ML20199H725

Safety Evaluation Supporting Amends 133 & 125 to Licenses DPR-42 & DPR-60,respectively

ML20199H725
Person / Time
Site:	Prairie Island
Issue date:	11/18/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20199H722	List: ML20199H717 ML20199H725
References
NUDOCS 9711260206
Download: ML20199H725 (13)

Text

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NUCLEAR REGULATORY COMMISSION WASHINGTON, O C. 30666-0001 o,s,...../

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS.133AND125TO FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 NQBT.HERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NOS.1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By letter dated May 15,1997, as supplemented August 29, October 20, October 24, and October 28,1997, the Nodhem States Power Company (NSP or the licensee) requested amendments to the Technical Specifications (TS) appended to Facility Operating License Nos.

DPR-42 and DPR-60 for the Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2. The proposed amendments would implement voltage-based altemate repair criteria for steam generator tubes in the TS. The proposed attemate repair criteria would allow steam generator tubes having outside diameter stress-corrosion cracking (ODSCC) that is predominately axially oriented and confined within the tube support plates to remain in service on the basis of bobbin coil voltage response. The NRC guidance on the attemate repair criteria is specified in Generic Letter (GL) 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking."

b The August 29, October 20, October 24, and October 28,1997, supplements provided clarifying information, proposed license conditions, and updated TS pages. This information was within the scope of the original application and did not change the staffs initial proposed no significant hazards consideration determination (62 FR 43371).

2.0 BACKGROUND

The acceptance criteria (i.e., plugging limits) for degraded steam generator tubes are specified in the plant TS. The traditional strategy for achieving adequate structural and leakage integrity of the degraded tubes has been to establish a minimum wall thickness requirement in accordance with NRC Regulatory Guide (RG) 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes." The minimum wall thickness requirement was developed with the assumption of a uniform thinning of the tube wall. This assumed degradation mechanism is inherently conservative for certain forms of tube degradation. Conservative repair limits may lead to removing degraded tubes from service that may ctherwise have adequate structural and leakage integrity for further service.

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l To reduce unnecessary conservatism in the minimum well thickr'ess requirement for certain 1

degradation, the industry proposed voltage-based repair critaria for ODSCC confined within the

- thickness of the tube support plates. The staff published several conclusions regarding voltage-based repair criteria in draft NUREG-1477, " Voltage-Based interim Plugging Criteria for Steam Generator Tubes," and in a draft GL titled "Voltags-Based Repair Criteria for Westinghouse Steam Generator Tubes." The latter document was published for public comment in the FederalRegister on August 12,1994 (5g FR 41520). On August 3, iggv he staff issued GL 95-05 that took into consideration public comments on the draft GL cited awe, domestic-operating experience under the voltage-based ropeir criteria, and additional data made l

available from European nuclear power plants.

The guidance of GL 95-05 does not set depth-based limits on predominantly axially oriented ODSCC at tube support plate locations; rather it relies on empirically derived correlations between a nondestructive inspection parameter, the bobbin co!I voltage, and tube burst

_ pressure and leak rate. The staff recognizes that although the total tube integrity margins may be reduced following application of a voltage-based repair criteria, the guidance in GL g5-05 ansures structural and leakage integrity continue to be maintained at acco,9 table levels consistent with the requirements of 10 CFR Part 50 and 10 CFR Part 100. Since the voltage-based repair criteria do not require minimum tube wall thicknees, there is the possibility for tubes with through-wall cracks to remain in service. - Because of the increased likelihood of such flaws, the staff included provisions for augmented steam generator tube inspections and restrictive operational leakage limits.

GL 95-05 specifies, in part, that (1) the repair criteria are only applicable to predomhantly axially oriented ODSCC located within the bounds of the tube support plates, (2) licensees perform an evaluation to confirm that the degraded steam generator tubes will retain adequate structural and leakage integrity from cycle to cycle, (3) licensees edhere to specific inspection criteria to ensure consistency in methods between inspections, (4) tubes must be penodically removed from the steam generators, examined, and destructively tested to verify the -

morphology of the degradation and provide burst and leakage data for structural and leakage integrity evaluations, (5) the operational leakage limit in the plant TS be reduced, (6) licensees implement an operational leakage monitoring pregram, and (7) specific reporting requirements

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be incorporated into the plant TS.

