64 SURVEILLANCE REQUIREMENTS 4.20-1 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4-66 SURVEILLANCE REQUIREMENTS Proposed Amendment 122, Revision 1 8507090165 850617 PDR ADOCK 05000312 k __

itANCHOSECOUNIT1 TECHNICAL SPECIFICATIONS Surveillance Standards -

4.17 STEAM GENERATORS Applicability Applies to inservice inspection of the steam generator tubes.

Objective To verify the operability of each steam generator and ensure the structural integrity of the tubes as part of the reactor coolant boundary.

Specification Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 1.3.

4.17.1 Steam Generator Sample Selection and Inspection Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting steam generators as specified in Table 4.17-1.

4.17.2 Steam Generator Tube Sample Selection and Inspection The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.17-2A. The inspection result classification and the corresponding action required-for inspection of " specific limited areas" (see Paragraph 4.17.2e) shall be as specifed in Table 4.17-28. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.17.3 and the inspected tubes shall be verified acceptable per Specification l 4.17.4. Tne tubes selected for these inspections shall include at least 3% of the total number of tubes in both steam generators and be selected on a random basis except:

a. If experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

b. The first sample inspection during inservice inspection (subsequent to the first inservice inspection) of each steam generator shall include:

1 All nonplugged tubes that previously had detectable wall penetrations (>20%), and

2. Special area tubes pursuant to Specification 4.17.4a6 Proposed Amendment 122, Revision 1 4-51

l liANCHOSECOUNIT1

. TECHNICAL SPECIFICATIONS

Surveillance Standards l I

4.17.2 (continued)

c. The second and third sample inspections during each inservice inspection may be less than a full tube inspection by concentrating (selecting at least 50% of the tubes to be ,

inspected) the inspection on those areas of the tube sheet array and on those portions of the tubes where tubes with imperfections were previously found.

d. A tube ' inspection (pursuant to Specification 4.17.4.5) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection. (" Adjacent" is interpreted to mean the nearest tube capable of being inspected.) Tubes which do not permit passage of the eddy current probe will be considered as degraded tubes when classifying inspection results.

e. Tubes in specific limited areas which are distinguished by unique operating conditions and/or physical construction (for example, tubes adjacent to the open inspection lane or tubes whose 15th tube support plate hole is not broached but drilled) may be excluded from random samples if all such tubes in the specific area of a steam generator are inspected. No credit will be taken for those tubes in meeting minimum sample size requirements.

The results of each sample inspection shall be classified into one of the following three categories: c Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10%

of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: . In all inspections, previously degraded tubes must exhibit significant (>10 % ) further wall penetrations to be included in the above percentage calculations.

4.17.3 Inspection Frequencies The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

Proposed Amendment 122, Revision 1 4-52 k

IkANCHOSECOUNIT1

. TECHNICAL SPECIFICATIONS

' Surv:111ance Standards -

4.17.3 Inspection Frequencies (Continued)

a. The first inservice inspection shall be performed during the first refueling outage. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months af ter the previous inspection. If two consecutive inspections following service result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no significant additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.17-2A and/or Table 4.17-2B l at 40-month intervals falls in Category C-3, the inspection frequency shall be increased to at least once per 20 months.

The increase in inspection frequency shall apply until a subsequent inspection meets the conditions specified in 4.17.3a and the interval can be extended to a 40-month period.

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.17-2A during the l shutdown subsequent to any of the following conditions:

1. A seismic occurrence greater than the Operating Basis Earthquake,

2. A loss-of-coolant accident requiring automatic actuation of the engineered safeguards, or

3. A main steam line or feedwater line break as defined in the USAR.

d. Af ter primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.1.6.9, an inspection of the affected steam generator will be performed in accordance with the following criteria.

1. If the leaking tube falls within one of the Special Area groups designated in either Table 4.17-3 or Table 4-17-4, all of the tubes in this group in this steam generator will be inspected.

Proposed Amendment 122, Revision 1 4-53 k - - - . - .. .

D RANCHO SECO UNIT 1

- TECHNICAL SPECIFICATIONS Surveillance Standards 4.17.3 Inspection Frequencies (Continued)

2. If the leaking tube does not fall within one of the Special Area groups, an inspection will be performed on the affected steam generator in accordance with Table 4.17-2A with an initial inspection sample size of 6% of the tubes in the affected steam generator.

4.17.4 Definitions l

a. As used in this Specification:

1. Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications of less than 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2. Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.

3. Degraded Tube means a tube containing imperfections >20% .

of the nominal wall thickness caused by degradation.- l

4. Defective Tube means a tube containing an imperfection i

>40% of the nominal tube wall thickness unless higher Timits are shown acceptable.by analysis. Defective tubes shall be plugged.

5. Tube Inspection means an inspection of the steam generator tube from the point of entry completely to the point of exit (except as noted in 4.17.2c and 4.17.4a6). l

6. Special Area Tubes include those tubes which have shown a significant amount of degradation throughout their operating history. These include peripheral tubes and lane region tubes designated in Tables 4-17-3. and 4-17-4.

b. The steam generator shall be determined OPERABLE after completing the corresponding actions required by Table 4.17-2A (and Table 4.17-28 if provisions of Paragraph 4.172e are utilized.)

4.17.5 Reports

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 30 days. l Proposed Amendment 122, Revision 1 4-54

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.17.5 Reports (Continued)

b. The results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period l in which this inspection was completed. This report shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

c. Results of steam generator tube inspections wnich fall into Category C-3 and require notification of the Commission shall be reported pursuant to Specification 6.9.5P prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

Bases The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The surveillance requirements of steam generator tubes are based on a modification of B+W - Standard Technical Specifications dated June 1, 1976. Inservice inspection of steam generator tubing is essential in order to maintan surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. For example, historical data has shown that certain areas of the steam generators are particularly susceptible to corrosion and degradation. Consequently, the inservice inspecton now includes provisions for a more indepth inspection of a Special area group of tubes specified in Table 4.17-3 and Table 4.17-4.

Proposed Amendment 122, Revision 1 -

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RANCHU SECU UlJIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Bases (Continued)

Operational experience has shown that tube defects can be the result of unique operating conditions or physical arrangements in certain areas of the steam gene rato rs. A full inspection of all of the tubes in such limited areas will provide complete assurance that degraded or defective tubes in these areas are detected. Because no credit is taken for these distinctive tubes in the constitution of the first sample or its results, the requirements for the first sample are unchanged. Tnis requirement is essentially equivalent to and meets the intent of the requirements set forth in IJRC Regulatory Guide 1.83, Revision 1 and does not reduce the margin of safety provided by those requirements.

Wastage-type defects are unlikely with AVT chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for defective tubes. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20 of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.9.5P prior to resumption of plant operation. Such cases l will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection and revision of the Technical Specifications, if nece ssa ry. ;

Proposed Amendment 122, Revision 1 4-5Sa ...

