ML20024J069
| ML20024J069 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/26/1994 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19353C445 | List: |
| References | |
| CON-NRC-94-058, CON-NRC-94-58 VPNPD-94-082, VPNPD-94-82, NUDOCS 9409080076 | |
| Download: ML20024J069 (10) | |
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Wisconsin Electnc POWER COMPANY 231 W McNgan. Po Bcm 2046. Mihuaukea wi 53201-2046 (414)2202345 VPNPD-94-082 10CFR50.4 NRC-94-058 10CFR50.90 August 26, 1994 Document Control Desk U.S.
NUCLEAR REGULATORY COMMISSION Mail Station P1-137 Washington, DC 20555 Gentlemen:
DOCKETS 50-266 AND 50-301 i
TECHNICAL SPECIFICATIONS CHANGE REOUEST 175 MODIFICATIONS TO SECTION 15.4.2, "IN-SERVICE INSPECTION OF SAFETY CLASS COMPONENTS" l
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 In accordance with the requirements of 10 CFR 50.4 and 50.90, Wisconsin Electric Power Company (Licensee) hereby requests l
amendments to Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2, respectively, to incorporate changes to the plant Technical Specifications.
The i
j proposed changes modify Technical Specifications Section 15.4.2, 4
"In-Service Inspection of Safety Class Components," by
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incorporating the use of acceptance criteria as described in Westinghouse WCAP-14157, " Technical Evaluation of Hybrid Expansion Joint (HEJ) Sleeved Tubes Containing Indications Within th, Upper Joint Zone," to allow sleeved tubes with certain upper sleeve joint parent tube indications to remain in service.
The basis for Technical Specifications Section~15.4.2 is also being revised to support the above changes.
Marked-up Technical Specifications pages, Westinghouse WCAP-14157, a safety evaluation, and a no 4
significant hazards consideration are enclosed.
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Approval of the proposed acceptance criteria and associated license amendment is necessary to support in-service inspection of the PBNP j
Unit 2 steam generator tube sleeves during the upcoming Unit 2 fall
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refueling outage.
Therefore, we request NRC approval of this Technical Specifications Change Request by September 24, 1994.
DESCRIPTION OF CURRENT LICENSE CONDITION Technical Specifications Section 15.4.2, "In-Service Inspection
~i of Safety Class Components," provides the in-service inspection requirements for safety class components to assure continued integrity of safety class systems.
Specification 15.4.2.A provides the in-service inspection requirements for steam generator tubes.
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Document Control Desk August 26, 1994 Page 2 Technical Specifications Section 15.3.1,
specifies those limiting conditions for operation of the reactor coolant system which must be met to ensure safe reactor operation.
DESCRIPTION OF PROPOSED CHANGES 1.
Existing Specification 15.4.2.A.6 is revised to incorporate the use of NRC-approved acceptance criteria to maintain sleeved tubes with certain parent tube through-wall indications in service as follows:
15.4.2 A.
Steam Generator Tube Inspection Requirements 6.
Corrective Measures All tubes that leak or have degradation exceeding the plugging limit shall be plugged or repaired by a process such as sleeving
- prior to retuin to power from a refueling or in-service inspection condition.
Sleeved tubes having sleeve degradation exceeding 40% of the nominal sleeve wall thickness shall be plugged.
' Sis'sysdJth6as[shlehshAvsTiiEfsntVtUbh ihd ic5Elons ?iniths?uppersj ointfregiont shallf ba asses. sed (fonshfficienthst'ructurallinteghitpdid accordance?withEacceptsd1iiethodologytthTubesslefE]IH servids/with thisjtypes.offindicationJshall:(bs]re-examinedyduringithelnextfin servicetlinspectiony a
2.
To support the above changes, the basis for Section 15.4.2 is revised as follows:
Basis
...However, the basis requirements and intent will be met, to the extent practical.
WCAPF1'4157E"Ts0hhic51TEvsluation?6f.
Hpbfid:iEsp$nsionfJ6ihES(HEJ)/ Sleeved (TubestContaining"
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l I nd ica t ibn s F Within ithe EUpper YJoinO Z one y'idontains3 acae
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j criteri af f odd.isp~osit;idninb3$ybrid iexpansiorsj oint$("HEJ)' ^ ~
l indicationssinfthelslsevelupperyointsregioni^^
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l 3.
