ML20141F501

Proposed Tech Specs 3.1.C, Reactor Coolant Sys Leakage & 4.12, SG Tube Surveillance

ML20141F501
Person / Time
Site:	Prairie Island
Issue date:	05/15/1997
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20141F481	List: ML20141F475 ML20141F485 ML20141F501
References
GL-95-05, GL-95-5, NUDOCS 9705210343
Download: ML20141F501 (44)

Text

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F Exhibit B

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Prairie Island Nuclear' Generating Plant i

License Amendment Request Dated May'- 15, 1997 i

Proposed Changes Marked Up

-On Existing Technical Specification Pages i

4 n

. Exhibit B consists of existing and new Technical Specification pages with the proposed changes highlighted on those pages. The pages affected by this License Amendment Request are listed below:

t TS-vi TS.3.1-9 TS.4.12-3 TS.4.12-4 TS.4.12-5 I

TS.4.12-6 (new)

TS.4.12-7 (new)

B.3.1-7 B.4.12-1 j'

B.4.12-3 B.4.12-4 (new) 4 l

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l 9705210343 970515 4

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PDR ADOCK 05000282 P

PDR y

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i TS-vi l

REV 432 1/2':/96 TABLE OF CONTENTS (Continued)

TS SECTION TITLS PAGE 4.12 Steam Generator Tube Surveillance TS.4.12-1 A. Steam Generator Sample Selection and TS.4.12-1 Inspection B. Steam Generator Tube Sample Selection TS.4.12-1 and Inspection C. Inspection Frequencies TS.4.12-3 D. Acceptance Criteria TS.4.12-4 E. Reports TS.4.12-4]

4.13 Snubbers TS.4.13-1 4.14 Control Room Air Treatment System Tests TS.4.14-1 4.15 Spent Fuel Pool Special Ventilation System TS.4.15-1 4.16 Deleted 4.17 Deleted 4.18 Reactor Coolant Vent System Paths TS.4.18-1 A. Vent Path Operability TS.4.18-1 B. System Flow Testing TS.4.18-1 4.19 Auxiliary Building Crane Lifting Devices TS.4.19-1 4

1

d TS.3.1-9 REV 91 10/27/89 i

3.1.C.2 e.

If the total reactor coolant system to secondary coolant system leakage through hth-lanfEbneisteam generators of a unit exceeds 2.0 gclier per =ir.ute ' gems $,0]ir,a11bssipptZdsf((GPD)l, within one hour initiate action to place the unit in HOT SHUTDOWN and be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in COLD SHUTDCWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and perform an inservice steam generator tube inspection in accordance with Technical Specification 4.12.

3.

Pressure Isolation Valve Leakane Leakage through the pressure isolation valves shall not exceed the maximum allowable leakage specified in Specification 4.3 when reactor coolant system average temperature exceeds 200*F.

If the maximum allowable leakage is exceeded, within one hour initiate the action necessary to place the unit in HOT SHUTDOWN, and be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

1

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TS.4.12-3 l

REV 30 S/25/78 i

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Inspection Freauencies-The above required in-service inspections of r. team generator tubes shall be performed at the following frequencies:

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1. In-service inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.

If 4

two consecutive inspections following service under AVT condition, not including the preservice inspection, result in all inspection results falling into the C-1 categre6?y or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

2. If the results of the inservice inspection of a steam generator conducted in accordance with Table TS.4.12-1 at 40 month intervals fall

[

in CateSory C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.12.C.1; the interval may then be extended to a maximum of once per 40 months.

3, Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table TS.4.12-1 during the shutdown subsequent to any of the following conditions.

(a)

Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.1.C.6.

(b)

A seismic occurrence greater than the Operating Basis Earthquake.

l (c)

A loss-of-coolant accident requiring actuation of the engineered safeguards.

(d)

A main steam line or feedwater line break.

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TS.4.12-4 j

REV 118 5/15/95.

D.

Acceptance Criteria s.

1. As used in this Specification:

(a)

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 below 20% of the nominal tube wall thickness, if detectable, may be considered as-imperfections..

(b)

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

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(c) -Derraded Tube means a tube containing imperfections 220% of the

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nominal wall thickness caused by de5radation.

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I (d) t Derradation means the percentage of the tube wall thickness 4

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affected or removed by degradation.

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4 (e)

Defect means an imperfection of such severity that it exceeds the 4

repair limit. A tube containing a defect is defective.

l (f)

Repair Limit means the imperfection depth at or beyond which the i

tube shall be removed from service by plugging or repaired by sleeving because it may become unserviceable prior to the next i

inspection and is equal to 50% of the nominal rub. usil thickness, i

If significant general tube thinning occurs, this criteric will be reduced to 40% wall penetration.

This definition does not apply l

to the portion of the tube in the tubesheet below the F* distance-provided the tube is not degraded (i.e., no indications of cracks) 2 within the F* distance for F* tubes.

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hpplyT t oTtubb rauppor tTplatsE intWsed tiEnN fo rlwhichithsf'fsitas,ey$

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Unserviceable describes the condition of a tube if i'c leaks or i

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contains a defect large enough to affect its structural integrity i

in the event of an Operating Basis Earthquake, a loss of-coolant accident, or a steam line or feeduater line break.

(h)

Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold Icg.

3 (i)

Sleevine means that tube sleeving is permitted only in areas where the sleeve spans the tubesheet area and whose lower joint is at j

the primary fluid tubesheet face.

j (j)

F* Distance is the distance from the bottom of the hardroll i

transition toward the bottom of the tubesheet that has been conservatively determined to be 1.07 inches (not including eddy l

current uncertainty).

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1 TS.4.12-5 j

REV L18 5/15/95 j

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(k). F* Tube is a tube with degradation, below the F* distance, equal to or greater than 40%, and not degraded (i.e., no indications of cracking within the F* distance.

2. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair by sleeving all tubes exceeding the reps.ir limit and all tubes containing through-wall cracks or-

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classify as F* tubes) required by Table TS.4.12-1.

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TS?4]1.23 REyj E.

Feports

1. Following each in-service inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.

2. The results of steam generator tube inservice inspections shall be included with the summary reports of ASME Code Section XI inspections submitted within 90 days of the end of each refueling outage.

Results of steam generator tube incervice inspections not associated with a refueling outage shall be submitted within 90 days of the completion of the inspection.

These reports shall include: (1) number and extent of tubes inspected, (2) location and percent of wall-thickness penetration for each indication of an imperfection and (3) identification of tubes plugged or sleeved.

3. Results of steam generator tube inspections which fall into Category C-3 require notification to the Commission prior to resumption of plant operation, and reporting as a special report to the Commission within 30 days.

This special report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measu es taken to prevent recurrence.

4. The results of inspections performed under Specification 4.12.B for all tubes that have defects below the F* distance, and were not plugged, shall be reported to the Commission within 15 days following the inspection.