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards

. TABLE 4.17-2A STEAM GENERAToet TUBE INSPECTION 2ND SAMLE INSPECTION 3RD SAMPLE INSPECTION IST SAMLE INSPECTION Sample size Result Action Require 1 Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A 5 of the ";e. e Tubes per C-2 Plug defective C-1 N/A N/A 5.G. tubes and inspect additional 25 of i.-4 Plug defective tubes and G-1 mone the tubes in this inspect additional 45 of 5.G. the tubes in this 5.G. C-2 Plug defective tubes C-J Plug eefective tubes and l .

perform action for C-3 result of first sample C-3 Plug defective tubes and l perform action for C-3 N/A N/A result of first sample C-3 Inspect M tubes The other None N/A N/A in this 5.G. plug . S.G. is defective tubes C-1

and inspect 25 tubes in the other 5.G. The other Perform action for C-2 N/A N/A Perform follow-on 5.G. 1s F. result of second sample inspections in the C-2 other 5.G. in accordance with results of the above The other (a) II defects can be N/A N/A

- Inspection as applied to 5.G. is localized to an affected

! Tatie 4.17-2A. ' C-3 area. inspect all tubes in affected area and L Nottffcation to NaC (2) plug defected tubes.

l pursuant to specification (b) If defects cannot be 6.9.5P localized to an affected area, inspect all tubes in this 5.G. and plug ,

defective tubes b

(1) 3*n Where n is the number of steam generators inspected during an inspection Proposed Amendment 122. Revision 1 4-57 k

i RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Survel11ance Standards TA8LE 4.17-25 STEM GENERATOR TUSE INSPECTION 2ND S mPLE INSPECTION OF A " SPECIFIC LIMITED AREA" IST SAN LE INSPECTION OF A "SFECIFIC LIMITED AREA

Action Required Result Action .7equired Sample Size Result s

None N/A N/A 100L of Area C-1 in both OTSGs Plug defective tubes N/A N/A

.C-2 Plug defective tubes. M/A N/A C-3

Notification to NRC pursuant to specification 6.9.5P C-1 None N/A N/A 100% of Area in one OTSG C-2 Plug defective tubes and inspect C None 100% of corresponding area in other OTSG.

C-2 P. lug defective tubes h ,' C-3 Plug defective tubes and notify NRC pursuant to specification 6.9.5P Plug defective tubes and inspect C-1 None C-3 100% of corresponding area in other OTSG. Notification to NRC pursuant to specification C-2 Plug defective tubes 6.9. 5P C-3 Plug defective tubes 4-57a Proposed heendment 122. Revision 1

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.- RANCHO SECO UNIT 1 -

TECHNICAL SPECIFICATIONS Surveillance Standards SPECIAL PERIPHERAL GROUP TUBES

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(Above arid including the 13th Tube Support Plate ONLY)

TABLE 4.17-3 Row Tubes 1 1-16 2 1-27 3 1-34 4 1-41 5 1-17, 30-46 6 1-14, 38-51 7 1-12, 43-54 8 1-10, 48-57 9 1-10, 53-62 10 1-9, 57-65 11 1-9, 60-68 12 1-9, 63-71 13 1-8, 67-74 14 1-7, 69-75 15 1-7, 72-78 16 1-7, 75-81 17 1-6, 77-82 18 1-7, 79-85 19 i- 1-6, 81-86 20 1-6, 80-85 21 1-6, 85-90 22 1-6, 88-93 23 1-6, 89-94 24 1-5, 91-95 25 1-6, 93-98 26 1-5, 95-99 27 1-5,96-100 28 1-5,97-101 29 1-5, 100-104 30 1-5, 101-105 31 1-5, 102-106 32 1-5, 103-107 33 1-4, 105-108 34 1-4, 104-107 35 1-4, 105-108 36 1-5, 109-113

,J, 37 1-5, 110-114 38 1-5, 111-115 39 1-5, 112-116 40 1-4, 114-117

Proposed Amendment 122 Revision 1 4-57b

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',"; RANCHO SECO UNIT 1

.i-TECHNICAL SPECIFICATIONS Surveillance Standards s

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, 0 ;. SPECIAL PERIPHERAL GROUP TUBES 3 , .

(Above and including the 13th Tube Support Plate ONLY) h TABLE 4.17-3 (Continued)

$ I Row Tubes i/ 41 1-3, 114-116

'< 42 1-3, 115-117 A

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.43 1-3, 116-118

) . 44 1-3, 117-119

\ 45' 1-4, 117-120

; f46 1-4, 116-119

, '47 1-4, 119-122

' 48 1-4, 120-123

,49 1-4, 121-124 l 50 1-3, 121-123 l 151 1-3, 122-124 52 1-3, 123-125 53 1-3, 124-126 l f / 54 1-4, 124-127

, 55 1-3, 124-126

'5 56 1-3, 125-127 57 1-3, 126-128

. 58 1-3, 127-129 59 c 1-3, 122-124 60 1-3, 127-129 61 1-3, 124-126 62 1-2, 128-129 63 3, 126-128 64 124-126

/ 65 123-124 66 125-127 127-128 67

f. ; - 68 126-127 69 128-130 70 128-129 71 e 127-129 72 ,

127-128 73 126-128 74 122-123 75 .' 123-124 76 58 77

122J124 78 . 122-124

. 79 -

126-128 80 128-129 Proposed Amendment 122, Revision 1 4-57c

f RANCHO SECO UNIT 1 -

TECHNICAL SPECIFICATIONS Survefilance Standards SPECIAL PERIPHERAL GROUP TUBES (Above and including the 13th Tube Support Plate ONLY)'

TABLE 4.17-3 (Continued)

Row Tubes 81 127-129 82 127-128 83 128-130 84 126-127 85 125-126 86 125-127 87 123-124 88 124-126 89 3, 126-128 90 1-2, 128-129 91 1-3, 124-126 92 1-3, 127-129 93 1-3, 122-124 94 1-3, 127-129 95 1-3, 126-128 96 1-3, 125-127 97 1-3, 124-126 98 1-4, 124-127 99 i- 1-3, 124-126 100 1-3, 123-125 101 1-3, 122-124 102 1-3, 121-123 103 1-4, 121-124 104 1-4, 120-123 105 1-4, 119-122 106 1-4, 116-119 107 1-4, 117-120 108 1-3, 117-119 109 1-3, 116-118 110 1-3, 115-117 111 1-3, 114-116 112 1-4, 114-117 113 1-5, 112-116 114 1-5, 111-115 115 1-5, 110-114 116 1-5, 109-113 117 1-4, 105-108 118 1-4, 104-107 119 1-4, 105-108 120 1-5, 103-107 Proposed Amendment 122, Revision 1 4-57d a

- RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards SPECIAL PERIPHERAL GROUP TUBES (Above and including the 13th Tube Support Plate ONLY)'

TABLE 4.17-3 (Continued)

Row Tubes 121 1-5, 102-106 122 1-5, 101-105 123 1-5, 100-104 124 1-5,97-101 125 1-5,96-100 126 1-5, 95-99 127 1-6, 93-98 128 1-5, 91-95 129 1-6, 89-94 130 1-6, 88-93 131 1-6, 85-90 132 1-6, 80-85 133 1-6, 81-86 134 1-7, 79-85 135 1-6, 77-82 136 1-7, 75-81 137 1-7, 72-78 138 1-7, 69-75 139 i 1-8, 67-74 140 1-9, 63-71 141 1-9, 60-68 142 1-9, 57-65 143 1-10, 53-62 144 1-10, 48-57 145 1-12, 43-54 146 1-14, 38-51 147 1-17, 30-46 148 1-41 149 1-34 150 1-27 151 1-16 l Proposed Amendment 122, Revision 1 4-57e k -__- -

.- RANCHO SECO UNIT 1 -

TECHNICAL SPECIFICATIONS Surveillance Standards SPECIAL LANE REGION GROUP TUBES (Above and including the 14th-Tube Support Plate ONLY)'

TABLE 4.17-4 Row Tubes 63 1-2 64 1-4 65 1-6 66 1-9 67 1-13 68 1-20 69 1-20 70 1-23 71 1-27 72 1-30 73 1-60 74 1-60 75 1-60 77 1-60 78 1-60 79 1-60 80 1-30 81 1-28 82 t- 1-23 83 1-20 84 1-20 85 1-13 86 1-9 87 1-6 88 1-4 89 1-2 l

Proposed Amendment 122, Revision 1

. 4-57f I

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls Special Reports

'6.9.5 Special reports shall be submitted to the Director of the Regulatory Operations Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

A. A one-time only, " Narrative Summary of Operating . Experience" will be submitted to cover the transition period (calendar year 1977).

B. A Reactor Building structural integrity report shall be submitted within ninety (90) days of completion of each of the following tests covered by Technical Specification 4.4.2 (the integrated leak rate test is covered in Technical Specification 4.4.1.1).

1. Annual Inspection

2. Tendon Stress Surveillance

3. End Anchorage Concrete Surveillance %

4. Liner Plate Surveillance C. Inservice Inspection Program D. Reserved for Proposed Amendment N0. 43 E. Reserved for Proposed Amendment flo. 83 Supplement 1, Revision 2 F. Reserved for Proposed Amendment No.125 G. Radioactive Liquid Effluent Dose 30 days (3.17.2)

H. Noble Gas Limits 30 days (3.18.2)

1. Radioiodine and Particulates 30 days (3.18.3)

J. Gaseous Radwaste Treatment 30 days (3.19)

K. Radiological Monitoring Program 30 days (3.22)

L. Monitoring Point Substitutions 30 days (3.22)

M. Land Use Census 30 days (3.23)

N. Fuel Cycle Dose 30 days (4.25)

0. Liquid Haldup Tanks 30 days (3.17.3)

P. Steam Generator Tube Inspection 30 days (4.17.5)

Proposed Amendment 122, Revision 1 6-12f

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t Revision 1 N i ' ' ' O.04 NUCLEAR REGULATCRY COMMISS13N '

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, A. INTRODUCTION containment. His guide describes a method acceptable to the NRC staff for implementing these General Design General Design Criteria 14, " Reactor Coolant Pressure Criteria by reducing the probability and consequences of l

i Boundary," and 31, " Fracture Prevention of Reactor steam generator tube failures through periodic inservice Coolant Pressure Boundary," of Appendix A " General inspection for early detection of defects and deteriora.

. Design Criteria for Nuclear Power Plants" to 10 CFR tion. His guide applies only to pressurized water'~ i f Part 50, "Ilcensing of Production and Utilization reactors (PWRs). He Advisory Committee on Reactor p Facilities," require that the reactor coolant pressure Safeguards has been consulted conceming this guide and l

U boundary have an extremely low probability of has concurred in the regulatory position.

abnormal leakage, of rapidly propagating failure, and of gross rupture. General Design Criterion IS, " Reactor B. DISCUSSION Co'olant System Design," requires that the reactor cwlant mtem ud mociated auxillay, control, and ne heat transfer area cf the steam genemier: in .

protection systems be designed with sufScient margin to pressurized water reactors can comprise well over 50% of !

p - ensure.that the design conditions of the reactorcoolant the area of the total primary system pressure. retaining d pressure boundary are not exceeded during any condi-tion of normal operation, including anticipated opera-boundary. The thin. walled steam generator tubing is an important part of a major barrier against fission product l tional occurrences. Furthermore, General Design Criter- release to the environment. The steam generator tubing ion 32, " Inspection of Reactor Coolant Pressure also acts as a barrier against steam release to the Boundary," requires that components that are part of containment in the event of a LOCA. To act as an the reactor coolant pressure boundary be designed to effective barrier, this tubing must be free of cracks, permit periodic inspection and testing of critical areas to perforations, and general deterioration. He design O .

assess their structural and leaktight integrity.* criteria used to establish the structuralintegrity of the steam generator tubing should also define the minimum Failure' of steam generator tubes, which can be tube wall thickness required to sustain the pressure and caused by cracking, wastage, and fretting, will release thermalloading caused by the worst postulated LOCAin radioactive materials to the secondary coolant system. combination with a safe shutdown earthquake.2 Furthermore, serious weakening of these tubes from similar causes could, in the event of a loss of coolant Inadequate control of the secondary coolant chem- l accident (LOCA), result in tube failures that would istry has been identified as one of the principal sources release the energy of the secondary system into the i .

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,As defined in Appendix A. " Seismic and Geologic Siting

, Failure is defined as full penetration of the pressure boundary Criteria for Nuclear Power Plants," to 10 CFR Part 100

with subsequent leakage. " Reactor Site Criteria."

