BSEP-94-0285, Submits Results of Insp Re Core Spray Spargers & Associated Piping,Per IE Bulletin 80-13

Submits Results of Insp Re Core Spray Spargers & Associated Piping,Per IE Bulletin 80-13

ML20071Q482
Person / Time
Site:	Brunswick
Issue date:	07/26/1994
From:	Lopriore R CAROLINA POWER & LIGHT CO.
To:	Ebneter S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
BSEP-94-0285, BSEP-94-285, IEB-80-13, NUDOCS 9408110242
Download: ML20071Q482 (19)

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July 26,1994 Serial: BSEP 94-0285 Mr. Stewart D. Ebneter Regional Administrator United States Nuclear Regulatory Commission 101 Marietta Street, N. W., Suite 2900 Atlanta, GA 30323 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-324 / LICENSE NO. DPR-62 RESPONSE TO IE BULLETIN 80-13 INSPECTION RESULTS OF BRUNSWICK UNIT 2 CORE SPRAY SPARGERS

Dear Mr. Ebneter:

Pursuant to NRC IE Bulletin 80-13, Carolina Power & Light Company (CP&L) hereby submits the results of the inspections performed for Brunswick Unit 2 core spray spargers and associated piping. The reactor pressure vesselinternal piping and spargers associated with the Core Spray (CS) System were visually examined with a remotely operated underwater camera during the B211R1 refueling outage which ended on June 26,1994.

Enclosure 1, provides relevant portions of the Engineering Evaluation Report (EER), which documents the analysis of the examination.

The analysis concludes that the as found condition of the Brunswick Unit 2 core spray spargers and associated piping is acceptable and no postulated scenario will affect the safe operation of the plant, and design margins for the core spray system will be maintained during the cycle 11 operation. Therefore, the condition of the internal core spray spargers and associated piping does not impose any restrictions on plant operations for the next operating cycle.

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• 4 Mr. S. D. Ebneter BSEP 94-0285 / Page 2 Please refer any questions regarding this submittal to Mr. G. Honma at (910) 457-2741.

Yours very truly,

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R. P. Lopriore Manager Regulatory Affairs Section SHC/shc (corespar.u2)

Enclosure cc: NRC Document Control Desk Mr. P. D. Milano, NRC/NRR Senior Project Manager - Brunswick Mr. R. L. Prevatte, NRC Senior Resident inspector - Brunswick

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ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2 NRC DOCKET NO. 50-324 OPERATING LICENSE NO. DPR-62 RELEVANT PORTIONS OF ENGINEERING EVALUATION REPORT (EER) 94-0137 UNIT 2 CORE SFRAY INVESSEL PlPING EVALUATION FOLLOWING IVVI EXAMINATIONS

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6 EER No. 94-0137 Revision No. 0 ENGINEERING EVALUATION Page No. , 4 f-This EER documents the Unit 2 Core Spray Sparger in-Vessel Visual Inspections (IVVI) performed during Refuel Outage 10 (B211RI); bounds the inspection results by previous evaluations performed on Unit 1 & 2 Core Spray piping and spargers by General Electric (GE); and providesjustification to use the spargers for another operating cycle in the as-found condition. This EER is identified as Q-list on Form 2. The internal core spray piping and spargers are designed to ANSI B31.1.0 - 1%7 Power Piping Code and as a reactor vessel internal component, are in compliance with the applicable portions of ASME Section III,1965 through Summer 1967 Addenda. They are not pressure boundary components, however, they are essential to the safety of the plant according to the definition of Criteria 1 of the NRC General Design Criteria,10 CFR 50, Appendix A. This classification is based on the function of the core spray spargers to provide a flow path to direct water to the core region during a inss of Coolant Accident (LOCA). OPT 90.1 (Core Spray /Feedwater Visual Examination)is performed each refueling outage to satisfy the requirements of ASME Boiler and Pressure Vessel Code,Section XI,1980 Edition through the Winter 1981 Addenda, Table IWB-2500-1, Category B-N-2, B13.21 (as applicabic), Technical Specification 4.0.5 for Unit 1 or Unit 2 (as applicable), and Provisions of Inspection and Enforcement Bulletin 80-13.

l 1.0 HISTORY OF CORE SPRAY PlPING & SPARGER NON-DESTRUCTIVE EXAMINATIONS 1.1 Introduction 1.1 ! In accordance with IE Bulletin 80-13 (Ref.1) the reactor pressure vessel internal piping l and spargers associated with core spray (CS) system are visually examined with a {

remote ograted underwater camera during each refueling outage as part of Periodic j Test PT-90.1 (Ref. 2). The inspection is recorded on video tape for a documentation i record.