The report shall include:

a.

Identification of F* tubes, and b.

1.ocation and extent of degradation.

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B.3.1-7 REV '^r e*

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3.1 ftEACTOR C001 ANT SYSTEM Bases continued l

C.

Reactor Coolant System Leakage Leakage.from the reactor coolant system is collected in the containment or by other systems. These systems are the main steam system, conden-sate and feedwater system and the chemical and volume control system.

+

J Detection of leaks from the reactor coolant system is by one or more of the followins (Reference 1):

1.

An increased amount of makeup water required to maintain normal level in the pressurizer.

2.

A high temperature alarm in the leakoff piping provided to collect reactor head flange leakage.

i 3.

Containment sump water level indication.

o 4

4.

Containment pressure, temperature, and humidity indication.

If there is significant radioactive contamination of the reactor coolant, the radiation monitoring system provides a sensitive indica-tion of primary system leakage. Radiation monitors which indicate primary system leakage include the containment air particulate and gas monitors, the area radiation monitors, the condenser air ejector monitor, the component cooling water monitor, and the steam generator blowdown monitor (Reference 2).

I AThs}hJaplips1leakrateglimigof1gpmcorrespondssdtoathroughwall crack less than 0.6 inches long based on test data.

Steam generator tubes having a 0.6-inch long through-wall crack have been shown to resist failure at pressures resulting from normal operation, LOCA, or steam line break accidents (Reference 3).

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• • Pprpsed {idQery@yt[wilkbsdAsstedMndithejlant[shdt[dhwr(in j a timelylmannery Specification 3.1.C.3 specifies actions to be taken in the event of failure or excessive leakage of a check valve which isolates the high pressure reactor coolant system from the low pressure RHR system piping.

References i

1.

USAR, Section 6.5 2.

USAR, Section 7.5.1 3.

Testimony by J Knight in the Prairie Island public hearing on

)

January 28, 1975, pp 13-17.

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B.4.12-1 REV 91 10/??/99 4.12 STEAM GENERATOR TUBE SURVEILIANCE Bases The Surveialance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

In-service inspection of steam generator tubing is essential in order to maintain 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 in service conditions that lead to corrosion.

In-service 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.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameters found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these parameters, localized corrosion would most likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator leakage between the primtry coolant system and the secondary coolant system (primary-to-secondary leakage -

1,0-gim150f gs116'ns @sriddy;fr6aC6nsist'sas%4insystor). Ilist6V1Eally?

gracks nggaddressedj byjyoltagejbasedhiteptjatsjepaityriteriajand having a primary-to-secondary leakage less than 1.0 gpm.(1440]ga11cusipef day)]during operation w444would have an adequate margin of safety against failure due to loads imposed by design basis accidents (Reference 1).

Operating plante ha ec0peratilonhl*exper;ienssihss demonstrated that primary-to-secondary leakage as low at 0.1 gp:$]gsll6nilpsF[dsf will be detected by secondarf{ system $ radiation monitors cf ctec generater M cede c. Tofprovide.defenss[if(depth]forf th6W615fgsjbaseld?repaif criteriagbleakage in exceos of 1.0 ap=150jal.lonsiper[daygfromjonefsteam generator will require plant shutdown and an unscheduled eddy current inspection, during which the leaking tubes will be located and plugged or sleeved.

Wastage-type defects are unlikely with proper chemistry treatment of secondary coolant. However, even if this type of defect occurs it will be found during scheduled in-service steam generator tube inspections.

i i

Repair or plugging will be required of all tubes with imperfections that could develop defects having less than the minimum acceptable wall thickness prior to the next inservice inspection which, by the definition of Specification 4.12.D l (f), is 50% of the tube or sleeve nominal wall thickness. Wastage type defects having a wall thickness greater than 0.025 inches will have adequate margins of safety against failure due to loads imposed by normal plant operation and design basis accidents (Reference 1).

Steam generator tube inspections of operating i

B.4.12-3 REV 11? S/15/95 4.12 STEAM CENERATOR TUBE SURVEILLANCE Bases continued Descriptions of other future sleeve designs shall be submitted to the NRC for review and approval prior to their use in the repair of degraded steam generator tubes. The submittals related to other sleeve designs shall be made at least 90 days prior to use.

I,gh e ' Suotidt t' Pl a to "Repa i~r7Li Qt; The" voltage-based'repsir limits of Specificatiod^4'.12;D:3 implemedt' the guidance in Generic' Letter 95-05 and are applicable only to Westinghouse'-

designed steam generators with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections., The voltage-bssed repair, limits are not applicable to other forms of steam generator tube degradation nor are they applicable to ODSCC that occurs'at other locations within the steam generator. Additionally, the repair criteria apply only to indications where the degradat. ion mechanism is dominantly axial DDSCC with no significant cracks extending outside the thickness of the support plate. Refer to Generic Letter,95 05 for additional description of the degradation morphologyj Implementation 7of5Sp~scifi6stiddf4!12 D137 req 01fesis?dsriVatibHT6fiths voltsage; atructuralhl'isitifrom$thef burstfWirsus(v01tage[empisidhi correlation sndtthenIthn subsequend derivation GfJtheivoltage nep~aiK limit f r'om? the fs t fucttikalilimit1(whichl isj thendimplemente di;bysthis " ' '

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References 4

1.

Testimony of J Knight in the Prairie Island Public Hearing on 1/28/75 2.

Testimony _of L Frank in the Prairie Island Public Hearing on 1/28/75 f

-n..

e n

Exhibit C Prairie Island Nuclear Generating _ Plant License Amendment Request Dated May 15, 1997 Revised Technical Specification Pages Exhibit c' consists of revised and new pages for'the Prairie Island Nuclear Generating Plant Technical Specifications with the proposed changes

' incorporated.

The revised and new pages are listed below:

'TS-vi

'TF.3.1-9.

TS.4.12-3 TS.4.12-4 TS.4.12-5 TS.4.12-6 TS.4.12-7 B.3.1-7 B.4.12-1 4

B.4.12-3 i

B.4.12-4 1

J.

I e

4 4

i I

l 4

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l.l'

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TS-vi REV l

TABLE OF CONTENTS (Continued)

TS SECTIM I.11LE PAGE i

l 4.12 Steam Generator Tube Surveillance TS.4.12-l' A. Steam Generator Sample-Selection and TS.4.12-1 Inspection

.B. Steam Generator Tube Sample Selection TS.4.12-1 and Inspection.

C. Inspection Frequencies

.TS.4.12-3.