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+4,g ,a of steam pnerator tube degradation and failure. Dere is During reactor operation, steam pnerator t'ube leaks ] '

are detected by monitoring the secondary system for evidence that excessive steam side corrosion attack l

d- occurs in restricted flow ueas that permit high local radioactivity and the presence of boron through instru.

concentration of free caustic, phosphates, and impurities ment analysis of steam and blowdown samples. Ifleaks 0' /:

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that may enter the steam generator through condenser .are present, they can usually be located by eddy current inleakap. 'Iherefore, secondary water chemistry specifi. examination of suspect tubing. Eddy current examina-tion is effective because it detects the presence of cations must reflect the limitation of the materials in the secondary system, and the supporting auxiliary chemical defect-caused variations in effective electrical conduc.

feed system must ,be designed to maintain desired tivity and/or magnetic permeability of the material being ,

4 feedwater quality to each steam pnerator. Effective tested. Because the eddy current probing technique has

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monitoring of water chemistry with in-line continuous excellent sensitivity in nonmagnetic materials, decreases analytical instrumentation supplemented by plant in effective conductivity due to a discontinuity in a tube laboratory ' sampling analysis of steam, condensate wall can be measured directly by increases in coil voltay return, and feedwater is necessary at all times during in the probe. Special eddy current probes designed for operation to ensure that water quality is not degraded scanning tubing from the inside have proved very- !

below acceptable limits by such events as condenser effective in locating defect areas in steam pnerator l inleakage or chemical feed system maloperation. tubes and for assessing the overall condition of the I

tubingin numerous operating PWRs.

Mechanical or Sow-induced vibration can cause fret-ting or fatigue damap to steam generator tubes, which Radiography is a supplemental method for inservice i could alsolead to tube failures.

inspection of steam generator tubing. Although radi- l 4

ography does not provide the speed and flexibility of l eddy current methods,it can supplement eddy current j A program of periodic inservice inspection of steam testing for defect characterization on a limited basis. !

pnerators is essential to monitor the integrity of the tubing, particularly if there is evidence of mechanical Leaking tubes, defective tubes,and tubes that exceed !

damap or progressive deterioration caused by inade. the plugging limit should be taken out of service by l quate design, martufacturing erron, or chemical imbal. plugging both ends of the tube at the tube sheet with l

- ance. Inservice inspection of steam generator tubing can welded plugs.Various methods are used for plugging and l also provide useful information regarding the nature and we!&ng. I1up may bc installed mcchsnically er explo-cause of any tube degradation, thereby asusting the sively, and welding may be performed manually, auto- J operator in taking proper and timely corrective matically,or explosively. i

' measures. .

Xp"iena has indicaud eat each suarn gertmtw Inspection and repairs of steam generator tubing in esign as amas 4g, wins, low now ateas, operating plants cause some radiation exposure to and regions that allow steam blanketing) where attack personnel. Careful pre-job planning can assist in main- and degradadon d the stearn genasta tubu mamur j taining radiation ~ exposures as low as is reasonably even if see adary water chemistry is properly main-achievable. Temporary shielding, decontamination, dMechanical damage to swam genuatw tubes may ,

I special tooling,jip and fixtures for remote inspection also occur in areas subject to flowinduced vibrations.

and repair, and other design and procedural considera- Typically .the number of tubes in these critical areas is

_t ions such as are outlined in Regulatory Guide 8.8, im 6an M of 6e td* )

"Information Relevant to Maintaining Occupational Radiation Exposure as Low as Practicable (Nuclear The usual shop examination of tubing can be consid- j ered to serve as an adequate baseline examination. An i Reactors)," should be used to the extent practical.

onsite preservice inspection of the steam generator tubing should be performed in the absence of a "Ihe recommendations in this gdide are applicable to' documented shop or field examination. For plants now current " typical" once.through and Ubend steam opunting, the initialinspection should sample tubes on pneraters that have NiCr Fe or stainless steel tubing. a random basis unless experience with similar designs The steam generator tubing is usually seamless, cold anil chemistry indicate entical areas. Subsequent inspec.

drawn, and annealed and is manufactured and tested in tions should concentrate on any entical areas identified accordance with specifications of the American Society so that most defective tubes will be found. This selection I

of Mechanical Engineers and the American Society for method can be expected to result in the ratio of tube

. d< Testing and Materials, defects found to total tubes inspected being consider-i ably higher than the ratio of defective tubes to total The initial quality of manufactured tubing is deter- tubes in the steam generator.

mined by hydrostatic, eddy current, and ultrasonic tests.

The tube-to-tube-sheet welds are inspected visually an'd

Lines indicate substantin chanses from previous issue.

by dye penetrant, then finally leak tested.

1.83 2

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6,. KEGULATORY POSITIO'N' g. The equipment used for eddy current testing should be designed so that operators may be shielded or f

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'g A program for inservice inspection of steam generator the equipment may be operated remotely to limit

'FICF tubing should be established and should include the operator exposure to radiation.

[ following:

h' Personnel engaged in data taking andinterpret-ing the results of the eddy current inspection should be

1. Access for Inspection tested and qualified in accordance with American

~ Society for Nondestructive Testing Standard

s. Steam generators of pressurized water reactors SNT-TC-IA and supplements.* l should be designed to facilitate inspection of all tubes.

L ne examinations should be performed accord.

b. Sufficient access should be provided to perform ing to written procedures.

l '

these inspections and to plug tubes as required.

! 3. BaselineInspection j .

c. Pre. job planning should be undertaken to make

a. All tubes in the steam generators should be provisions for inspections that ensure that personnel radiation exposure is maintained as low as is reasonably inspected by eddy current or alternative techniques prior

achievable. to service to establish a baseline condition of the tubing.

b. For operating plants without an initial baseline l ,

2. Inspection Equipment and Procedures inspectior., the first inservice inspection performed j ' according to regulatory positions C.4 and C.5 will define

a. Inservice inspection should include nondestrue. the baseline condition for subsequent inspections.

i tive examination by eddy current testing or equivalent

( ; techniques. De equipment should be capable oflocating c. Operating plants instituting a major change in

. and identifying stress corrosion cracks and tube. wall their secondary water chemistry (e.g., phosphate to j thinning by chemical wastage, mechanical damage, or volatile treatment) should conduct a baseline inspection i .

other causes. before resumption of power operation.

.- n

b. De inspection equipment should be senstuve 4. Sample Selection and Testing i ssure ' enough to dete:t imperfections 20% or more through ;

h.,' h e th. ;

the tube wall. Selection and testing of steam generator tubes

/ 4 should be made on the following basis:

! c. A suitable eddy current m.spection system j could consist of (1) an intemal sensing probe, (2) a a. De preservice inspection should include all the i two channel eddy current tester, (3) a viewing oscillo- tubes in the steam pnerators.

I scope, (4) a conventional twochannel strip chart j recorder and(3)a magnetic tajie data recorder. b. ' rubes for the Mspection of' opug ple ~

- . should be selected on a random basis except where (c : 1

d. Examination results and reports should be stored and maintained for the operating life of the experience in similar plants with similar secondary water chemistry indicates critical areas to be inspected.