1.1.2 Cracking in the in-vessel CS spargers and piping is an industry concern. The first instance of cracking in the CS spargers occurred between 1978 and 1980 at Oyster Creek and Pilgrim nuclear power stations, which eventually resulted in the issuance of IE Bulletin 80-13 in 1980.

1.1.3 In 1982 CP&L notified the NRC that during the IE Bulletin 80-13 visual inspection of Unit 2 a crack was located on the A-loop upper clockwise sparger arm near the weld l

I to the T-box at 170 azimuth. The crack was approximately 20 mils wide and extended cricumferentially approximately 120 around the pipe. Although continued ,

operation without corrective action wasjudged to be safe, a clamp was installed over l

the cracked area as a precaution. Installation of the clamp provided full structural reinforcement to the sparger equivalent to a welded joint.

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EER No. 94-0137 Revision No. 0 ENGINEERING EVALUATION Page No. ,

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Inspections continued on Unit 2, in accordance with IE Bulletin 80-13, without evidence of further core spray piping and sparger cracking until 1988. During the performance of a 1988 visual inspection in accordance with IE Bulletin 80-13 (PT 90.1), a crack indication was identified on the north core spray line outside the shroud on the piping adjacent to the junction box in the heat affected zone near the weld.

Upon completion of the IVVI, the reactor pressure vessel was deflooded and supplemental liquid penetrant (LP) and ultrasonic tests (UT) were performed on the crack. These examinations revealed the crack was through wall and approximately 3.5 inches in length along the inside diameter and 1.75'along the outside diameter. An analysis concluded that the unit could be safely operated during the next fuel cycle with no operational changes or restrictions (Ref. 5). The unit operated for two refuel cycles before brackets were installed during the Refuel Outage No. 9 (B210RI). The brackets provide full structural reinforcement to the core spray piping. ,

The Unit 2 piping has previously been analyzed (Refs. 3 & 5) for both structural adequacy and the effect of potential leakage through the cracks on the ability of the CS system to deliver cooling water to the core. The conclusion was that the existing cracks with the addition of the clamp and brackets was acceptable for continued operation. The amount of leakage through the maximum predicted crack size was within the design margin of the core spray system. Core spray piping outside the shroud has been analyzed for postulated cracks in all four T-box welds and the predicted leakage is still within the design margin of the CS system.

1.1.4 The BNP Unit 1 CS piping had no reported indications until Refuel No. 8 (B109Rl).

During the performance of PT 90.1 two linear indications were found in the Unit 1 in-vessel core spray piping (B-loop), by visual examination using a remote operated underwater camera. One indication is in the heat aflected zone of a circumferential weld which is located in the in-vessel piping between B-loop inlet nozzle and the sparger, approximately 18" downstream of B-loop T-box. This linear indication is approximately 4" long. The other indication is located on a tee-to-sparger arm circumferential weld on one of the lower B-loop spargers. This indication is approximately 3" long. Both indications were evaluated by GE (Ref. 4) and determined to be acceptable in the as-found condition. The Unit I core spray piping will be routinely examined during refuel Outage No. 9 (B110RI) as part of I'r 90.1, and be re-evaluated and/or repaired based upon the results of the examination.

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EER No.'944137 Revision No. 0 Page No. 6 ENGINEERING EVALUATION i

i 2.0 REFUEL No.10 (B211R1) EXAMIN ATION RESULTS 2.1 Inspection Results 2.1.1 During the performance of IT 90.1 on the A-loop Core Spray spargers a crack was found at a seal weld to flow nozzle coupling, extending radially from the toe of the weld to approximately 1/2" into the sparger base material. The nozzle coupling is the third fitting counter-clockwise from the 240 sparger support bracket. See Figures 1 & 2.