D. Acceptance Criteria TS.4.12 4 E. Reports TS.4.12 7-l 4.13 Snubbers TS.4.13-l' i

4.14 Control Room Air Treatment System Tests TS.4.14-1 4.15. Spent Fuel Pool Special Ventilation System TS.4.15-1 4'16 Deleted 4.17 Deleted 4.18 Reactor Coolant Vent System Paths TS.4.18-1

'I A. Vent Path Operability TS.4.18-1 B. System Flow Testing TS.4.18-1 4.19 Auxiliary Building Crane Lifting Devices TS.4.19-1.

B

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TS.3.1-9 REV 3.1.C.2

e. If the total reactor coolant system to secondary coolant system leakage through any one steam generator of a unit exceeds 150 gallons per day (GPD), within one hour initiate action to place the unit in HOT SHUTDOWN and be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and perform an inservice steam generator tube inspection in accordance with Technical Specification 4.12.

3.

Pressure Isolation Valve Leakage Leakage through the pressure isolation valves shall not exceed the maximum allowable leakage specif!ed in Specification 4.3 when reactor coolant system average tempetature exceeds 200*F.

If the maximum allowable leakage is exceeded, within one hour initiate the action necessary to place the unit in HOT SHUTDOWN, and be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

i I

TS.4.12-3 i

REV

5. Indications left in service as a result of application of tube support j

plate voltage-. based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.

6. Implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot leg and i

cold leg tube support plate intersections down to the lowest cold leg l

i l

tube support plate.with known outside diameter stress corrosion l

cracking (ODSCC) indications. The determination of the lowest cold leg l

4 tube support plate intersections having ODSCC indications shall be I

based on the performance of at least a 20 percent random sampling of.

tubes inspected over their full length.

l l

C.

J.nspection Freauencies-The abovo required in-service inspections of steam generator tubes shall be performed at the following frequencies:

1. In-service inspections shall be' performed at intervals of not less than l

12'nor more than 24 calendar months after the previous inspection.

If two consecutive inspections following service under AVT condition, not including the preservice inspection, result in all inspection results

+

falling into the C-1 category or if two consecutive inspections l

demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be E

extended to a maximum of once per 40 months.

2. If the results of the inservice inspection of a steam generator i

conducted in accordance with Table TS.4.12-1 at 40 month intervals fall 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 the subsequent inspections satisfy the criteria of Specification 4.12.C.1; the interval may then be extended to a maximum of once per 40 months.

3. Additional, unscheduled inservice inspections shall be performed on l

each steam generator in accordance with the first sample inspection i

specified in Table TS.4.12-1 during the shutdown subsequent to any of j

the following conditions.

l (a)

Primary-to-secondary tube leaks (not including leaks originating i

from tube-to tube sheet welds) in excess of the limits of j

Specification 3.1.C.6.

I (b)

A seismic occurrence greater than the Operating Basis Earthquake.

(c)

A loss-of coolant accident requiring actuation of the engineered

?

safeguards.

(d)

A main steam line or feedwater line break.

]

i l

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7 TS.4.12-4 f

REV D.

Accentance Criteria'

1. As used in this Specification:

r (a)

Imperfection means an exception to the dimensions, finish or j

contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the

[

4 nominal. tube wall thickness, if detectable, may be considered as imperfections.

(b).Derradation means a service-induced cracking, wastage, wear or

.{

general corrosion occurring on either inside or outside of.a tube, i

4 f-(c)

Degraded Tube means a tube containing imperfections it20% of the nominal wall thickness caused by degradation.

(d)

% Degradation means the percentage of the tube wall thickness f

affected or removed by degradation.

l (e)

Defect means an imperfection of such severity that it exceeds the l

repair limit. A tube containing a defect is defective.

I a

(f)

Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by l

sleeving because it may become unserviceable prior to the next

.[

inspection and is equal to 50% of the nominal tube wall thickness.

4 If significant general tube thinning occurs, this criteria will be i

reduced to 40% wall penetration.

This definition does not apply to the portion of the tube in the tubesheet below the F* distance f

provided the tube is not degraded (i.e., no indications of cracks) i within the F* distance for F* tubes. This definition does not

[

apply to tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to Specification j

4.12.D.3 for the repair limit applicable to these intercections.

j (g)

Unserviceable describes the condition of a tube if it leaks or f

contains a defect large enough to affect its structural integrity i

in the event of an Operating Basis Earthquake, a loss-of-coolant i

accident, or a steam line or feedvater line break.

(h)

Tube Insvection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

(i)

Sleevine means that tube sleeving is permitted only in areas where the sleeve spans the tubesheet area and whose lower joint is at l

-the primary fluid tubesheet face.

i (j)

F* Distance is the distance from the bottom of the hardroll I

transition toward the bottom of the tubesheet that has been conservatively' determined to be 1.07 inches (not including eddy current uncertainty).

I J

l l

j l

TS.4.12-5 REV

)

i (k)

F* Tube is a tube with degradation, below the F* distance, equal to or greater than 40%, and not degraded (i.e., no indications of cracking) within the F* distance.

I

2. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair by sleeving all tubes exceeding the repair limit and all tubes containing through-wall cracks or classify as F* tubes) required by Table TS.4.12-1.

3. Tube Support Plate Repair Limit is used for the disposition of a steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator serviceability as described below; a.

Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 3 volts will be allowed to remain in service, b.

Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts, will be repaired or plugged, except as noted in Specification 4.12.D.3.c below, c,

Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit, may remain in service if a rotating pancake coil (or comparable examination technique) inspection does not detect degradation.

Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit will be plugged or repaired, t

r

.m

TS.4.12-6 REV d.

If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instecd of the limits in Specifications 4.12.D.3.a. b and c.

The mid-cycle repair limits are determined from the following equations:

V,gt Va=

1.0+NDE+Grl\\

CL i

~

-2.0)

Vm=Vm (Vm where:

Vma - upper voltage repair limit Vua - lower voltage repair limit Vgma - mid-cycle upper voltage repair limit based on time into cycle vnua - mid-cycle lower voltage repair limit based on V na and a

time into cycle at - length of time since last scheduled inspection during which Vaz and Vua were implemented CL - cycle length (time between two scheduled steam generator inspections)

V3t - structural limit voltage Gr - average growth rate per cycle length NDE - 95 percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by the NRC)

Implementation of these mid-cycle repair limits should follow the same approach as described in Specifications 4.12.D.3.a, b and c.

Note: The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.

TS.4.12-7 REV E.

Reports

1. Following each in-service inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.

2. The results of steam generator tube inservice inspections shall be included with the summary reports of ASME Code Section XI inspections submitted within 90 days of the end of each refueling outage. Results of steam generator tube inservice inspections not associated with a refueling outage shall be submitted within 90 days of the completion of the inspection.

These reports shall include: (1) number and extent of tubes inspected, (2) location and percent of wall-thickness penetration for each indication of an imperfection and (3) identification of tubes plugged or sleeved.