Existing Technical Specification 15.3.1.D.4 specifies a maximum primary to secondary leak rate of 500 gpd in either steam generator.
We will administratively reduce this limit to 150 gpd.
BASIS AND JUSTIFICATION In April 1994, Kewaunee Nuclear Power Plant (KNPP) discovered 77 parent tube indications in the upper sleeve joint regions of sleeved tubes.
Seventy-four of these indications were subsequently judged to be circumferential in nature.
These indications created concerns regarding the structural integrity of the associated upper sleeve joint.
Westinghouse evaluated the structural integrity of l
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Document Control Desk i
j August 26, 1994 Page 3 4
4 the sleeve upper hybrid expansion joint (HEJ) and determined that a j
through-wall circumferential crack with a maximum end-of-cycle arch j
length of 240* could be present without affecting the structural load carrying capability of the upper joint.
Kewaunee Nuclear a
i Power Plant presented the acceptance criteria to the NRC on l
l April 19, 1994.
Point Beach Nuclear Plant plans to inspect the Unit 2 steam l
generator sleeves during the upcoming fall refueling outage to assure a similar mechanism is not present.
Since June 1994, i
personnel from Wisconsin Electric Power Company-(WEPCO), Wisconsin Public Service Corporation (WPS), Commonwealth Edison Company (CECO), American Electric Power Corporation (AEP), and Westinghouse j
Electric Corporation have been coordinating efforts to resolve NRC concerns regarding disposition of the HEJ indications.
The method 3
proposed to disposition the HEJ indications is similar to that i
proposed by WPS for Kewaunee Nuclear Power Plant.
Electric Power Company (or PBNP Unit 2) is serving as lead plant j
for submitting this proposed acceptance criteria due to the l
upcoming fall 1994 outage.
An overview of the WEPCO inspection 1
j plan and submittal was presented to NRC staff on July 28, 1994.
1 Westinghouse WCAP-14157 is enclosed and contains the proposed j
acceptance criteria.
This WCAP incorporates the results of leak i
rate and structural testing programs performed for Kewaunee in I
j April 1994 and supplemented by results of additional testing which is currently being conducted.
This WCAP will be revised as
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additional test results are available.
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l We have determined that the proposed amendments do not involve a
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significant hazards consideration, authorize a significant change in the types or total amounts of any effluent release, or result in
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any significant increase in individual or cumulative occupational j
exposure.
We therefore conclude that the proposed amendments meet j
the requirements of 10 CFR 51.22 (c) (9) and that an environmental i
impact statement or negative declaration and environmental impact appraisal need not be prepared.
Please contact us if you have any questions.
Sincerely, s
[
Bob Link Subscribed and sworn before me on Vice President this 264A day of Mua usd 1994.
Nuclear Power DAW /jg b.
a 011 %
Enclosures
( paff Public, State of Wisconsin M commission expires 10-29 96 cc:
NRC Resident Inspector NRC Regional Administrator Public Service Commission of Wisconsin
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TECHNICAL SPECIFICATIONS CHANGE REOUEST 175 SAFETY EVALUATION INTRODUCTION Wisconsin Electric Power Company (Licensee) is applying for amendments to Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant, Units 1 and 2.
The proposed changes modify Technical Specifications Section 15.4.2, "In-Service Inspection of Safety class components," by incorporating WCAP-14157 which contains acceptance criteria to allow sleeved tubes with certain upper joint parent tube through-wall indications to remain in service.
The WCAP defines:
1.
The area of the steam generator tube length affected by sleeving which should be subject to eddy current i
inspection and plugging / repair criteria, 2.
The end of cycle (EOC) circumferential crack extent within the lower hardroll transition of the upper hybrid expansion joint (HEJ) which will maintain sleeved tube joint integrity during all normal _and l
accident plant conditions at EOC conditions, 3.
The location below which any existing degradation will not affect HEJ integrity / leakage and allow the tube to remain in service, 4.