CONCEPTUAL LAYOUT OF CORE SPRAY PIPING WITHIN THE REACTOR VESSEL N

AREA OF INTEREST 5If EE""l

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's N x si g u p,- {g1T r Figure 1 - Internal Core Spray Piping i

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f EER No. 94-0137 Revision No. 0 ENGINEERING EVALUATION Page No. ,7 5- i LOCATION OF AN INDICATION APPROXIMATELY )f" !N LENGTH IN SPARGER, ORIGINATING FROM NOZZLE WELD 2W t COfME SPRAY 'A' LOOP ~

UPPER SPARGER NOZZLE NOZZLE NOZZLE -

0 NOZ,AZLE 4 -( 424 i 43A 444 , , 0 7 Jg% '

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OW

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l l l BRACKET m# 1 I NOZZLE NOZZLE

( NOZZLE }

I NOZZLE 418 42B 43B 44B _-

CORE SPRAY 'B' LOOP -

LOWER SPAROER i

VIEW FROM INSIDE THE REACTOR VESSEL LOOKING OUT Figure 2 - 12> cation of Linear Indication

EER No. 94-0137 +

Revision No. O Page No.

ENGINEERING EVALUATION ,

8 3.0 EVALUATION FOR BOUNDING INDICATION BASED ON PREVIOUS ANAL.YSES t 3.1 Similar indications in the Unit 1 & 2 Core Spray piping and spargers have been previously analyzed by General Electric Company (Refs. 3,4 & 5) and determined to be acceptable in their as-found condition based on (1) the effect of the structural integrity of the in-vessel core spray piping; (2) the effect of leakage through assumed through-wall cracks and the impact to the ECCS analysis and (3) the elTects of any postulated loose parts on the safety related equipment in the reactor pressure vessel or the effect on in-vessel components.

3.2 For conservatism the crack is assumed to be through wall. The crack does not follow typical I ICSCC cracking, i.e., the crack does not follow a circumferential path around the nozzle in the area of the heat affected zone, rather extends radially outward into the base material.

The non-typical crack configuration is most likely the result of cold working in the base material during manufacturing of the sparger and the result of IGSCC. Thermal stresses in  !

this part of the core are very low and irradiation Assisted Stress Corrosion Cracking (IASCC) is not expected to be a contributor in the core spray spargers. An estimate of Oux at the I.D. t of the shroud wall for different elevations was performed for Unit 1 (Ref. 8). The estimated  :

Cux at +353.88" above vessel zero is less than 1.5 E 19 n/cm2 . The industry accepted - l 1ASCC initiation threshold value is 5.0 E 20 n/cm2 for low stress. Even though Unit 2 i 2

Duence is expected to be 25% higher (1.875 E 19 n/cm ) than Unit 1, the Unit 2 value is I well below the threshold value required for IASCC initiation.

3.3 The new crack is considered bounded by previous General Electric analyses as follows:

f Structural Effects 3.3.1 i

Effects on structural integrity of the in-vessel core spray piping have been previously analyzed l by GE for cracks in higher stressed areas of the core spray piping (Refs. 3,4 & 5). These analyses showed that all identified stresses expected during normal reactor operation were small. Based upon the review of the stresses, it was concluded that the structural integrity of  !

the piping and spargers with the existing cracks would be maintained during core spray I injection. Stresses considered included those due to downcomer Cow impingement loads, ,

seismic loading, pressure, weight and thermally induced loads. )

I The analyses determined that normal operating loads by themselves do not result in stresses which are sufficient to cause IGSCC initiation. however, the addition of weld stresses coupled with the local cold work could support initiation. Once IGSCC has been initiated, the normal load stresses and the residual stresses could cause subsequent growth of the induced crack.

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EER No. 94-0137 Revision No. O ENGINEERING EVALUATION Page No. ,

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in order to determine the integrity of the core spray line and spargers with the previously existing cracks, crack arrest evaluations were performed. The stresses due to pipe restraint were included in the evaluations. Because the applied normal loading of the components are predominantly displacement controlled, the stresses were found to relax as the cracks grow and the compliance (or Dexibility) of the pipe and sparger increased. The compliance was reduced sufHciently to relieve almost all of the displacement controlled stresses below the threshold to sustain IGSCC crack growth when the crack reached 180 of the circumference.

Therefore, the crack growth is expected to be negligible or at virtual arrest prior to reaching 180. The current extent of cracking in the A-toop core spray sparger is less than 15 of the sparger circumference.

It is expected that the new crack will arrest once it has reached the end of the cold working zone. Cold working of surface material by grinding or machining accomplishes a number of material changes, such as: tears,6ssures and smears or laps that can enhance corrosion.

Without the additional stresses introduced by welding or primary loads, the stress levels drop below the threshold required to sustain ICSCC crack growth.