3. Results of steam generator tube inspections which fall into Category C-3 require notification to the Commission prior to resumption of plant operation, and reporting as a special report to the Commission within 30 days. This special report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

4. The results of inspections performed under Specification 4.12.B for all tubes that have defects below the F* distance, and were not plugged, shall be reported to the Commission within 15 days following the inspection. The report shall include:

a.

Identification of F* tubes, and b.

Location and extent of degradation.

5. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:

a.

If estimated Icakage based on the projected end-of-cycle (or if a

not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle, b.

If circumferential crack-like indications are detected at the tube support plate intersections, c.

If indications are identified that extend beyond the confines of the tube support plate, d.

If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

e.

If the calculated conditional burst probability based on the projected end of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2, notify the NRC and provide an assessment of the safety significance of the occurrence.

B.3.1-7 REV 3.1 REACTOR COO 1 ANT SYSTEM Bases continued C.

Reactor Coolant System Leakage Leakage from the reactor coolant system is collected in the containment or by other systems.

These systems are the main steam system, conden-sate and feedwater system and the chemical and volume control system.

Detection of leaks from the reactor coolant system is by one or more of the following (Reference 1):

1.

An increased amount of makeup water required to maintain normal level in the pressurizer.

2.

A high temperature alarm in the leakoff piping provided to collect reactor head flange leakage.

3.

Containment sump water level indication.

4.

Containment pressure, temperature, and humidity indication.

If there is significant radioactive contamination of the reactor coolant, the radiation monitoring system provides a sensitive indica-tion of primary system leakage. Radiation monitors which indicate primary system leakage include the containment air particulate and gas monitors, the area radiation monitors, the condenser air ejector monitor, the component cooling water monitor, and the steam generator blowdown monitor (Reference 2).

The historical leak rate limit of 1 gpm corresponded to a through wall crack less than 0.6 inches long based on tent data.

Steam generator tubes having a 0.6-inch long through-vall crack have been shown to resist failure at pressures resulting from normal operation, LOCA, or steam line break accidents (Reference 3).

The leakage limits incorporated into Specification 3.1.C for implementation of the Steam Generator Voltage Based Alternate Repair Criteria are more restrictive than the standard operating leakage limits and are intended to provide an additional margin to accommodate a crack which might grow at a greater than expected rate or unexpectedly extend outside the thickness of the tube support plate.

Hence, the reduced leakage limit, when combined with an effective leak rate monitoring i

program, provides additional assurance that should a significant leak be experienced in service, it will be detected, and the plant shut down in a timely manner.

Specification 3.1.C.3 specifies actions to be taken in the event of failure or excessive leakage of a check valve which isolates the high pressure reactor coolant system from the low pressure RHR system piping.

References 1.

USAR, Section 6.5 2.

USAR, Section 7.5.1 3.

Testimony by J Knight in the Prairie Island public hearing on January 28, 1975, pp 13-17.

B.4.12-1 REV 4.12 STEAM GENERATOR TUBE SURVEILIANCE 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 program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

In-service inspection of steam generator tubing is essential in order to maintain 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 in-service conditions that lead to corrosion.

In-service 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.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameters found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these parameters, localized corrosion would most likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage -

150 gallons per day from one steam generator). Historically, cracks not l

addressed by voltage-based alternate repair criteria and having a primary-to-secondary leakage less than 1.0 gpm (1440 gallons per day) during operation would have an adequate margin of safety against failure due to loads imposed by design basis accidents (Reference 1).

Operational experience has demonstrated that primary-to-secondary leakage as low as 5 gallons per day will be detected by secondary system radiation monitors.

To provide defense in depth for the voltage based repair criteria, leakage in excess of 150 gallons per day from one steam generator will require plant shutdown and an unscheduled eddy current inspection, during which the leaking tubes will be located and plugged or sleeved.

Wastage-type defects are unlikely with proper chemistry treatment of secondar.y coolant. However, even if this type of defect occurs it will be found during scheduled in-service steam generator tube inspections.

Repair or plugging will be required of all tubes with imperfections that could develop defects having less than the minimum acceptable wall thickness prior to the next inservice inspection which, by the definition of Specification 4.12.D.l.(f), is 50% of the tube or sleeve nominal wall thickness. Wastage type defects having a wall thickness greater than 0.025 inches will have adequate margins of safety against failure due to loads imposed by normal plant operation and design basis accidents (Reference 1).

Steam generator tube inspections of operating i

B.4.12-3 REV i

4.12 STEAM CENERATOR TUBE SURVEILLANCE Bases continued e

9 Descriptions of other future sleeve designs shall be submitted to the NRC for review and approval prior to their use in the repair of degraded steam generator tubes. The submittals related to other sleeve designs shall'be made at least 90 days prior to use.

Tube Support Plate Repair Limit The voltage-based repair limits of Specification 4.12.D.3 implement the guidance in Generic Letter 95-05 and are applicable only to Westinghouse-designed steam generators with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections.

The voltage based repair limits are not applicable-to other forms of steam r

generator tube degradation nor are they applicable to ODSCC that occurs at other' locations within the steam generator. Additionally, the repair criteria apply only to indications where the degradation mechanism is dominantly axial ODSCC with no significant cracks evtending outside the thickness of the support plate. Refer to Generic Letter 95-05 for additional description of the degradation morphology.

Implementation of Specification 4.12.D.3 requires a derivation of the voltage.sttuctural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this i

surveillance).

The voltage structural limit is the voltage from the burst pressure / bobbin voltage correlation, at the 95 percent prediction interval curve reduced to account for the lower 95/95 percent tolerance bound for tubing caterial properties at 650*F (i.e., the 95 percent LTL curve).

The volta 6e structural limit must be adjusted downward to account for potential flaw growth during an operating interval and to account for NDE uncertainty.

The upper voltage repair limit; Vma, is determined from the structural voltage limit by applying the following equation:

i Vmu.

• Vst - Vo, - Vgg where Va, represents the' allowance for flaw growth between inspections and Vmm represents the allowance for potential sources of error in the measurement of the bobbin coil voltage.

Further discussion of the assumptions necessary to determine the voltage repair limit are discussed i

in Generic Letter 95-05.

i The mid-cycle equation in Specification 4.12.D.3 should only be used

)

'during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.

l i

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B.4.12-4 REV 4.12 p] LAM GENERATOR TUBE SURVEILIANCE Bases continued Specification 4.12.E.5 implements several reporting requirements recommended by Generic Letter 95-05 for situations which the NRC wants to be notified prior to returning the steam generators to service.

For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle voltage distribution (refer to Generic Letter 95-05 for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to returning the steam generators to service. Note that if leakage and conditional burst probability _were calculated using the measured EOC voltage distribution for the purposes of addressing Generic Letter 95-05 Section 6.a.1 and 6.a.3 reporting criteria, then the results of the projected EOC voltage distribution should be provided per Generic Letter 95-05 Section 6.b (c) criteria.

Reference 2 i

i 1.