Maximum joint leakage for various zones in the upper HEJ region.
The basis for Section 15.4.2 is also being revised to support the above changes.
EVALUATION In order to assess the adequacy of the HEJ pressure boundary, the structural integrity and leakage characteristics of the HEJ were evaluated under certain plant conditions.
Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes,"
i was used to define acceptable criteria for circumferential indication involvement in the lower hardroll transition of the upper HEJ.
Steam Generator Tube Intearity Regulatory Guide 1.121 incorporates safety factors for tube burst during normal operating and faulted conditions which are consistent with Section III of the ASME Code.
Regulatory Guide 1.121 recommends that degraded tubes meet a burst requirement of the most limiting of either 3 times the normal operating pressure differential (3AP o) or 1.43 times the maximum faulted conditions u
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pressure differential (1.43APg3).
For PBNP Unit 2, the 3APso case produces the largest pressure differential.
This pressure j
differential creates a pressure end cap load on the parent tube which would act to pull the parent tube out of the joint.
- Hence, the degraded joint must exhibit sufficient strength to counteract the end cap forces and maintain the sleeve / tube joint intact.
For PBNP Unit 2, defining an allowable circumferential crack extent based on 3AP end cap loads is extremely conservative.
Some of the conservatisms are:
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1.
Resistance provided by dented / packed tube support plate intersections is neglected.
Tube removal efforts at similar plants have shown the removal forces to be in the range of 1,000 to 3,000 lbs for non-dented tube support plate intersections.
2.
Nearly 3 inches of tube motion is required to separate the tube from the sleeve and create steam generator tube i
j rupture-type leak rates.
Tube proximity in the steam generator U-bend region will restrict tube motion to e maximum of approximately 1 inch.
j 3.
The structural model assumes that the rotating pancake coil (RPC) measured angle is through-wall over its entire indicated length and contains uniform, 40% deep non-detected degradation over the remaining portion of the tube.
Experience has shown that the through-wall crack lengths are likely to be less than the indicated angle associated with 50 percent of the peak RPC amplitude and much less than the full measured angle of the indication.
4.
Lower tolerance limit parent tube material property values were used.
5.
A minimum frictional force of 300 pounds is assumed between l
the parent tube and sleeve hardroll.
No benefit from the non-degraded " ligament" is provided for in the test.
In field sleeved cubes, the tube ligament would supply a bending strength component which would add significantly to the friction force of the joint.
The structural model of the parent tube hardroll transition has determined that a 224 single through-wall circumferential crack with 40% non-detected degradation penetration in the remaining ligament can withstand a tube loading of the pressure end cap load at 3AP. minus the minimum 300 lb hardroll friction u
resistance without suffering plastic overload.
Using a crack l
growth rate of 25% per year (or operating cycle for PBNP) is J
consistent with outer diameter stress corrosion cracking (ODSCC) growth rates.
Therefore, the maximum circumferential crack extent which could be left in service within the lower hardroll transition has been determined to be 179' at BOC.
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Leakace Assessment Conservative leak rate testing was initially used to assign bounding leak rates to indications in various zones of the HEJ region.
For sleeved tube joints where the tube was machined away below the bottom of the lower hardroll transition of the upper HEJ, the maximum main steam line break (MSLB) leak rate was bounded by.003 gpm at 2516 psid.
For tubes machined away at the bottom of the lower hardroll transition, the leak rate was bounded by.008 gpm.
For tubes machined away at the inflection point of the lower transition, the leak rate was bounded by
.016 gpm.
For tubes machined away at the top of the transition (representing a tube / sleeve separation condition), leakage was approximately.082 gpm at normal operating conditions.
Relevant SLB leak rate data could not be gathered from these specimens because the leakage exceeded the make-up capacity of the test apparatus.
However, it is estimated that the leakage would not exceed 2.5 gpm.
The conservatisms in these tests include:
1.
The parent tube was completely machined away in all test specimens, representing a 360 100% through-wall crack.
2.
The tortuous nature of ODSCC cracking would reduce leakage to levels less than the measured leakage of the test specimens.
3.
The leakage testing was performed at 600"F.
Normal operating hot leg temperature at PBNP Unit 2 13 597"F.