Each Prairie Island unit has two Westinghouse Model 51 steam generators, which use mill-annealed alloy 600 tubing. These steam generators use carbon steel drilled-hole tube

. cupport plates and do not have flow distribution baffle plates. The outside diameter and nominal wall thickness of each tube are 7/8 inch and 0.050 inch, respectively.

3.0 EVALUATION

- The licensee stated that it will comply with the guidance in GL 95-05 when implementing its voltage-based ' alternate repair criteria. In addition, the licensee proposed to incorporate verbatim the model TS in GL g5-05 into the TS for Prairie Island Units 1 and 2. The major issues related to the licensee's imp'ementation of the attemate repair criteria are discussed below.

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3.1 Tube Repair Limits The proposed repair criteria will (1) permit degraded tubes having indications confined to within the thickness of the tube support plates with bobbiri voltages less than or equal to 2.0 volts to

- remain in service, (2) permit degraded tubes having indications confined to within the thickness

. of the tube support plates with bobbin voltages greater then 2.0 volts but less than or equal to the upper voltage limit to remain in service.lf a motorized rotating pancake coil probe or-acceptable altomative inspection does not detect degradation, and (3) require degraded tubes having indications confined to within the thickness of the tube support plates with bobbin voltages greater than the upper voltage limit be plugged or repaired.

The proposed lower voltage limit of 2.0 volts is derived based on the use of a correlation between the burst pressure and the bobbin coil voltage of pulled tube and model boiler data l.

and is consistent with the recommended value specified in GL 95 05 for 7/8-inch steam

generator tubing. The upper voltage limit is derived based on the lower 95-percent prediction interval of the burst pressure versus bobbin voltage correlation, adjusted for lower bound

- material properties evaluated at the 95-percent confidence level. The upper voltage limit is -

further reduced to account for uncertainty in the nondestructive examination techn,que and flaw growth over the next operating cycle. The industry periodically updates the database for burst pressure and bobbin voltage when the destructive test data from pulled tubes are available; therefore, the upper voltage limit may vary as additional data are incorporated into the database.

3.2 Irispection lasues Section 3.c.3 of Attachment i to GL 95-05 specifies guidance for probe wear. The licensee proposed to use an attemative to section 3.c.3. The attemative approach, developed through the Nuclear Energy Institute, specifies that if the probe does not satisfy the voltage variability criterion for wear of

• 15-percent limit before its replacement, all tubes that exhibited flaw signals with voltage responses measured at 75 percent or greater of the lower repair limit must ic be reinspected with a bobbin probe satisfying the i 15-percent wear standard criterion. The voltages from the reinspection should be used as the basis for tube re;. air. The staff completed a review of the Nuclear Energy Institute proposed attemative method and concluded that the approach is acceptable as discussed in a letier from Brian Sheron of the NRC to Alex Marion of the Nuclear Energy institute dated March 18,1996. The licensee's proposal to follow the industry approach to address probe wear is acceptabia.

In the laboratory and field studies supporting the attemative probe wear critena, the correlation of voltages measured by wom probes and new probes shows that for all significant voltage.

t levels, the wom probe voltages are never less than 75 percent of the new probe voltage as discussed in the letter from Alex Marion of ttee Nuclear Energy Institute to Brian Sheron of the NRC dated January 23,1996. However, in the 60-day inspection report for Byron Unit 1 dated September 9,1996, Commonwealth Edison, the licensee for Byron, compared the wom probe voltage to the new probe voltage and found that the wom probe voltage was substantially less than 75 percent of the new probe voltage for a few indications. Commonwealth Edison evaluated these indications and concluded that the criteria to retest tubes with wom probe voltages above 75 percent of the repair limit is adequate and generally conservative due to the y-

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4-trend for wom probe voltages to exceed new pobe voltages. Comparison of the actual and projected end-of-cycle voltages did not show anphing unusual ettributable to the alternate probe wear criteria. The staff concludes that the stu%nentioned probe w,er results do not indicate an immediate need to modify the probe wear criteria developed by the industry.