Testimony of J Knight in the Prairie Island Public Hearing on 1/28/75 2.

Testimony of L Frank in the Prairie Island Public Hearing on 1/28/75 j

Exhibit D Prairie Island Nuclear Generating Plant License Amendment Request Dated May 15,199'l Plan for Implementation of the Voltage-Based Repair Criteria k

I b

I J

i Exhibit D Pcg31 PLAN FOR IMPLEMENTATION OF VOLTAGE-BASED REPAlR CRITERIA INTRODUCTION This exhibit describes the Prairie Island plan for implementation of the voltage-based repair criteria and the requirements of Generic Letter 95-05 (Reference 1). Sections 1 to 6 of Attachment 1 to Generic Letter 95-05 are addressed herein to summarize the Prairie Island plan, which complies with the requirements of Generic Letter 95-05.

Clarifications to the Section 3 Inspection Criteria are identified in the corresponding section of this attachment. These clarifications relate to implementation of the bobbin probe variability requirements (Section 3.c.2), the probe wear requirements (Section 3.c.3) and the use of alternate probes to the rotating pancake coil (RPC) probe.

PRAIRIE ISLAND DESIGN FEATURES Prairie Island is a Westinghouse 2-loop pressurized water reactor plant with Model 51 steam generators featuring carbon steel drilled hole tube support plates and 7/8 inch diameter Alloy 600 tubing.

Section 1:

APPLICABILITY The repair criteria will be applied to predominantly axial ODSCC indications at tube-tU-tube support plate (TSP) intersections (within the TSP thickness) of the steam generator tube bundle. Tube specimens will be removed and examined to determine if ODSCC is occurring at the tube-to-TSP intersections. Any other type of degradation or any other location in the tube bundle shall continue to be evaluated in accordance with existing Prairie Island Technical Specifications. The observation of circumferential cracks, or primary water stress corrosion cracking associated with TSP indications, or ODSCC beyond the TSP thickness will be reported to the NRC prior to returning the steam generators to service.

The voltage-based repair criteria will not be applied at the following atypical tube-to-TSP intersections of the steam generator tube bundle:

1. At tube-to-TSP intersections having dent signals exceeding 52 volts as measured with the bobbin probe. Any indications confirmed by RPC will be repaired.

2. At tube-to-TSP intersections where mixed residuals could mask a 1.0 volt bobbin coil ODSCC indication. Any indications confirmed by RPC will be repaired.

3. At tube-to-TSP intersections where copper deposits interfere with bobbin coil signals. Any indications confirmed by RPC will be repaired.

There are no tube-to-flow distribution baffle plate intersections at Prairie Island and therefore section 1.b.5 does not apply.

I l

Exhibit D -

I Page 2

)

Section 2:

TUBE INTEGRITY EVALUATION There are three principal engineering analyses that shall be performed during each voltage-based repair process at Prairie Island.

a) Prediction of steam generator bobbin voltage population distribution.

b) Calculation of steam generator tube leakage during a postulated main steamline break (MSLB) c) Calculation of steam generator tube burst probability during a postulated MSLB.

I The latest approved EPRI database (7/8 inch diameter tubing) utilizing NRC approved data exclusion criteria will be applied in the voltage correlations used for the MSLB leak -

4 rate, MSLB burst probability and upper voltage repair limit calculations. The upper i

voltage repair limit will be determined prior to each outage, using the most recently approved data base.-

i-The methodology for the performance of these analyses, including correlations which i

relate bobbin voltage amplitudes, free span burst pressure, probability of leakage and i

associated leak rates will be consistent with the methodology of Attachment 1, Section 2 of Generic Letter 95-05 and Westinghouse WCAP-14277. In addition, the upper voltage repair limit used to repair bobbin indications independent of RPC confirmation will be determined at each outage based on the guidance of Section 2.a.2 of Generic j

Letter 95-05.

l To support reporting criteria of Section 6 of Generic Letter 95-05, calculations of MSLB leakage and probability of burst may be determined using the actual measured bobbin i

voltage distribution instead of the projected EOC voltage distribution in accordance with paragraph 2.c of Generic Letter 95-05.

J Allowable MSLB Tube Leak Rate The calculated maximum allowable tube leak rate for Prairie Island during a hypothetical MSLB event shall not exceed 6.4 gpm in the faulted loop. The 6.4 gpm t

leak rate is calculated consistent with current accepted licensing basis assumptions as specified in Paragraph 2.b.4 of Generic Letter 95-05. The analyses do not take credit j

for reduced reactor coolant system iodine activity, as allowed by this paragraph. The i

analyses of offsite doses and control room doses used the analytical methods and assumptions outlined in the Standard Review Plan (Exhibit E). The 6.4 gpm leak rate value in the faulted loop (with leakage in the intact loops equal to the technical specification normal operation leakage limit of 150 gpd) will not result in offsite doses exceeding 10% of 10 CFR 100 requirements or control room doses exceeding 10 CFR

- 50, Appendix A, GDC-19 requirements, and therefore, is consistent with the Prairie

____j

Exhibit D Page 3 Island Licensing basis (the 10% of 10 CFR 100 requirements is applicable only to the dose assessment for the voltage based repair criteria at Prairie Island). If it is determined that this leakage limit might be exceeded during the operating cycle, the reporting requirements of Paragraph 6.a.1 of Generic Letter 95-05 will be followed.

Calculation of Projected MSLB Leakaae The projected MSLB leakage will be dotermined using the probability of leakage model and the conditional leak rate model in accordance with paragraphs 2.b.3(1) and 2.b.3(2) of Generic Letter 95-05. Consistent with the guidance of Section 2.c of Generic Letter 95-05, the Prairie Island MSLB leak rate analysis performed prior to returning the steam generators to service may be performed based on the projected next end-of-cycle (EOC) voltage distribution or the actual measured bobbin voltage distribution at a given outage. The method selected at a given outage will be based on outage scheoule constraints, particularly the ability to complete the growth rate analysis prior to restart.

Distribution of Bobbin Indications as a Function of Voltaae at BOC.

The frequency distribution of bobbin indications found during inspection will be determined in accordance with paragraph 2.b.1 of Generic Letter 95-05.

Projected EOC Voltaae Distribution The projected EOC voltage distribution will be determined in accordance with paragraph 2.b.2 of Generic Letter 95-05.

Eddy Current Voltaae Measurement Uncertainty.

The eddy current voltage measurement uncertainty will be determined in accordance with paragraph 2.b.2 (1) of Generic Letter 95-05.

Voltaae Growth Due to Defect Proaression Voltage growth rate will be determined in accordance with paragraph 2.b.2 (2) of Generic Letter 95-05. Steam generator chemical cleaning has not been performed; however, if chemical cleaning is scheduled in the future, the impact on voltage growth rates will be evaluated.