13ased on the above, the new crack is considered bounded by previous structural analyses performed by GE.

3.3.2 Effects of Leakaae Throuch the Crack The crack in A-loop core spray sparger is located inside the shroud, therefore, any leakage through the crack would be delivered directly to the core region. A crack of this size is not expected to signiGcantly affect the spray distribution. According to an analysis performed by GE (Ref. 6), if a crack or multiple cracks (in a single sparger) were to grow to the extent that all Dow passed through the crack (s) and none through the nozzles, the ECCS performance during a postulated LOCA is not expected to suffer degradation. This conclusion was based on large scale test which conGrm that Counter Current Flow Limiting (CCFL) breaks down soon after spray initiation. This breakdown causes downnow of the water inventory from the upper plenum with subsequent rapid delivery and rapid re0ooding of the core. Following this, a residual pool of water remains in the upper plenum ensuring uniform coolant delivery to.the ,

individual bundles. Therefore, adequate core cooling from the core spray system is expected to be maintained as long as the spray water is injected into the upper plenum, regardless ofits distribution through the spray nozzles.

Ilased on the above, the new crack is considered bounded by previous analyses performed by GE on the effects of crack leakage.

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- t EER No. 94-0137 Revision No. 0 ENGINEERING EVALUATION Page No. ,10  ;

' l 3.3.3 Effects of Loose Parts in the Reactor Vessel i Based on previous structural analysis performed by GE (Ref. 3,4 & 5), no breaks are expected in the core spray line or core spray sparger piping and consequently, no loose pieces in the reactor. However, analyses of the possible consequences of a potential loose piece was previously performed by GE (Ref. 3,4 & 5). The analyses evaluated two different types of loose pieces postulated for the core spray line (a section of core spray pipe and a small piece of core spray pipe) and three different types for the core spray sparger (a section of sparger  ;

pipe, a small piece of sparger pipe and an outlet nozzle). The analyses concluded that the probability for unacceptable corrosion or other chemical reaction due to loose pieces is zero.

The potential for unacceptable Anw blockage or other damage to the fuel assemblies was negligible. The potential for unacceptable control rod interference was negligible. Therefore, the evaluations concluded that no safety concern was posed by postulated loose parts.

Based on the above and the crack geometry, the new crack is considered bounded by previous loose parts analyses performed by GE. 1 4.0 JUSTIFICATION FOR USING PIPING /SPARGERS AS-FOUND  !

i 4.1 It has been determined from previous General Electric Company analyses that the crack found on the Unit 2 A-loop Core Spray sparger piping during the performance of FT 90.1 is bounded by the analyses.  !

4.2 The crack is in a non-typical location (extending radially into the base material) and is believed to be a result of cold working in the area. The crack is not expected to grow l beyond the cold worked area since stress levels drop below the threshold to sustain crack  ;

growth.

- 4.3 Previously performed ar.alyses have concluded that cracks of a similar nature do not degrade the piping such that the structuralintegrity of the piping would be compromised during core l spray injection.

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Revision No. 0 ENGINEERING EVALUATION Page No. 11 c i i

4.4 The crack is located within the shroud area, therefore, any leakage from the crack will be directed to the core. Analyses have concluded that if all of the Dow from core spray header is passed through the cracks and none through the nozzles, the ECCS performance during a postulated LOCA is not expected to suffer degradation.  ;

4.5 Analyses have been performed to evaluate the effect of loose parts as a result of structural integrity degradation of the core spray piping and spargers. Results concluded that the ,

probability for unacceptable corrosion or other chemical reactions due to loose nieces is zero.

The potential for unacceptable Dow blockage or other damage to fuel assemblies and unacceptable control rod interference is negligible.  ;

5.0 DISPOSITION OF PIPINGISPARGERS AS-FOUND IN REFUEL OUTAGE NO.10 5.1 Ilased on previous analyses performed by General Electric Company (GE), the Unit 2 internal core spray piping is acceptable in the as-found condition for the next operating' cycle of Unit  !

2. There is no postulated scenario involving the internal core spray piping that wi.ll affect the safe operation of the plant and all design margins for the core spray system will be maintained '

during the operating cycle.

5.2 The condition of the internal core spray piping does not impose any restrictions to plant operation for the next operating cycle. ,

5.3 The core spray piping is routinely examined every refueling outage as part of PT 90.1,  ;

therefore the piping /spargers will be reevaluated and/or repaired in Refuel Outage No.11 (B212Rl) based upon the results of the next examinatinn.