However, the staff will continue to monitor probe wear in the licensees' 90 day inspection -

tepocts.

With respect to probe variability, the licensee proposed to follow an attemative approach developed through the Nuclear Enorgy Institute. The proposed procedures and methodology are described in an October 15,1996, letter from A. Marion, Nuclear Energy Institute, to B. Sheron, NRC. The approach spe::ifies that the voltage responses from the primary frs:iuency and mix frequency channe(s of new probes be within *10 percent of the nominal a

voltage responses when voltages are normalized to the 20-percent flaw values. The nominal

. voltage responses were established as the average voltages obtained from the American Society of Mechanical Engineers (ASME) standard drilled-hole flaws for at least 10 production probes. The licensee's proposal to follow the industry approach to address probe variability is acceptable.

3.3 Structural and Leakage integnty Assessments GL 95-05 guidance for the voltage-bar,ed repair criteria focuses on maintaining tube structural integrity during the full range of normal, tra,.: lent, and postulated accident conditions with adequate allowance for eddy-current test uncertainty and flaw growth projected to occur during the next operating cycle. RG 1.121 recommends that a margin of safety of 1,43 against tube failure under postulated accident conditions and a margin of safety of 3 against burst during normal operation be maintained for steam generator tubes. Because GL 95-05 addresses tubes affected with OUSCC confined to within the thickness of the tube support plate during normal >peration, the staff concluded that the structural constraint provided by the tube support plate ensures all tubes to which the voltage-based criteria apply will retain a margin of 3 with respect to burst under normal operating conditions. For a postulated main steamline break accident, however, the tube support plate may displace axially during steam generator blowdown such that the ODSCC-affected portion of the tubing may no longer be fully constrained by the tube support plate. Accordingly, it is appropriate to consider the ODSCC affected regions of the tubes as free standMg tubes for the purpose of assessing burst integrity under postulated main steamline break conditions.

In order to confirm the structural and leakage integrity of the tube until the next scheduled inspection, GL 95-05 specifies a methodology to determine the conditional burst probabihty and the total primary-to-secondary leak rate from an affected steam generator during a postulated main steamline break event. To complete GL 95-05 prescribed assessments, the licensee proposes to follow the methodology described in WCAP-14277, Revision 1, "SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections," date6 December 1996 (See E. Fitzpatrick, Indiana Michigan Power Co., letter to NRC dated December 20,1996).' The staff finds the methodology in WCAP-14277, Revision 1, acceptable for Prairie Island.

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_w 5-4 GL 95-05 specifies that the structural and leakage integrity assessments should use the latest available database from destructive examinations of tubes removed from Westinghouse-3 designed steam generators. A protocol is being established between the industry and the NRC to formalize the requirements for updating the industry database. The licensee indicated that it

= will follow the protocol as documented in a letter to G. Laines, NRC, from D. Modeon of the Nuclear Energy institute, dated Janwy 15,1997, until the final version of the protocol is completodi in addition, the licensee will describe, in the GL 95-05 90-day reports, the database that was used for GL 95-05 specified calculations. The staff finds that the licensee's commitment to follow the prcocol and to use NRC approved database to perform structural and leakage assessments are acceptable.

3.3.1 Conditional Probability of Burst The licensee will use the methodology described in Revision 1 of WCAP-14277 for performing a probabilistic analysis to quantify the potential for steam generato-tube ruptures given a main steamline break event. The results of the probabilistic analysis will be compared to a threshold

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.value of 1x10-8 per cyc,le in accordance with GL 95-05. This threshold value provides assurance that the probability of burst is acceptable considering the assumptions of the calculation and the results of the staff's generic risk assessment for steam generatorc contained in NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." Failure to meet the threshold value

. Indicates ODSCC confined to within the thickness of the tube support plate could contribute a significant fraction to the overall conditional probability of tube rupture from all forms of degradation assumed and evaluated as acceptable in NUREG-0844. The NRC staff concludes the licensee's proposed methodology for calculating the conditional burst probability is i

consistent with the guidance in GL 95-05 and is acceptable.