Section 3:

INSPECTION CRITERIA All steam generator tubes will be inspected with a bobbin coil during each normally scheduled refueling cutage. The inspection will include all hot leg side tube-to-TSP intersections and all cold leg side tube-to-TSP intersections to the extent of any known ODSCC. Data acquisition and analysis guidelines and procedures will be developed

-. -.. -.. - - - ~

c Exhibit D Pcg3 4 j

and performed consistent with the methodology of Generic Letter 95-05. Data analysts' I

j will be trained and qualified in the use of the voltage based repair criteria analysts guidelines and procedures and in the identification of other potential indications at tube' 3

support plates. Any indication with bobbin voltage exceeding 2.0 volts shall be inspected with a rotating pancake coil or equivalent technology and shall be repaired if the bobbin indication is confirmed by RPC. Any indication will be plugged or repaired regardless of any RPC inspection results if the bobbin voltage exceeds the upper voltage repair limit as obtained per Section 2.a.2 of Generic Letter 95-05. The supplementary guidance of Section 3 of Generic Letter 95-05 will be applied with the clarifications noted below.

v Clarifications to Section 3, inspection Criteria Probe wear will be monitored to ensure compliance with the NRC approved industry procedure (Reference 3). The NEl letter identifies the following exception to the j

referenced NRC letter regarding probe wear:

i l

Reinspection of all intersections in low row tubes (rows 1-9) where entry from both the cold leg as well as the hot leg would be required will not be performed. All tubes with indications above 75% of the repair limit will be reinspected with an acceptable probe when the probe fails the wear check.

j The affected intersections in those subject tubes will be reinspected as permitted by access from the hot leg side unless cold leg entry is required i

to reinspect indications of the cold leg side that are 75% of the repair limit.

Since data supporting this exception were obtained in similar steam generators, this procedure and exception is applicable to Prairie Island.

i Additionally, the industry methodology will be utilized for new probe variability as approved by the NRC.

i Where rotating pancake coil inspection is specified, comparable or improved nondestructive examination techniques, such as the + Point coil may be used.

l i

l Section 4:

TUBE REMOVAL AND EXAMINATIONITESTING l

The Prairie _ Island program for tube removal and examination will comply with the guidance of Section 4 of Generic Letter 95-05 in order to determine if axial ODSCC is i

occurring at the tube support plate intersections at Prairie Island. In order to implement the voltage-based repair criteria, the corrosion morphology of the degradation at the i

tube support plate intersections in the Prairie Island steam generators will be demonstrated to_ be consistent with that attributable to ODSCC via tube pulls and destructive analysis. Destructive analysis will provide additional data for the voltage i

+

i based burst pressure, probability of leakage, and leak rate correlations. In addition, l

future tube pulls will monitor the degradation mechanism over time, and will assess

)

l

. Exhibit D Pcg3 5 inspection capability. The number and frequency of subsequent tube pulls will be consistent with Generic Letter 95-05 requirements, or Prairie Island will participate in 2

l an NRC endorsed industry program per Section 4.a of Generic Letter 95-05.

Section 5:

LEAKAGE The operational leakage limit will be limited to 150 gpd through each steam generator.

Steam generator tubes with known leaks will be repaired prior to returning to service.

Leakage monitoring measures will follow the guidance of the industry /EPRI report of Reference 2 or later revisions. In accordance with the existing Prairie Island Technical Specifications 4.12.D.1(f) and (g), tubes with known leaks will be repaired prior to restoring the steam generator to service following a steam generator inspection outage.

Section 6:

REPORTING REQUIREMENTS Prairie Island will comply with the reporting requirements of Section 6 of Generic Letter 95-05.

REFERENCES

1. NRC Generic Letter 95-05, " Voltage-Based Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking, "

August 3,1995.

2. EPRI TR-104788, "PWR Primary-to-Secondary Leak Guidelines," Final Report, May 1995.

l

3. NRC letters form Brian Sheron (NRR) to Alex Marion (NEI) dated February 9,1996 and March 18,1996, and NEl letters from Alex Marion to Brian Sheron dated January 23,1996 and February 23,1996.

i

4. Westinghouse WCAP-14277 Revision 1, "SLB Leak Rate and Tube Burst i

Probability Analysis Methods for ODSCC at TSP Intersections". December 1996 I

6 4

e

+

w m-

+

+---

4

(

Exhibit E r

Prairie Island Nuclear Generating Plant License Amendment Request Dated May 15,1997 4

Westinghouse Main Steam Line Break Allowable Leak Rate Analysis a

t t

s d

I i

4 r

7) 9 a'-

=

5 WeSlingil0use-Enery Systems Nuclear Services Division Electric Corporatioil Box 15e Mamson Pennsylvania 15663 0158 NSD-E-TAP-0032 May 13,1997 Mr. Richard Pearson Prairic Island Nuclear Generating Plant 1717 Wakonade Drive East Welch,'MN 55089

Reference:

(1) NSD-E-TAP-0003, Hennann Lagally to Richard Pearson; dated January 28,1997 (2) E-mail, Richard Pearson to Hermann Lagally, dated 4/8/97 (3) Northem State Power Company, Purchase Order PJ6150SQ

Dear Mr. Pearson:

Reference I transmitted a draft of the MSLB Allowable Leak Rate analysis report together with the final input assumptions for the analysis for your review and comment. Reference 2

- provided your comments via E-mail. Your comments have been addressed and appropriate responses have been included in the fmal analys:s.

i Attachment I to this letter is the final input assumptions used in the steamline break radiological consequences analysis to calculate the maximum allowable primary to secondary leakage. Attachment 2 to this letter contains the fmal letter report which documents the maximum allowableleakage.

This action completes Item 1.2 of the Technical Description incorporated into the contract,

-i Reference 3.

l l

Sincerely,

=

< Hermann Lagally Principal Engineer, SG Degradation Management "The mission of,VSib is to pruride our customers with people, equipmen:, and services that set Ihr standards of excelknce in the nuclear industry."

i

\\

ATTACHMENT 1 Final input Assumptions for the Maximum Allowable Leakage Calculations t

Parasmeter Proposed Value Basis for n 4 :ad Value Core power, MWT -

1721 Consistent with core and coolant activities provided in Appendix D of the USAR Prunary coolant activities by nuclide Kr-85 1.11 USAR, Appendix D Table D.4-1, Revision 4 based on 1% fuel defects, (pCi/cc at Kr-85m 1.46 Consider dose to thyroid and gamma whole body only.

578'F)

Kr-87 0.87 Kr-88 2.58 Xe-133 174 Xe-133m 1.97 Xe-135 4.95 Xe-135m 0.14 Xe-138 0.36 I-131 2.11 1-132 0.62 1-133 2.55 I-134 0.39 I-135 1.4 Primary coolant activity prior to 1 Ci/gm of dose equiv. I-Consistent with Technical Specifications, 3.1.D.1 and accident 131 for the accident initiated Standard Review Plan 15.1.5, Appendix A, Rev. 2 iodine spike case. 60 Ci/gm DE I-131 for the pre-accident spike case.