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4 EER No. 94-0137 Revision No. 0 ENGINEERING EVALUATION Page No. 12 s

REFERENCES

1. USNRC IE-Bulletin 80-13, " Cracking in Core Spray Spargers", hiay 12,1980.

2. Ol'T-90.1 for Unit 2, R.efuel Outage No.10 (Outage B211Rl), " Core Spray /Feedwater Visual Examination"

3. General Electric Company Report No. EAS-03-0190 (Supplement 1 of EAS-14-0388), Core Spray Line Crack Growth Analysis Update for Brunswick Steam Electric Plant Unit 2, January 1990.

4. General Electric Company Report No. GE-NE-523-97 0793, " Core Spray Crack Analysis for Brunswick Steam Electric Plant, Unit 1, July 1993.

5. General Electric Company Report, " Core Spray Sparger Crack Analysis at BSEP - Unit 2, NEDO 22171, July 1982.

6. General Electric Company Analytical hiodel for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K, NEDO-20566, Volume 11, 1975.

7. Design Basis Document (DDD)- 18, Core Spray.

8. EER 93-0536 - Evaluation of Unit i Core Shroud Indications and Operability Assessment of Unit I and Unit 2.

__ _ . . ._ _ _ _ ~ . _ _ _ _ _ _

ATTACHMENT 2 (Cont'd)

REVISION 3 10CFR50.59 PROGRAM MANUAL Page 55 I ATTACHMENT A '

CP&L SAFETY REVIEW PACKAGE Page _14 of N/A c ,

SAFETY REVIEW COVER SHEET  !

DOCUMENT NO. EER 94-0137 REV. NO. O I DESCRIPTION OR TITLE: Unit 2 CS Invessel Pinino Evaluation followino IVVI Exam.

1. Assigned Responsibilities:

Safety Analysis Preparer: Georoe L. Frick l Lead let Safety Reviewer: Georce L. Frick 2nd Safety Reviewer: Steve Bertz  !

i

2. Safety Analysis Preparer: mplete RT I SAFETY ANALYSIS ,

Safety Analysis Preparer 7f' - N< 1I di / d' / 994 h

~-'STGNATURE ~ " DATE l

1

3. Lead let Safety Reviewers Complete Part II, Item Classification.  ;

4. Lead let Safety Reviewer: III may be completed. If either question 1 or 2 is "yes," then Part IV is not required.

5. Lead 1st safety Reviewers Determine which DISCIPLINES are required for review of this item (including own) and mark the appropriate block (s) below. .

DISCIPLINES Recuired (Print Name) Sionature/Date (Steo 7)  !

[ ] Nuclear Plant Operations

[ } Nuclear Engineering _ m ,_

[/) Mechanical Georae L. Frick A fM 5-+-%

[ ] Electrical 6/

[ ] Instrumentation & Control

[ } Structural - , _ . . , 3

[/] Metallurgy L o/ut.u/,6arau 7/[4And/ # Bh 5-r-# [

[ ] Chemistry / Radiochemistry

[ ] Health Physics

[ ] Administrative Controls  !

[

6. A QUALIFIED SAFETY REVIEWER will be assigned for each DISCIPLINE marked in step 5 and his/her name printed in the space provided. Each person shall .

perform a SAFETY REVIEW and provide input into the Safety Review Pcckage. },

7. The Lead let safety Reviewer will assure that a Part III or Part IV is i completed (see step 4 above) and a Part VI if. required (see 9.d of Part II).

f Each person listed in step 5 shall sign and date next to his/her name in step  !

5, indicating completion of a SAFETY REVIEW.

8. 2nd Safety Reviewer erform a SAFETY REVIEW accordance with Section 8.0.

2nd Safety Reviewer rr 7N - v Date 7 /hM -

DISCIPLINE: Mechanical e

Les No' i

9. PNSC review required? If "yes" attach Part V and mark reason [ ] (/)

• below ,

[ ] Potential UNREVIEWED SAFETY QUESTION

[ } Question 9 of Part IV answered "Yes"

[ ] Other (specify):

1 I

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0 AI-109 Rev. 002 Page 72 of 86 t

ATTACHMENT 2 (Cont'd)

REVISION-3 10CFR50.59 PROGRAM MANUAL Page 56 ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page 2 of E/A PART Is SAFETY ANALYSIS s (See instructions in Section 8.4.1)

(Attach additional sheets as necessary)

DOCUMENT NO. EER 94-0137 REV. No. O DESCRIPTION OF CHANGE: The invessel core sprav oicino and searcers were visually examined per PT 90.1 durina the current refuelina outaae No. 10 fB211RI). EER 94-0137 was written to document the examination results and bound the as-found condition usino previous analyses to allow the olant to operate another refuelino evele.