Therefore, initial testing results indicate that faulted condition leakage of sleeved tubes which may remain in service due to application of the proposed criteria with indications I
below the top of the lower hardroll upper transition can be conservatively assigned a bounding leak rate of.02 gpm per l
degraded sleeved tube.
This assessment is subject to change as l
further testing results are obtained.
Off-Site Dose EstIsation Durina Postulated MSLB Event The licensing basis of PBNP limits off-site dose during a MSLB l
event to the guidelines identified in 10 CFR 100.
We have i
performed an evaluation of the allowable primary to secondary leakage during a MFLB according to NUREG-0800, "Stanaard Review Plan," methodology.
The results of this evaluation indicate that a maximum faulted primary loop leakage of 25.0 gpu will not result in off-site doses which exceed a small fraction (10 percent as defined in NUREG-0800) of the 10 CFR 100 requirements.
l The 25.0 gpm leakage limit was calculated using the following assumptions:
1.
Pre-accident primary-to-secondary leakage of 150 gpd (.1 gpm).
The current Technical Specifications limit is 500 gpd and will be administratively reduced to 150 gpd.
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2.
Pre-accident primary coolant iodine activity of 1 pci/g dose equivalent I-131 (Technical Specifications limit).
3.
ICRP-30 dose conversion factors for the thyroid.
Any sleeved tubes which remain in service due to implementation of the proposed criteria must not leak at a rate such that the 25.0 gpm limit is exceeded.
Therefore, up to 1250 sleeved tubes with parent tube indications may be maintained in service in accordance with the proposed acceptance criteria.
This is based on postulated post-MSLB primary-to-secondary leakage of.02 gpm per degraded sleeved tube.
General Desian Criteria The General Design Criteria (GDCs) adopted for PBNP and documented in the PBNP Final Safety Analysis Report (FSAR) are similar in content to'the proposed GDCs published by the Atomic Energy Commission (AEC) in 1967 and revised by the Atomic Industrial Forum (AIF).
The applicable PBNP GDCs as defined in the PBNP FSAR for the reactor coolant system are GDC-9, " Reactor Coolant Pressure Boundary," GDC-16, " Monitoring Reactor. Coolant Leakage," GDC-33, " Reactor Coolant Pressure Boundary Capability,"
GDC-34, " Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention," and GDC-36, " Reactor Coolant Pressure.
Boundary Surveillance."
The analysis and test programs which form the foundation of the proposed acceptance criteria incorporate sufficient margin to preclude any of the referenced GDCs from being challenged during normal and faulted plant conditions.
Therefore, under the provisions of the referenced GDCs and proposed acceptance criteria, the reactor coolant system will continue to:
1)
Provide a boundary for containing the coolant under operating temperature and pressure conditions, 2)
Confine radioactive material and limits to acceptable values any release to the secondary system and to other
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parts of the plant under conditions of either normal or abnormal reactor operation, and 3)
Accommodate coolant volume changes within the protection system criteria during transient operation.
CONCLUSIONS The proposed. revisions will ensure the safe and reliable i
operation of Point Beach Nuclear Plant.
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TECHNICAL SPECIFICATIONS CHANGE REQUEST 175 "NO SIGNIFICANT HAZARDS CONSIDERATION" i
In accordance with the requirements of 10 CFR 50.91(a), Wisconsin Electric Power Company (Licensee) has evaluated the proposed changes against the standards of 10 CFR 50.92 and has determined that the operation of Point Beach Nuclear Plant, Units 1 and 2, in accordance with the proposed amendments, does not present a significant hazards consideration.
The analysis of the requirements of 10 CFR 50.92 and the basis for this conclusion are as follows:
1.
Operation of this facility under the proposed Technical Specifications change will not create a significant increase in the probability or consequences of an accident previously evaluated.
l The limiting EOC crack angle of 224 (determined by using the conservative structural model and no allowance for tube / tube j
support plate friction) would not represent a potential for failure of a tube during recovery from a MSLB.
While some slippage of the joint may occur, the maximum forces developed by the joint would support the hardroll joint slipping past the
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degraded area.