1% defect level for noble Typical value,1% defects approx. equal to 100/E bar gases for noble gases in Technical Specifications Secondary coolant activity prior to 0.1 Ci/gm of DE I-131 Technical Specifications, 3.4.C.D accident RHR cut-in time, hr 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Standard Value used in steam release calculations, USAR 11.9.2 Primary to secondary leakage, intact 150 gpd or ~ 0.1 gpm Technical Specifications for alternate tube plugging SG cuteria per GL 95-05.

i

~...-..m.

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

Parameter Proposed Vahse Basis for Fi-p_j Value

~

Steam release to environment from 254400 lb, (0-2 hours)

Steam release from intact SG values taken from recently intact SG, Ib 486000 lb, (2-8 hours) calculated values for Point Beach (another 2 loop plant) and increased by 20%

SG iodine partition factor

- faulted SG 1.0 Faulted SG boils dry & all iodine released

- intact SG 0.01 Standard Review Plan 15.6.3, Rev. 2 Breathing rate, m'/sec Reg. Guide 1.4 and USAR Appendix D Table D.8-3

- Control Room 3.47E-4 (0-24 hr)

- Offsite 3.47E-4 (0-8 hr) 2 Site boundary X/Q, sec/m 6.49E-4 (0-2 hrs.)

USAR Appendix H, Rev. 4 & Flour Daniel Calc. No.

M-834532-ZC-002, p.12 Iow population zone X/Q, sec/m' l.77E-4 (0-8 hrs.)

USAR Appendix H Rev. 4 & Flour Daniel Calc. No.

M-834532-ZC-002, p.12 Control Room X/Q, sec/m' 5.58E-3 (0-8 hrs.)

Flour Daniel Calc. No. M-834532-ZC-002, p. 2-1 Iodine DCFs 1-131 1.07E6 ICRP30 allowed per GL 95-05 I-132 6.29E3 1-133 1.81E5 I-134 1.07E3 I-135 3.14E4 i

Cmtrol Room Parameters Prairie Island Calculation, ENG-ME-314 for control

- Normal Mode room volume.

- Control Room Volume, ft3 165000 6

- Total Flow, cfm 12000 i

Other parameters

- Recirculation Flow, cfm 10000

- Total Flow, cfm rs from Fluor Daniel Calc. No

- Filtered Recirculation, cfm 0

- Unfiltered Recirculation, cfm 10000 M-834532-ZC-002

- Fresh Air laict Flow, cfm 1835 Fluor Power Services Letter FN-3713

- Filtered inlet Flow, cfm 0

USAR Section 10.3

- Unfiltered inlet Flow, cfm 1835 r

em m w

w-M

+-

=-o-t-9e-we e--

-,-mm+-

sym--

sb-yv v

ww-m g ewu v

v meg

.-s-n y

~

Parnaseter Proposed Value Basis for Proposed Vahne

- Unfikerec leakage Flow, cfm 165 r Nncy %

Unfiltered inleakage based on 6.1 % of control room

- Particulate 95 volume

- Elemental 90

~

- Organic ~

90 Control Room Parameters' Prairie Island Calculation, ENG-ME-314 for control I

- Emergency Mode room volume.

- Control Room Volume, ft' 165000

- Total Flow, cfm 12000

- Recirculation Flow, cfm 12000 Other parameters from Fluor Dan. l Calc. No.

e

- Filtered Recirculation, cfm 3600 M-834532-ZC-002

- Unfiltered Recirculation, cfm 8400 Fluor Power Service Letter FN-3713.

- Fresh Air Inlet Flow, cfm 0

USAR Section 10.3

- Filtered Inlet Flow, cfm 0

- Unfiltered Inlet Flow, cfm 0

Unfiltered inleakage based on 0.1 % of control room t

- Unfiltered Leakage Flow, cfm 165 volume. Filtered recirculation Iased on TS 4.14.B

- Fiher Efficiency, %

- Particulate 95

- Elemental 90

- Organic 90 Liquid Volumes At 100 % Power, NSP Fax (No. 6123307603),1/3/97,

- Secormiary, Ibs 109155 per SG R. Pearson to J. Monahan, p. 3

[

- Primary, ft' 5227.39 Dose acceptance criteria

- pre-accident iodine spike 25 remwhole body 10CFR100 300 rem thyroid

- accident initiated iodine spike 2.5 remwhole body 10% of 10CFR100 30 rem thyroid

- fuel failure case 25 remwhole body 10CFR100 300 rem thyroid -

Percentages consistent with Standard Review Plan 15.1.5, Appendix A, Rev. 2

.~

.c.

ATTACHMENT 2 Final Letter Repon for the Maximum Allowable Primary to Secondary Leak Rate t

1 i

l 1

Ee===M=. Break Radiological Ccesequences Introduction For this analysis the complete severance of a main steamline outside containment is assumed to occur.

De affected SG will rapidly depressurize and release radiciodines initially contained in the secondary coolant and primary coolant activity, transferred via SG tube leaks, directly to the outside atmosphere.

A portion of the iodine activity initially contained in the intact SGs and noble gas activity due to tube leakage is released to atmosphere through either the atmospheric dump valves (ADV) or the safety valves (MSSVs). This analysis evaluated the maximum permissible primary to secondary leak rate which could exist in the faulted SG without exceeding the allowable dose rates at the site boundary, the low population zone or the control room. This section describes the assumptions and analyses performed to determine the amount of radioactivity released and the offsite and control room doses resulting from this release.

Input Par==**rs and Assumptions The analysis of the steam line break (SLB) radiological consequences uses the analytical methods and assumptions nutlined in the Standard Review Plan (Reference 1). For the pre-accident iodine spike it is assumed that a reactor transient has occurred prior to the SLB and has raised the RCS iodine -

concentration to 60 pCi/gm of dose equivalent (DE) 1-131. For the accident initiated iodine spike the reactor trip associated with the SLB creates an iodine spike in the RCS which increases the iodine release rate from the fuel to the RCS to a value 500 times greater than the release rate corresponding to the maximum equilibrium RCS Technical Specification concentration of 1.0 pCi/gm of DE I-131. The duration of the accident initiated iodine spike is 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The noble gas activity concentration in the RCS at the time the accident occurs is based on a fuel defect level of 1.0%. This is approximately equal to the Technical Specification value of 100/E bar Cilgm for gross radioactivity. De iodine activity concentration of the secondary coolant at the time the SLB occurs is assumed to be equivalent to the Technical Specification limit of 0.1 pCi/gm of DE I-131.