ANALYSIS: The invessel core scrav searaers are desianed to ANSI B31.1.0 1967 Power Ploina Code and as a reactor vessel internal component meet the aoolicable portions of ASME Section III. 1965 throuah Summer 1967 Addenda. They are classified as Safety Related because the core sorav system is part of ECCS. The searcers are inspected in accordance with the recuirements of ASME Boiler and ELqssure Vessel Code,Section XI, 1980 Edition throuah the Winter 1981 Addenda.

Table IWB-2500-1 Cateoory B-N-2, B13.21 f as aoolicable), Technical Specification 4.0.5 for Unit 1 or Unit 2 fas applicable) and Provisions of Insoection and Enforcement Bulletin 80-13.

The crack has been evaluated by comparison to previous analyses performed by GE for similar indications for, (1) the ef fects of structural intecrity of invessel pioino (2) the effect of leakaoe throuch the indication and the ability of the core sprav scaraers to deliver coolina water to the core durina core sprav initiation and (3) the ef fect of any costulated loose parts on safe operation of the Unit. For conservatism the crack is considered to be throuch wall.

Comparisons show that A-loon scaraer is bounded by the previous analyses and based on those analyses, the crack does not dearade the searcers such that the int earity of the scarcers would be compromised durina a core iniection. All structural desian marains wil1 still be met at the end of another eichteen month operatina evele.

O AI-109 Rev. 002 Page 73 of 86

+

ATTACHMENT 2 (Cont'd)

REVISION 3 10CFR50.59 PROGRAM MANUAL Page 56 ATTACHMENT A .

c CP&L SAFETY REVIEW PACKAGE Page _11 of HL&

PART I: SAFETY ANALYSIS (See instructions in Section 8.4.1)

(Attach additional sheets as necessary)

DOCUMENT NO. EER 94-0137 REV. NO. O ANALYSIS (Cont. );The location of the crack is within the shroud recion, therefore any leakaoe throuch the crack would be to the core recion. The crack will not sionificantiv af fect the sprav distribution. Analyses have concluded that if all the flow from the core sorav header is passed throuch cracks and none throuch the nozzles, the ECCS performance durino a postulated LOCA is not expected to suffer dearadation.

The effect of oostulated loose parts from the invessel core scrav searcers and pipino has been previoucly analvred by GE and there was no safety concerns identified. The effect of any loose parts on the safety related reactor vessel control rod drive components, fuel essemblies, or any other reactor vessel internal components was considered neolioible.

Ih_e conclusion reached by review of orevious analyses is that the Unit 2 Core Sorav Sparcers, in their as-found condition, will be acceptable for another 18 month operatino evele.

REFERENCES:

I MFSAR Sect. 3.6. 3.7, 3.9, 3.11, 5.2, 5.3. 5.3A, 6.1, 6.2, 6.3, 7.0 and 15.0 Tech. Specs. 3/4.3.3. 3/4.4.3, 3/4.4.8, 3/4.5.3 and associated Bases. j l

O AI-109 Rev. 002 Page 73 of 86

l ATTACHMENT 2 (Cont'd)

REVISION 3 10CFR50.59 PROGRAM MANUAL Page 57 ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page _17 of Efg PART II: ITEM CLASSIFICATION O DOCUMENT NO. EER 94-0137 .. REV. NO. 0 Yes Eo

1. Does this item represent:

a. A change to the facility as described in the SAFETY [ ] [/)

ANALYSIS REPORT 7

b. A change to the procedures as described in the [ ] [/)

SAFETY ANALYSIS REPORT 7

c. A test or experiment not described in the SAFETY [ ] (/)

ANALYSIS REPORT 7

2. Does this item involve a change to the individual plant [ ] [/)

Operating License or to its Technical Specifications?

3. Does this item require a revision to the FSAR7 [ ] [/)

4. Does this item involve a change to the Offsite [ ] [/)

Dose Calculation Manual?

