ML063470557

Collection of Documents for the ACRS License Renewal Subcommittee Meeting on January 18, 2007 Associated with the Oyster Creek Drywell

ML063470557
Person / Time
Site:	Oyster Creek
Issue date:	12/13/2006
From:	Lund A NRC/NRR/ADRO/DLR/RLRA
To:
Donnie J. Ashley, 301-415-3191
References
%dam200701
Download: ML063470557 (127)

Text

Table of Contents December 13, 2006 Attached is a collection of documents for the ACRS License Renewal Subcommittee meeting on January 18, 2007. The documents are associated with the Oyster Creek drywell.

DATE SUBJECT TAB 4/15/66 NRC Inspection - QA Program for Drywell Reinforced Concrete 1

12/06/66 NRC Inspection - Pouring of Concrete Around Drywell 2

4/28/89 NRC Letter - Actions to Ensure Drywell Integrity 3

3/14/90 NRC Inspection - Drywell Thinning 4

4/11/90 GPU Letter - Commitments Associated with Drywell Thickness 5

7/10/90 NRC Letter - Review of Drywell Containment Structural Integrity 6

10/3/90 NRC Meeting Summary-Drywell Corrosion 7

10/16/90 NRC Letter - Clarification Regarding Corrosion of Drywell Shell 8

12/11/90 NRC Inspection - Drywell Corrosion Problem Activities 9

2/14/91 NRC Letter - RAIs on GE Drywell Stress and Stability Analysis 10 2/14/91 NRC Information Notice - Degradation of Steel Containments 11 5/23/91 NRC Internal Memo - RAIs on Corroded Drywell Analysis 12 9/3/91 NRC Letter - Staff Position on Evaluation of Steel Containment 13 11/19/91 NRC Letter - Staff Position on Evaluation of Steel Containment 14 4/9/92 NRC Internal Memo - Evaluation Report Structural Integrity of Drywell 15 4/24/92 NRC Letter - Evaluation Report on Structural Integrity of Drywell 16 5/26/92 GPU Letter - Plan for Drywell UT Thickness Measurement 17 6/2/92 NRC Inspection Report - Inadvertent Spray of the Drywell 18 6/30/92 NRC Letter - UT Inspection of Drywell Containment is Acceptable 19 4/19/94 GPU Letter - Agrees to Develop Program to Monitor Concrete 20 9/15/95 GPU Letter - Assessment of Drywell and Submitting an Inspection Plan 21 11/1/95 NRC Letter - Change in Drywell Corrosion Monitoring Program 22 12/15/05 GPU Letter - Drywell Corrosion Monitoring Program 23 2/15/96 NRC Memo - Change in Drywell Corrosion Monitoring Program 24

' i'"UNITED) STATES GOANMEN"I"

, *Memorandum Td

R. S. Boyd, Chief, Research & Power ReactorDATh:

Safety Branch, Division of Reactor Licensing, Ho

._k~. q.

o"keN.

.iko P.e O 21'6enior 11teactor Inspector rgion I, DivMIE n of Compliance (IAp41 15, 1966

SUBJECT:

JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219 The attached report by our field inspector of a visit to the subject facility on March 22 and 23, 1966, is forwarded for information.

The construction activities at the site are estimated to be 38% complete, based on 'money expended.

The present status, according to GE personnel, indicates that they are 2 to 3 months behind schedule.

The contributing causes, jurisdic-tional labor disputes, a steel shortage and an alignment problem with a vent header, are outlined in the attached report.

Attachment:

CO Rpt. No. 219/66-.

by R. T. Carlson dtd 4/15/66 I

• .1 CC:

L.E.

R.

CO M.

H.

J.

R.

Kornblith, Jr., CO:HQ G. Case, DRL G. Page, SLR

HQ File L. Ernst, CO:II:

D. Thornburg, C6:III W. Flora, CO:IV G. Engleke-A, COy*V 4

• 1 8.,.-,-

Buy U.S. Savings Bo7dr Regularly on the Payroll Satings Plan 9508030278 950227 PDR FOIA DEKOK95-36 PDR.

f

U.

S.

ATOMIC ENERGY COMMISSION REGION I DIVISION OF COMPLIANCE April 15, 1966 CO REPORT L;O.

219/66-1

Title:

JERSEY CLNTRAL POWER & LIGHT COMPANY LICENSE NO.

CPPR-15 pDates of Vi it:

March 22 and 23, 1966 By

%CT-\\ T. Ca'~o R&Nor Inapector

SUMMARY

The status of construction activities is discussed in the report.

Overall construction is estimated to be 38% corn-plete, based on money expended.

The installationi overload and initial leak rate tests of the dry well and torroidal chamber were completed satisfactorily.

A problem with an expansion joint located in one of the vent headers that joins the dry well and torroidal chamber, that resulted in both the replacement of the joint and a repeti-tion of the overload test on the dry well, is discussed in the report.

Adequate quality control measures appear to be in effect for reinforced concrete.

A 400' meteorclogical tower has been installed and data are being accumulated.

A fatality, the first at this site, resulted from injuries received by a construction worker in a fall.

(continued) 9508030200 950227 PDR FOlA DEKOK95-3 6 PDR

0 DETAILS I.

Scope of Visit Mr.

R. T.

Carlson, Reactor Inspector, Region I, Division of Compliance, visited the construction site of the Jersey Central Power & Light Company's reactor facility at Oyster Creek, New Jersey, on March 22 and 23, 1966.

The visit included the following:

A.

A review of the construction organization.

B.

A review of the status of the containment syscem.

C.

A review of the quality control measures in effect for reinforced concrete.

D.

A review of the status of construction and the timetable of significant events.

E.

A tour of the construction site.

The principal persons contacted were as follows:

Jersey Central Power & Light Company (Jersey Central)

Mr.

Ivan Finifrock, Nuclear Project Engineer Mr.

Norb...=n M. Nelson, Plant Maintenance Supervisor, Designee General Electric Company (GE)

Mr. Willard C.

Royce, Resident Manager Mr.

Abel B. Dunning, Construction Engineer, Mechanical Mr.

Glen C. brockmeir, Construction Engineer, Civil (continued)

0 II.

Results of Visit A.

Organization

1. Jersey Central Jersey Central currently has two people at the site on a full-time basis -

Mr. Nelson, the designated Plant Maintenance Supervisor, and Mr. Fred Kossatz, the designated Plant Mechanical Maintenance Foreman under Mr.

Nelson.

Both are present for on-the-job training relating to plant construction and operation.

Mr. Finfrock, the Nuclear Project Engineer, operates out of the Company Office in Morristown, New Jersey, and spends much of his time at the

site, 3 to 4 days per week.

His principal con-cern at this time relates to site meteorology.

Both Messrs. Nelson and Finfrock report to Mr. Donald Rees, the Project Engineer, who is located in the Company Office in Morristown.

2.

General Electric GE, the prime contractor for the Oyster Creek Project, currently has six people at the site.

These personnel are:

Mr.

Royce; Messrs.

Dunning and Brockmeir - the men most actively engaged in followingdayf.to-day construction; Mr. Stibers, Office Engineer; Mr.

Ryan, Site Auditor; and a clerical worker.

According to Mr.

Royce, the staff will be increased to eight in the near future.

Mr.

Royce reports to Mr.

R.

A. Huggins, Project Engineer, Atomic Power Equipment Department (APED),

San Jose, California.

(continued)

X

0

) Results of Visit (continued)

3.

Burns and Roe, Inc.

(B&R)

B&R is the Architect-Engineer and the direct Supervisor of Construction for this project.

The senior site representative for B&R is Mr. Giles Willis, w1o reports to Mr. David Kregg, the Project Manager.

The principal channel of communication between GE and B&P is through Messrs. Huggins and Kregg.

4.

Other Principal Contractors Other principal contractors associated with this project, and their responsibilities, are listed below:

Contractor American Bridge American Dewatering Corp.

Chicago Bridge & Iron Co.

Eastern Transit Mix Co.

Hatzel & Buehler, Inc.

McBride Plumbing Co.

Poirier & McLane Corp.

Pesponsibility Structural steel on Turbine Building, and on b-idge crane Site dewatering Containment system Concrete Miscellaneous electrical work Miscellaneous piping Superstructure (continued)

0.

Results of Visit (continued)

)

Contractor United Roofing &

Waterproofing U. S. Testing Laboratory Responsibility Concrete waterproofing Construction related testing White Construction Co.

Reactor Building Worthington Corp.

Turbine condensers B. Construction Status Overall construction was estimated by Mr. Dunning to be 38% complete, based on expenditures, as of March 1, 1966.

A picture reflecting the construction status as of early February is shown in Figure 1 of this report.

The reported status of the major subdivisions of the facility, as of March 1, 1966, is provided below:

Subdivision Percent Complete Containment system 100%

Reactor Building, structural portion 35%

Turbine Building, structural portion 80%

Intake and discharge structures, structural portions 98%

Intake and discharge canals, excavation 5%

Waste Disposal Building, excavation 90%

(continued)

Results of Visit (continued)

Construction activities at the site are estimated by GE to be 2 to 3 months behind schedule.

The principal delay being the result of labor jurisdictional disputes.

Mr.

Royce told the inspector that this was not a current cause for delay; however, it was still a sensitive subject area and could result in further delays in the future.

C.

Containment System The installation, overload and initial leak rate tests of the containment system, the dry well and torroidal pressure suppression chamber, by CB&I have been completed.

Significant aspects of these operations were reviewed by the inspector and are discussed in the following paragraphs:

1.

General The installation and testing of the system was completed several months behind schedule.

Mr.

Dunning told the inspector that a major con-tributing factor, in addition to the problem of labor jurisdictional disputes, was the upset in material delivery schedules caused by the then impending strike in the steel industry.

Late deliveries of large quantities of material necessitated the hiring of additional welders, a shortage of which resulted in the acceptance of some welders that would not have been hired otherwise.

As a result, the percentage of welds requiring repair increased from 0.5% to 50 -

75%.

When asked by the inspector what assurance he had that all faulty welds were repaired, Mr.

Dunning stated that this assurance was pfovA4ed by the fact that all welds on the containment system were 100% X-rayed, and that the results were reviewed by qualified representatives of the following or-ganizations:

CB&I, B&R, The Hartford Steel Boiler Inspection and Insurance Company, and GE.

(continued)

Results of Visit (continued)

2.

Expansion Joint Problem The expansion joint in one of the ten vent lines that join the dry well to the torroidal chamber, the fourth going clockwise from the personnel airlock, was found to be distorted when a temporary protective cover was removed from the joint during the initial phase of post-installa-tion testing*,

i.e.,

a low pressure soap bubble test immediately preceding the pneumatic overload test on the dry well.

The faulty joint was sub-sequently replaced.

According to Mr. Dunning, the distortion in the joint, the last to be installed, was the result of torsional and radial stresses imposed during installation when compensating for misalignment between the vent line and the torroidal chamber.

He said that the distortion was inadvertently overlooked by construction supervision at the time of installation and that its discovery was delayed because of the presence of the protective cover.

Mr.

Dunning told the in-spector that the original misalignment problem was corrected by proper mitering during replace-ment of the joint.

He said that the remaining joints were subsequently inspected and found to be satisfactory.

The decision to replace the joint was made sub-sequent to the completion of the pneumatic over-load and leak rate tests on both the dry well and the torroidal chamber.

Post-replacement pressure testing included a repeat of the pneumatic overload test on the dry well, and the performance of hydro-pneumatic overload and leak rate tests on the tor-roidal chamber as originally planned.

(continued)

• Containiaent testing, including results, discussed further in paragraph II.C.3.

0

)

Results of Visit (continued)

Mr. Dunning told the inspector that a report of the expansion joint problem was being pre-pared by him and would be submitted to Jersey Central.

The inspector's zeview of the expansion Joint problem indicated that the corrective measures taken were adequate and in accordance with good engineering practice.

3.

Overload and Leak Rate Test Proqram The inspector discussed with Mr. Dunning the scope and results of the overload and leak rate test programs.

The sequence of significant tests conducted, as told to the inspector, was as follows:

a.

Pneumatic overload test of dry well and vent system at 71.3 psig, 1.15 times the design pressure of 62 psig*.

b.

Pneumatic leak rate test of dry well and vent system at design pressure.

c.

Pneumatic overload test of torroidal chamber at 40.25 psig, 1.15 times the design pressure of 35 psig.

d.

Pneumatic leak rate test of torroidal chamber K

at design pressure.

e.

Repeat of the test described in paragraph 3.a.

because of the replacement of the faulty ex-pansion joint.

(continued)

• .itnessed performance and results discussed in CO REPORT No.

219/65-3, paragraph II.A.

0

) Results of Visit (continued)

f.

Hydro-pneumatic overload test of torroidal chamber at 40.25 psig.

The chamber contained 91,000 cubic feet of water to simulate operating conditions.

g.

Hydro-pneumatic leak rate test of torroidal chamber at design pressure, with the same water present as described in paragraph 3.f.

The preliminary results of the leak rate tests were stated by Mr. Dunning to be as follows:

Test Leak Rate, % Per Day Dry well and vent system at 62 psig 0.064 Torroidal chamber at 35 ppig, dry 0.078 Torroidal chamber at 35 psig, wet

iii0.l (computations incomplete)

According to Mr.

Dunning, Jersey Central repre-sentatives were present throughout the significant phases of containment testing and will be provided with a report of the test results from CB&I, the group responsible for the performance of the tests, through GE.

D.

Reinforced Concrete - Quality Control Program The inspector reviewed the quality control program for reinforced concrete.

Included in the review were the following:

An examination, on a selective basis, of pertinent (continued)

01 10 -

Results of Visit (continued) records including contracts and specifications, testing programs and results; a visual examination of construction field activitiesl; and discussions with cognizant site per-sonnel.

It appears to the inspector, as a result of the review, that adequate measures are in effect to assure that the reinforced concrete will meet the minimum requirements of applicable American Society for Testing and Materials (ASTM) and American Concrete Institute (ACI) codes.

E.

Site Meteorology A 400' meteorological tower has been erected about 1500' southwest of the facility stack.

Mr. Finfrock is over-seeing this aspect of the Oyster Creek Project.

According to Mr. Finfrock, the accumulation of data was started on February 14, 1966, and includes the following:

1.

Wind velocity and direction at 75' and 400'.

2.

Ambient temperature at 10'.

3.

Thermal stability data as reflected by the differences between the temperature at 10' and at 75',

200' and 400'.

4.

Rainfall.

Mr.

Finfrock said that the tower installation was completed ten months behind schedule because of delays encountered in his dealings with State officals, FAA officials, and the contractor.

He said that as a result, the submission to DRL of the desired one year's accumulation of data from the site will be made subsequent to the sub-mission of the Final Safety Analysis Report (FSAR),

tentatively scheduled for July 1966.

(continued)

0

)

11 -

Results of Visit (continued)

F.

Miscellaneous

1. E..pansion Gap, Dry Well - Bioloqical Shield The inspector reviewed a letter from Mr. Kregg to Mr.

Huggins, dated October 26, 1965,.in which a method of attaining the desired ex-pansion gap between the dry well and its sur-rounding biological shield was discussed.

The method discussed proposed the application to the exterior of the dry well, prior to the pouring of the biological shield, of a layer of an inelastic, compressible, asbestos-magnesite cement product.

A layer of polyethylene sheeting would then be installed as a bond breaker at the concrete interface, and the concrete pours made.

The letter stated that the material would com-press about 0.150" during the pouring and curing of the concrete.

Subsequently, the dry well would be filled with steam and heated to 280 0 F.

The resultant pressures from the expansion of the dry well would be sufficient to compress the heated cement product an additional amount sufficient enough to attain the desired gap, 3/8".

This subject area will be reviewed further during future inspection visits.

2.

Proqress Revorts The inspector reviewed monthly progress reports from GE to iJersey Central for the period since September 1965.

One item of interest noted, as extracted from the report for January 1966, is as follows:

(continued)

"* ' "0

')

',A....

,.... Results of Visit (continued)

"An informal meeting was held with AEC Licensing and Regulatory Staff repre-sentatives in Washington to update that group on-the details of the metal-water reaction design basis and cooling system design approach to be utilized in the Oyster Creek Station.

The information presented was well received by the Staff, and as a result of these discussions, the

.. cooling system design was firmed on the basis of providing 4-loop cooling.

Space allocation previously held for the 5th and 6th loops was released for other sys-tem requirements."

3.

Timetable for Significant Events The latest timetable for significant events as obtained from site personnel. and supplemented by information obtained at an information meeting* between DRL and Jersey Central, and attended by the inspector, is as follows:

Event Date Initiation of erection of turbine-generator 4/66 Submission of PSAR 7/66 Submission of technical specifications 9/66 Receipt of reactor pressure vessel at 0ite 9/66 (continued)

• Meeting held at Headquarters on March 24, 1966.

HI-. Results of Visit (continued)

Event Date Completion of installation of reactor pressure vessel within dry well 11/66 Issuance of Notice of Proposed Issuance of Operating Permit 3/67 Initiation of significant preoperational tests 3-4/67 Initiatio.A of loading 7/67 Attainment of full power and I

plant turnover 12/67 G.

Exit Interview A formal exit interview was not held because of the nature of the visit.

Significant comments by those inter-viewed during the course of the visit are contained within the body of the report.

Attachment:

Figure 1

0 JERSEY CENTRAL POWER & LIGHT COMPANY (CO REPORT NO, 219/66-1)

Figure 1 Picture Showing Construction Status as of February 1966

SIem rpm* (a.. 1 M.,

UNITED WTATES GOVERNMENT Memorandum B. H. Grier, Senior Reactor

• ivision of Compliance, HQ FROM:

J ý 0P1O'Reill Senior React Region I, Divi *n of Compli*

SuBJECT: JERSEY CENTRAL POWER & LIGHT DOCKET NO.

50-219 0

Inspector or Inspector anc e COMPANY Decmber 6, 1966 DATE:

The attached report,.by our field the subject facility on November for information.

Attachment:

CO Rpt. r.

219/66-5 by J.

R. 6ears dtd 12/6/66 cc:

L. Kornblith, Jr.,

CO:HQ E.

G. Case, DRL (2)

R. S.

Boyd, DRL (2)

R. G. Page, SLR CO:HQ File inspector of a visit to 15, 1966, is forwarded A

9508030282 PDR FOIA DEKOK95-36 itB U.S. So 950227 PDR li'- Bomb R,&4arlY on Ib Pa A"9 MIN.

MUMN'.,

P

U. S.

ATOMIC ENERGY COMMISSION REGION I DIVISION CF COMPLIANCE December 6, 1966 CO REPORT NO.

219/66-5

Title:

JERSEY CENTRAL POWER & LIGHT COMPANY LICENSE NO.

CPPR-15 ote f Visit:

November 15, 1966 By.

Sears or Inspector

SUMMARY

The pouring of concrete in the reactor building around the dry well has progressed to the next-to-the-top floor level.

The compressible material between the dry well and the concrete shield was observed.

Major mechanical equipment in the turbine building is in place.

The operating: staff is now on-site.

DETAILS I.

Scope of Visit A visit was made to the Jersey Central Power & Light Company reactor, under construction at Oyster Creek, New Jersey, by Mr.

John R. Sears, Reactor Inspector, Region I, Division of Compliance, on November 15, 1966.

The visit included a tour of the constructicn site and discussions with the following:

(continued) 9508030264 9.0227 PDR FOIA DEKOK95-36 PDR

0 Scope of Visit (continued)

Mr.

Abe Dunning, Site Representative, General Electric (GE)

Mr.

Tom McCluskey, Plant Superintendent, Jersey Central Power & Light Company (Jersey Central)

Mr.