. facility. -

c. At least 3% of the total number of tubes in

e. Standards consisting of similar as. manufactured each steam generator to be inspected should be tested ,

i steam generator tubing with known imperfections during each inspection (see regulatory positions C.3 and l

! should be used to establish sensitivity and to calibrate C.6).

j the equipment. Where practical, these standards should l include reference flaws that simulate the length, depth, d. All of the steam generators in a given plant

! ; and shape of actual imperfections that are characteristic should be inspected at the first inservice inspection.

i of past experience. Subsequent inspections may be limited to one steam i .

generator on a rotating schedule encompassing 3% of the

f. The equipment should be capable of examming total tubes of the steam generators in the plant if the

^

(% *SNT-1C-IA and Supplements, " Recommended Practice for 1 8 For U-bend designs, entry for the hot-leg side with examination Nondestructive Testing Personnel Qualification and Certifica-from the point of erstry completely around the (Rend to the tion." Copies may be obtained from the American Society for j- ".a top support of the cold leg is considered sufficient to constitute Nondestructive Testing. 914 Clucaso Avenue, Evanston, j .. a tube inspection. Illinois 60202. .

1.83-3 I,

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"- results of the first inspection indicate that all steam c. Inspections may be made coincident with

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pnerators are performing in a like manner. (Note: refueling outages or any shutdown for plant repair and ,

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\ Under some circumstances, the operating conditions in maintenance in accordance with the American Society one or more specific steam generators may be found to of Mechanical Engineers Boiler and Pressure Vessel be more severe than those in the other generators. Under Code,Section XI.*

such circumstances, the sample sequence should be modified to inspect the steam generator with the most d. If two consecutive inspections, not including

- severe conditions.) the preservice inspection, result in less than 10% of the

(, tubes with detectable wall penetration p20%) and no

e. Every inspection subsequent to the preservice significant p10%) further penetration of tubes with inspection should include all nonplugged tubes that previous indications, the inspection frequency should be ,

previously had detectable wall penetrations p20%)and extended to 40-month intervals. If it can be demon-- l should also include tubes in those areas where experi- strated through two consecutive inspections that ',

ence has indicated potential problems. previously observed degradation has not continued and no additional degradation has occurred, a 40-month '

5. Supplementary Sampling Requirements inspection interval may be initiated.

a. If the eddy current inspection pursuant to , e. Unscheduked inspections should'be conducted regulatory position C.4.d indicates any tubes with in the event of primary.to secondary leaks exceeding previously undetected imperfections of 20% or greater technical specifications, a seismic occurrence greater depth, additional steam generators, if any, should be than an operating basis earthquake,2 a loss of-coofant inspected. If previously degraded tubes exhibit signifi- accident requiring actuation of engineered safeguards, or cant p10%) further wall penetration, additional steam a major steam line or feedwater line break.

pnerators should be inspected.

b. If the eddy current inspection pursuant to 7. Acceptance Limits 5

regulatory position C.4.c indicates that more than 10%

a. 'As used in this rephtnry pide i of the insyrted tubes have detectable wall penetation p20%) or that one or more of the inspected tubes have an indication in excess of the plugging limit (see (1) Imperfection means an exception to the --

f. df regulatory position C.7.a), an additional 3% of the tubes should be inspected, concentrating on tubes in those dimensions, finish, or contour required by dra' wing or specification.

i areas of the tube sheet array where tubes withimperfec. f tions were found. In addition, the rest of the steam -

(2) Defect means an imperfection of such pnerators should be inspected according to regulatory severity that the tube is unacceptable for contin 2ed position C.4.c. service.

c. If this additionalinspection inditates that more (3) Rugging limit means the imperfection than 10% of these additionally inspected tubes have depth at or beyond which plugging of the tube must be detectable wall penetration p20%) or one or more of performed. (Note that the plugging limit is not a depth !-

these additionally inspected tubes has an indication in of penetration within the defect range but rather an .,

excess of the plugging limit, additional tubes (no less imperfection depth with conservative allowances. 'Ihese than 6% of the total tubes in the steam generator)iri the allowances include such considerations as general corro-area of the tube sheet array where tubes withimperfec- sion and measurement error.)

tions were found should be inspected. 3 (4) Rugging criteria means those calculational -

6. InspectionIntervals and analytical procedures used to arrive at the plugging (

limit. 'Ihese currently may be submitted by a licensee

a. The first inservice inspection of steam genera- for approval by NRC.

tors should be performed after 6 effective full power months but before 24 calendar months. b. If, in the inspection performed under regula-tory position C.4,less than 10% of the tubes inspected

( .

have detectable wall pentration p20%) and no tube has

b. Subsequent inservice inspections should be not less than 12 nor more than 24 calendar months after the imperfections that exceed the plugging limit defect, previous inspection. plant operation may resume.

'In au ' Inspections. previously degraded tubes that exhibit ' Copies may be obtained from the American Society of significant (>10%) further wan penetration must be included in Mechanical Engineers. United Engineering Center. 345 East 47th Street. New York, New York 10017. l the 10E 1.83 4

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"- c. If,in the' inspections performed under regula. D. IMPLEMENTATION tory position C.5, less than 10% of the total tubes f

j. b inspected have detectable wall penettstion P20%) and The purpose of this section is to provide information i OFFlf no more than three tubes exceed the plugging limit, to applicants and licensees regarding the NRC staff's plant operation may resume after required corrective plans for utilizing this regulatory guide, measures have been taken. -

d. If, in the inspections performed under regula-(" tory position C.5, more than 10% of the total tubes This guide reflects current regulatory practice.There-inspected have detectable wall penetration p20%) or fore, except in those cases in which the applicant more than three of the tubes inspected exceed the proposes an acceptable attemative method for comply.

plugging limit, the situation should be immediately ing with specified portions of the Comfrussion's regula-reported to the Commusion. In accordance with the tions, the methods described herein will be used by the facility license for resolution and approval of the NRC staff in evaluating an applicant's program for proposed remedial action. Additional sampling and more inspection of steam generator tubes.

frequent inspections may be required.

8. Corrective Measures All lealdng tubes, defective tubes, and tubes with Technical specifications for ensuring inspection as imperfections exceeding the plugging limit should be recommended in regulatory position C should be incor-plugged. porated in operating licenses.

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NUCLEAR REGULATORY COMMISSION

r. j WASHINGTON, D. C. 20656 March 14,1985 k**.*/

Docket No. 50-312 LICENSEE: Sacramento Municipal Utility District FACILITY: Rancho Seco Nuclear Generating Station

SUBJECT:

SUMMARY

OF MEETING HELD ON NOVEMBER 20, 1984 WITH SACRAMENTO MUNICIPAL UTILITY DISTRICT (SMUD) TO DISCUSS STEAM GENERATOR TUBE FAILURE OF RANCHO SEC0 NUCLEAR GENERATING STATION The Rancho Seco Nuclear Generating Station had experienced three shut downs in recent months due to steam generator failures. As a result of these failures the NRC staff requested a meeting with SMUD to discuss the cause of the failures and the licensees program for addressing the steam generator failures. The meeting was held on November 20, 1984 in Bethesda, Maryland.