5. Does this item constitute a change to the Process Control [ ] [/)

Program?

6. Does this item involve a major change to a Radwaste Treatment [ ] [/)

System?

7. Does this item involve a change to the Technical [ ] [/)

Specification Equipment List (BSEP and SHNPP only)?

8. Does this item impact the NPDES Permit (all 3 sites) or [ ] [/]

constitute an "unreviewed environmental question" (SHNPP Environmental Plan Section 3.1) or o "significant environmental impact" (BSEP)?

9. Does this item involve a change to a previously accepted:

a. Quality Assurance Program [ ] [/)

b. Security Plan (including Training, Qualification, and [ ] [/]

Contingency Plans)?

c. Emergency Plan? [ ] [/)

d. Independent Spent Fuel Storage Installation license? [ ] (/)

(If yes, refer to Section 8.4.2, " Question 9," for special considerations. Complete Part VI in accordance with Section 8.4.6)

SEE SECTION 8.4.2 FOR INSTRUCTIONS FOR EACH "YES" ANSWER.

REFERENCES. List FSAR and Technical Specification references used to anewer questions 1-9 above. Identify specific reference sections used for any "Yes" l answer.

UFSAR Sect. 3.6. 3.7. 3.9. 3.11. 5.2. 5.3, 5.3A. 6.1. 6.2. 6.3. 7.0 and 15.0 i Tech. Specs. 3/4.3.3. 3.4.4.3, 3/4.4.8. 3/4.5.3 and associated Bases. l I

I O AI-109 Rev. 002 Page 74 of 86 l

ATTACHMENT 2 (Cont'd)

REVISION'3 10CFR50.59 PROGRAM MANUAL Page 58

. ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page 18 of Efh PART III: UNREVIEWED SAFETY QUESTION DETERMINATION SCREEN c DOCUMENT NO. EER 94-0137 REV. NO. O XES NO

1. Is this change fully addressed by another completed [ ] [/)

UNREVIEWED SAFETY QUESTION determination? (See Section 7.2.1, 7.2.2.5, and 7.9.1.1)

REFERENCE DOCUMENT: REV. NO.

YES NO

2. For procedures, is the change a non-intent change which oniv [ ] [/)

(check all that apply): (See section 7.2.2.3)

[ ] Corrects typographical errors which do not alter the meaning or intent of the procedure; or,

[ } Adds or revises steps for clarification (provided provided they are consistent with the original purpose or applicability of the procedure); or,

[ ] Changes the title of an organizational position; or,

[ ] Changes names, addresses, or telephone numbers of persons; or,

[ ] Changes the designation of an item of equipment where the equipment is the same as the original equipment or is an authorized replacement; or,

[ ] Changes a specified tool or instrument to an equivalent substitute; or,

[ ] Changes the format of a procedure without altering the meaning, intent, or content; or

[ ] Deletes a part or all of a procedure, the deisted portions of which are wholly covered by approved plant procedures?

If the answer to either Question 1 or Question 2 in PART III is "Yes," then PART IV need not be completed.

O AI-109 Rev. 002 Page 75 of 86

ATTACHMENT 2 (Cont'd)

REVISION 3 10CFR50.59 PROGRAM MANUAL Page 59

. ATTACMMENT A CP&L SAFETY REVIEW PACKAGE Page _12 of Efb PART IV: UNREVIEWED SAFETY QUESTION DETERMINATION c DOCUMENT NO. EER 94-0137 REV. NO. O Using the SAFETY ANALYSIS developed for the change, test or experiment, as well as other required references (LICENSING BASIS DOCUMENTATION, Design Drawings, Design Basis Documents, codes, etc.), the preparer of the Unreviewed Safety Question Determination must directly answer each of the following seven questions and make a determination of whether an UNREVIEWED SAFETY QUESTION exists.

A WRITTEN BASIS IS REQUIRED FOR EACH ANSWER Xgg No

1. May the proposed activity increase the probability of [ ] [/)

occurrence of an accident evaluated previously in the SAFETY ANALYSIS REPORT 7 Use of the core scrav scarcers in their present condition does not in-crease the probability of occurrence of any previously evaluated accident.

For the next operatino evele the ability of the searcers to function as oriainally desianed is unaffected by the indication and the structural intecrity of the soarcers is not ieonardized.