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3.3.2 Accident Leakage The licensee will use the methodology described in Revision 1 of WCAP-14277 for calculating the steam generator tube leakage from the faulted steam generator during a postulated main steamline break event. The model consists of two major components: (1) a model predicting the probability that a given indication will leak as a function of voltage (i.e., the probability of leakage model); and (2) a model predicting leak rate as a function of voltcge, given that leakage occurs (i.e., the conditional leak rate model). The staff concludes that the licensee's proposed mett Mology for calculating the tube leakage is consistent with the guidance in GL 95-05 and is acceptable.

l 3.3.7 Primary-to-Secondary Leakage fu j

Because the voltage-based repair criteria would allow degraded tubes to remain in service, the j '

transients, or postulated accidents. Therefore, as a defense-in-depth measure, GL 95-05 degraded tubes may develop through-wall cracks which may leak during normal operation,

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. specifies that the operational leakage limits of the plant TS be limited to 150 gallons per day

'- (gpd) from any one steam generator. The licensee proposed to change the leakage limits in the L

-plant TS to 150 gpd through any one steam generator, in addition, the licensee has i incorporated the guidelines in Electric Power Research Institute (EPRI) Report TR-104788,

- "PWR Primary-to-Secondary Leak Guidelines," into the Prairie Island plant operating procA,s. The staff concludes that the proposed operstional leakage limit of 150 gpd for Prr.e Island TS !s consistent with GL 95-05 and, ;herefore, is acceptable.

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3'3.4; Polantial for Tube Collapse There is a potential for tube collapse in the steam generator at some plants during a i

loss-of-coolant accident (LOCA) in combination with a safe shutdown earthouake (SSE). This is the case as the tube support plates may become deformed as h result of lateral loads at the wedge supports at the periphery of the plate due to the combined effects f.,f the LOCA rarefaction wave and SSE loadings. Then, the resulting pressure differential on the deformed tubes may cause some of the tubes to collapse. Them are two issues associated with the steam generator tube collapse. F_irst, the collapse of steam generator tubing reduces the RCS :

[ reactor coolant system] flow area through the tubes. The reduction in flow area increases the resistance to flow of steam from the core during a LOCA which, in tum, may potentially.

increase the peak clad temperature. Second, there is a potential that partial through-wall cracks in tubes could progress to complete through-wall cracks during tube deformation or collapse.

Tubes that are susceptible to collapse during accident conditions will be excluded from application of the voltage-based repair criteria. Since the leak-before-break methodology is applicable to the reactor coolant loop piping at Prairie Island, the probability of breaks in the primary loop piping is sufficiently low that they need not be considered in the structural design of the plant. However, review of Westinghouse seismic umbrella spectra for Model 51 steam generators shows that Prairie Island is bounded by these spectra, such that no tubes will undergo deformation due to the combined effects of LOCA plus SSE, and, therefore, no tubes will be excluded from application of the criteria due to loading from LOCA plus SSE.

Based on the staff's review, as discussed above, the staff concurs with the licensee's assessment that no tubes need be excluded from application of the voltage-based repair criteria s

due to combined LOCA plus SSE loads.

3.4 Degradation Monitonng e

To confirm the nature of the degradation at the tube support plate elevations, tubes are periodically removed from the steam generators for destructive tests. The test data from removed tubes can confirm that the nature of the degradation observed at these locations is predominantly axially. oriented ODSCC, provide data for assessing the reliability of the

- inspection methods, and supplement the existing databases (e.g., burst pressure, probability of leakage, and leak rate). GL 95-05 specifies that at least two tubes be removed from steam generators with the objective of retrieving as many intersections as practical (minimum of four.