Ivan Finfrock, Project Engineer, Jersey Central

4r. Donald Hettrick, Project Engineer, Jersey Central II.

Results of Visit A.

Tour The inspector toured the construction site in company with GE and Jersey Central representatives.

It was observed that major pieces of equipment had been installed in the turbine building, e.g., the turbine shell, the condenser, some tanks.

The installation of some larger sized piping is in progress.

During the tour, a concrete floor slab was being poured for the next-to-the-top floor of the reactor building.

The concrete biological shielding around the dry well had been poured to this level.

The inspector observed that compressible material, which appeared to be similar to mineral asbestos insulation, had been applied to the sides of the dry well.

This was covered by thin polyethylene sheets.

Mr.

Dunning stated that after all the concrete is placed around this material and has set, the atmosphere in the dry well will be raised to 280°F and 20 psig in order to compress the compressible covering.

He stated that GE engineers have calculated that when the dry well atmosphere then returns to ambient conditions, the shrinkage should leave a one half inch gap between the dry well and the concrete.

Mr.

Dunning described the alternate methods being used at Niagara :Mohawk and at Tarapur to allow for dry well expansion at MCA conditions, and said that simple economics of installation costs will determine which method will ba used for future facilities.

(continued)

I ~

I 9'

.4 0

) Results of Visit (continued)

B.

Interview Mr. Tom McCluskey, Plant Superintendent, stated that the following Jersey Central people are now resident at the site:

Mr.

Tom McCluskey, Plant Superintendent, Mr. Richard Doyle, Chemistry Supervisor; Mr. Dbnald Kaulback, Radiation Protection Engineer; Mr. Norman Nelson, Maintenance Engineer; Mr. Woody Riggle, Electrical Foreman; Mr. Fred Cassady, Mechanical Foreman; and also two chemical technicians, two radiation protection technicians, ten operators and four shift foreman.

He stated that the Technical Engineer and two assistants are at GE, San Jose, California, for training.

Mr. McCluskey said that the operators had been chosen from conventional plant operators who had bid for the job.

Successful candidates were selected on the basis of a series of screening tests.

They were given a six week course in reactor engineering at Morristown, followed by ten months of practical on-the-job training at Saxton.

Each operator on-site has been assigned a system of the plant and is presently reading manufacturer's literature on system components toward the goal of writing operating proceiures and cautions.

Mr. McCluskey stated that it is not standard practice in Jersey Central's conventional plants to operate via written procedure, but he affirmed that the Oyster Creek plant will be operated via written procedure because of its newness and the operator's lack of familiarity with such a facility.

J i

UNITED STATES

) wpr y.'

NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 April 28, 1989 Docket N~o.

50-219 Mr. E. E. Fitzpatrick Vice President and Director Oyster Creek Nuclear Generating Station Post Cffice Eox 388 Forked River, New Jersey 08731

£ear Vr. Fitzpatrick:

SUBJECT:

DRYWELL CONTAIN(MENT - OYSTER CREEK NUCLEA; GENERATING STATION (TAC NO.

72C29)

In a letter dated Septenber 12, 1988, GPU Nuclear Corporation (GPUN/the licensee) coirritted to provide an assessment of the drywell corrosion to etE (12 FRefuelling outage) ar, the projected corrosion rate for the following operating cycle.

In a letter dated February 9, 1989, CPUI:

provided the staff with this information.

The pertinent information as given by GPUt' is summarized in Table I (enclosed).

On the basis cf the corrosion rate listed in Table 1, the licersee concluded that the.nccst limiting condition is in the sand bed region of the drywell shell arid the drywell shell thickness is projected to be acceptable until June 1992.

In an attempt to reduce the corrosion rate, the licensee has (1) installed cathodic protection in selkcted sand bed locations, (2) taken steps to eliminate water leakage from reactor building equipment and refueling cavity, and (3) drained water from sand bed region.

In order to assure the structural integrity tilth.c*r-y_*e 11the-- el!*.n see; ha*s* conm, it-ted--pe~ri-dl ec UT t1hi ckn~s s**a ti *i 6f*

tht drywEll shell at all uLte.*s of opporturity.

1l1th licersee er'.1 sized that the projection to June 1992 was based on consErvative approachEs.

Cased on our review of the information provided by GPUN, we concur with the licensee that with the actions taken and to be taken by the licensee to ensure drywell integrity, and that plant uperation car, contilue tL tht 13R refueling Lutage.

In the event that efforts to arrest corrosion are not successful the licensee has argued that existir;ng conservatism would still allow operation.

However, the staff has reservations due to the fact that such conservatisms are not easily quantifiable and are required in assuring drywell adequacy for the protection of public health and safety.

The licensee is required to perform I;3?989042.ot PDR

"*bC';CK CP.QO421 FrDC

'Mr.E. E. Fitzpatrick

-2 April 28, 1989 thickness measuremrets and reconfirm the adequacy of the containment integrity at future cutages of opportunity, incluaing forced cutages requiring dryell entry during the next cycle, but no later than prior to the resumption of power operation following the 13R refueling outage.

Sincerely,

/s/

Alexander W. Dromerick, Project Manager Project Directorate 1-4 Division of Reactor Projects I/I1 Office of Nuclear Reactor Regulation

Enclosure:

Table I cc w/enclosure:

See next page DISTRIBUTION Docket, **.I.

NRC & 'Ucal PDRS PDI-4 Reading SVarga, 14/E/4 BBoger, 14/A/2 SNorris ADromerick OGC (for info. only)

Ei.

EJordan, 3302 MNBB B. Grimes, 9/A/2 ACRS (1.0)

[OYST CRK DRYWELL CONT 72029]

S is ri ck:bd 04/01/89

ý F4 '/10 9o4/AT/89

ENCLOSURE TABLE 1 Location (Elevation)

Nominal Thckness(Inch UT Measured Corrosion Rate (MPY*

8-11 3/4" to 12'-3" 1.15 0.700 0.838

-27.6 + 6.1 (sand bed region) 50'

- 2" 0.77

.725 0.750

-4.3 +.03 87'

- 5" 0.64

.639 0.620*

0

'Aceqpted on the basls of data from certifitc material test reports (CMTRs) amid r, corroslin afttr plant uperatluu (cormutiu, ctLwrree duri:s *t;Itirl).

U. S. NUCLEAR REGULATORY COMMISSION REGION I Report No.

Docket No.

License No.

Licensee:

50-219/90-03 50-219 DPR-16 GPU Nuclear Corporation I Upper Pond Road Parsippany.

New Jersey 07054 Facility Name: Oyster Creek Nuclear Generating Station Inspection Conducted:

January 7, 1990, - February 17, 1990 Participating Inspectors:

M. Banerjee, Resident Inspector E. Collins, Senior Resident Inspector

0. Lew, Resident Inspector Approved By:

R. Hernan, Acting Section Chief, Reactor Projects Section 4B S

1Date Inspection Summary:

Inspection Report No. 50-219/90-03 for January 7, 1990 - February 17, 1990 Areas Inspected:

The inspection consisted of 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> of direct inspection hours by resident inspectors.

The areas inspected included observation and review of plant operational events (paragraph 1.0), the fire protection deluge system (paragraph 2.0), main steam isolation valve leak repair (paragraph 3.0),

drywell wall thinning.measurements (paragraph 4.0), recirculation pump discharge valve failure (paragraph 5.0), recirculation pump "A" seal failure (paragraph 6.0), isolation condenser steam leak (paragraph 7.0), core spray keep fill pumps (paragraph 8.0), engineered safeguard feature system walkdown (paragraph 9.0), monthly maintenance observation (paragraph 10.0), monthly surveillance observation (paragraph 11.0), review of the Fitness For Duty Initial Training Program (paragraph 12.0), and onsite review of Licensee Event Reports (paragraph 15.0).

F.L

_fsý 1.

FC)

Results:

The plant was operated in a safe manner during this inspection period.

Licensee discovery of an inoperable deluge system 16 days after a trouble alarm is an unresolved item.

The absence of documentation of material used in a valve repair is an unresolved item.

Licensee evaluation of recirculation pump seal problems was thorough, and the subsequent removal of the pump from service was well planned and executed.

Recirculation pump discharge valve problems may have contributed to seal failure.

The Standby Gas Treatment System (SGTS) was evaluated as able to perform its intended safety function.

Initial training sessions for the Fitness For Duty Program were well presented.

The licensee changed the date for their estimate to reach minimum code wall thickness in the drywell from June 1992 to June 1991.

The steam leak was also repaired during the 12U-8 unplanned outage by Leak Repair Company.

The inspector reviewed the work package.

A clamp was installed around the bonnet flange, the inside of which was injected with Fermanite 2X material to seal the body to bonnet area leak.

The installation of the clamp was not considered a temporary variation as the installation of the clamp did not affect system function or operation.

The licensee determined that because of the weight of the clamp, the additional loading was acceptable and seismic qualification was not affected.

A final injection of Fermanite was made during startup at 1000 psig reactor pressure.

The leak was minimized to a very small value.

The licensee evaluated it as acceptable.

A permanent repair is scheduled to be made during 13R outage.

The work package did not include any QA paperwork documenting the acceptability of the vendor supplied Fermanite material.

The licensee later identified that a QA receipt inspection was not performed before the Fermanite material was accepted for installation.

This item is unresolved.

(UNR 50-219/90-03-02).

4.0 Drywell Wall Thinning During outage 12U-8, the licensee performed ultrasonic measurements of the drywell wall thickness.

The results showed that the most limiting portion of the drywell had shifted from the sand bay area to the 51-foot elevation and the most conservative estimate of the time when mirirnj, code wall thickness would be reached had changed from June 1992g 7o Fwne 1991.

A telephone conference was initiated by the licensee to inform the NRC about their preliminary findings.

During the conference, the licensee stated that a copy of the revised safety evaluation will be provided to the NRC Project Manager and the resident inspectors.

5.0 "A" Recirculation Pump Discharge Valve On 1/10/89, a plant shutdown was commenced when the "A" recirculation discharge valve failed to close and the recirculation loop was placed in an isolated condition.

When the licensee was able to place the loop in an idle configuration, the plant shutdown was secured and the plant returned to full power.

Technical specifications allow continued plant operation with one loop in an idle configuration.

In an idle loop configuration, the recirculation pump is stopped with the discharge valve shut and the discharge bypass valve and the suction valve open.

If the suction valve is shut, the recirculation loop is considered isolated and the plant must be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The "A" recirculation loop was isolated during an evolution to remove the "A" recirculation pump motor generator from service for maintenance.

The sequence to remove the motor generator from service required shutting the

GPU Nuclear Corporation One Upper Pond Road N uclear Parsippany. New Jersey 07054 201-316-7C00 TELEX 136-482 Writer's Direct Dial Numoer:

(201) 316-7246 April 11, 1990 5000-90-1910 U. S. Nuclear Regulatory Commission Mail Station PI-137 Washington, DC 20555 Attention:

Document Control Desk

SUBJECT:

Oyster Creek Nuclear Generating Station Docket No.

50219, Licensing No.

OPR-16 Oyster Creek Drywell Containment

References:

GPUN Letters 5000-88-1633 dated September 12' 1988, 5000-89-1717 dated February 9,

1989, NRC Letter dated April 28, 1989, and GPUN Letter

.5000-89-1820, dated September 29, 1989 In our conference call on March 8, GPUN committed to provide the staff with information on our recent findings regarding the Oyster Creek drywell wall thickness and our plans to assess and maintain the structural integrity of that vessel.

This letter satisfies that commitment.

It has been our practice for several years to monitor the drywell thickness at selected representative locations by periodic ultrasonic (UT) inspection during plant outages where a drywell entry is made.

One such UT inspection was made during the brief Oyster Creek outage in February of this year.

The wall thickness data obtained during that inspection suggested that corrosion rates in some locations were higher than previously projected.

As a result of those findings, we prepared an update (Revision 4) to our safety evaluation, concluding that, based on present analyses and observed corrosion rates, the drywell's service life can be conservatively confirmed to extend beyond our current operation cycle, that is, mid-1991.

Because the February database was somewhat limited, we decided also to conduct a more extensive examination at the next opportunity.

That examination was conducted during the 12UJ outage (March 26 through April 3, 1990),

and the results are reported below.

We chose also to expand significantly our other ongoing activities to abate the drywell corrosion, to analyze the drywell, and to 'develop methods for any needed drywell repair.

Our plans in these areas are also discussed in this letter.

9004230245 900411 PDR ADOCK 09000219 P

PNU GPU Nuclear Corporation is a subsidiary of General Public Uhilites Corporation

NRC April 11, 1990 Page Tb'o 12UJ INSPECTION PROGRAM For the 12UJ outage, GPUN conducted the following inspections:

1.

We re-inspected areas in the sand bed and the sphere where the February 1990 data indicated apparent changes in corrosion rate.

These additional confirmatory data were obtained from Bay 5 in the Sphere (El.

50'-2") and from Bay 13A in the sand bed region, the same locations where the February 1990 data was taken.

2.

We inspected additional areas in the sand bed region and El. 50'-2" which previously had exhibited lower corrosion rates.

This included eight locations in the sand bed region which had not been inspected since October 1988 and performance of an' A-scan of the accessible portions of the drywell circumference at El. 50'-2".

We also took grids of 7 x 7 UT readings at the three thinnest points found by A-scan of the 50'-2" circumference.

3.

We re-inspected the three regions in the upper cylinder (El.

87'-5")

where we had previously not observed ongoing corrosion. (Had this examination showed ongoing corrosion, the plan called for an A-scan of accessible segments of the drywell circumference at El.

87'-5" followed by taking grids of 7 x 7 UT readings at the three thinnest locations found by, A-scan.

This expanded examination at 87'-5" proved unnecessary.)

An evaluation o1 the data taken in 12UJ as described above indicates that the conclusions of the safety evaluation (SE 000243-002, R4) are unchanged.

This is based upon the determination that corrosion rates in the sand bed and sphere are about the same as those rates calculated from the February 1990 data.

The areas~inspected in the sand bed which had shown low corrosion when last examined in October 1988 have not changed with the exception of one location in Bay 13D.

We plan to redesignate this location as a Priority i location for frequent monitoring pending completion of our evaluation of the 12UJ data.

Our evaluation of the three thinnest locations on El.

50'-2" found by the A-scan shows that the minimum thickness around the circumference is consistent with that at the Priority 1 location currently being monitored.

Finally, our evaluation of the three regions in the upper cylinder showed no ongoing corrosion.

In addition to the UT inspections during 12UJ, we also extracted a 2" diameter sample (core plug) from drywell Bay 13A in the sand bed.

Bay 13A is an area of apparent significant corrosion (based on February 1990 data) which is not cathodically protected.

This core plug was removed and will be chemically and metallurgically examined to determine if significant corrosion is occurring and to identify the corrosion mechanism.

Removal of the plug also permitted removal of a sample of sand for chemical analysis to assess the condition of the sand bed.

NRC April 11, 1990 Page Three While the lab results of the core plug and surrounding sand are not yet in hand, by visual inspection the core plug looked similar to those removed in 1986, and the surrounding sand appeared relatively dry.

ONGOING WORK Based on our conclusions from the drywell inspection activities in February and March of this year, we are proceeding with several parallel work paths on a very high priority.

Our ultimate objective is to ensure that the drywell is, and remains, structurally adequate to meet its intended safety function.

Our workplan includes several main elements, as follows:

o Augmented data acquisition o

Corrosion mitigation:tasks o

Structural analysis o

Drywell modification/repair.

Our plan of attack in each of these areas is outlined in the following sections.

Augmented Data Acquisition Our approach here is to build on the existing database of UT wall thickness measurements and other examinations already conducted, and to continue on an aggressive data acquisition program.

We are considering an augmented effort to include measurements at locations not yet interrogated in order to provide high statistical confidence that our program does in fact characterize the entire drywell vessel.

Our feasibility study of the expanded plan will take into account both accessibility and radiation exposure implications.

Our target is to complete our evaluations by September 1990.

Until implementation of any augmented program, we will continue the current program.

Corrosion Mitigation This involves several activities.

The primary one is to evaluate the effectiveness of the existing cathodic protection (CP) system and to consider design and/or operational changes to enhance its performance.

The system installed at Oyster Creek is quite extensive and was the result of a major engineering effort.

We have been monitoring the effectiveness of this system since placed in operation in 1988.

So far it appears to be less effective than we had hoped.

A system performance test has recently been concluded, and the results are currently being evaluated by GPUN and Corrosion Services Co.,

Ltd.

(the consultant who designed the system).

The results of this evaluation should indicate the level of protection being afforded the drywell and potential enhancements to the operating system.

.~vI

NRC April 11, 1990 Page Four Also, actions will continue to prevent or retard intrusion of water into the gap and the sand bed.

During the 12R outage, a strippable coating was anplied to the refueling cavity prior to refueling in order to eliminate this sorce of water into the sand. bed.

For thie 13R outage, the applicaLion of this type of coating will be expanded to include both the refueling cavity and the equipmcnt storage pool, which is another presumed source of water".

In parallel with the above evaluation of CP and becauseý of the uncertainties in its effectiveness, we are reconsidering other mitigation niethods we previously evaluated, including the use of drying systems, addition of chemical corrosion inhibitors, and chemical inhibitors in combination with CP.

Our target in corrosion mitigation is to develop a course of action by October 1990 with implementation as soon as possible thereafter.

Over the long term, the effectiveness of the installed CP system or any Other selected methcds wilt be monitored by ongoing UT measurements.

Structural Analysis Our objective is to develop a more comprehensive understanding of the dynamic structural performance of the' -rywell vessel under varying conditions in order to ensure that the drywell is structurally adequate for continued use.

This will include application of state-of-the-art techniques for modelling and analyzing the vessel, review,-of the design basis loading conditions, and consideration of the actual material properties of the Oyster (reek vessel.

Our target in this activity is to conclude our structural analysis work by September 1990.

Drywell Modification/Repair Our approach here is to build on previous evaluatiuos of potential structural repair of corrosion damaged areas of the drywell.

This will include review of the previous study performe6 by CB&I Services to define conceptually various options for structural repair in the sand bed region.

This study evaluated selected plate replacement, doubler plates, weld overlay, and stiffener structures as potential repair methods.

This study will be expanded to consider Oyster Creek plant-specific constructability rquirtmi.nts, radiation dose estimates, decontamination and contamination control requirements, radwaste disposal requirements, schedule development, cost estimates, and locations in the drywell most likely to require repair.

Options for repair of elevations above the sand bed will also be evaluated.

Our target is to select a preferred repair option before the 13R outage, and then take steps to be ready to implement that option if and when it is required.

During the 13R outage, drywell walkdowns will be performed to assess physical aspects of the job and to compile the information required to complete selection of and planning for a repair option.

NRC April 11, 1990 Page 'Five In suimary, GPUN's l.*fety E,,luation 000243-002 (Rev.

4) has conservatively confirmed safe operation of Oyster C-'eek through the 13R outage until August 1991.

Our current actions, including ontinued inspections, structural analysis, and corrosion mitig.7tion will establish the basis for continued operation until the 14R outage.

Over the lor.ger term, the repair contingency plan will be devoloped to the extent that it is available to support a timely decision by GPUN regarding stl.r-s necessary to ensure drywell serviceability.

We will continue to keep you informed of our progress in this area.

If you have any questions or you wish to schedule a meeting for further discussion,,

please contact M. W. Laggart, Manager, BWR '.icensing at (201) 316-7968.

j.

C. DeViine, Jr.

Vice President, Technical Functions JCD:mes cc: Administrator Region I U.

S.

Nuclear regulatory Commission 475 Allendale Road King of Prussia, PA 19406.

NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, NJ 08731 Mr.

Alex Uromerick, Jr.