The amount of primary to secondary SG tube leakage in the intact SG is assumed to be equal to the Technical Specification limit of 150 gpd. The allowable leak rate for the faulted SG was determined to be 6.4 gpm.

No credit for iodine removal is nken for any steam released to the condenser prior to reactor trip and concurrent loss of offsite powct.

The SG connected to the broken steamline is assumed to boil dry within the initial 15 minutes following the SLB The entire liquid inventory of this SG is assumed to be steamed off and all of the iodine initially in this SG is released to the environment. Also, iodine carried over to the faulted SG by SG I

tube leaks is assumed to be released directly to the environment with no credit taken for iodine retention in the SG.

1

'1

_ - _ _ = _.

d.

An iodine partition factor in the intact SG of 0.01 (curies'l /gm steam)/(curies 1/gm water) is used (Reference 1).

All noble gas activity carried over to the secondary through SG tube leakage is assumea to be immediately released to the outside atmosphere.

Eight hours after the accident, the RHR System is assumed to be placed into service for heat removal, and there are no further steam releases to atmosphere from the secondary system.

l The thyroid dose conversion factors, breathing rates, and atmospheric dispersion factors used in the q

dose calculations are given in Table 1. The parameters associated with the control room HVAC modes are summarized in Table 2. The remaining major assumptions and parameters used specifically in this analysis are itemized in Table 3.

l l

Control Room Model i

The Prairie Island control room HVAC system operates in one of two modes. Mode 1 is the ponnal HVAC mode, in which 1,835 cfm of air flow is outside air and 10,000 cfm is recirculated air all of

. which is unfiltered. Mode 2, which consists of 100% recirculated air within the control room a portion of which is filtered. In each mode there is 165 cfm of unfiltered in-leakage. The parameters associated with the control room HVAC modes are summarized in Table 2.

l For the steam line break accident it is assumed that the HVAC system begins in Mode 1.

Following the steam line break it is assumed that the system is shifted to Mode 2 within 2 minutes.

Description of Analyses Performed The analysis of the steam line break (SLB) radiological consequences uses the analytical methods and l

assumptions outlined in the Standard Review Plan (Reference 1). Both the pre-accident iodine spike and accident initiated lodine spike models are analyzed for these release paths.

i Acceptance Criteria The offsite dose limits for a SLB with a pre-accident iodine spike are the guideline values of 10CFR100. These guideline values are 300 rem thyroid and 25 rem 7-body. For a SLB with an accident initiated lodine spike the acceptance criteria are a "small fraction of" the 10CFR100 guideline values, or 30 rem thyroid and 2.5 rem y-body. The criteria defined in SRP Section 6.4 (Reference 2) will be used for the control room dose limits: 30 rem thyroid,5 rem whole body and 30 rem beta skin.

Results e

Jihe offsite and control room thyroid, y-body, and beta skin doses due to the SLB are given in Table 4.

l 2

.l i

Conclusions I

It was determined that for a primary to secondary leak rate in the faulted SG of less than or equal to 6.4 gpm, the offsite thyroid and whole body doses are within the current NRC acceptance criteria for a steamline break accident based on the low population zone dose for the accident initiated spike case.

t

- The control room thyroid, whole body and beta skin doses are also within the current NRC acceptance criteria for the control room.

References 1.

NUREG-0800, Standard Review Plan 15.1.5, Appendix, A, " Radiological Consequences of Main Steam Line Failures OutsiJe Containment of a PWR, Rev. 2 July 1981, 2.

NRC SRP Section 6.4, " Control Room Habitability System", Rev 2, July 1981, NUREG-0800.

i I

i F

t 3

i

TABLE 1 DOSE CONVERSION FACTORS, BREATHING RATES AND ATMOSPHERIC DISPERSION FACTORS Thyroid Dose Conversion Factors (D Isotope (rem / curie)'

I-131 1.07 E6 I-132 6.29 E3 1-133 1.81 ES I-134 1.07 E3 i

I-135 3.14 E4 Time Period Breathing Rate

(hr)

(m'/sec) 0-8 3.47 E-4 O'

Atmospheric Dispersion Factors 3

(sec/m )

Site Boundary 0-2 hr 6.49E-4 Low Population Zone 0-8 hr 1.77E-4 Control Room 4

i 0-8 hr 5.58E-3

0) ICRP Publication 30 Regulatory Guide 1.4 (2i 0'USAR Appendix D i

4

e

e, ;,

i TABLE 2 CONTROL ROOM PARAMETERS h

Volume.

165,000 ft' i

Unfiltered Inleakage 165.0 cfm Total Flow Rate 12000 cfm i

i Unfiltered Makeup /In-leakage I

Mode 1 2000 cfm Mode 2 165.0 cfm i

Filtered Makeup Mode 1 0 cfm Mode 2 0 cfm Filtered Recirculation Mode 1 0 cfm Mode 2 3600 cfm Filter Efficiency Elemental 90 %

Organic 90 %

Particulate 95 %

Occupancy Factors 0-8 hours 1.0

)

i 5

~

V,'

l TABLE 3 l

ASSUMPTIONS USED FOR SLB DOSE ANALYSIS J

Power 1721 M Wt Reactor Coolant Noble Gas Activity Prior to Accident 1.0% Fuel Defect Level l

Reactor Coolant Iodine Activity Prior to Accident Pre-Accident Spike 60 Ci/gm of DE I-131 Accident Initiated Spike 1.0 Ci/gm of DE I-131 Reactor Coolant k> dine Activity increase Due 500 times equilibrium

- to Accident Initiated Spike release rate from fuel for initial 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after SLB 5

Secondary Coolant Activity Prior to Accident 0.1 pCi/gm of DE l-131 SG Tube Leak Rate During Accident 4

Intact SG 150 gpd l

. Faulted SG 6.4 gpm lodine Partition Factor i

Faulted SG 1.0 (SG assumed to steam dry)

Intact SG 0.01 t

Duration of Activity Release Secondary System 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Offsite Power Lost Steam Release from Intact SG 254,400 lb (0-2 hr) 486,000 lb (2-8 hr) 4.

f 6

i

TABLE 4 SLB OFFSITE & CONTROL ROOM DOSES k

Site Boundary (0-2 hr) ppg Acceptance Criteria Thyroid: Accident Initiated Spike 14.11 rem 30 rem Thyroid: Pre-Accident Spike 20.15 rem 300 rem y-body 0.068 rem 2.5 rem Low Population Zone (0 8 br) i Thyroid: Accident Initiated Spike 20.47 rem 30 rem 4

Thyroid: Pre-Accident Spike 20.18 rem 300 rem y-body 0.058 rem 2.5 rem Control Room (0 8 hr)

Thyroid: Accident Initiated Spike 28.26 rem 30 rem Thyroid: Pre-Accident Spike 29.71 rem 30 rem i

y-body 0.047 rem 5 rem Beta skin 0.116 rem 30 rem i

t 7

I