. Intersections) during the plant steam generator inspection outage preceding initial application of the voltage-based repair criteria. - On an ongoing basis, additional tube specimens (minimum of two intersections) should be removed at the first refueling outage following 34 effective full power months of operation or at the maximum interval of three refueling outages after the previous tube pull. Altematively, the licensee may participate in an industry-sponsored tube pull program endorsed by the staff as described in GL 95-05. The licensee pulled two tubes during the current Unit 1 outage and confirmed that tube degradation was consistent with that attributable to ODSCC. The licensee plans to pull future tubes consistent with the tube removal

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. guidelines in GL 95-05 until an NRC-endorsed industry program is available as described in Section 4.s of GL 95-05.~

3.5 Aasessment of Radiological Consequences f

in its license amendment submittel dated May 15,1997, the licensee requested that the specific

- activity limits of dose-equivalent iodine-131 ('8'l) in the primary coolant be established at 1.0 ~

pCl/g for the 48-hour limit and at 60 pCi/g for the maximum hotantaneous limit (in accordancs with GL 95-05), The allowable activity level of dose-equivale,t '8l1 in the secondary coolant was assumed to be' equivalent to the TS limit of 0.1 pCi/g. This license submittal also requested that Prairie island be approved to operate based upon a 6.4 gallons per minute (gpm) primary-to-secondary leak initiated by an accedent in the fauttod steam generetor and the TS allowable value fo primary-to-secondary leakap from the intact steam generator of 150 gpd.

As part of the request for license amendment, the licensee performed an assessment of the radiological dose consequences of a main steamline break accident. The licensee's calculations assumed that 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 />.

in a telephone conversation with the licensee on October 1,1997, the licensee stated that it had found an error in its calculations and based on the revised calculations, NSP would like to revise the requested allowable leak rate for the faulted steam generator by reducing the -

leakage from 6.4 gpm to 4.64 gpm. Using this revised leakage figure, the licensee stated that it L had calculated doses that met the NRC acceptance criteria for doses at the Exclusion Area l

Pandary (EAB), the Low-Population Zone (LPZ), and the control room.

i The staff reviewed the licensee's calculations and performed confirmatory calculations to check

- the acceptability of the licensee's methodology and resulting doses. As part of their review, the staff calculated the doses resulting from a main steamline break accident using the methodology associated with Standard Review Plan (SRP) 15.1.5, Appendix A. The staff performed two separate assessments. One was based upon a pre-existing iodine spike actLity level of 60 pCilg of dose equivalent '*'l in the primary coolant and the other was based upon an accident-Initiated iodine spike. For the accident-Initiated spike case, the staff assumed that the primary coolant activity level was 1.0 uCi/g of dose-equivalent I. The accident initiated an increase in the release rate of iodine from the fuel by a factor of 500 over the normal release rate to maintain an activity level of 1.0 pCl/g of dose-equivalent l in the primary coolant. For these two cases, the staff calculated the thyroid doses for individuals located at the EAB and the LPZ and thyroid doses to the control room operator. The parameters that were utilized in the staff's assessment are presented in Table 1.

Using the licensee's estimated allowable leak rate for the faulted steam generator of 4.64 gpm, the staff calculated that the thyroid doses to the control room operator wouid exceed the guidelines of SRP 6.4 of NUREG-0800 (acceptance criteria of 30 rem thyroid to the control room operator), in performing its calculations, the staff assumed that the accident-initiated iodine spike continued for the duration of the accident (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />). In its October 20,1997, supplemental submittal the licensee revised its calculation so that the duration of the iodine

spike would last for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> which resulted in a 1.66 gpm allowable primary-to-secondary -

' leakage. The values the staff used for the efficiencies of the control room ventilation filters were reduced by 1 percent (below the licensee's numbers) to account for bypass flow. For the 4

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i thyroid dose to the control room operator to be within the 30-rom limM, the _ staff calculated that -

1 the maximum allowable _ leakage from the faulted steam generator would have to be less than or.

n equal to 3.4 gpm for the pre-existing case (giving a control room dose of 2g.7 rem thyroid) or 1.42 gpm for the accident 'nitiated case (giving a dose of 2g.6 rem thyroid)(both of these E

' leakage values are for an RCS temperature of 578 degrees F). Since the lower of these two calculated leakages is 1.42 gpm, the accident-initiated case is the limiting case.