U. S.

Nuclear Regulatory Commission Mail Station PI-137 Washinqton, DC 20555 M SC/ 17

• 0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 July 10, 1990 Docket No.

50-219 Mr. E. E. Fitzpatrick Vice President and Director Oyster Creek Nuclear Generating Station P.O. Box 388 Forked River, New Jersey 08731

SUBJECT:

REVIEW OF OYSTER CREEK DRYWELL CONTAINMENT STRUCTURAL INTEGRITY (TAC No.

75064)

References:

1.

GPU Nuclear "Safety Evaluation on Steel Shell Plate Thickness Reduction" SE No. 00243-002, Rev. 4, Dated February 16, 1990

2.

Letter to NRC from J. C. Devine of GPU Nuclear, Dated April 11, 1990 In February 1990 GPU Nuclear Corporation (GPUN),

the licensee of the Oyster Creek Nuclear Generating Station informed NRC staff that based on UT measurements made in February there is evidence of possible ongoing corrosion at elevation 50'

- 2", which is above the sand bed region, at a rate greater than previously estimated.

As a result GPUN projected that the minimum thickness of the drywell shell will be reached in August of 1991 instead of in June 1992 as previously projected.

In addition the licensee has found the cathodic protection system (CPS) is less effective than expected.

In view of these findings the staff was concerned with the continuous deterioration of the drywell shell and its effect on the structural integrity of the drywell.

On March 8, a conference call between GPUN representatives and NRC staff was held.

The staff requested GPUN to submit a plan for short and long term actions to address the degraded condition of the drywell.

In response to the staff's request GPUN submitted Reference 2 which is summarized as follows.

(A) During the 12 UJ outage (March 26 through April 3, 1990),

GPUN conducted a more extensive examination than than performed in February 1990.

It consisted of inspecting areas in the sand bed and at elevation 50'

- 2" and additional areas in these same regions previously found to exhibit lower corrosion rates, and areas in the upper cyclinder at elevation 8 7' - 5".

In addition to UT measurements, a 2" diameter core plug was taken from drywell bay 13A in the sand bed together with the removal of a sand sample.

From the evaluation of i* 'k:

00 i ~ j V

Mr. E. E. Fitzpatrick

-2 July 10, 1990 the UT measurements taken in the UJ outage, GPUN found that the safe operation of Oyster Creek through August 1991 can be assured as indicated in Reference I is still valid.

(B) GPUN has formulated a work plan which consists of the following elements:

(1) Augmented Data Acquisition -Measurements will be made at locations not yet inspected in order to augment the data acquired as a measure to provide high statistical confidence that the inspection program instituted does in fact characterize the entire drywell.

(2) Corrosion Mitigation - The existing cathodic protection system is being evaluated for its effectiveness and for possible enhancement.

Other methods of mitigation are under consideration.

Measures to prevent or retard intrusion of water into the gap and sand bed are being taken.

(3) Structural Analysis - Use of state-of-the-art techniques for modelling and analyzing the vessel is being considered in conjunction with the use of the actual material properties.

(4)

Drywell Modification/Repair - A study has been made for various options such as selected plate replacement, doubled plates, weld overlay or stiffeners for structural repair in the sand bed region and other areas.

Factors such as constructability and radiation exposure are to be taken into consideration.

A preferred repair option will be selected before the 13R outage, and steps will be taken to be ready for implementing that option if and when it is required.

In accordance with GPUN, the effort outlined above under (A) is to confirm safe operation of Oyster Creek through the 13R outage until August 1991 as indicated in Reference 1, and the effort under (B) (1),

(2) and (3) is to establish the basis for continued operation until the 14R outage.

The effort in (B)(4) above for longer term is formulated as a repair contingency plan.

From the information provided by GPUN under (A) above it can be stated with reasonable confidence that the Oyster Creek drywell minimum thickness will not be violated until at least August 1991 as indicated in Reference 1. However the staff has some reservations on GPUN's use of the effort in (B) (1),

(2) and (3) as a basis for continued operation until the 14R outage, especially GPUN's intention to use strength values of the drywell steel in the certified material test reports (CMTRS).

GPUN's rationale for such an approach is that in the evaluation of the cylinder portion (EL 8 7' - 5") GPUN used the allowable stress derived from CMTRs with the approval of NRC.

However, from an AISI survey of test results for thousands of individual product samples, it has been found that strength levels vary as much as 20% from the CMTR test values.

Therefore it'is the staff's position that minimum specified strength values (e.g, ASME Code minimum

Mr.

E. E. Fitzpatrick 3 -

July 10, 1990 strength values) should be used as the basis for allowable stresses in the stress re-evaluation of degraded components.

Consequently GPUN cannot predicate drywell integrity on CMTR values.

We believe that plans should be made for the implementation of the drywell repair by August 1991 not relying on favorable results of the effort in (B)(1),

(2) and (3) to justify continued operation until the 14R outage.

If the reanalysis effort includes considering changes to the design basis of the plant, a license amendment will be required.

At the same time GPUN should continue the inspection as presently instituted.

We will arrange a meeting with your staff during late August or early September at our Rockville office to hear the status of your work plan and discuss the NRC staff's concerns noted in this. letter.

Sincerely, Original signed by Alexander W. Dromerick, Senior Project Manager Project Directorate 1-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Reguiation cc:

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OFFICIAL RECORD COPY Document Name:

REVIEW OC 75064

A "UNITED STATES 4

• NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 4

,e' October 3,. 1990 Docket No. 50-219 LICEN-2E GPU NUCLEAR CORPORATION JERSEY CENTRAL POWER & LIGHT COMPANY FACILITY:

OYSTER CREEK NUCLEAR GENERATING STATION

SUBJECT:

Summary of September 19, 1990 Meeting With GPU Nuclear Corporation (GPUN) to Discuss Matters Related to Oyster rI Creek DrywellCorrosion.

On Wednesddy, September 19, 1990, a meeting was held at the NRC, One White Flint North, Rockv4lie, Maryland with GPUN, the licensee, to discuss the drywell corrosion problem at the Oyster Creek Nuclear Generating Station. is the list of participants that attended the meeting. is the licensee's morning session agenda. is the licensee's afternoon session agenda.

The following is a summary of the significant items discussed.

The Licensee indicated that the Oyster Creek Drywell (1) has been examined thoroughly:

its present condition and the ongoing corrosion problem are well understood, (2) is a rugged, conservatively designed pressure vessel; it has ample margin to permit continued safe plant operation for several years while corrective action is being taken, and (3) program is a very high priority, resource intensive, and multifaceted one and that GPUN intends to arrest the drywell corrosion by positive means and ensure containment integrity for the full licensed life of the plant.

During the discussion the licensee described a three phase program to address the drywell corrosion problem.

The licensee stated that based on analysis performed during the first phase of the program, GPUN concluded that:

1) current best estimates of corrosion rates at the worst areas of the drywell sphere indicate Code allowable stresses will not be exceeded for at least three years, even if corrosion extended over its entire surface.

2) Taking into account actual conditions, the Oyster Creek Drywell will be in full compliance with the ASME.code for at least three years even at very conservatively projected (95% confidence level) corrosion rates.

3) The Oyster Creek design basis pressure (62 psig) is conservative by a significant margin.

r4)1R The licensee stated that he will submit the details of his program including the structural analysis by December 1990.

The staff advised the licensee that GPUN should expedite the submittal including plans to arrest corrosion.

Alexander W. Dromerick, Senior Project Manager Project Directorate 1-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/enclosures:

See next page WN

t a/

OYSTER CREEK DRYWELL CORROSION GPU Nuclear I NRC Meeting September 19, 1990

MEETING OBJECTIVES

• To communicate the scope, depth, objectives, and expectations of GPUN's Oyster Creek Drywell Program.

• To permit technical exchange among GPUN and NRC technical staff membesm on engineering and analysis Issues.

• To obtain feedback from NRC on GPUN's course of action

• To agree on subsequent steps and their timing.

THREE KEY

1. The and Oyster Creek Drywell has been examined the ongoing corrosion problem are well u Its present condition

2. The Oyster Creek Drywell Is a rugged, conservatively vessel; It has ample margin to permit continued safe several years while corrective action Is being taken. -lr opwonk fcw

3. GPUN's drywell program Is a very high priority, resource Intensive, and multifaceted one; we Intend to arrest the drywell corrosion by positive means and ensure containment Integrity for the full Ucensed life of the plant

OYSTER CREEK DRYWELL PROGRAM Phase:

Phase 1 Phase II Phase III Objective:

Develop Success Path' Solve the Problem Keep It Solved Timing:

Through 1990 Through 1992 Long Term Focus:

0 Examine all Information in-hand.

• Confirm shell Integrity through Phase II.

• Develop detailed plan and engineering for full solution.

• Continue corrosion prevention activities.

• Implement plans/engineering developed In Phase I to:

Fully characterize shell.

Complete analysis of shell strength and margin.

Arrest corrosion.

e Implement life-of.

plant monitoring program.

• Other work as needed.

BASES FOR OYSTER CREEK DRYWELL SAFETY DETERMINATION DURING PHASE II

1. Based on current best estimates of corrosion rates at the worst areas of the drywell sphere, Code allowable stresses will not be exceeded for at least three years, even if that corrosion extended over its entire surface.

2. Taking Into account actual conditions, the Oyster Creek Drywell will be In full compliance with the ASME Code from at least three years, even at very conservatively projected (95 percent confidence level) corrosion rates.

3. The Oyster Creek design basis pressure (62 psig) is known to be conservative by a significant margin.

SUMMARY

OF DRYWELL ACTIVITIES OCT/

NOV '87 JUL' OCT '88 NOV '88 FEB '89 (12R)

DEC '86 CYCLE 12

'90 SANDBED REGION SPHERICAL REGiON (50' -2")

CYUNDRICAL REGION (87' - 5")

- UT Readings

- UT Readings

- Installed

- Energzed CP

- UT Readings cathodic system 3/89 (Feb, March &

Ceproecion (CP), UT adi n

,Apri)

-Removed a"9

- Corm Samples water UT i UReadings.

- Core Samples

- UT Readings

- UT Readings

- UT Readings

- UT Readings

- UT Readings

- Steps taken 6/89 (Feb March &

to nduce

- UT Readngs Apri) water sources 9/89

- Expanded to elevation 51'-10" (Apri)

- UT Readings

- UT Readings

- UT Readings

- UT Readings

- Steps taken 6/89 (March) to reduce water sources I

w Projections Based On Inverse Regression (SCHEMATIC)

A T

H C

K N MIN Thck E L S

S Now Curve Fit Lower 96%

,t nf I 4An &-m Upper 96%

Confidence Limit Bound i

It IB I I Limit Bound Beat Estlmiate I

nK p

TIME (Years)

Projection Based On 95% Confidence

SHELL EXAMINATION

SUMMARY

Sand Bed Region

• ., Extensive measurement data inwhand.

• Corrosion rate is highest of 3 regions (%=39 mils/yr.)

• This is the thickest part of the shell (initially 1.154").

• Several years margin remain based on:

Best estimate corrosion rate In worst area.

Original design basis, Code requirements.

SHELL EXAMINATION

SUMMARY

(Cont'd)

Spherical Region

• 2.5 years data In~hand (although less extensive than sandbed).

• Observed corrosion rate is low (=4.6 milslyr.)

• Initial shell thickness is.722" and.770".

• Several years margin remain based on:

-- Best estimate corrosion rate in worst area.

-- Original design basis.

-- Code requirements.

SHELL EXAMINATION

SUMMARY

(Cont'd)

Cylinder Region

• 2.5 years data in-hand (although less extensive than sandbed).

• No ongoing corrosion observed..

• Environmental conditions make region less prone to corrosion.

• Area.of least margin.

ASME III - SUBSECTION NE EVALUATION

• Code defines "local primary membrane stress intensity" to be greater than 1.1 Smc and less than 1.5 Smc.

• This 10% variation in allowable stress was provided because of the "beam on elastic foundation; effects, i.e., stress decays but remains greater than zero for significant distances.

.Clearly not intended to design for 1.1 Smc, however, given a design that satisfies the code intent, it is not a violation of the code for the membrane stress to be between 1.0 Smc and 1.1 Smc for significant distance.

• Largest exceedance of Smc is 3%. Therefore, drywell currently complies with the code.

A DETAILED INSPECTIONS

• Sandbed Region, EL. 11' - 3" (1986)

• Cylindrical Region, El. 87' - 5" (1987)

• Spherical Region, El. 50'- 2" (1987 & 1990)

• Spherical Region, El. 51'- 10,, (1990)

MIWN ONGOING INSPECTION PROGRAM 0 Outage of opportunity (and)

• Drywell entry for reasons other than program. Inspection

• Priority #1 Locations- _ 3 month frequency

• Priority #2 Locations -

=_ 18 month frequency

CORROSION RATE CALCULATION

• Mean of 49 points

• Mean Is plotted over time 0 Unear regression model/curve fit?

Slope of curve - calculated corrosion rate

• Mean model/curve fit?

No slope - No corrosion rate

Curve Fit Based On Linear Regression T

H I

C N

E S

S TIME (Years)

Adawwm&býmlml 0"imi i

ýW-M6 I

,qW Curve Fit Based On Mean Model A

T H

I C

K N

E S

S Now L

Mean Thickness I

0 r

W Curve Fit W

TIME (Years)

-~

Projections Based On Inverse Regression (SCHEMATIC)

T H

I C

K N MIN Thck EL S

S Now L

Curve Fito~

Lower 95 6%n rnf l i4n.em Upper 96%

Confidence

/

Limit Bound Limit Bound Beat Eetlmýite

<1 I

TIME (Years)

Projection Based On 95% Confidence

SIGNIFICANT CORROSION RATE CONCLUSION (AS OF APRIL, 1990)

• Spherical region, elevation 50' - 2"

- Bounding calculated corrosion rate = 4.6 +/-1.6 MPY

• Sandbed region, elevation 11' - 3" Bounding calculated corrosion rate = 39.1 +/-3.4 MPY

• Sandbed, cathodically protected regions No significant corrosion rate reduction (15 to 25 MPY)

• Cylindrical region elevation 87' - 5" No observed ongoing corrosion

• Spherical region elevation 51' - 10" Calculated corrosion rate not available

Current Projections (Based on Data Up To April 1990)

T H

I C

K N

E S

S Best Estimate-MIN Thck L

Sandbed Elev. 50'-2" Elev. 51'-109 Jul 1993 Jun 1992 Oct 1991

• Mar 1994 Sep 1994 Jul 1993

-NOTE: Projection Based on E!ev. 50'-20 Corrosion Rate i -,

SUMMARY

This augmented Inspection plan, using 60 locations selected at random, provides a statistically based characterization of the drywell. The Inspection plan provides a sensitive test for unacceptable observations. Measurements of the region adjacent to a low area, should one be found, will be made In order to show that the condition of the plate Is, in general, much better.

CONTROLLING LOAD CASES FOR VARIOUS LOCATIONS (Reported from analyses completed from 1986 to 1988)

• CYLINDRICAL REGION - (Design t=0.640", min. as found t=0.619") Accident Condition - Primary membrane stress caused by design pressure dominates

• SPHERICAL REGION (Design t=0.722") - Accident Condition - Primary membrane stress caused by design pressure dominates

• SPHERICAL REGION (Design t=0.770") - Accident Condition - Primary membrane stress caused by design pressure dominates

• SPHERICAL REGION SANDBED (Design t=1.154", min. as found t=0.808",

assumed t=0.700") - Refueling Condition - Buckling due to compressive stresses caused by deadweight and water in refueling cavity + 2 psi external pressure dominate

REVIEW OF RESTART EVALUATIONS (198611987)

• CYLINDRICAL REGION:

Established minimum as found thickness of 0.619" accepted using CMTR data and the fact that there is no ongoing corrosion.

• SPHERICAL REGION SANDBED:

The stress analysis was performed to ensure structural Integrity for the shell assumed to be 0.700" thick. This configuration subjected to the combined load cases yielded the following conclusions:

The tensile stresses were less than the specified allowable stress from the 1962 issue of the ASME Code,Section VIII, Including the Summer 1964 Addendum plus Code Cases 1270N-5 and 1272N-5 (1.1 Sm= 19,250 psi).

The compressive stresses were less than the specified allowable stress computed according to rules of Code Case N-284.

ADDITIONAL STRUCTURAL ANALYSIS 0 COMPARISON WORK FOR BUCKLING EVALUATION:

Capacity margin (buckling) in the sandbed is improved by Including the details of vent piPeand its reinforcing plates.

Stability analysis comparing 3-D FEM methods and BOSOR techniques (shell of revolution) using a similar Mark I drywell has been performed using the same percentage reduction in wall thicknesses as observed at Oyster Creek In the sandbed region.

Loads were adjusted to produce a stress state at the midpoint of the sandbed equal to that computed for the Oyster Creek stability analysis.

The ratio of the FEM results divided by the BOSOR results was computed and is equal to 2.1. Hence, the previously computed capacity margin of 1.00 is very conservative.

CORROSION ASSESSMENT CONCLUSIONS

-Different local environments most likely exist within the drywell annular space which would explain various corrosion rates observed.

P-Aqueous corrosion is primarily responsible for the metal loss.

Galvanic action, oxygen, pH and temperature are most likely Influencing the rate.

,-Corrosion mitigators must be aimed at changing local environments as well as global environments, le. we must utilize a mitigative scheme which deals with the bulk environment in the sandbed or insulation material and with the environment in the oxide crust.

P-Corrosion rates are within the bounds discussed in the literature for aqueous corrosion. Therefore, we do not expect to find regions of the drywell with more extensive metal loss than that already observed.

UNITED STATES

, oNUCLEAR REGULATORY COMMISSION WASHINGTON. 0 C. 20555 October 16, 1990 Docket No.

50-219 Mr.

J. D. DeVine, Jr.

Vice President and Director Technical Functions GPU Nuclear Corporation One Upper Pond Road Parsippany, New Jersey 07054

SUBJECT:

DRYWELL CORROSION PROGRAM -

OYSTER CREEK NUCLEAR GENERATING STATION On September 29, 1990, GPU Nuclear Corporation (GPUN) met with the NRC staff to discuss the Oyster Creek Nuclear Generating Station's Drywell Corrosion Program.

During the meeting, GPUN requested that the staff provide feedback regarding the Drywell Corrosion Program.

As a result of the discussions held during the meeting the staff so far has identified the following aspects of GPUN's presentation that call for staff feedback.

These are:

1) sampling of shell surfaces for UT measurements,

2) appropriateness of the use of ASME Section III Subsection NC, and 3) the need for detailed review of preliminary results of the stress analysis presented by GPUN.

The Enclosure provides

'details of the required clarification.

If during our ongoing review of your program additional items requiring further clarification are identified we will notify you.

If you have any questions regarding the above, please contact me.

Sincerely, Alexander W. Dromerick, Senior Project Manager Project Directorate 1-4 Division of Reactor Projects : I/I1 Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/enclosure:

See next page

ENCLOSURE REQUESTED CLARIFICATION REGARDING OYSTER CREEK CORROSION OF DRYWELL SHELL

ýDOCKET NO.

50-219 There are several aspects of the licensee presentation that call for staff feed back, these are:

i) sampling of shell surfaces for UT measurements, ii) appropriateness of the use of ASME Section III Subsection NC, and iii) need for detailed review of preliminary results of the stress analysis presented by the licensee.

i)

Sampling plan for monitoring drywell corrosion:

The licensee presented a statistically based inspection program of the entire shell surface not embedded in concrete.

However, based on the results of observation so far, the licensee presented a correlation between corrosion and presence of moisture for example, in the sand region the plug samples 15A and 1hA-H were dry and had corrosion rates equal to zero.