Therefore, only a limited leakage potential l
exists.
For degradation below the hardroll region in the upper HEJ, sufficient structural and leakage integrity is provided such that a complete circumferential separation of the parent tube will not affect either structural or leakage integrity of the joint during all plant conditions.
Based on leakage testing, all tubes with indications in the hardroll lower transition region would have leakage bounded by the current Final Safety Analysis Report primary to secondary leakage assumption of 1 gpm.
Wisconsin Electric Power Company has performed an analysis in accordance with NUREG-0800 guidelines which shows a maximum primary to secondary leakage of 25.0 gpm in the faulted loop with reactor coolant activity of 1 pCi/g dose equivalent I-131 will l
not result in off-site dose exceeding a small fraction (10 percent as defined in NUREG-0800) of the 10 CFR 100 requirements.
In addition, limiting the maximum primary to secondary leakage during operation to 150 gpd will help to identify tubes with rapid growth rates not bounded by the 25 percent growth allowance that result in leakage which could possibly affect tube integrity during a MSLB.
As previously stated, the postulated degraded tube would have to experience about 3 inches of axial motion prior to tube rupture-type release rates being achieved.
Due to tube proximity in the U-bend region and bending restraint provided by the tube support plates, tube motion would be limited to 0.4 to a maximum of 1 inch.
Therefore, even if a tube were to experience rapid crack growth not bounded by the structural model assumptions or
circumferential1y separate and experience slippage, the amount of slippage and subsequent leakage would be limited.
In this case, 1 inch of tube motion would still provide intimate contact between the tube and sleeve in the hardrolled region.
If it is further postulated that the remaining length of tube-to-sleeve hardroll interference (about.25 to.5 inch) for a tube which has slipped 1 inch is deformed to a " bell" shape due to the pressure difference, or the tube were to experience slippage up to 2 inches, leakage would be limited by the thin gap between the sleeve hydraulically expanded region and the tube hardrolled region.
The maximum leakage would be expected to be approximately 30 percent to 50 percent of the normal Unit 2 makeup capacity.
The results show that we remain within the acceptance criteria of the aforementioned FSAR Chapter 14 accident analyses.
Therefore, the proposed changes will not create a significant increase in the probability or consequences of an accident previously evaluated.
2.
Operation of this facility under the proposed Technical Specifications change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
Implementation of the proposed acceptance criteria does not introduce any significant changes to the plant design basis.
Use of the criteria does not provide a mechanism which could result in a tube rupture during a faulted event outside of the PBNP design basis.
Neither a single or multiple tube rupturo event would be expected in a steam generator in which the sleeved tube plugging criteria has been applied.
Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Operation of this facility under the proposed Technical Specifications change will not create a significant reduction in a margin of safety.
The use of the proposed acceptance criteria is demonstrated to mainthin steam generator tube integrity commensurate with the criteria of Regulatory Guide 1.121.
Regulatory Guide 1.121 describes a method acceptable to the NRC staff for meeting reactor coolant system general design criteria (GDCs) by reducing l
the probability or the consequences of steam generator tube rupture.
This is accomplished by determining the limiting conditions of degradation of steam generator tubing, as established by in-service inspection, for which tubes with unacceptable cracking should be removed from service or repaired.
Upon implementation of the circumferential crack acceptance criteria for sleeved tubes, even under worst case conditions, the occurrence of circumferential cracks at the lower hardroll i
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transition of the upper HEJ elevation is not expected to lead to a steam generator tube rupture event during normal or accident plant conditions.
Addressing the considerations in Regulatory Guide 1.83, "In-service Inspection of Pressurized Water Reactor Steam Generator Tubes," Revision 1, implementation of the proposed acceptance criteria is supplemented by enhanced eddy current inspection using a probe capable of detecting parent tube circumferential flaws in sleeve joints.
i In addition, implementation of the proposed acceptance criteria will decrease the number of tubes which must be plugged.
The installation of steam generator tube plugs reduces the RCS flow margin.
Thus, implementation of the proposed acceptance criteria will maintain the margin of flow that would otherwise be reduced in the event of increased tube plugging, thereby maintaining departure from nucleate boiling (DNB) margins.
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