Using a leak rate of 1.42 gpm for both the pre-existing and accident-initiated cases (and maintaining the specific activity limits of dose-equivalent *l in the primary coolant at 1.0 pCi/g --

for the 48-hour limit and at 60 pCi/g for the maximum instantaneous limit), the staffs calculations showed that the thyroid doses at the EAB and LPZ would be less than the

_ guidelines established by SRP 15.l.5, Appendix A of NUREG-0800 (acceptance criterion of 300 rom thyroid dose at the EAB and LPZ for the pre-existing spike case and 30 rem thyroid dose at

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the EAB and LPZ for the accident-initiated spike case). The control room operator thyroid dose would be less than the guidelines of SRP 6.4 of NUREG-0800 (acceptance criterion of 30 rem thyroid to the control room operator)(see Table 2).

In order for the licensee not to exceed the 30 rem thyroid dose limit to the control re >m operator (while maintaining the specific activity limits of dose-equivalent *l in the primary coolant at 1.0 pCi/g for the 48-hour limit and at 60 pCilg for the maximum instantaneous limit), the staff has determined that the licensee must limit the maximum allowable leakage in the faulted steam generator to 1.42 gpm (at an RCS temperature of 578 degrees F). In order to comply with this limit, by letter dated October 24,1997, the licensee proposed tae following license condition.

NSP will assure that during the implementation cf steam generator repairs r

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utilizing the voltage based repair criteria, the total calculated primary to secondary side leakage from the faulted steam generator, under main stmm line break conditions (outside containment and upstream of the main steam isolation valves), will not exceed 1.42 gallons per minute (based on a reactor coolant system temperature of 578'F).

This license condition will ensure that doses to the control room operator will be maintained within the 30 rem thyroid limit under main steamline break conditions by ensuring that the

. leakage from the faulted steam' generator will not exceed 1.42 gpm (based on an RCS temperature of 578 degreee F). Therefore, the staff finds this license condition to be acceptable -

and is including it in Appendix B to the licenses.

3.6 Proposed TS Changes

in order to incorporate a voltage-based steam generator repair criteria, the licensee has proposed the following changes to the TS.

-- 1. Proposed Changes to TS 3.1.C.2.e.

L The limit for total reactor coolant system to secondary coolant system leakage through both L

. steam generators of 1.0 gallon per minute will be changed to a limit of 150 gallons per day j

of primary-to-secondary leakage through any one steam generator.

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9-12.L Proposed new TS 4.12.8.5.

• lndications left in service as a result of application of tube support plate voltage-based

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

3. Proposed new TS 4.12.B.6.

implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot leg and cold leg tube support pl ate intersections down to the lowest cold leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications.' The determination of the lowest cold leg tube support plate intersections having ODSCC indicat

ons shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length."

- 4. Proposed Changes to TS 4.12.D.1.(f).

An exemption is added for the voltage-based repair criteria "This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applieo. Refer to Specification 4.12.D.4 for the repair limit applicable to these intersections."

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5. Proposed new TS 4.12.D.4.

This new sec'!cn provides the detailed requirements for the voltage-based repair criteria.

6. Proposed new TS 4.12.E.5.

This new section provides the detailed reporting requirements for the voltage-based repair criteria.-

7. Correction to TS 4.12.C.1.

A typographical error in the word " category" has been corrected in Section 4.12.C.1.

8.- Proposed change to Bases The Bases for TS 3.1.C and 4.12 are revised to incorporate the voltage-based repair criteria.

A typographical omission of the word "or" was corrected in the last sentence of the second paragraph of Bases page B.4.12-1.

~ g. Proposed change to Index

. The Index has been revised to reflect pagination changes.

The' staff has reviewed the TS chan0es discussed above and finds that they consistently incorporate the voltage-based repair criteria per the requirements of NRC Generic Letter 95-05 e-

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Therefore, the proposed changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Minnesota State official was notified of the preposed issuance c'the amendments. The State official had no comments.