It is not clear to the staff how the licensee plans to locate sensors for on-line monitoring of drywell corrosion rate at those places where the presence of moisture is likely.

The staff needs to review the statistically based sampling plan.

ii)

The original design code for the Oyster Creek shell is ASME Section VIII.

Should the licensee choose to use a more recent code, there will be a burden on the licensee to clearly establish that the material selection, design, fabrication, inspection and surveillance in service are all in accordance with the requirements of the current code which should be the ASME Section 1II, Subsection NE, and Section XI.

iii) It is clear that through the corrosion process, the margin for over pressure capacity of the containment has been reduced (see GDCSO and 51).

Therefore, the staff judgment as to the adequacy of the drywell shell margin must be based on a detailed review of the stress calculations and the stress allowables.

iv)

In your presentation you indicated that there has been leakage from refueling cavity liner, equipment pool and spent fuel pool.

Describe the actions you will take to prevent leakage from these structures into the drywell gap and the effect of the leakage on other structures or equipment.

U. S. NUCLEAR REGULATORY COMMISSION REGION I Report No.

50-219/90-21 Docket No.

50-219 License No.

DPR-16 Licensee:

GPU Nuclear Corporation P.O.

Box 388 Forked River, New Jersey 08731 Facility Name:

Oyster Creek Nuclear Generating Station Inspection At:

Forked River, New Jersey Inspection Conducted:

October 29-31, 1990 Inspector:

Approved by:

H.

Raeactor Engineer, Materials and Processes Section, EB, ORS E. H. Gray, Chieef, Waierials and Processes

Section, EB, DRS date date Inspection Summary:

Ln_*q._ecion on October 29-31 1990_(Report No.

5o-219/90-?1)

Areas Inspected:

An announced Inspection of the licensee's activities involving the drywell corrosion problem activities.

The scope of this inspection included review of ultrasonic thickness procedures and records, inspection and repairs of suspected sources of leakage, review of metallurgical reports and a facility tour.

Results:

On the basis of this inspection, it was concluded that the licensee's program for monitoring, repairing and evaluating the corrosion problem was comprehensive and was being conducted in a systematic manner in accordance with prescribed procedures.

Of the area inspected, no violations were identified.

The licensee has presented substantial evidence that the plant can be operated safely until the 14R refuel outage provided that thickness measurements are taken in the prescribed intervals, and show no significant loss in wall thickness.

9101020041 901219 PDR ADOCK 05000219 Q

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DETAILS 1.0 Persons Contacted 1.1 GPU Nuclear Cor ration"

• E. E. Fitzpatrick, Vice President and Director

• J. A. Martin, Mechanical Engineer

• J. 0. Amramovici, Manager, Pressure Vessels

• R. Zak, Licensing Engineer

• S. Gicobbi, Manager, Materials Engineering 1.2 U.S. Nuclear Regul atory Commis sion NRC_

• G. Bagchi, Office of Nuclear Reactor RegulaLion (NRR),

ESGB

• E. Collins, Sr. Resident Inspector

• Denotes attendance at exit meeting on October 30, 1990.

2.0 Scope The objective of this inspection was to review the licensee's continuous on site activities regarding the drywell corrosion problem.

The results of a plant walkdown of accessible areas and an evaluation of the licensee's analytical methodology by NRR will be reported separately by Mr.

Goutam Bagchi.

The overall strategy to monitor and control drywell corrosion had been presented by the licensee in a meeting held in Headquarters on September 19, 1990.

3.0 History Corrosion was initially discovered by the licensee on the outside surface of the drywell in the sand cushion region of the drywell in late 1986.

Since then, the licensee has carried out an extensive program to ensure the short and long term integrity of the drywell.

The program includes continuous monitoring of the corrosion as reflected by frequent thickness measurements, inspection and repair of suspected sources of leakage which are believed to be responsible for the leaks, reanalysi.s of the drywell stresses, and a study of feasible corrective actions.

The corrosion apparently was caused by moisture trapped inside the thermal insulation surrounding the drywell and in the sand cushion around its base..

The highest corrosion rate has occurred in the sand bed area (39 mils/year) followed by the spherical region (4.6 mils/year).

No recent corrosion has been observed in the upper cylinder region.

Although the calculated stresses based on thickness measurements and corrosion rates indicate a marginal condition from the standpoint of code allowable stresses, the licensee has concluded that the drywell will still be in compliance with the code at refuel outage 14R on the basis of assuming that the major source of leakage has been eliminated.

3

4.0 Findings

4.1 Ultrasonic Thickness Measurements The inspector reviewed the methods and appropriate records associated With ultrasonic thickness determinations.

The measurements are obtained from the inside of the dry'well using a calibrated ultrasonic instrument (0 METER) in accordance with GPUN Procedures 6150-QAP-7209.07 Rev.

0 and 1S-322227-004 Rev.

2.

Forty-nine (49) individual readings are taken in 11 discrete areas using a 6 inch x 8 inch grid template.

The 11 arctas covered 7 areas in the sand bed area, 3 :in the cylinder region (87' level) and I in the spherical (51')

level.

To assure validity of the data, the instrument is calibrated before each set of data is taken.

In the presence of the inspector, the licensee demonstrated the accuracy of the instrument using the specified stepped calibration standard.

The inspector reviewed 2 recent data sheets 87-026-135 and 87-026-143 representing Bay No.

19 Area C (sand bed) and Bay No.

13 Area 6 (52'). Except for three anomalous points in 87-026-135, the inspector found no discrepancies.

The three points were subsequently attributed to a welded plug in an area in which a core bar had been previously removed.

The data is subsequently sent to GPU Engineering in Parsippany, New Jersey for analysis.

Basically, the data points for each sector are averaged, statistically analyzed and compared with previous data to calculate conservative stress values as determined by corrosion rates and wall thickness measurements.

In addition to performing wall thickness measurements during the last outage (12R),

the licensee removed a core sample from the sand bed Area 13A as part of his continuous effort to monitor the drywell corrosion.

The inspector reviewed the GE metallurgical report covering evaluation of core bar 13A.

The report concluded that the findings were similar to those generated in previous core bar evaluations and that no basic changes occurred in the conditions driving the corrosion of the drywell.

4.2 Repair Activities The inspector reviewed certain aspects of the licensee's activities regarding the inspection and/or repair of the suspected sources of leakage.

The major source of leakage which appears to be responsible for the cnrrosion of the drywell shell is the reactor cavity liner.

The cavity is filled with demineralized water during refueling and thus provides a direct leak path to the outside surface of the drywell if there were defects in the liner.

The inspector reviewed comprehensive visual and liquid penetrant inspection reports as documented in Material Nonconformance Report 87-240 which showed that the.109" thick type 304 stainless steel liner exhibited numerous cracks on its I.D. surface in additio~n to 2 severely damaged areas which-were reported have been caused by movement of equipment used in refueling.

The cracks showed no preferred orientation or preferred location with regard to base metal or welds.

The inspector reviewed a metallurgical report (General Electric 88-178-006) which covered an evaluation of two

4 through-wall samples which were removed from the cavity liner to include the cracks.

The investigation did not disclose any material deficiencies or anomalies associated with the failure.

Although the cracks were found to be transgranular, no detrimental anions such as Cl or F which are known to cause transgranular stress corrosion cracking were found to be associated with the cracking.

I The report concluded that because of the wetted surface and thermal fluctuations, the most likely cause of failure was corrosion fatigue.

The source of stress was believed to have occurred during initial welding and the restraint caused by welding to backing strips embedded in the concrete.

The fluctuations may have been higher than anticipated because the liner was found to be.109" instead of the specified.250".

The conclusions in the subject report appear to be valid.

Because of the excessive number of defects found in the cavity liner, the licensee opted to employ a unique, temporary system that covered 100% of the I.D. surface.

The system consisted of a combination of stainless steel adhesive tape covered by two coats of a Latex barrier (ISOLOCK 300).

The licensee provided the inspector a report (TOR-938) which showed that the tape-coating had been qualified for 125' F-10 week immersion service using both adhesion, pressure and leachate testing.

The system is designed to be removed after refueling and is applied with the reactor head in place.

The inspector reviewed other documents pertaining to the inspection and repair of the suspected sources of leakage.

These are listed below:

IS-328 257-001 - Repair of Reactor Cavity Concrete Trough Material Nonconformance Report 85-034 Weld Repair and Inspection of Weld Defects in'Equipment Storage Pool Technical Specification - SP-1302-22-006 of Reactor Cavity - Repair of Reactor Cavity and Storage Pool Lining Material Nonconformance Report 87-240 Installation Specification for Replacement of Drywell Vessel Core Sample Plugs The Inspector's review of these documents indicated that the prescribed activit12 s were performed in accordance with appropriate procedures:

Repair welds were inspected 1using various NDE procedures (magnetic particle, liquid penetrant and vacuum box).

Documents included Quality Assurance require-ments including inspection points and records.

A sampling of welding activities Indicated the use of appropriate ASME Section IX qualified procedures, The licensee is currently exploring methods for removing the wet sand and possible repairs to reinforce the drywell ii required.

The cathodic

5 protection system which has been in operation for several years has not been effective apparently because the major source of leakage has been eliminated.

5.0 Conclusions On the basis of the above findings, the inspector concluded that the licensee's program for monitoring, repairing and evaluating the corrosion problem was being conducted in a systematic manner in accordance with prescribed procedures.

Since the major sources of leakage has been found and corrected, no significant leakage has been observed as indicated by frequent inspections of five sand bed drains.

6.0 Management Meetings Management was informed of the scope and purpose of the inspection at the entrance meeting at the start of the inspection.

The findings of the inspection were discussed with licensee representatives during the course of the inspection and presented to licensee management at the October 30, 1990 exit interview (see Paragraph 1 for attendees).

At no time during the inspection, was written material provided to the licensee by the inspector.

The licensee did not indicate that proprietary information was involved within the scope of this inspection.

February 14, 1991 Docket No.

5 0-219 Distribution:

Docket File BDLiaw NRC & Lucz,l PDRs ACkS (10 PD 1-4 Plant CWHehl Mr. John J. Barton, Director SVarga Oyster Creek Nuclear Gererating Station EGGreenman P. C. Uox 388 SNorris Forked Piver, New Jersey 08731 ADromerick 00C

Dear Mr. [.artG,

[:

FJordan SUPJFCT:

I'EQUEST F0P. ADDITIONAL INFORMATIC40!

ON OYSTER CREEK DRYWEIL.

STRESS AND STAF1L.ITY ANALYSIS (TAC 1iO.

79166)

The staff has reviewed the GE reports Index rio. 9-I and 9-2, "An ASME Section VII Evaludtion ol the Oyster Creek Drywell Stress and Stability Analysis" and our coinments ?nd request for additior;al information are contained in the enclosure.

We request that the information be provided within 30 days of receipt cf this letter.

If you have any questions regarding this request, please contact me.

The requircments of this letter affect fewer than 10 respondents and therefore, arc not subject to Office of Management and Budget review under P.L.

97-51].

Sincerely,./.

Alexander W. Dromerick, Senior Project Maiajer Project Directorate 1-4 Division of Reactur Projects -/1i Office of Muclear Pcitctor Regulalion Fnclo=_~r e:

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Comments on GE Reports Index No. 9-1 and 9-2 An ASMR Section VIII Evaluation of the Oyster Creek Drywall Stress and Stability Analysis PART I - Stress Analysis

1.

Page 2-3, first paragraph Reference is made to Table 2-1 which shows the 95 percent confidence thickness values in the locally corroded areas of the drywell.

The basis and method of calculating these projected thicknesses should be explained.

Furthermore, the anticipated date for reaching these projected thicknesses should be specified.

2.

Page 2-5, first paragraph The last sentence states that "given a design which satisfies the general code intent, as the Oyster Creek drywell does as originally constructed, it is not a violation of Subsection NE requirements for the membrane stress to be between 1.OS.

and 1.iS, over significant distances."

Further justification for the licensee's position should be provided.

Under what conditions would this become a code violation?

In other words, at what point does the "local" region become a "general membrane" region? Has the opinion of the Code Committee been solicited regarding this matter?

If reference is to be made to Code Case N-480, the specific portions of the Code Case as it applies to the Oyster Creek drywell situation should be fully explained.

3.

Page 5-2, Section 5.4 This section states that "the membrane stresses for the degraded thickness condition were obtained by scaling upwards the calculated stresses for the nominal thickness case (Table 5-2) by the thickness ratio." It should also be explained how the primary membrane plus bending stresses shown in Table 5-3 were obtained.

It appears that the combined stress was scaled upwards linearly by the thickness ratio.

However, the bending portion of the stress should be scaled by the square of the thickness: ratio.

Also, the effect of stress concentrations due to the change of thickness should be addressed.

4.

Appendix A, page 21, second paragraph The last sentence states that "impact testing would not be required by the present code rules unless the LST (lowest metal service temperature) were less than 300F, and the Oyster

Creek drywell material would not require impact testing."

Earlier in this section it is stated that an LST of 30-F was used for the Oyster Creek design basis.

Is the LST for the dryvelL-monitored by any plant operating procedures or the Technical Specifications?

Have studies and plant operating history demonstrated that the drywell shell temperature is not expected to be lower than 30"F for all loading conditions?

5.

Appendix F, page 1, first paragraph What is the basis for performing the sand sensitivity study with a nominal sand stiffness of 366 psi/inch and a sand stiffness of 80 percent of the nominal value?

Were studies and/or tests performed to support these assumptions?

Otherwise, the sensitivity study should be conducted further with lower stitfnes* values.

The licensee's letter of December 5,

19'00 indicates that structural calculations assuming the sand removed would be completed by December 31, 1990.

The results of these studies should be provided to demonstrate the sensitivity of the stresses to the assumed sand stiffness.

PART 2 - Stability Analysis

6.

Page 2-3, Section 2.3 This section states that the method described in Reference 2-5 was used to quantify the effect that the orthogonal tensile stress has on reducing the effect of imperfections on the buckling strength.

The sensitivity of the results should be studied by using other methods which also address this effect.

7.

Page 2-4, Section 2.4 This section states that Reference 2-6 was used to calculate the plasticity reduction factor for the meridional direction elastic buckling stress.

Since this approach apparently has not been incorporated into Code Case N-284, the sensitivity of the results should be studied by using other methods which address this effect.

8.

Page 3-3, second paragraph For the stability analysis the stiffness for the sandbed was assumed to be 366 psi/inch and no sensitivity studies are reported.

As described in Question 5, the results of the stability analysis with the sand removed should be provided.

9.

Page 3-6, Section 3.5.3 The first sentence states that "the 2 psi external pressure load for the refueling case is applied to the external faces of all of the drywell and-vent shell elements."

Unless it can be demonstrated that this pressure actually is present at all times during normal operation and refueling, the effect on the buckling analysis results of assuming no external pressure for these two load cases should be reported.

Furthermore, is it possible to have an external pressure greater than 2 psi on the drywell shell?

If so, an enveloping pressure case should be considered in the analysis.

PART 3 - General

10.

Justification for the use of ASME Section 1I1, Subsection NE has been pro-vided to evaluate the Oyter Creek Steel drywell, taking into consideration DESIGN, material's, fabrication inspection and'testing with exception of the comments indicated above, the justification appears to be reasonable.

Since the present-day quality assurance and quality control requirements for the design and construction of nuclear power were in the formative stage at the time when the Oyster Creek Plant was designed and constructed.

indicate what quality assurance and quality control programs were imple-mented for the Oyster Creek drywell.

Ini,-te if documentation of the programs is available.

11.

In GPU's presentation to the staff in September, 1990, it was indicated that GPU would have an on-line thickness measurement capability in the critical areas of thickness measurement.

GPU has i current commitment toý make UT measurements at outages of opportunity.

t'ate clearly what on-line thickness measurement program GPU will have 'uring the fuel cycle starting in early 1991.

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.

20555 February 14, 1991 NRC INFORMATION NOTICE NO. 86-99, SUPPLEMENT 1:

DEGRADATION OF STEEL CONTAINMENTS Addressees:

All holders of operating licenses or construction permits for nuclear power reactors.

Purpose:

This supplement to Information Notice (IN) 86-99 Is intended to alert addressees to additional information about a potential degradation problem regarding corrosion in steel containments.

It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.

However, suggestions contained in this supplement to the information notice do not constitute NRC requirements; therefore, no specific action or written response is required.

Discussion:

IN 86-99 was issued on December 8, 1986, in response to the discovery of signif-icant corrosion on the external surface of the carbon steel drywell in the sand bed region of the Oyster Creek plant.

This supplement updates the status of Oyster Creek containment corrosion and the licensee's mitigation program.

Since drywell corrosion was detected in 1986, the licensee instituted periodic wall thickness measurements by the ultrasonic testing (UT) technique to deter-mine corrosion rates.

The most severe corrosion was found in the sand bed region at a nominal elevation of 11'-3".

The highest corrosion-rate-determined was 35.2+/-6.8 mils per year.

To mitigate the corrosion in the sand bed region, water was drained from the sand bed and cathodic protection (CP) was installed in the bays with the greatest wall thinning in early 1989.

Subsequent UT thickness measurements in these bays indicated that CP was ineffective.

The licensee's consultants indicated that it would be necessary to flood the sand bed and to install CP in all the bays to make the CP system effective.

The licensee decided that large amounts of water in the sand bed would be counterproductive.

9102080329

IN 86-99, Supplement I February 14, 1991 Page 2 of 3 In the spherical portion of the drywell above the sand bed region, the highest corrosion rate determined was 4.6+/-1.6 mils per year at a nominal elevation of 51'.

In the cylindrical portion of the drywell above the spherical portion, where minor corrosion was discovered and was thought to have originated mostly during construction, no significant wall thinning was detected (at a nominal elevation of 87').

However, this is the region in which the nominal thickness of the wall has the least margin, thus requiring periodic monitoring of actual thickness.

The licensee has instituted a drywell program to arrest corrosion and to ensure containment integrity for the full licensed term of the plant.

The licensee has taken action to investigate, identify, and correct leak paths into the drywell gap and plans to take more action to survey leakage and prevent it.

The stainless steel liners in the refueling cavity and the equipment pool developed cracks along the perimeter of the liner plates where they were welded to embedded channels.

For the refueling cavity, all potential leakage pathways have been thoroughly checked and liner cracks are sealed with adhesive stain-less steel tape before a strippable coating is applied.

Since the refueling cavity is flooded only during refueling, no leakage concerns exist at other times.

At the end of an outage, the refueling cavity is drained, and the tape and strippable coating are removed.

The licensee found leaks related to the equipment pool and stopped them with liner weld repairs.

The equipment pool also will be protected with a strippable coating during flooded periods of operation.

The licensee believes that a thorough program has been established for managing leakage that could affect drywell integrity due to corrosion from moisture ingress into the drywell gap.

Recent surveillance of the sand bed drains indicates that the sand bed is free of water.

To further mitigate drywell corrosion, the licensee is considering removing the sand, insulation, gap filler material, and corrosion film and applying a protective coating to the exterior drywell surface.

The licensee is proceeding with the analysis, engineering and planning to support removing the sand from the drywell sand bed region in the near future.

Removal of the insulation and gap filler material

-from-the-dryweil--gap-t-s--being--

eva-luated-for-future-cons-iderati-on...........................

The BWR Owners Group is surveying its members to determine whether other plants are experiencing water leakage into the drywell gap and possible corrosion of the exterior surfaces in the sand bed region as well as in the spherical and cylindrical parts of the drywell.

IN 86-99, Supplement 1 February 14, 1991 Page 3 of 3 This supplement requires no specific action or written response.

If you have any questions about the information in this notice, please contact one of the technical contacts listed below or the appropriate NRR project manager.