The attendees list (Enclosure 1) and copies of the viewgraph used at the meeting (Enclosure 2) are attached.

The tube failures have been occuring in the lane tube area of the steam generators from the 15th support plate to the upper tube sheet. The failure occurrs initially as a small leak probably due to micro cracks that grow slowly. On cool down the cracks open up as the tube goes from compression to tension. The licensee hypothesized that the micro cracks are formed by a combination of corrosion due to concentrated chemic:ls carried by the moisture and by normal tube loadings. The cracks are then propagated by high cycle fatigue at low alternating stresses. The licensee noted that of the 31,000 I Rancho Seco Steam Generator tubes, to date 66 have been plugged and of these 6 were leaking.

The licensee described the approaches being considered to correct the problem of lane tube failures in B&W plants. Methods being considered include sleeving the lane tubes from the primary face of the upper tube sheet through the 15 tube support plate, and lane flow blockers.

The licensee then discussed the eddy current tube inspection program at Ranch i Seco. The licensee noted that during the last outage caused by a steam generator leak they went to an improved eddy current inspection program. In addition a more conservative plugging criteria was used in establishing which tubes should be plugged. As a result more tubes were plugged on both steam generators. The licensee noted that during the refueling outage scheduled for early in 1985 two tubes would be removed for examination. Finally the licensee described their secondary water chemistry control program.

The staff noted that the current Technical Specification for additional

' unscheduled inservice inspections of the steam generator is based on 1131 j

S g activity in the secondary IJ I side of the steam generators.

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MAR 251985 l

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concentration depended ThestafffurtherstatedthatthesecondarysideI}Mrytosecondaryside on the primary side coolant activity as well as pr tube leakage. Therefore primary to secondary side coolant leakage could be well in excess of the Technical Specification leakage limit for reactor operation before the Technical Specifications would require the additional unscheduled inservice inspection. The staff indicated that the current Technical Specification is not an adequate indication of tube degradation and noted that when plant shut down is required because the Technical Specification leakage limit is reached that the additional unscheduled inservice inspection should be conducted during the shut down. The licensee stated that when the plant is shut down because of tube leakage in the tube lane area, the lane tubes are inspected as well as some additional tubes on either side of the lane tubes (wedge tubes). The licensee comitted to propose a revision the Technical Specifications to specify that if the leak occurs in the lane tube area, the lane tubes and wedge tubes will be inspected. If the leak occurs in other areas, inspection will be in accordance with table 4.17-2 in the Technical Specifications. We stated that this would be acceptable.

Sincerely, Sydney Miner, Project Manager Operating Reactors Branch #4 Division of Licensing Enclosures Meeting Attendees cc: See next page 4

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MEETING

SUMMARY

DISTRIBUTION Licensee: Sacramento Municipal Utility District

Copies also sent to those people on service (cc) list for subject plant (s).

Docket File NRC PDR L PDR ORBf4 Rdg Project Manager - Sydney Miner

.1Stolz BGrimes (Emerg. Preparedness only)

OELD NSIC EJordan, IE PMcKee, IE ACRS-10 NRC Meeting

Participants:

H. F. Conrad B. D. Liaw J. H. Eckhardt Mary S. Wegner C. Y. Cheng Louis Frank -

W. V. Johnston i

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Sacramento Municipal Utility Rancho Seco, Docket No. 50-312

. . , District ,

cc w/ enclosure (s):

fj 8"jj

,, ooic1 Ealt Branch Sacramento Municipal Utility State Department of Health Services District 714 P Street, Office Building #8 6201 S Street Sacramento, California 95814 P. 0. Box 15830 .

Sacramento, California 95813 Ms. Eleanor Schwartz California State Office Sacramento County 444 North Capitol Street, N.W., Suite 305

! Board of Supervisors Washington, D.C. 20001 827 7th Street, Room 424 Sacramento, California 95814, Docketing and Service Section Office of the Secretary Mr. John B. Martin, Regional U.S. Nuclear Regulatory Commission Administrator Washington, D. C. 20555 U.S. Nuclear Regulatory Commission Resi3ntInspector/RanchoSeco Region Y c/o U. S. N. R. C.

1450' Maria Lane, Suite 210 14410 Twin Cities Road Walnut Creek, California 94596 Herald, CA 95638 Director Energy Facilities Siting Division Energf Resour.ces Conservation &

Development Commission Regional Radiation F.epresentative 1516 - 9th Street EPA Reaion IX Sacramento, California 95814 215 Fremont Street San Francisco, California 94111 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue I Bethesda, Maryland 20814

Thomas Baxter, Esq.

l Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.

Washington, D. C. 20036 1

Helen Hubbard P. O. Box 63 -

Sunol, California 94586

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o Enclosure 1 MEETING NOVEMBER 20, 1984 RANCHO SECO STEAM GENERATOR MEETING ATTENDEES Sydney Miner NRC H. F. Conrad NRC B. D. Liaw NRC L. H. Whl SMUD R. E. Marlow CONAM Jim Field SMUD R. P. Wichert SMUD C. A. Creacy B&W F. W. Kellie SMUD Ron Colombo SMUD J. H. Eckhardt NRC J. H. Taylor B&W R. W. Ganthner B&W Paul Guill Duke Power Co.

R. L. Gill Duke Power Co.

C. W. Hendrix, Jr. Duke Power Co.

Gary Abell B&W Mary S. Wegner NRC Lynn Connor Doc-Search Associates C. Y. Cheng NRC Louis Frank NRC Ron Rodriguez SMUD Jerry Delezenski SMUD W. V. Johnston NRC l

L

Enclosure 2 SMUD/NRC RANCHO SECO STEAM GENERAT0R STATUS MEETING

.b PRESENTED BY:

~

~" - - - -

-SACRAENTO. MUNICIPAL

~ '

UTILITY DISTRICT

-B&W

-B&W OWNERS GROUP

-CONAM AT NRC--BETHESDA, MD NOVEMBER 20, 1984 s

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

PURP0SE ADDRESS NRC STAFF QUESTIONS ON:

4 o RECENT HISTORY OF THE RANCHO SECO OTSG'S o SMUD MANAGEMENT POSITION TOWARDS

-INSPECTION

-REPAIR

-0PERATION

~~~ '

.....T~ ~~ 0F THE RANCHO SECO OTSG'S~~ ~~-Z '

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C- -_ _ _ . _ _ _ _ . _ _ _ . . . _ ___ .

SMUD/NRC MEETING 11-20-84 BETHESDA, MD AGENDA 4 INTRODUCTION

. SMUD

(]0 MINUTES)

- PURPOSE

- OVERVIEW

- PARTICIPANTS 4

THE ONCE-THROUGH STEAM GENERATOR--0TSG BWOG (20 MINUTES)

- 0TSG INDUSTRY OPERATING EXPERIENCE

- DESIGN a OPERATING CHARACTERISTICS

- AREAS OF SPECIAL INTEREST e

~

RANCHO SECO OTSG OPERATING $iPElilENCE ' SMUD/CONAM- - ~

(60 MINUTES)

- MECHANICAL

- NDE

~

1- - CHEMISTRY 0

DISTRICT

SUMMARY

AND ACTIONS SMUD (30 MINUTES)

- DISCUSSION 3

TUBE PLUGGING

SUMMARY

THROUGH NOVEMBER 15,1984 CATEGORIZED BY NDE CATEGORY

=_ -_ __

PLANT / GEN. 1 2 3 TOTAL 4 OVERALL CAT. 1,2,&3 TOTAL

- _ _ _ --- _-___=_

"A" A-0TSG 3 31 1 35 42 77 B-OTSG B 235 13 256 59 -

315 "B"

A-OTSG O 2 0 2 3 5 B-OTSG 3 20 2 25 17 42 "C"

A-0TSG 9 14 5 28 79 107 B-0TSG S 12 14 31 4 35 "D"

A-0TSG 6 169 27 202 4 206 B-0TSG O 82 3 85 7 92 RANCHO SECO A-0TSG 3 21 5 29 21 50 B-0TSG 3 10 13 26 8 34 ap.

A-0TSG 1 0 1~ 2 4 6 - --

B-0TSG O O O- 0 27 27 - '

"G" A-DTSG 2 3 2 7 12 19 B-0TSG O 2 1 3 6 9 43 601 87 731 293 1024 CATEGORY EXPLANATION 1 TUBE LEAK 2 EXCEEDED 40% TUBE PLUGGING CRITERIA (NDE) 3 CUSTOMER OPTION (0-40% THROUGH WALL) 4 OTHER (INSTRUMENTATION, ERROR, SHOP PLUGS, ETC.)

i

TUBE PLUGGING

SUMMARY

THROUGH NOVEMBER 15,1984 CATEGORIZED BY DAMAGE MECHANISM SECONDARY SULFUR LANE EROSION AFW PLANT / GEN. IGA REGION CORROSION HEADER OTHER TOTAL

_ = -- -

>40% CUST. OPT.

"A" A-0TSG O 6 1 24 -

46 77 B-0TSG O 11- 6 199 -

99 315

.. B "

A-0TSG O O O 2 -

3 5 B-OTSG O 5 2 17 -

18 42 "C"

A-0TSG O 7 1 6 2 91 107 B-0TSG O 8 12 5 0 10 35 "D"

A-0TSG 142 23 17 1 -

23 206 B-0TSG 70 1 O O -

21 92 RANCHO SECO A-0TSG O 16 5 1 2 26 50 B-0TSG O 8 12 0 6 8 34

.pu A-0TSG O O 'O O -

6 6 B-0TSG O O O O -

27 27 "G"

A-0TSG O O O 1 6 12 -19 B-OTSG O O O O 3 6 9 DEFINITIONS IGA - UPPER TUBESHEET MIDSPAN LANE REGION - ROWS 73,74,75,77,78,79 & 15th TSP-UTS FACE EROSION / CORROSION - 11th,12th,13th, or 14th TSP (PERIPHERY)

AFWH - 15th TSP-UTS FACE & OUTERMOST TUBES f

I L __. _ _ _ _ _ _ _ _ _

LANE REGI0N FAILURES FAILURE HYP0 THESIS

. Corrosion attack in the upper spans is due to concentrated chemical species carried by moisture

. Microcracks are formed by the combination of surface damage due to corrosion and by normal tube loadings

. Crack propagation is caused by'high cycle fatigue at low alternating stress I

db

LANE REGI0N FAILURES FACTS ,

. Moisture is present'along the lane at the upper span

. Tube samples reveal deposits and surface damace

. Surface damage has been reproduced in the laboratory by concentrated solutions of silicates and sulphates

. Micro cracking has been observed in the surface damage area

. Vibration measurements indicate low amplitude vibration

. .::.. ::n :.utc.nc..n3. A concentrated silicate:environmnet hasr _ .~u=rn been shown to-substantially-lower the~ ' - - - - ~ ~ ~ " ' ~

fatigue strength of inconel 1

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OTSG SLEEVE OBJECTIVE: Provide a new pressure boundary from primary face of upper tubesheet through 15th TSP.

DESIGN FEATURES: 80 inch long sleeve

. Mechanical Sleeve 1600 - thermally treated STATUS: Engineering Qualification Complete Prototype Tooling Developed _. _

Demonstration Sleeving Program in

- early December . . .

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RANCHO SECO OPERATIONAL TUBE INTEGRITY A OTS6 B OTSG FIRST REFUELING INSPECTED INSPECTED 9-77 8 TUBES PLUGGED No TUBES PLUGGED SECOND REFUELING INSPECTED INSPECTED 12-78 No TUBES PLUGGED No TUBES PLUGGED THIRD REFULLING INSPECTED INSPECTED 2-80 1 TUBE PLUGGED 2 TUBES PLUGGED FOURTH REFUELING INSPECTED NOT INSPECTED 3-81 No TUBES PLUGGED TUBE LEAK NOT INSPECTED 1 LEAKER PLUGGED 5-81 3 ADDITIONAL TUBES PLUGGED aux. FEEDWATER MODS 6 TUBES PLUGGED 6 TUBES PLUGGED 6-82 DUE To PROXIMITY DUE To PROXIMITY 4 ADDITIONAL TUBES 3 ADDITIONAL TUBES PLUGGED PLUGGED TUBE LEAK 1 LEAKER PLUGGED NOT INSPECTED 11-82 1 ADDITIONAL TUBE lj PLUGGED FIFTH REFUELING INSPECTED INSPECTED 4-83 No TUBES PLUGGED 2 TUBES PLUGGED

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TUBE LEAK 1 LEAKER PLUGGED NOT INSPECTED 9-83 TUBE LEAK NOT INSPECTED 1 LEAKER PLUGGED 7-84 TUBE LEAK NOT INSPECTED 1 LEAKER PLUGGED 8-84 TUBE LEAK 1 LEAKER PLUGGED INSPECTED 10-84 12 ADDITIONAL TUBES 12 TUBES PLUGGED PLUGGED

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SUMMARY

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V ECT TECHNIQUES USED AT RANCHO SECO

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SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO .

NUCLEAR' POWER PLANT 7, -

OCTOBER *1984 .