2. May the proposed activity increase the consequences of an [ ] [/]

accident evaluated previously in the SAFETY ANALYSIS REPORT 7 Based on a review of previous analyses performed by GE, the core sorav scarcers will function as oriainally desianed durina the next operatina cycle and there is no postulated impact on any safety related reactor vessel components, therefore, usino the sparaers in their cresent condition will not increase the consecuences of an accident previousiv evaluated in the UFSAR.

3. May the proposed activity increase the probability of [ ] [/]

occurrence of a malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT 7 Usino the core sorav scarcers in their present condition does not increase the probability of occurrence of a malfunction of eauipment imoortant to safety evaluated previousiv in the UFSAR. The oostulated effects of loose parts was considered and the consecuences on safety related eauipment in the reactor vessel is neoliaible. The ability of the core scrav scarcers to perform within their desian marain of safety is unaffected.

4. May the proposed activity increase the consequence of a [ ] (/)

malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT 7 The consecuences of a malfunction of safety related eauioment is not increased since there is no impact to any sa f ety related system or component i in the plant as a result of usina the core sorav scarcers in their existina condition for the next eichteen month operatino cycle.

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1 ATTACHMENT 2 (Cont'd)

• REVISION'3 10CFR50.59 PROGRAM MANUAL Page 59 ATTACRMENT A CP&L SAFETY REVIEW PACKAGE Page _10 of Hig PART IV: UNREVIEWED SAFETY QUESTION DETERMINATION .,

DOCUMENT NO. EER 94-0137 REV. NO. O

5. May the proposed activity create the possibility of an [ ] [/)

accident of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT 7 The notential effect of anv loose earts on any other safety related clant eauipment is neolioible. Since there is no credicted adverse effect on any safety related comoonents, there is no possibility to create a new tvoe of accident. Therefore, no nossibility of an accident of a dif ferent tvoe than any previously evaluated in the UFSAR exists.

6. May the proposed activity create the possibility of a () [/)

malfunction of equipment important to safety of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT 7 The core sprav searcers in their present condition can be used for another operatino evele without causino the malfunction of any invessel components.

The scareers are within their structural and hydraulic desion maroins and will not create the possibility of a malfunction of ecuipment imoortant to safety of a different tvoe than any previousiv evaluated.

7. Does the proposed activity reduce the margin of safety as [ ] [/)

defined in the basis of any Technical Specification?

No other system or eculoment will be af fected by the core sprav scarcers and all maroins of safety for the system will be maintained. In the event of a core scrav iniection durino the next operatino evele, no water will be lost from the shroud area. The structural intecrity is not reduced below the desion maroin and crack crowth is not expected to be any creater than crowth experienced durino the cast. Therefore, no reduction of the maroin of safety as defined in the bases of any Tech. Spec, will occur.

8. Based on the answers to questions 1 - 7, does this item () [/)

result in an UNREVIEWED SAFETY QUESTION 7 If the answer to any of the questions 1-7 is "Yes", then the item is considered to constitute an UNREVIEWED SAFETY QUESTION.

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REVISION 3 10CFR50.59 PROGRAM MANUAL Page 59 ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page _21 of EfA PART IV: UNREVIEWED SAFETY QUESTION DETERMINATION c DOCUMENT NO. EER 94-0137 REV. NO. 0

9. Is PNSC review required for any of the following reasons? [ ] (/)

If, in answering questions 1 or 3 "No", it was determined that the probability increase was small relative to the uncertainties; or, in answering question 2 or 4 "No", it was determined that the doses increased, but the dose was still less than the NRC ACCEPTANCE LIMIT; or in answering question 7 "No", a parameter would be closer to the NRC ACCEPTANCE LIMIT, but the end result was still within the NRC ACCEPTANCE LIMIT; then PNSC review is required.

REFERENCES :

UFSAR Sects. 3.0, 5.0, 6.0, 7.0, and 15.0 Tech. Soecs. 3/4.3, 3/4.4. 3/4.5 and associated Bases.

This Unreviewed Safety Question Determination is for the following DtSCIPLINE(s):

(Additional Part IV forms may be included as appropriate.)

( ) Nuclear Plant Operations [ ] Structural

[ ] Nuclear Engineering [/) Metallurgy

(/) Mechanical [ ] Chemistry / Radiochemistry

( ) Electrical ( ) Health Physics

[ ] Instrumentation & Control ( ) Administrative Controls 0 AI-109 Rev. 002 Page 77 of 86