Charles E. ROgst, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical Contacts:

Frank J. Witt, NRR (301) 492-0767 C.P. Tan, NRR (301) 492-3315

/

Attachment:

List of Recently Issued NRC Information Notices 7Z

MAY 2 3 1991 MEMORANDUM FOR:

FROM:

SUBJECT:

Plant Name:

Licensee:

Request Status:

Tac No.:

John F. Stolz, Director Project Directorate 1-4 Division of Reactor Projects I/I1 Goutam Bagchi, Chief Structurdl and Geosciences Branch Division of Engineering Technology REQUEST FOR ADDITIONAL INFORMATION REVIEW OF OYSTER CREEK CORRODED DRYWELL ANALYSIS Oyster Creek Nuclear Power Plant GPU Nuclear Corporation Request for Additional Information M79166 The staff of the Structural and Geosciences Branch has reviewed the licensee's

'esponses to the staff's previous request for information (CPU Mdrch 20, 1991 Letter) ard the information provided on the drywell analysis with the sand removed (GPU March 4, 1991 Letter).

In order to complete uur review, we find more information is required.

The required infurmation is contained in the enclosure.

The review was performed by C. P. Tan of the Geosciences Section with the assistance of consultants from Brookhaven National Laboratory.

/1/

Goutar. Eagchi, Chief Structural and Geosciences Urdnch Division uf Engineering Technoflogy

Enclosure:

As stated cc:

J.

E. Richarason B. D. Liaw A. Dromerick DISTRIBUTION:

Central File ESGB R/F CPTan RRothman GEagchi 9106040 4 5 0 910523 CF ADOCPK 05000219 ESGB:DET CPTan e r 0 5 /./19 1

0 5/-'7-91

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ENCLOSURE ESGB REQUEST FOR ADDITIONAL INFORMATION ON OYSTER CREEK CORRODED DRYWELL ANALYSES

1.

Your response to question 2 states "For a vessel that originally complied with the code, increases beyond 1.0 Smc in localized areas of undefined size are acceptable."

This statement is loose and you have applied it throughout the drywell as shcwn in tables 5-1b, 5-2a and 5-2b of GE Report Index No. 9-3.

One may conclude from what you have stated and implemented that a corroded drywell has increased its structural capability.

Your interpretation of section NE 3213.10 of the ASME Code is questionable.

Without corrosion, you consider the drywell when subjected to internal pressure to be under general primary membrane stress (tensile) and with corrosion you consider it to be under local primary membrane stress (tersile).

NE 3213.10 considers a membrane stress to be local primary if it is produced by pressure or other mechanical loading and associated with a primary or discontinuity effect, resulting in excessive distortion in the transfer of load to other portions of the structure.

NE 3213.10 specifies the region to be considered local over which the membrane stress intensity exceeds I.i Snic.

The code gives an example of the discrete regions of local primary membrane stress.

We realize that there is no code limit for the extent of the region in which the membrane stress exceeds 1.0 Smc but is less than 1.1 Si;c.

Lugical judgement is to be exercised in the interpretation, and the basis for your judgment should be clearly defined.

Even if your interpretation of NE 3213.10 for application to the corroded Oyster Creek drywell is acceptable for localized areas, it should be demonstrated that the present and projected corroded condition of the Oyster Creek drywell falls within the boundaries established in accordance with NE. 3213.10.

Unless and until the staff's concerns as indicated dbove are satisfactorily resolved, the staff has reservations on, your use of 1.1 Smc as indicated in GE4Report Index No. 9.3.

This means that the allowable stresses indicated in tables 5-1b, 5-2a and 5-2b shuuld be based on 1.0 Smc for primary membrane.

2.

The response to Question 3 coes not fully address Lhe question regarding possible stress concentrations resultiny from the corroded condition of the drywell.

This issue should be fully discussed, noting that at corrosion locdtions the change it, the plate thickness is tut likely to be tapered as assumed in your analyses.

3. In GE Report Index No."9-3, Section 5.2.2, compariscns 0f circuw;ferential and reridional stress magnituoes with tte large arid :,wall displacement options should be provided from the Sandbcd regior up tu the kruckle region of the drywell.

The amount of stress reduction obtained t-.3d result of the large displacement method appears to be too high for the small deflection calculated; the results of these calculations should be further investigated.

Also show mathematically as in the case of beams and flat plates, that consideration of large deflection decreases the stress in the drywell shell which is in membrane tension under internal pressure for regions of the -shell away from the discontinuity.

4.

In GE Report Index No. 9-3, Tables 3-3 ar, 3-4 indiLate the large concen-trated loads considered in the analysis; however, these loads are uniformly distributed along the circumference of the pie slice finite element model.

at variouselevations.

Since the stresses in the corroded regions of the drywell are close to the allowables, what effect would a more refined treatment uf these Ioads have on the stress evaluation?

This question should be addressed for all drywefl regions (i.e., cylinder, knuckle, upper sphere, middle sphere, lower sphere, and sandbed).

The response should consider stresses directly under the lcad (if corrosion in this area is prestnt), as well as the effect on the stress distribution at further distances from the load.

5.

In GE Report Index No. 9-3, Section 3.2..3 indicates that the seismic loads are imposed or the pie slice model by applying forces at four elevations of the model and matching stresses at selected elevations with those from the axisymmetric model.

how sensitive are the calculations to the location and number of elevations chosen to match the stresses?

How well do the stresses compare-at other elevations in the drywell?

6.

In order to examine your analysis in more detail, the staff requests that you provide the ANSYS input file for both the axisymmetric and pie slice models.

This irfurmation should be provided on a high density 5 1/4 in. floppy disc forian IBM PC.

September 3, 1991 Docket No. 50-219 Mr. John J. Barton Vice President and Director Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731

Dear Mr. Barton:

Distribution:,

NRC & Local PDRs PD 1-4 Plant SVarga JCalvo SNorris ADromerick OGC FJordan GBagchi ACRS (10)

CWHeh 1

SUBJECT:

OYSTER CREEK NUCLEAR GENFRATING EATION EVALUATION OF STRUCTURAL INTEGRITY OF A CONTAINMENT (TAC NO.

79166)

At a meeting held on July 24,

1991, the NRC staff Corporation (GPUN) that they would inform GPUN on application, of the ASME Code in the evaluation of STAFF POSITION ON DE;RADED STEEL advised GPU Nuclear the staff's position on the degraded steel containments.

Enclosed is the staff's position regarding this matter.

We request that you respond within 21 days of receipt of this letter indicating,our intent to comply with our position.

The requirements of this letter affect fewer than I0 respondents, and therefore, are not subject to Office of hanagement. and,udget review under P.L.96-511.

Sincerely, Alexander 1. Drowerick, Senior Project. 1.1ranager Project. Diret(torate 1-4 Division of Peactor Pirojecfs -

1/il Office of Nu!ledr Peactor Reyulat.ion

Enclosure:

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ENCLOSURE STAFF POSITION ON EVALUATION OF STRUCTURAL INTEGRITY OF A DEGRADED STEEL CONTAINMENT OYSTER CREEK NUCLEAR GENERATING STATION ASME Section XI Subsections IWE-3519.3 and IWE-3122.4 state that a steel containment is acceptable:'if the thickness of the area of degradation discovered is reduced by not more than 10%.

This is acceptable only on the basis of considering the area of degradation as a form of discontinuity as stipulated in ASME Section III Division I Subsection NE-3213.10.

The area of degradation, where the stress intensity exceeds 1.1 Smc, is stipulated in NE-3213.10 in terms of the square root of the product of R and t as defined therein.

The code requires such a discontinuity be localized.

This is due to the fact that the load on a highly stressed and localized area will be transferred to the adjacent area.

If the area of degradation is localized, the effect on the overall behavior of the containment will be minimal or negligible.

The code does not specify the limit of the extent of the support region in which the stress intensity varies from 1.0 Smc to 1.1 Smc.

However, the limit can be determined from the analysis for load combinations with the inter'ral pressure as the major load.

On the basis of the above observation, the staff has established the following position:

1.

The corroded or degraded area with a reduction in thickness of not more than IQ% should be considered in accordance with NE-3213.10 as aldiscontinuity with the limits of its extent as prescribed therein.

2.

For a corroded containment shell where the thicknesses of the corroded areas'are obtained through UT measurements, the extent of each corroded area should be determined as accurately as practical.

3.

Except in the support zone of the discontinuity where the stress intensity value may vary from 1.0 Smc to 1.1 Smc, the primary membrane stress should be in accordance with the stress intensity limits as stipulated in Table NE-3221-1, Summary of Stress Intensity Limits.

-Doc./(4A 17t (t2

1. 4 0

UNITED STATES ItA 2",

NUCLEAR REGULATORY COMMISSION g

WASHINGTON. D. C. M055 November 19, 1991 Docket No. 50-219 Mr. John J. Barton, Vice President and Director GPU NuclEir Corporation Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731

SUBJECT:

CLARIFICATION OF STAFF POSITION ON EVALUATION OF STRUCTURAL INTEGRITY OF A DEGRADED STEEL CONTAINMENT (TAC NO/A79166)

References:

1. Letter to J. J. Barton from A. W. Dromerick providing the subject staff's position dated September 3, 1991.

2.

Letter to NRC from GPU Nuclear Corporation providing the response to staff's position dated October 9, 1991.

In a letter of October 9, 1991 (Reference 2),

GPU Nuclear Corporation (GPUN) provided responses to the staff position on the evaluation of the structural integrity of a degraded steel containment.

It appears from the responses that GPUN differs with the staff's position, specifically on the application of ASME subsection NE-3213.10.

Enclosed is the staff's review of GPUN's response.

It clarifies the staff's position and requ-ires GPUN to provide additional information to aid in a final resolution of staff's concerns.

We request that the information be provided within 30 days of receipt of this letter.

If you have any questions regarding this request, please contact me.

9112050128 911119 PDR ADOCK 05000219 P

PDR

Mr. John The requirements of this letter affect fewer than 10 respondents, and therefore, are not subject to Office of Management review under P.L.97-511.

Sincerely,

/s/

Aleyarder W. Dromerick, Sr. Project Manager Project Directorate 1-4 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation

Enclosure:

As stated cc:

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REVIEW OF GPUN'S RESPONSE OF OCTOBER 9, 1991 RELATED TO THE STAFF'S POSITION ON EVALUATION OF DEGRADED STEEL CONTAIMMENT AT OYSTER CREEK The staff has reviewed GPU Nuclear Corporation's (GPUN) response of October 9, 1991 to the staff's position on the evaluation of the structural integrity of a degraded steel containment.

It is to be noted that this staff position is to be applied generically in the evaluation of steel containments which are degraded, not specifically to the Oyster Creek steel drywell.

The staff's position is based on technical criteria that conform to the spirit and intent of ASME subsection NE-3213.10.

NE is the design part of the ASME code and cannot be directly applied to the situation of inservice degradation without the exercise of engineering judgment.

By considering the corroded area as equivalent to a discontinuity as indicated in NE-3212.10, great caution must be exercised.

It should be understood that the discontinuity as created by corrosion is not the same as the "designed" discontinuity such as a change in shell thicknesses, the presence of a bracket or a penetration as envisioned in the code.

The basic characteristic of the discontinuity due to corrosion is irregularity, e.g. variation in thickness and extent of corroded areas.

In view of the above observation, the NE 3312.10 stipulation cannot be applied indiscriminately to a corroded steel containment.

NE-3312.10 specifies the limit of the discontinuity region in which the stresses can be greater than 1.1 Smc.

The code does not specify the outside limit of the region which is contiguous to and supports the discontinuity and in which the stresses vary from 1.1 Smc to 1.0 Smc.

This should be expected because this outside limit varies with the configuration of the discontinuity and the loading.

Therefore, the lack of specific stipulation in the code in this respect should be understood and should not be construed to allow the stress limit of 1.1 Smc to be applied universally throughout the containment shell.

The staff position is not, in any way, more restrictive than the stipulation in the ASME Code.

The staff is well aware of the extensive examinations and analysis performed on the Oyster Creek drywell as reported by GPUN.

GPUN has repeatedly claimed that the Oyster Creek drywell has been examined thoroughly and the condition of the drywell is fully understood with a 95% confidence level.

On the basis of this claim, the staff has requested GPUN to determine the extent of each corroded area.

The staff is not requesting any additional physical examination.

However, on the basis of the information available, GPUN should present in a figure the known areas of corrosion with the critical stresses (general primary membrane stress or local primary membrane stress) identified.

The purpose of such an action is to determine the behavior of the drywell especially at and around the corroded areas.

By comparing the calculated stresses of the drywell shell at and around corroded areas with the code allowables the staff can reasonably determine the adequacy of the licensee's proposed actions.

I a

APR 0 9 IV MEMORANDUM FOR:

FROM:

SUBJECT:

Plant Name:

Applicant:

Docket No.:

Review Status:

Tac No.:

John F. Stolz, Director Project Directorate 1-4 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Goutam Bagchi, Chief Structural and Geosciences Branch Division of Engineering Technology Office of Nuclear Reactor Regulation EVALUATION REPORT ON STRUCTURAL INTEGRITY OF THE OYSTER CREEK DRYWELL Oyster Creek Nuclear Generating Station GPU Nuclear Corporation 50-2:19 Complete M79166 The Structural and Geosciences Branch (ESGB) has completed the review and evaluation of the stress analyses and stability analyses reports of the corroded drywell with and without the sand bed.

Our evaluation report together with a SALP is contained in the enclosure.

The licensee used the analyses to justify the removal of the sand from the sand bed region.

Even though the staff, with the assistance of consultants from Brookhaven National Laboratory (BNL),

concurred with licensee's conclusion that the drywell meets the ASME Section III Subsection NE requirements, it is essential that the licensee continue UT thickness measurements at refueling outages and at outages of opportunity for the life of the plant.

The review is performed by C. P. Tan of Geosciences Section of ESGB with the assistance of BNL.

Goutam Bagchi, Chief Structural and Geosciences Branch Division of Engineering Technology

Enclosure:

As stated cc:

J.

E. Richardson B.

D.

Liaw A.

Dromerick DISTRIBUTION:

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• SEE PREVIOUS CONCURRENCE

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SAFETY EVALUATION REPORT OYSTER CREEK NUCLEAR GENERATING STATION DRYWELL STRUCTURAL INTEGRITY STRUCTURAL AND GEOSCIENCES BRANCH I.

INTRODUCTION In 1986 the steel drywell at Oyster Creek Nuclear Generating Station (OCNGS) was found to be extensively corroded in the area of the shell which is in contact with the sand cushion around the bottom of the drywell.

Since then GPU Nuclear, the Licensee of OCNGS, has instituted a program of periodic inspection of the drywell shell sand cushion area through ultrasonic testing UT thickness measurements.

The inspection has been extended to other areas of the drywell and some areas above the sand cushion have been found to be corroded also.

From the UT thickness measurements, one can conclude that corrosion of the drywell shell in the sand cushion area is continuing.

In an attempt to eliminate corrosion or reduce the corrosion rate, the licensee tried cathodic protection and found it to be of no avail.

An examination of the results of consecutive UT measurements, confirmed that the corrosion is continuing.

There is concern that the structural integrity of the drywell cannot be assured.

Since the root cause of the corrosion in the sand cushion area is the presence of water in the sand, the licensee has considered sand removal to be an important element in its program to eliminate the corrosion threat to the drywell integrity.

In the program, the licensee first established the analysis criteria and then performed the analyses of the drywell for its structural adequacy with and without the presence of the sand.

The licensee performed stress analyses and stability analyses for both with and without the sand cases and concluded the drywell with or without the sand to be in compliance with the criteria established for the reevaluation.

It is to be noted that the original purpose of the sand cushion is to provide a smooth transition of stresses from the fixed portion to the free-standing portion of the steel drywell.

II.

&EVALUATION The staff with the assistance of consultants from Brookhaven National Laboratory (BNL) has reviewed and evaluated the information (Refs. 1,2,3,4,5) provided by the licensee.

2

1.

Re-Analysis Criteria The drywell was originally designed and constructed to the requirements of ASME Section VIII code and applicable code cases, with a contract date of July 1, 1964.

The section VIII code requirements for nuclear containment vessels at that time were less detailed than at any subsequent date.

The evolution of the ASME Section III code for metal containments and its relation with ASME Section VIII code were reviewed and evaluated by Teledyne Engineering Services (TES).

The evaluation criteria used are based on ASME Section III Subsection NE code through the 1977 summer addenda.

The reason for the use of the code of this vintage is that it was used in the Mark I containment program to evaluate the steel torus for hydrodynamic loads and that the current ASME Section III Subsection NE Code is closely related to that version.

The following are TES's findings relevant to Oyster Creek application:

a)

The steel material for the drywell is A-212, grade B, Firebox Quality (Section VIII),

but it is redesignated as SA-516 grade in Section III.

b)

The relation between the allowable stress (S) in Section:VIII and the stress intensity (Smc) in Section III for metal containment is 1.1S = Smc.

c)

Categorization of stresses into general primary membrane, general bending and local primary membrane stresses and membrane plus bending stresses is adopted as in Subsection NE.

d)

The effect of a locally stressed region on the containment shell is considered in accordance with NE-3213.10.

In addition to ASME Section III Subsection NE Code, the licensee has also invoked ASME Section XI IWE Code to demonstrate the adequacy of the Oyster Creek drywell.

IWE-3519.3 and IWE-3122.4 state that it is acceptable if either the thickness of the base metal is reduced by no more than 10% of the normal plate thickness or the reduced thickness can be shown by analysis to satisfy the requirements of the design specification.

The staff has reviewed the licensee's adoption of ASME Section III Subsection NE and Section XI Subsection IWE in its evaluation of the structural adequacy of the corroded Oyster Creek drywell, and has found it to be generally reasonable and acceptable.

3 By adopting the Subsection NE criteria, the licensee has treated the corroded areas as discontinuities per NE-3213.10, which was originally meant for change in thicknesses, supports, and penetrations.

These discontinuities are highly localized and should be designed so that their presence will have no effect on the overall, behavior of the containment shell.

NE-3213.10 defines clearly the level of stress intensity and the extent of the discontinuity to be considered localized.

A stress intensity limit of 1.1 Smc is specified at the boundary of the region within which the membrane stress can be higher than 1.1 Smc.!

The region where the stress intensity varies from 1.1 Smc to 1.0 Smc is not defined in the code because of the fact that it varies with the loading.

In view of this, the licensee rationalized that the 1.1 Smc can be applied beyond the region defined by NE-3213.10 for localized discontinuity without any restriction throughout the drywell.

The staff disagreed with the licensee's interpretation of the code.

The staff pointed out that for Oyster Creek drywell, stresses due to internal pressure should be used as the criterion to establish such a region.

The interpretation of Section XI Subsections IWE-3519.3 and IWE-3122.4 can be made only in the same context.

It is staff's position that the primary membrane stress limit of 1.1 Smc not be used indiscriminately throughout the drywell.

In order to use NE-3213.10 to consider the corroded area as a localized discontinuity, the extent of the reduction in thickness due to corrosion should be reasonably known.

UT thickness measurements are highly localized; however, from the numerous measurements so far made on the Oyster Creek drywell, one can have a general idea of the overall corroded condition of the drywell shell and it is possible to judiciously apply the established re-analysis criteria.

2.

Re-analyses The re-analyses were made by General Electric Company for the licensee, one reanalysis considered the sand present and the other considered the drywell without the sand.

Each re-analysis comprises a stress analysis and stability analysis.

Two finite element models, one axisymmetric and another a 360 pie slice model were used for the stress analysis.

The ANSYS computer program was used to perform the analyses.