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,8/G A 681 Tubes !

8/G B 674 Tubes EXAMINATION AREA --

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: INSTRUMENT: MlZ-18 Digital Eddy Current instrument !

PROBE SIZE: .520" ' Diameter EXAMINATION FREQUENCIES AND MO' DES l 800 kHz Differential and Absolute 400 kHz Differential and Absolute 200 kHz Differential and Absolute 45 kHz Differential and Absolute I

I!L_. . . .. _ _ _ 8 X 1 COILS .. _. __ .. _ . ... ..._ . _ .._. . . _ _

lNSTRUMENTf MlZ- 18:Digitil' Eddy CuWint: Inst'ruininT~ ~

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PROBE SIZE: .8C .520 "

l EXAMINATION FREQUENCIES AND MO' DES 400 kHz Absolute '

100 kHz Absolute

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PLUGGING CRITERIA.

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PLUGGING LIST 8 DEVELOPED FROM FOLLOWING : -

f. -

1. Discontinuity signals in excess of 39% thru wall as measured by either bobbin coil or 8 x 1 coils examination.

2. Discontinuity signals at edge of. support plate

> or upper tube . sheet inhich were short and did not appear as wear and werd crack like only in either bobbin coil'or 8 iicoils ~ ~ ~ ~

examinations.

i NOTE:

Plugging Ests were developed with a' conservative attitude to help prevent future leaks. '

9

.i CONAM INSf8ECTION $acramento Municipal Utility District Rancho Seco. Unit 1 Steam Generators A & 8 8/G A ,

October 1984 -

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CONAM M88W3 TION Sacramento Itunicipal Utility District Rancho Seco. Unit 1 Steam Generators A & B 8/G A -

October 1984 i

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wucos vants uww

.f..,

OTSG SECONDARY SIDE INSPECTIONS

1. LAST INSPECTION 3-17-84. l 1

2. OBSERVED LEVELS OF DEPOSITS ARE SIMILAR TO THOSE SEEN AT OTHER B&W OPERATING PLANTS WITH SIMILAR SERVICE LIFE. .

.. . ;r. cz.- -- 3;. THEr6ENERAL COMPOSITION 0F DEPOSITS IS-THErt

-_ r;---- :---

SAME.AS.0BSERVED AT.ALL B&W PLANTS. . _ _ _ . . . _ . .._..

4

4. IRON IS THE MAJOR CONSTITUENT.

5. NICKEL WAS FOUND IN SOMEWHAT HIGHER LEVELS (1-12%). 1981 LEVELS, 0.5-4%. HIGHEST LEVEL AT AFW N0ZZLE PENETRATION:

A. NICKEL-CHROMIUM RATIO IS 2.25:3.4. RATIO IN INCONEL IS 5:1. =

.B: THIS SUGGESTS INCONEL CORROSION IS NOT

! THE SOLE SOURCE OF NICKEL.

l

C. MAY BE FROM THE FEEDWATER SYSTEM.

6. LEVELS OF MANGANESE AND MOLYBDENUM (~0.5%

RESPECTIVELY) ARE CONSISTENT WITH VALUES 1

} FOUND IN 1981 INSPECTION.

I

}

L -

7

. r., ,

.. OTSG SECONDARY SIDE INSPECTIONS !

7. ~90% OF ALL SULFUR (S0-2) LEVELS REPORTED FROM B&W OPERATING PLANTS ARE LESS THAN OR EQUAL TO 1.0 MILLIGRAMS PER SQUARE FOOT.

8. ' SULFUR. VALUES REPORTED FOR RANCHO SECO .. .

(0.21 TO i.32 MILLIGRAM PER SQUARE FOOT)

.ARE : CONSISTENT-WITH OTHER PLANTS. LEVEL.- : --

. . .r: . - : . :::

. -0F . SULFUR -I-N-DEROSITS HAS REMAINED CONSTANT. ..-. . - -.... __ _.

9. THE AMOUNT OF DEPOSITS FOUND IN BOTH OTSG'S -

ARE APPR0XIMATELY EQUIVALENT.

10. ALL PERIPHERAL (FLOW PATHS) BROACHED OPENINGS ARE 95% CLEAR AT THE 9TH, 10TH, AND 1STH TUBE SUPPORT PLATES. SUPPORT PLATES BELOW TIE 9TH _

WERE NOT EXAMINED.

11. SILICON AS SIO2 HAS DECREASED FROM 1977, 5.5%,

TO 1981, 0.5%, FROM DEPOSIT SAMPLES IN THE LOWER PORTION OF TIE OTSG.

e z

.g.'*

OTSG SECONDARY SIDE INSPECTIONS

12. PRIMARY SIDE DEPOSITS, JULY 1982:

CHLORIDE 70-110 MICR0 GRAM PER FOOT SQUARE SOLFATE <275 MICROGRAM PER FOOT SQUARE OTHER B&W PLANTS RANGED FROM <70 TO 340 .

MICR0 GRAM PER FOOT SQUARE CHLORIDE AND

.- .*275-T0 5600 MICR06 RAM PER FOOT SQUARE --- - - - - - - -

SULFATE. .

4 i

a 6

L _

, .; a, r DISTRICT ACTIONS / PROGRAMS TO MINIMl2E TUBE LE NEW ACTIONS 8

INSPECTION UPGRADES

- TECHNICAL SPECIFICATION CHANGES

- STATE-0F-THE-ART TECHNIQUES 8 CHEMISTRY UPGRADES

- DISPOSABLE RESINS

- ACTION TO MINIMIZE AIR INLEAKAGE -

- POSSIBLE WET LAYUP PROCEDURE CHANGES S

CONSERVATIVE PLUGGING CRITERIA 8

PULL TUBE FOR EXAMINATION i

ONGOING ACTIONS / PROGRAMS 8

L B8W OWNERS GROUP S' TEAM GENERATOR PROGRAM i

e B8W 0WNERS GROUP NDE COMMITTEE 8

EPRI STEAM GENERATORS GROUP II s e

> SULFUR ANALYSIS AND MONITORING PROGRAM ,

n e e B

TUBE SLEEVING TRIAL PROGRAM

!1 0 SECOND PARTY CHEMISTRY PROGRAM OVERVIEW q

m

j I; 5,,,

t....

_S U M M A R Y e o B&W HAS THE BEST INTEGRITY .. RECORD PWR MANUFACTURERS.

o PROBLEM AREAS IN THE OTSG HAVE BEE DOCUMENTED.

o THE NUMBER OF TUBE LEAKS IS CONSISTEN EXPERIENCE OF OTHER B&W UNITS.

o SECONDARY CHEMISTRY IS ROUTINELY KEPT INDUSTRY GUIDELINE LIMITS.

3 o THE DISTRICT HAS DEVELOPED AND IMP MULTIFACETED SOLUTION TO OTSG AVAIL 2

ML20128G361

Contents

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SUMMARY

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Participants:

SUMMARY

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SUMMARY

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