5.0 ENVIRONWiiNTAL CONSIDERATION The amend'ments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative -

occupational radiation exposure. The Commission has previously issued a proposed finding -

that the amendments involve no significant hazards consideration and there has been no public comment on such finding (62 FR 43371). In addition, the amendments change reporting requirements. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10). Pursuant to 10 CFR 51.22(b), no environmental

- impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by

operation in the proposed manner, (2) such activities will be conducted in compliance with the

. Commission's regulations, and (3) the issuance of the amendments will not be inimical to the comrwn defense and security or to the health and safety of the public.

Principal Contributors: J.Tsao C. Hinson J. Rajan

- Date: November 18,'1997 Attachments: 11. Table'1

- 2.' : Table 2 4

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' INPUT PARAMETERS FOR PRAIRIE ISLAND UNITS 1 AND 2 EVALUATION OF MAIN STEAMLINE BREAK ACCIDENT

1. Primary Coolant Concentration of 60 pCilg of Dose Equivalent "'I Pre-existina Soike Value (uCl/a)

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"'I = 14.6

"'I = 60.2

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2. Primary and Secondary Coolant Specifications Primary Coolant Volume (ft')

5,227.39 Primary Coolant Temperature ('F) 578 Secondary Coolant Steam Mass (Ibm) 107,000 Secondary Coolant Liquid Mass (Ibm) 5,700 Secondary Coolant Steam Temperature ('F) 510 Secondary Coolant Feedwater Temperature ('F) 427.3

3. TS Limits for D9! "'l in the Primary and Secondary Coolant Maximum Instantaneous DE "'I Concentration (pCl/g) 60.0 Primary Coolant DE *l Concentration (pCi/g) 1.0 Secondary Coolant DE *l Concentration (pCilg) 0.1

4. TS Value for the Primary to Secondary Leak Rate Primary to secondary leak rate, maximum any SG (gpd)

' ISO Primary to secondary leak rate, both SGs (gpd) 300

5. Maximum Primary to Secondary Leak Rate to the Faulted and intact SGs Faulted SG (gpm) 1.42 Intact SG ;gpm/SG) 0.1

6. lodine Partition Factor Faulted SG 1.0 Intact SG 0.01

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7. Steam Released to the Environment

[

FauNed SG (0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 109,155 lbs intact SGs (0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 254,400 lbs intact SGs (2 8 hoars) 486,000 lbs

8. Letdown Flow Rate (9pm) 40

9. Release Rate for 1.0 pCilg of Dose Equivalent '8'l Release Rate (Cl/hr) 500X Release Rate (Cl/hr)

• l =

5.62 2810

• l =

9.06 4530 t

• 1 =

9.9 4950

• l =

13.3 6640

• l =

9.55 4770 10.

Atmospheric Dispersion Factors sec/m8 CAB (0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 9.0 x 10d

,t'Z (0-8 hours) 1.77 x 10d Control Room (0-8 hours) 5.58 x 10

11.

Control Room Parameters t

Filter Efficiency (%)

Air intake filter 0

Air recirculation filter.

Elemental 90 Organic 90

-l Particulste 95 Volume (ft*)

165,000 Makeup flow (cfm)

Mode 1 (0-2 minutes) 1,835 Mode 2 (after 2 minutes) 0 Recirculation Flow (cfm)(filtered)

Mode 1 (0 2 minutes)

O Mode 2 (after 2 minutes) 3,600 r

Unfiltered inleaka9e (cfm)'

Mode 1 (0 2 minutes) 165 Mode 2 (after 2 minutes).

165 F

Occupancy Factors -

01 day 1.0 NRC staff calculated'value

.a -

a,

l i !

TABLE 2 THYROID DOSES FROM PRAIRIE ISLAND UNITS 1 AND 2 MAIN STEAM LINE DREAK ACCIDENT (REM)(VALUES CALCULATED BY NRC STAFF)

DOSE LOCATION Predalsting Spike Accident initiated Spike" l

EAB 10.0*

7.34 LPZ 7.0*

17.1 j

Control Room "

,12.4 29.6

• Acceptance Criterion = 300 rem thyroid

• • Acceptance Criterion = 30 rem thyroid f

- - -. -