The axisymmetric model was used to determine the stresses for the seismic and the thermal gradient loads.

The pie slice model was used for dead weight and pressure loads.

The pie slice model includes the vent pipe and the reinforcing ring, and was also used for buckling analysis.

The same models were used for the cases with

4 and without sand, except that in the former, the stiffness of sand in contact with the steel shell was considered.

The shell thickness in the sand region was assumed to be 0.700" 1for the with-sand case and to be 0.736" for the without-sand case.

The 0.70" was, as claimed by the licensee, used for conservatism and the 0.736" is the projected thickness at the start of fuel cycle 14R.

The same thicknesses of the shell above the sand region were used for both cases.

For the with-sand case, an analysis of the drywell with the original nominal wall thicknesses was made to check the shell stresses with the allowable values established for the re-analyses.

The licensee used the same load combinations as specified in Oyster Creek's final design safety analysis report (FDSAR) for the re-analyses.

The licensee made a comparison of the load combinations and corresponding allowable stress limits using the SRP section 3.8.2 and concluded they are comparable.

The results ofi the re-analyses indicated that the governing thicknesses are in the upper sphere and the cylinder where the calculated primary membrane stresses are respectively 20,360 psi and 19,850 psi vs. the allowable stress value of 19,300 psi.

There is basically no difference, in the calculated stresses at these levels, between the with and without sand cases.

This should bel expected, because in a steel shell structure the local effect or the edge effect is damped in a very short distance.

The stresses calculated exceed the allowable by 3% to 6%,

and such exceedance is actually limited to the corroded area as obtained from UT measurements.

However, in order to perform the axisymmetric analysis and analysis of the pie slice model, uniform* thicknesses were assumed for each section of the drywell.

Therefore, the calculated over-stresses may represent only stresses at the corroded areas and the stresses for areas beyond the corroded areas are less and would most likely be within the allowable as indicated in results of the analyses for nominal thicknesses.

The diagram in Ref.

6 indicated such a condition.

It is to be noted that the stresses for the corroded areas were obtained by multiplying the stresses for nominal thicknesses by the ratios between the corroded and nominal thicknesses.

The buckling analyses of the drywell were performed in accordance with ASME Code Case N-284.

The analyses were done on the 360 pie slice model for both with-sand and without-sand cases.

Except in the sand cushion area where a shell thickness of 0.7" for the with-sand

5 case and a shell thickness of 0.736" for the without-sand case were used, nominal shell thicknesses were considered for other sections.

The load combinations which are critical to buckling were identified as those involving refueling and post accident conditions.

By applying a factor of safety of 2 and 1.67 for the load combinations involving refueling and the post-accident conditions respectively, the licensee established for both cases the allowable buckling stresses which are obtained after being modified by capacity and plasticity reduction factors.

It is found that the without-sand, case for the post-accident condition is most limiting in terms of buckling with a margin of 14%.

The staff and its BNL consultants concur with the licensee's conclusion that the Oyster Creek drywell has adequate margin against buckling with no sand support for an assumed sandbed region shell thickness of 0.736 inch.

A copy of BNL's technical evaluation report is attached to this SER.

III. CONCLUSION With the assistance of consultants from BNL, the staff has reviewed and evaluated the responses to the staff's concerns and the detailed re-analyses of the drywell for the with-sand and without-sand cases.

The reanalyses by the licensee indicated that the corroded drywell meets the requirements for containment vessels as contained in ASME Section III Subsection NE through summer 1977 addenda.

This code was adopted in the Mark I containment program.

The staff agrees with the licensee's justification of using the above mentioned code requirements with one exception, the use of 1.1 Smc throughout the drywell shell in the criteria for stress analyses.

It is the staff's position that the primary membrane stress limit of 1.1 Smc not be used indiscriminately throughout the drywell.

The staff accepted the licensee's reanalyses on the assumption that the corrodled areas are highly localized as indicated by the licensee's UT measurements.

The stresses obtained for the case of reduced thickness can only be interpreted to represent those in the corroded areas and their adjacent regions of the drywell shell.

In view of these observations, it is essential that the licensee perform UT thickness measurements at refueling outages and at outages of opportunity for the life of the plant.

The measurements should cover not only areas previously inspected but also areas which have never been inspected so as to confirm that the thicknesses of the corroded areas are as projected and

the corroded areas are localized.

Both of these assumptions are the bases of the reanalyses and the staff acceptance 'f the reanalysis results.

References:

1.

"An ASME Section VIII Evaluation of the Oyster Creek Drywell Part 1, Stress Analysis",..GE Report No.

9-1 DRF #00664 November 1990, prepared for GPUN (with sand).

2.

"Justification for use of.Section III, Subsection NE, Guidance in Evaluating the Oyster.Creek Drywell" TR-7377-1, Teledyne Engineering Services, November 1990 (Appendix A to Reference 1).

3.

"An ASME Section VIII:evaluation of the Oyster Creek

Drywell, Part 2,,.Stability Analysis",

GE Report No.

9-2 DRF

1. 00664, Rev. 0, & Rev. I.

November 1990, prepared for GPUN (with sand).

4.

"An ASME Section VIII Evaluation of Oyster Creek Drywell for without sand case, Part I, stress analysis," GE Report No.

9-3 DRF. #00664,* Rev.

O, February 1991.

Prepared for GPUN.

5.

"An ASME Section VIII Evaluation of Osyter Creek Drywell, for without sand case, ýPart.2 Stability Analysis",

GE Report No.

9-4, DRF #00664 Rev.O, Rev. 1 November 1990, prepared for GPUN.

6.

Diagram attached to a letter from J.

C.

Devine Jr. of GPUN to NRC dated January 17, 1992 (C321-92-2020, 5000-92-2094).

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~ 4~2SALP INPT FACILITY NAME:: OysterC k Nucla Generating Station IEPU NN learCorporation K

SUMMARY

-OF REVIEW~

'", Since the discovery:..of!lcorrosion in the sand cushion area of the

,.drywell, the licenseeihas Iperformed UT..thickness measurements at outage of opportunity-and 'at.refueling outages from the results of the UT measurements it ý,can be concluded that corrosion is still continuing.in,ýview of this,,the licensee has considered sand removal to be an important-element in its program to eliminate the corrosion threat.to the drywell integrity.

Since removal of the sand may affect !.the behavior of the drywell, the.

-licensee had General:Electric performed.stress and stability analyses of the drywell for both with and without sand conditions taking into consideration the reduction in thickness in the sand cuhinregion.,ý The. criteria,for the re-analyses are based on ASME Section VI.:Code* SubsectionNE...The use of subsection NE was Sexamined and justified by~the licensee's hconsultant from Teledyne

_-ý,:,Engineering W"4 Services*

The staffwith the assistance of consultants from Brookhaven National Laboratory reviewed the reanalyses and the criteriaused and found them to tbe acceptable.

NARRATIVE DISCUSSION OF LICENSEE PERFORMANCE

-FUNCTIONAL AREAS ENGINEERING/TECHNICAL SUPPORT Since the discovery of, the corrosion of the drywell, the licensee has been working diligently to monitor the state of the

,,,corrosion,, to stop the, source of leakage and to eliminate further aggravation.

Even though in the review process differing opinion and disagreement with staff's position arose, the licensee has been co-operative and forthcoming in striving to resolve staff's

,_,concerns.

TECHNICAL EVALUATION REPORT ON STRUCTURAL ANALYSES, OF THE CORRODED OYSTER CREEK STEEL DRYWELL

1.

Introduction An inspection of!the steel drywell at the Oyster Creek Nuclear Generating Station in November 1986 revealed that some degradation due to corrosion had occurred in the sandbed region of the shell.

Subsequent inspections also identified thickness degradations in the upper spherical and cylindrical sections of the drywell.

The

licensee, GPU Nuclear Corporation, has performed structural analyses to demonstrate the integrity of the drywell for projected corroded conditions that may exist at the start of the fourteenth refueling outage (14R)."

This outage is expected to start in October 1992.

In an attempt to arrest the corrosion, the licensee plans to remove the sand from the sandbed region.

Consequently, they have submitted structural analyses of the drywell both with and without sand for drywell wall thicknesses projected to exist at the start of 14R outage.

2.

Summary of Licensee's Analyses The analyses performed by the licensee utilized the drywell.

wall thicknesses summarized in Table 1.

Table 1 Drywell Wall Thicknesse.s Projected 95%

As-Designed Confidence Thicknesses 14R Thicknesses Drywell Region (in.)

(in.)

Cylindrical Region 0.640 0.619 Knuckle 2.5625*

2.5625*

Upper Spherical Region 0.722 0.677 Middle Spherical Region 0.770 0.723 Lower Spherical Region 1.154 1.154 Except Sand Bed Area Sand Bed Region V

1.154 0.736

• NOTE:

Table 2-1 of both References 1 and 3 indicates that the knuckle th~ickness is 2.625".

This appears to be a

mistake since the knuckle thickness is shown to be 2-9/16" in Figure 1-1 of the same report.

1

p, The stress analysis for the "with sand" case is described in Reference

1.

For this analysis the licensee utilized the as-designed thicknesses, except for the sandbed region where a

thickness of 0.70" was used.

The stress results were obtained from a

finite element analysis which utilized axisymmetric solid elements and the ANSYS computer program.

Later, the stress results were scaled to addreOs the local thinning in areas other than the sandbed region (the projected 95% confidence 14R thicknesses in Table 1).

The loads and load combinations considered in the analysis are based on the FSAR Primary Containment Design Report and the 1964 Technical Specification for the Containment.

Appendix E of Reference 1 compares the load combinations considered in the analysis with those given in Section 3.8.2 of the NRC Standard Review Plan, Rev.

1, July 1981.

The stress analysis for the "without sand" case is described in Reference 3.

For this analysis the licensee also utilized the as-designed thicknesses, except for the sandbed region where a

thickness of 0.736" was used.

In this case, two finite element models, an axisymmettic and a 360 pie slice model, were used.

The axisymmetric model tis essentially the same as that used in Reference 1; howevert the elements representing the sand stiffness were removed.

Thisimodel was used to determine the seismic and thermal stresses.

The pie slice model was used to determine the dead weight and pressure stresses, as well as the stresses for load combinations.

Thepie slice model included the effects of the vent pipes and the reinfoicing ring in the drywell shell in the vicinity of each vent pipe.

The drywell and vent shell were modeled using 3-dimensional elastib-plastic quadrilateral shell elements.

At a distance of 76 inches from the drywell shell, beam elements were used to model the remainder of the ventline.

The loads and load combinations are the same as those considered in Reference 1.

The code of reqord for the Oyster Creek drywell is the 1962 Edition of the ASME qode,Section VIII with Addenda to Winter 1963, and Code Cases 1270N-5, 1271N and 1272N-5.

The licensee utilized these criteria in evaluating the stresses in the drywell, but also utilized guidance from the NRC Standard Review Plan with regard to allowable stresses !for service level C and the post-accident condition.

The liceinsee also used guidance from Subsection NE of Section III of the ASME Code in order tc justify the use of a limit of

.l1s.,

in evaluating the general membrane stresses in areas of the drywell where reduced thicknesses are specified.

Based on these criteria the licensee has concluded that the stresses in the drywell shell are wi'thin code allowable limits for both the "with sand" and "without stand" cases.

The licensee also performed stability analyses of the drywell for both the "with sand" case (Reference 2) and the "without sand" case (Reference 4).

For the "with sand" case the licensee utilized the as-designed thicknesses shown in Table 1, except in the sandbed region where a thickness of 0.700 inch was used.

For the "without 2

sand" case the same thicknesses were used, except in the sandbed region where a thickiness of 0.736 inch was used.

The buckling capability of the drywell for both the "with sand" and "without sand" cases was evaluated by using the 360 pie slice finite element model discussed above.

For the "with sand" case spring elements were used in the sandbed region to model the sand support.

For the "without sand" case these spring elements were removed.

The most limiting load combinations which result in the highest compressive stresses in the sandbed region were considered for the buckling analysis.

These are the refueling condition (Dead Weight + Live Load + Refueling Water Weight + External Pressure + Seismic) and the post-accident condition (Dead Weight + Live Load + Hydrostatic Pressure for Flooded:Drywell + External Pressure + Seismic).

The buckling evaluations performed by the licensee follow the methodology described in ASME Code Case N-284, "Metal Containmernt Shell Buckling Design Methods,Section III, Class MC",

Approved August 25, 1980.

The theoretical elastic buckling stress is calculated by analyzing the three dimensional finite element model discussed above.

Then the theoretical buckling stress is modified by capacity and plasticity reduction factors.

The allowable compressive stress is obtained by dividing the calculated buckling stress by a factor of safety.

In accordance with Code Case 14-284 the licensee used a factor of safety of 2.0 for the refu' i g condition and 1.67 for the post-accident condition.

The capacity reduction factors were also modified to take into accouQ the effects of hoop stress.

Originally the licensee based the hoop stress modification 'on data related to the axial compressive strength of cylinders (References 2 and 4).

Later the licensee revised the approach based on a review of spherical shell buckling data and recalculated the drywell buckling capacities for both the "with sand" and "without sand" cases (Reference 8).

For the "with sand" case, the licensee reports a margin above the allowable compressive stress of 47% for the refueling condition and 40% for the post-accident condition.

For the "without sand" case, the licensee reports margins of 24.5% for the refueling condition and 14% for the post-accident condition.

3.

Evaluation of Licensee's Approach The analyses performed by the licensee as summarized in Section 2 and discussed more fully in References 1 through 4 have been reviewed and found to provide an acceptable approach for demonstrating the structural integrity of the corroded Oyster Creek drywell.

The finite element analyses performed for both the stress and stability evaluations are consistent with industry practice.

Except for the use of a limit of l.1S% in evaluating the general membrane stress in areas of reduced drywell thickness, the loads, load combinations and acceptance criteria used by the licensee are.

consistent with the :guidance given in Section 3.8.2 of the NRC Standard Review Plan,: Rev.

1, July 1981.

To further support their position, the licensee has provided two appendices to Reference 3

Appendix A provides a detailed justification for the use of Section

III, Subsection NE as guidance in evaluating the Oyster Creek drywell.

Appendix E compares the load combinations given in the Final Design Safety Analysis Report (FDSAR) with the load combinations given in SRP 3.8.2 and demonstrates that the load combinations used in the analysis envelop those given in the SRP.

In the areas of the drywell where reduced thicknesses are specified, the licensee has used a limit of l.iSc, to evaluate the general membrane stresses.

In support of this position the licensee has cited the provisions of NE-3213.1 of the ASME Code concerning local primary membrane stresses.

In

effect, the licensee's criteria!' would treat corroded or degraded areas as discontinuities.

For such considerations the code places no limit on the extent of the region in which the membrane stress exceeds 1.OSec but is less than l.11Sc.

In support of this position the licensee has provided the opinion of Dr. W.E. Cooper, a well known expert on the development of the ASME Code.

Dr. Cooper concluded that "given a design which satisfies the general Code intent, as the Oyster Creek drywell does as originally constructed, it is not a violation of Subsection NE requirements for the membrane stress to be between 1.0S., and l.lS.c over significant distances".

The licensee has also cited the provisions of IWE-3519.3 which accepts up to a 10% reduction in the thickness of the original base metal.

The licensee's position has merit, but great caution must be exercised to assure that such a

position is. not applied indiscriminately.

In the case of the Oyster Creek drywell the licensee has concluded that "there are very few locations where the calculated stress intensities for design basis conditions, would exceed 1.0SC,, and in these cases only slightly" (Reference 7).

The licensee has provided additional information in Reference 9

to support this conclusion.

Based on the information provided by the licensee which demonstrates that the use of the l.1S., criteria is limited to localized areas, it is concluded that the Oyster Creek drywell meets the intent of the ASME Code.

As discussed in Section 2, the capacity reduction factors used in the buckling analysis are modified to take into account the beneficial effects of tensile hoop stress.

As a result of a question raised during the review regarding this matter, the licensee submitted additional information in Reference 5 to support the approach.

This information included a report prepared by C.D.

Miller entitled "Effects of Internal Pressure on Axial Compression Strength of Cylinders" (CBI Technical Report No.

022891, February 1991).

The report presented a design equation which was the lower bound of the test data included in the report. It also demonstrated that the equation used in References 2

and 4 was conservative relative to the proposed design equation.

The report presented further arguments that the rules determined for axially compressed cylinders subjected to internal pressure can be applied to spheres.

Subsequently the iicensee has submitted Reference 8,

which 4

indicates that the original approach was not conservative with regard to its application to spherical shapes and recommends a new equation.

However, the documentation supporting the use of this equation is not inicluded in Reference 8,

but apparently is contained in a referenced report prepared by C.D. Miller entitled "Evaluation of Stability Analysis Methods Used for the Oyster Creek Drywell" (CBI Technical.

Report Prepared for GPU Nuclear Corporation, September 1991).

This report was subsequently submitted and reviewed by the NRC staff.

As discussed in Section 2,

the use of the revised equation still results in calculated capacities in compliance with the ASME Code provisions; however, the margins beyond those capacities are reduced from those reported by References 2 and 4.

It is noted that the licensee may have "double-counted" the effects of hoop tension, since the theoretical elastic instability stress was calculated from the finite element model using the ANSYS Code.

The elastic instability stress calculated by the ANSYS Code may have already taken into account the effects of hoop tensile stress.

However, by comparing the theoretical elastic instability stress and the corresponding circumferential stress predicted by the licensee for the. refueling and post-accident cases, it appears that the effect of hoop tension in the ANSYS calculations is small and there is suffici'nt margin in the results to compensate for the potential "double-coUnting".

Furthermore, it is judged that there is sufficient capacity in the drywell to preclude a significant buckling failure under the postulated loading conditions since the licensee's calculations: (a) incorporate factors of safety of 1.67 to 2.0, depending upon the load condition, and (b) utilize a

conservative assumption by considering the shell wall thickness to be severely reducedi-for the full circumference of the drywell throughout the sandbed region.

During the course of the review of the licensee's submittals, a number of other issues were raised regarding the approach.

These included:

(a) the basis and method of calculating the projected drywell thicknesses, (b) the scaling of the calculated stresses for the nominal thickness case by the thickness ratio, (c) the effect of stress concentrations due to the change of thickness, (d) monitoring of the drywell temperature, (e) sensitivity of stresses due to variations in the sand spring stiffness, (f) sensitivity of the plasticity reduction factor in the buckling analysis, (g) use of the 2

psi design basis external pressure in the buckling

analysis, (h) effect of the large displacement method, (i) the treatment of the large concentrated loads considered in the analysis, and (j) the method of applying the seismic loads to the pie slice model.

These issues were adequately addressed by the additional information provided by the licensee in References 5 and

6.

i 5

4.

Conclusions The licensee has demonstrated that the calculated stresses in the Oyster Creek drywell (both with and without the sandbed),

as a result of the postulated loading conditions, meet the intent of the ASME Code for projected corroded conditions that may exist at the start of the fourteenth refueling outage.

However, if the actual thickness in the sandbed region at 14R is close to the projected thickness of 0.736",

there may not be adequate margin left for further corrosion through continued operation unless it is demonstrated that removal of sand will completely stop further thickness reductions.

The licensee has also demonstrated that there is sufficient'margin in the drywell design (both with and without the sandbed') to preclude a buckling failure under the postulated loading conditions.

It should be recognized that the conclusions reached by th1 licensee have been accepted for this particular application with due regard to all the assumptions made in the analysis and thd available margins.

The use of the l.iS., criteria for evaluating general membrane stress in corroded or degraded areas should be investigated further, by the NRC staff and the ASME Code Committee and appropriate bounds established before it is accepted for general use.

The -licensee's buckling criteria regarding the modification of capacity reduction factors for tensile hoop stresis and the determination of plasticity reduction factors should also be investigated in a, similar manner.

5.

References

1.

GE Report Index No.

9-1, "An ASME Section VIII Evaluation olf the Oyster Creek Drywell - Part I - Stress Analysis", November 1990.

2.

GE Report Index. No.

9-2, "An ASME Section VIII Evaluation o:ft the Oyster Creek Drywell Part 2 Stability Analysis,!"

November 1990.-

3.

GE Report Index No.

9-3, "An ASME Section VIII Evaluation ot the Oyster Creek Drywell for Without Sand Case -

Part 1 Stress Analysis," February 1991.

4.

GE Report Index No.

9-4, "An ASME Section VIII Evaluation of the Oyster Creek Drywell for Without Sand Case -

Part 2 _

Stability Analysis," February 1991.

5.

GPU Nuclear letter dated March 20, 1991, "Oyster Creek Drywell1 Containment."

6.

CPU Nuclear letter dated June 20,

1991, "Oyster Creek Drywelýýl Containment".

4)]

7.

GPU Nuclear letter dated October 9,

1991, Drywell Containment"

8.

GPU Nuclear letter dated January 16,

1992, Drywell Containment".

9.

GPU Nuclear letter dated January 17,

1992, Drywell Containment".

"Oyster Creek "Oyster Creek "Oyster Creek, 7

April 24, 1992 Docket No. 50-219 Mr. John J. Barton Vice President and Director GPU Nuclear Corporation Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731

Dear Mr. Barton:

Distribution:

Docket File NRC & Local PDRs PD 1-4 Plant SVarga JCalvo SNorris ADromerick OGC CPTan ACRS (10)

CWHehl, RI

SUBJECT:

EVALUATION REPORT ON STRUCTURAL INTEGRITY OF THE OYSTER CREEK DRYWELL (TAC NO.

M79166)

The staff has completed the review and evaluation of the stress analyses and stability analyses reports of the corroded drywell with and without the sand bed.

Our evaluation report is contained in the enclosure.

GPUN used the analyses to justify the removal of the sand from the sand bed region.

Even though the staff, with the assistance of consultants from Brookhaven National Laboratory (BNL),

concurred with GPUN's conclusion that the drywell meets the ASME Section III Subsection NE requirements, it is essential that GPUN continue UT thickness measurements at refueling outages and at outages of opportunity for the life of the plant.

The measurements should cover not only areas previously inspected but also accessible areas which have never been inspected so as to confirm that the thickness of the corroded areas are as projected and the corroded areas are localized.

We request that you respond within 30 days of receipt of this letter indicating your intent to comply with the above requirements as discussed in the Safety Evaluation.

The requirements of this letter affect fewer than 10 respondents, and therefore, are not subject to Office of Management and Budget review under P.L.96-511.

Sincerely,

/s/

9204300078 920424 PDR ADOCK 05000219 E

PDR Alexander W. Dromerick, Sr. Project Manager Project Directorate 1-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosure:

As stated WIC LE CENTER COPY cc w/enclosure:

See next page OFFICIAL RECORD COPY Document Name: M79166 OFC

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NAME

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UNITED STATES NUCLEAR REGULATORY COMMISSION

' *WASHINGTON.

0. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ORYWELL STRUCTURAL INTEGRITY-OYSTER CREEK NUCLEAR GENERATING STATION GPU NUCLEAR CORPORATION DOCKET NO.

50-219 I.

INTRODUCTION In 1986 the steel drywell at Oyster Creek Nuclear Generating Station (OCNGS) was found to be extensively corroded in the area of the shell which is in contact with the sand cushion around the bottom of the drywell.

Since then GPU Nuclear Corporation, (GPUN, the licensee of OCNGS),.has instituted a program of periodic inspection of the drywell shell sand cushion area through ultrasonic testing (UT) thickness measurements..

The inspection has been extended to other areas of the drywell and some areas above the sand cushion have been found to be corroded also.

From the UT thickness measurements, one can conclude that corrosion of the drywell shell in the sand cushion area is continuing.

In an attempt to eliminate corrosion or reduce the corrosion rate, the licensee tried cathodic protection and found it to be of no avail.

An examination of the results of consecutive UT measurements, confirmed that the corrosion is continuing.

There is concern that the structural integrity of the drywell cannot be assured.

Since the root cause of the corrosion in the sand cushion area is the presence of water in the sand, the licensee has considered sand removal to be an important element in its program to eliminate the corrosion threat to the drywell integrity.

In the program, the licensee first established the analysis criteria and then performed the analyses of the drywell for its structural adequacy with and without the presence of the sand.

The licensee performed stress analyses and stability analyses for both with and without the sand cases and concluded the drywell with or without the sand to be in compliance with the criteria established for the reevaluation.

It is to be noted that the original purpose of the sand cushion is to provide a smooth transition of stresses from the fixed portion to the free-standing portion of the steel drywell.

11.

EVALUATION The staff with the assistance of consultants from Brookhaven National Laboratory (BNL) has reviewed and evaluated the information (Refs. 1,2,3,4,5) provided by the licensee.

9204300067 920424 PDR ADOCK.05000219 E

PDR I

1.

Re-Analysis Criteria The drywell was originally designed and constructed to the requirements of ASME Section VIII code and applicable code cases, with a contract date of July 1, 1964.

The Section VIII Code requirements for nuclear containment vessels at that time were less detailed than at any subsequent date.

The evolution of the ASME Section III Code for metal containments and its relation with ASME Section VIII Code were reviewed and evaluated by Teledyne Engineering Services (TES).

The evaluation criteria used are based on ASME Section III Subsection NE Code through the 1977 summer addenda.

The reason for the use of the Code of this vintage is that it was used in the Mark I containment program to evaluate the steel torus for hydrodynamic loads and that the current ASME Section III Subsection NE Code is closely related to that version.

The following are TES's findings relevant to Oyster Creek application:

a)

The steel material for the drywell is A-212, grade B, Firebox Quality (Section VIII), but it is redesignated as SA-516 grade in Section I11.

b)

The relation between the allowable stress (S) in Section VIII and the stress intensity (Smc) in Section III for metal containment is I.IS - Smc.

c)

Categorization of stresses into general. primary membrane, general bending and local primary membrane stresses and membrane plus bending stresses is adopted as in Subsection NF.

d)

The effect of a locally stressed region on the containment shell is considered in accordance with NE-3213.10.

In addition to ASME Section III Subsection NE Code, the licensee has also invoked ASME Section XI IWE Code to demonstrate the adequacy of the Oyster Creek drywell.

IWE-3519.3 and IWE-3122.4 state that it is acceptable if either the thickness of the base metal is reduced by no more than 10% of the normal plate thickness or the reduced thickness can be shown by analysis to satisfy therequirements of the design specification.

The staff has reviewed the licensee's adoption of ASME Section III Subsection NE and Section XI Subsection IWE in its evaluation of the structural adequacy of the corroded Oyster Creek drywell, and has found it to be generally reasonable and acceptable.

By adopting the Subsection NE criteria, the licensee has treated the corroded areas as discontinuities per NE-3213.10, which was originally meant for change in thicknesses, supports, and penetrations.

These discontinuities are highly localized and should be designed so that their presence will have no effect on the overall behavior of the containment shell.

NE-3213.10 defines clearly the level of stress intensity and the extent of.the discontinuity to be considered localized.

A stress intensity limit of 1.1 Smc is specified at the boundary of the region within which the membrane stress can be higher than 1.1 Smc.

The region where the stress intensity varies from 1.1 Smc to 1.0 Smc is not defined in the Code because of the fact that it varies with the loading.

In view of this, the licensee rationalized that the 1.1 Smc can be applied beyond the region defined by NE-3213.10 for localized discontinuity without any restriction throughout the drywell.

The staff disagreed with the licensee's interpretation of the Code.

The staff pointed out that for Oyster Creek.

drywell, stresses due to internal pressure should be used as the criterion to establish such a region.

The interpretation of Section XI Subsections IWE-3519.3 and IWE-3122.4 can be made only in the same context.

It is staff's position that the primary membrane stress limit of 1.1 Smc not be used indiscriminately throughout the drywell.

In order to use NE-3213.10 to consider the corroded area as a localized discontinuity, the extent of the reduction in thickness due to corrosion should be reasonably known.

UT thickness,measurements are highly localized;

however, from the numerous measurements so far made on the Oyster Creek

drywell, one can have a general idea of the overall corroded condition of the drywell shell and it is possible to judiciously apply the established re-analysis criteria.

2.

Re-analyses The re-analyses were made by General Electric Company for the licensee, one reanalysis considered the sand present and the other considered the drywell without the sand.

Each re-analysis comprises a stress analysis and stability analysis.

Two finite element models, one axisymmetric and another a 36° pie slice model were used for the stress analysis.

The ANSYS computer program was used to perform the analyses.

The axisymmetric model was used to determine the stresses for the seismic and the thermal gradient loads.

The pie slice model was used for dead weight and pressure loads.

The pie slice model includes the vent pipe and the reinforcing ring, and was also used for buckling analysis.

The same models were used for the cases with and without sand, except that in the former, the stiffness of sand in contact with the steel shell was considered.

The shell thickness in the sand region was assumed to be 0.700" for the with-sand case and to be 0.736" for the without-sand case.

The 0.70" was, as claimed by the licensee, used for conservatism and the 0.736" is the projected thickness at the start of fuel cycle 14R.

The same thicknesses of the shell above the sand region were used for both cases.

For the with-sand case, an analysis of the drywell with the original nominal wall thicknesses was made to check the shell stresses with the allowable values established for there-analyses.

The licensee used the same load combinations as specified in Oyster Creek's final design safety analysis report (FDSAR) for the re-analyses.

The licensee made a comparison of the load combinations and corresponding allowable stress limits using the Standard Review Plan (SRP) section 3.8.2 and concluded they are comparable.

The results of the re-analyses indicated that the governing thicknesses are in the upper sphere and the cylinder where the calculated primary membrane stresses are respectively 20,360 psi and 19,850 psi vs. the allowable stress value of 19,300 psi.

There is basically no difference, in the calculated stresses at these levels, between the with and without sand cases.

This should be expected, because in a steel shell structure the local effect or the edge effect is damped in a very short distance.

The. stresses calculated exceed the allowable by 3% to 6%,

and such exceedance is actually limited to the corroded area as obtained from UT measurements.

However, in order to perform the axisymmetric analysis and analysis of the pie slice model, uniform thicknesses were assumed for each section of the drywell.

Therefore, the calculated over-stresses may represent only stresses at the corroded areas and the stresses for areas beyond the corroded areas are less and would most likely be within the allowable as indicated in results of the analyses for nominal thicknesses.

The diagram in Ref. 6 indicated such a condition.

It is to be noted that the stresses for the corroded areas were obtained by multiplying the stresses for nominal thicknesses by the ratios between the corroded and nominal thicknesses.

The buckling analyses of the drywell were performed in accordance with ASME Code Case N-284.

The analyses were done on the 36° pie slice model for both with-sand and without-sand cases.

Except in the sand cushion area where a shell thickness of 0.7" for the with-sand case and a shell thickness of 0.736" for the without-sand case were used, nominal shell thicknesses were considered for other sections.

The load combinations which are critical to buckling were identified as those involving refueling and post accident conditions.

By applying a factor of safety of 2 and 1.67 for the load combinations involving refueling and thepost-accident conditions respectively, the licensee established for both cases the allowable buckling stresses which are obtained after being modified by capacity and plasticity reduction factors.

It is found that the without-sand, case for the post-accident condition is most limiting in terms of buckling with a margin of 14%.

The staff and its Brookhaven National Laboratory (BNL) consultants concur with the licensee's conclusion that the Oyster Creek drywell has adequate margin against buckling with no sand support for an assumed sandbed region shell thickness of 0.736 inch.

A copy of BNL's technical evaluation report is attached to this safety evaluation.

Ill.

CONCLUSION With the assistance of consultants from BNL, the staff has reviewed and evaluated the responses to the staff's concerns and the detailed re-.1nalyses of the drywell for the with-sand and without-sand cases.

The reanaifses by the licensee indicated that the corroded drywell meets the requirements for containment vessels as contained in ASME Section III Subsection NE through summer 1977 addenda.

This Code was adopted in the Mark I containment program.

The staff agrees with the licensee's justification of using the above mentioned Code requirements with one exception, the use of 1.1 Smc throughout the drywell shell in the criteria for stress analyses.

It is the staff's position that the primary membrane stress limit of 1.1 Smc not be used indiscriminately throughout the drywell.

The staff accepted the licensee's reanalyses on the assumption that the corroded areas are highly localized as indicated by the licensee's UT measurements.

The stresses obtained for the case of reduced thickness can only be interpreted to represent those in the corroded areas and their adjacent regions of the drywell shell.

In view of these observations, it is essential that the licensee perform UT thickness measurements at refueling outages and at outages of opportunity for the life of the plant.

The measurements should cover not only areas previously inspected but also accessible areas which have never been inspected so as to confirm that the thicknesses of the corroded areas are as projected and the corroded areas are localized.

Both of these assumptions are the bases of the reanalyses and the staff acceptance of the reanalysis results.

References:

1. "An ASME Section VIII Evaluation of the Oyster Creek Drywell Part 1, Stress Analysis" GE Report No.

9-1 DRF #00664 November 1990, prepared for GPUN (with sand).

2.

"Justification for use of Section Ill, Subsection NE, Guidance in Evaluating the Oyster Creek Drywell" TR-7377-1, Teledyne Engineering Services, November 1990 (Appendix A to Reference 1).

3.

"An ASME Section VIII evaluation of the Oyster Creek Drywell, Part 2, Stability Analysis" GE Report No.

9-2 DRF #00664, Rev. 0, & Rev.

1.

November 1990, prepared for GPUN (with sand).

4.

"An ASME Section VIII Evaluation of Oyster Creek Drywell for without sand case, Part I, stress analysis" GE Report No.

9-3 DRF #00664, Rev. 0, February 1991. Prepared for GPUN.

5.

"An ASME Section VIII Evaluation of Oyster Creek Drywell, for without sand case, Part 2 Stability Analysis" GE Report No. 9-4, DRF #00664 Rev.

0, Rev. 1 November 1990, prepared for GPUN.

6.

Diagram attached to a letter from J. C. Devine Jr. of GPUN to NRC dated January 17, 1992 (C321-92-2020, 5000-92-2094).

Principal Contributor:

C'.P. Tan Date:

April 24, 1992

Attachment:

BNL Technical Evaluation Report

GPU Nuclear Corporation One "pper Pond Road Parsippany. New Jersey 07054 201-316-7000 TELEX 136-482 Writer's Direct Dial Number:

May 26, 1992 5000-92-3026 C321-92-2163 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Gentlemen:

Subject:

Oyster Creek.Nuclear Generating Station (OCNGS)

Docket No. 50-219 Facility Operating License No.

DPR-16 Oyster Creek Drywell Containment

References:

(1) NRC Letter dated April 24, 1992, "Evaluation Report on Structural Integrity of the Oyster Creek Drywell (TAC No.

M79166)."

(2)

GPUN Letter C320-92-264 dated November 26, 1990, "Oyster Creek Drywell Containment."

In response to the Reference I request, GPU Nuclear commits to continue taking UT drywell measurements at refueling outages and at other outages of opportunity.

The measurements will be at areas previously inspected and also at other accessible areas not previously inspected.

Drywell thickness measurements will continue for the life of the plant.

The following is our current plan for Oyster Creek drywell UT thickness measurements.

(1) During the 14R outage, GPU Nuclear will take UT thickness measurements in the drywell sandbed region, from the torus room side (outside the drywell), at shell locations not readily accessible from inside tne drywell.

These are areas not previously inspected.

The specific locations selected for inspection will be identified once we have direct.

access to the sandbed region.

Assuming that these measurements confirm that we have bounded the corrosion problem with our current inspection locations, we currently do not plan to make repeat measurements at these specific locations.

9206010165 920526 PDR ADOCK 05000219 P

PDR GPU Nuclear Uorporation is a subsidiary of General Public Utilities Corporation

I C321-32-2163 Page 2 (2) Now through the 15R outage, GPU Nuclear will continue taking UT thickness measurements in accordance with the priority method described in Peference 2, Attachment I, "GPUN Specification IS-328227-004, FunLtional Requirements for Drywell Containment Vessel Thickness Examination".

(3) After the 15R outage, GPU Nuclear will assess the condition of the drywell by evaluating the then current UT thickness measurements and will formulate an extended inspection plan.

The plan will identify measurement locations including frequency of inspection for the remaining life of the plant.

If you have any questions or comments on this submittal or the overall drywell corrosion program, please contact Mr. Michael Laggart, Manager, Corporate Nuclear Licensing at (201) 316-7968.

Very truly yours, f

J. C. DeVine, Jr.

Vice President and Director Technical Functions J

CD! RZ/ amk cc:

Administrator, Region I Senior Resident Inspector Oyster Creek NRC Project Manager

-ý!74W- -,r I' -"" --" 7

Docket No. 50-2 19 Mr. John I. Barton Vice President and Director GPU Nuclear Corporation Oyster Creek Nuclear Generating Station P.O. Box 388 Forked River, New Jersey 08731

Dear Mr. Barton:

SU;BJECT: NRC INSPECTION REPORT NO. 50-219/92-08 This letter transmits the report of the resident safety inspection conductcd h% Nilr.

)' Vito utr the period March 29, 1992, through May 2, 1992, at the Oyster Creek Nuclear (;WneraTluig Station. The inspection consisted of document reviews, personnel interviews and observations of activities. Inspectors discussed the findings with Mr. D. Ranft. I'lant Engineering Director, and members of your staff after the inspection.

Inspector observations during this report period indicate that activities conducted wterc salc and conservative. However, we.?ýre concerned about the inadvertent actuation of the containment spray system and the spray of the drywell with approximately 825 galltton ot water. This event was caused by a licensed operator's failure to follow a containnmcnt spra.

system surveillance procedure and is a violation of NRC requirements as specified in tihe enclosed Notice of Violation.

You are required to respond to this letter and should IollAo the instructions specified in the enclosed Notice when preparing your response.

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice." a copy of thiislettcr and its enclosures will be placed in the NRC Public Document Room.

The responses directed by this letter and the enclosed Notice are nol subjecl t ti lhe cl:arancC procedures of the Office of Management and Budget as required hv the I,aperwork R*educ,,ti'n Act of 1980, Public Law No..96.511.

OFFICIAL RECOrD COpy ',0 9206090040 920602 j7, -

L.

PDR ADOCK 05000219 0

PDR

GPU Nuclear Corporation We appreciate your cooperation.

Sincerely..

A. Randolph Bl<ough, Chief Projects Branch No. 4 Division of Reactor ProjcticCs

Enclosures:

1. Notice of Violation

2.

NRC Report No. 50-219/92-08 cc w/encls:

M. Laggart, Manager, Corporate Licensing

-G. Busch, Licensing Manager, Oyster Creek K. Abraham, PAO, (2)

Public Document Room (PDR)

Local Public Document Room (LPDR)

Nuclear Safety Informatioij Center (NSIC)

NRC Resident Inspector State of New Jersey OFFICIAL RECORD COP'r

ENCL()SUtR1E I NOTICI-.-

VIOLATION GPU Nuclear Corporation Docket No. 5O 2 I-Oyster Creek Nuclear Generating Station L.icense No. l)l'R Ib, During an NRC inspection conducted March 29. 1992. through May 2. lQ*-2. at vtiokllttln of NRC requirements was identified.

In accordance with the "General Statement otf Pit.-*

iund Procedure for NRC Enforcement Actions," 10 C(:R Part 2. Appendix C. ( 19921. tle violation is listed below:

Technical Specification 6.8.1 requires that written procedures shall lie csthlisheld.

implemented, and maintained that meet or exceed the requirements of kcgul.atorN Guide (Reg Guide) 1.33, revision 2. "Quality Assurance lProgram Rcwi relments (Operation)."

Reg Guide 1.33, Appendix A requires that procedurcs be wr'nttn !or surveillance testing of the containment spray system.

Station procedure 604.4.007, revision 13, "Containment Spray and liiiergcncy Service Water System I Pump Operability and Inservice Test." step 6.20. requires the containment spray and emergency service water (ESW) pumps to be secured ii inservice testing (IST) is not required to be performed.

Contrary to the above, on April 20, 1992. the control roomn olprator failed to implement procedure 604.4.007 in that the containment spray and l-SW pumpf,

'crc not secured when performance of IST was not required before procecdint it, the ncI\\

step in the procedure. As a result of this action the system,w-as aligned to,praý Hi t-containment when the operator placed the system control. s% itch in the.I\\ Ir() I position and approximately 825 gallons of water were sprayed into the co'll.tnilncnil.

This is a Severity Level IV violation (Supplement I).

Pursuant to the provisions of 10 CFR 2.201, GPU Nuclear Corporation is hereby required it) submit a written statement or explanation to the U.S. Nuclear Regulatory (otmminsslon.

ATTN: Document Control Desk, Washington. D.C. 20555 with a copy to the Rcgionall Administrator, Region I, and a copy to the NRC Resident Inspectur, within 30 days of ihc date of the letter transmitting this Notice of Violation (Notice).

This reply should he cleirly marked as a "Reply to a Notice of Violation" and should include for each vitoltion: i I thec reason for the violation, or, if contested, the basis for disputing the violation. (2) the corrective steps that have been taken and the results achieved. (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance %ill he achieved. Where good cause is shown consideration will bc given to extending the r,,put*,,

time.

Dated at: King of ',ur-ia. PA this 2dday of.Jin 1-92 9206090045 920602 0}'O?*c.

O.....

PDR ADOCK 05000219 pnp

0 Report No.

Docket No.

License No.

Licensee:

Facility Name:

Inspection Period:

Inspectors:

U. S. NUCLEAR REGULATORY COMMISSION REGION I 92-08 50-2 19 DPR-16 GPU Nuclear Corporation I Upper Pond Road Parsippany, New Jersey 07054 Oyster Creek Nuclear Generating Station March 29, 1992 - May 2, 1992 David Vito, Senior Resident Inspector John Nakoski, Resident Inspector John Rogge, Section Chi *ef Reactor Projects Section 4B Approved By:

C. ~k:

,LJ-J.

i)at'e Inspection Summary: This inspection report documents the safety inspections conducted during day shift and backshift hours of station activities including: plant operations; radiation protection; maintenance and surveillance; engineering and technical support; security; and safety assessment/quality verification.

Results: Overall, GPUN operated the facility in a safe manner. A violation was identified as the result of an operator error which caused the inadvertent spray of the drywell with approximately 825 gallons of water from containment spray system 1. This operator error was contrary to the associated containment spray system surveillance procedure.

Two starting failures on emergency diesel generator (EDG) No. 2 were determined to be the result of a broken prop spring on the E-DG output breaker. The licensee's apparent lack of corrective action in response to generic correspondence related to this failure mechanism wW's addressed as part of a separate inspection of the prevcntive maintenance arca (sec Inspection(1 Report 50-219/92-07).

9206090049 920602 PDR ADOCK 05000219

3 Initially, the March 23, 1992, EDG No. 2 start failure was not considered to be reportable because it was believed that the problem was within the automatic synchronization portion of the EDG testing circuitry and did not affect the fast starting capability of the diesel. The results of the root cause assessment performed after the April 5 failure found that this was not the most probable cause. The most probable cause (the prop spring failure) would have affected diesel fast start capability.

The failure mode of the prop spring was originally identified in NRC Information Notice 90(-

41, "Potential Failure of General Electric Magne-Blast Circuit Breakers and AK Circuit Breakers," dated June 12, 1990. General Electric also distributed a service action letter (SAL) dated December 7, 1990, which discussed the prop spring failure mechanism and the availability of a newly designed spring with a considerably longer service life. The licenrsce had not taken correction action related to this generic correspondence prior to the discovery of the broken prop spring on the EDG No. 2 output breaker. The license:e's apparent laýck of effective corrective action related to this generic correspondence was reviewed in detail during a separate inspection of the preventive maintenance area (see Inspection Report 50-219/92-07).

It should.be noted that the EDG No. 2 start failures, caused by the broken prop spring.

provide additional information related to the generic correspondence.

While not spocitically stated, !he related Information Notice and GE SAL imply that the prop spring is necessary Ito ensure breaker closure. However, the results of the GE tests on the removed EDG No. 2 output breaker (breaker latched closed on 3 of 20 tries) with a broken prop spring and the March 23, 1992, and April 5, 1992, EDG No. 2 load test results (i.e., the output breaker successfully latched and remained closed on the second attempt in each casc) show that it is possible for the breaker to remain closed, even with a broken prop spring.

Thus. a breakcr closure test, by itself, may not necessarily reveal a spring failure.

The licensee's decision to call the March 23, 1992, EDG No. 2 start failure a reportable event after completion of their root cause assessment was appropriate.

However, these events could have been precluded had appropriate corrective action been taken on the related generic correspondence.

EDG No. 2 was declared operable on April 9, 1992 after successful completion of po't-maintenance testing. The EDG No. I output t..+/-ker was replace,] with a refurbished breaker on April 27, 1992 and was returned to service later the same day after successful post-maintenance testing. The EDG No. I output breaker had considerably fewer cycles on it (1700) than the EDG No. 2 output breaker (30W0).

The generic correspondence had indicated that prop spring failures were seen to occur at around 20(K) cycles.

1.4 Inadvertent Spray of the Dyywell On April 20, 1992, at 12:54 p.m., approximately 825 gallons of water were sprayed into the drywell during performance of containment spray system I surveillance testing. A control

4 room operator (CRO) was performing survdillance procedure ()7.4.(X)4.

revision 13.

"Containment Spray and Emergency Service Water System ! Pump ()Oprability and Inscrice Test."

The testing was required by technical specifications tlS) because system 2 w.as out of service for preventive maintenance. The plant was operating at 0(X)%

p.-wer and nitrogen makeup to the torus was in progress before the expected decrease in torus pressure duhrinlg the surveillance.

In the process of completing the surveillance the CRO inadverwntly rel.positioned the sysicill control switch from the DYNAMIC TEST I position to the AUTO I [xpsitlin hetfore sc.triiuL the containment spray and emergency service water (ESW) pumps as required. When the control switch was placed in the AUTO I pxosition, the discharge to containment spray val\\c (V-2 1-11) went open and the dynamic test flow return valve i V-2 1-17) w&,ent closed 'as designed. The CRO recognized that the pumps were still running and secured the pulmlps about 30 seconds after placing the system control switch in the AUTO I potIsition.

At 1:02 p.m. the DRYWELL HI LEAK RATIE alarm was received indicating that 0hc unidentified leak rate had increased substantially. The group shift stupervisor M(;SS) and."1A reviewed the emergency -plan implementing procedures (1l:l11s) to determine the nced to enter an emergency condition.

Based on their review and knowledge of the soturce of th.

water (the inadvertent spray of the drywell), they determined that entry intot1 an cmierncyIC action level based on excessive unidentified leakage rate was not required. TS 3.3.D) requires the licensee to reduce the leakage rate to within acceptable limit.,, w.ith i.n hoturs tor place the reactor in the shutdown condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and bc inI c0ld.,shtLItdk

• I within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The leak rate returned to normial levels (about 0141 gpnI unidentified leakrate) within 40 minutes of initiation of spray.

The licensee has experienced two other occasions when the containentc spray system was inadvertently used to spray the drywell. The first occurrence was In l)eceinbcr 1982 whnci ;I CRO mistakenly started a containment spray pump aligned to the drywell and sprayed *bo*t 2000 gallons of water into the drywell (see NRC inspection repo.rt 50-.2 19182 -29 scetion 7.5),

A more recent occurrence on August 6. 1990. involved the leikage of 313 gallo*ns Into t hc drywell during an automatic actuation test. A design configuration deficiency for the position indication of valve V-2 I-5 resulted in the operators leaving the valve partially opn even though it indicated closed (see NRC inspection report 50-21)/i0-12 section 1.2).

lhe licensee had conducted thorough reviews of the effects of the 1992 and 1990() dr, well spray events. Based on the testing and analysis performed in respmnse to the 1982 and 1990) Cv0m, and a review of the environmental qualification of the equipment in the drvwell. t lie Iice*ne*e determined that testing of the main steam line (MSI.) safety and electro-nitic relief v:le, (EMRV) acoustic monitors and thermocouple monitors was warrantcd. The remaining equipment was determined not to be adversely affected by the small aniount (t-f altcr introduced into the drywell.

Plant response was reviewed by a post transient revicw group (tl1*R(

)2-13f,\\iA c**l.ýstiu., W the shift technical advisor (STA) and plant operations and en*gitcerini deptartmnciit,x'rsonncl.

5 Drywell pressure, initially decreased by about 0. 13 psig during the containment spray.

)Once the spray was stopped and pumps were secured, drywell pressure began to increase as the water came into contact with hot components in the drywell and flashed to steam. Drywell pressure remained below any trip setpoints or required action levels, however the DRYWELL PRESSURE HI/LO alarm was received at 12:58 p.m. with a maximum drywell pressure reading of about 1.39 psig. As a result of the containment spray, a minor power transient, lasting about 2 minutes, was experienced causing reactor power to increase from 1930 MWth to a maximum of 1938 MWth. The cause of tile minor power transient was attributed to thermal effects on the sensing lines for reactor water level. When sprayed with the relatively cooler containment spray water, the differential pressure sensed by the reactor water level instrumentation increased due to the cooling of the reference legs. This resulted in an indicated reactor water level less than actual and caused a momentary increase in feedwater flow. The observed reactor power transient was the result of this feedwater ilow transient. No unexpected plant response was noted during the transient based uLpo ll'R(;

review of plant response data obtained from instrument traces and computer data.

The human performance issues identified during the surveillance test were the subject of an April 20, 1992, Operations Critique (number 2100-92-006). During the critique, the licensee determined that the CRO performing the surveillance did not perform tile surveillance as written. Specifically, he failed to secure the pumps before placing the system control switch in the AUTO I position as required by step 6.20 of procedure 607.4.(X)4. Contributing to the event was a weakness in the procedural instructions of step 6.20. This step required the operator to perform a specific set of actions if inservice test (IST) data was to be obtained and a different set of actions if no IST data was being taken. The intermixing of instructions enhanced the probability of the operator missing a required action before proceeding to the next step. Complicating the response to this event was that the CRO performing the surveillance did not inform the other CROs and the group shift supervisor (GSS) until about 10 minutes into the response that he had secured the pumps after placing the system control switch in the AUTO I position.

One of the short-term corrective actions to prevent this event from recurring was to issule ;a temporary procedure change (TPC) to both system surveillance procedures that separated the individual actions of the IST performance paragraph into discrete steps. The procedures had yet to be updated to the procedure writer's guide and were cumbersome to use. As a long-term corrective action, the licensee plans to submit the system I and 2 containment spray/ESW surveillance procedures for review and rewriting to meet the requirements of the procedure writer's guide.

Review of the involved CRO's response to his failure to follow the procedure and the time required for him to provide the information to the others on shift resulted in operations management removing him from licensed duties. The CRO was required to complete a requaliFication program before returning to licensed duties. By the end of tile inspection period the CRO had not yet completed his requalification program and was nol performing licensed duties. The requalification program involved retraining on self-checking; review of

6 procedure 106 "Conduct of Operations;" review of procedure compliance standards; a system checkout on the containment spray system that included a review of the system control logic:

participation in a crew teamwork and leadership session concentrating on individual and team self-checking and intra-crew communication; and interviews with Plant Operations Managers and Directors. Upon completion of the requalification program, the GSS must make a recommendation for requalification followed by an interview with the Vice President and Director, Oyster Creek who will make the determination to return the operator to licensed duties. When the involved CRO returns to licensed duties, he must interview other plant personnel affected by his actions and develop a presentation emphasizing the cost to the company due to the adverse effects on the plant and the potential for more severe adverse effects.

The inspector reviewed procedure 607.4.004; Operation Critique 2100-92-006; a draft version of PTRG report 92-136A; PTRG report for the 1990 event (PTRG 90-135A), NRC inspection reports 50-219/82-29 and 50-219/90-12; observed performance of the MSL safety/EMRV acoustic monitoring surveillance (see section 1.4 of this report); discussed the event with the involved CRO and operations supervision; and monitored plant conditions shortly after the transient had occurred. No abnormal plant response was noted. Control room response to the event was significantly hampered by the failure of the involved CRO to inform the rest of the operating crew of his actions. However, the response was appropriate by the other members of the crew based on the available information. Discussions with the involved CRO were unable to determine the reason for the 10 minute delay in providing the information on continued operation of the containment spray pump while the discharge valve was going open.

The inspector concluded that inadequate self-checking by the involved CRO, contrary to his training, resulted in the operator phissing the requirement to secure the containment spray and ESW pumps. The licensee's actibns to remove the operator from licensed duties and the development of a detailed individual requalification program were appropriate. The inspector was particularly concerned with the CRO's failure to inform the others on shift of the error he had made. The timely.and accurate communication of information between onshift crew.

members is vital to ensure the safe operation of the plant. The corrective actions specified by Operation Critique 2100-92-006 were adequate to address the immediate and long term concerns identified by this event. The PTRG was thorough in reviewing the effects on equipment in the drywell from this event.

Failure to secure the containment spray and ESW pumps as required by procedure 607.4.004, step 6.20 was determined to be a violation of NRC requirements. Specifically.

TS 6.8.1 requires procedures to be established, implemented, and maintained that meet the requirement of Reg Guide 1.33, revision 2, "Quality Assurance Program Requirements (Operation)." Appendix A, to Reg Guide 1.33 requires that procedures be written for surveillance testing of the containment spray system. This event was caused by the failire o*

the operator to adequately perform self-checking resulting in the procedural noncompliance.

Previous events have been cause4 by similar personnel errors (specifically closure of all five

7 recirculation loop suction valves that occurred in August of 1991). As sueh this violation does not meet the criteria for nori-citing as described in 10 CFR Part 2, Appendix C (1992).

(VIO 50-219/92-08-01).

1.5 Reactor Building Ventilalion Trips On April 17, 1992, at 2:30 p.m., a trip of the reactor building (RB) ventilation system occurred. Following the trip the licensee determined that a sticking relay (XC) associated with a high RB pressure sensor located on the 119 foot elevation resulted in the trip. With strong or gusting wind a pressure transient is sensed by this senso, which can cause a large enough spike to generate a trip condition. If the relay sticks the trip condition will not reset before the RB ventilation trips. However, prior to replacing the relay, a second RB ventilation trip occurred on April 22, 1992, at 7:10 p.m.

High RB pressures were not observed during both trips using other indications available for monitoring RB conditions. To correct the problem the licensee replaced the XC relay on April 26, 1992. Since the relay replacement no additional RB ventilation trips asses'iated with the XC relay or strong or gusting winds have occurred.

The inspector observed the response to the RB ventilation trip that occurred on April 17.

Control room operators (CROs) responded to the event by reviewing the RB pressure instrumentation in the control'room, determining which of the trip relays caused the trip, and starting the standby gas treatment system (SGTS) to ensure the RB pressure remained negative. The group shift supervisor (GSS) and electrical maintenance supervisor reviewed the electrical drawings to verify the source of the signal that had caused the trip.

Overall, the inspector' found the response by the CROs was good when the RB vcntilation1 tripped on April 17, 1992. Evaluation of the trip was adequate in determining that a faulty XC relay had caused the RB ventilation to trip on both occasions. The inspector concluded that the licensee's response and corrective actions were adcquamte to address ihe tripping of the reactor building ventilation system.

1.6 Facility Tours The inspectors observed plant activities and conducted routine plant tours to assess equipment conditions, personnel safety hazards, procedural adherence and compliance with regulatory requirements. Tours were conducted of the following areas:

control room intake area cable spreading room 0

reactor building diesel generator building 0

turbine building new ra dwaste building 0

vital switchgear rooms old radwaste building 9

access control points transformer yard

~ IT~ftl STA'TES,

NUCLEAR REGULATORY. COMMISSION I-

-WASHINGTON.

D.C. *

• -O:

J.une 30, 1992`

ocket No. 50-219 Mr. John J. Barton Vice President and Director..

GPU Nuclear Corporation

,Oyster Creek Nuclear Generating'Station I* Post Office Box 388 Forked River, New Jersey 08731

SUBJECT:

OYSTER CREEK DRYWELLCCONTAINMENT (TAC NO.

M79166)

In our letter of April 29,1992,:.regardi'ng Oyster Creek drywell containment,

.we requested that GPU Nuclear Corporationh.(GPUN), continue ultrasonic testing (UT) thickness measurements at refueling outages and at outages of opportunity for the life of the plant.

The measurements should cover not only areas previously inspected but also accessible areas which have never been inspected so as to confirm that the thicknesses of the corroded areas are as projected and the corroded areas are localized.

We also requested that you indicate

..your.intent to comply with the above requirements as discussed in the Safety Evaluation.

.In your letter of May 26, 1992, GPUN committed to continue taking UT drywell measurements at refueling outages and at other outages of opportunity.

The measurements will be at areas previously inspected and also at other accessible

.,-areas not previously inspected.- Drywell thickness measurements will continue for life.

You also indicated that the following is your current plan for Oyster Creek drywell UT thickness measurýement.

(1)

During the 14R outage, GPU Nuclear will take UT thickness measurements in the drywell sandbed region, from the torus room side (outside the drywell), at shell locations not readily accessible from inside the drywell.

These are areas not previously inspected.

The specific locations selected for inspection will be identified once GPU has direct access to the sandbed region.

Assuming that these measurements confirm that GPU has bounded the corrosion problem with current inspection locations, GPU does currently not plan to make repeat measurements at these specific locations.

(2) Now through the 15R outage, GPU Nuclear will continue taking UT thickness measurementis in accordance with the priority method described in Reference 2, Attachment I, "GPUN Specification IS-328227-004, Functional Requirements for Drywell Containment Vessel Thickness I

Examination."

9207060046 920630 PDR ADOCK 05000219 a,

C?..PDR

Mr. John J. Barto II (3) After the 15R outage, GPU Nuclear will assess the condition of the drywell by evaluating the then current UT thickness measurements and will formulate an extended inspection plan.

The plan will identify measurement locations'iAncluding frequency of inspection for the remaining life of the plant.

We have reviewed the above6 Information and find that your program commitments regarding UT inspection of-the Oyster Creek drywell containment are acceptable.

This closes TAC No. M79166.

Sincerely,

.Is/

Alexander W. Dromerick, Sr. Project Manager Project Directorate 1-4 Division of Reactor Projects - I/HI Office of Nuclear Reactor Regulation cc:

See next page Qistrkb uL~LQn Docket File NRC & Local PDRs PD 1-4 Plant SVarga JCalvo

• SNorri s ADromerick OGC GBagchi RHermann ACRS.(10)

RBlough, RI OFFICIAL RECORD COPY Document Name: M79166.INF

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6/11/92

GPU Nuclear Corporation One U~pper Ponld Road Parsippany New Jersey 07.054 201.316-7000 TELEX 136-4,82 Writer's Dirt-.-c Dial Nurnbr:);:

April 19, 19V4 C321-94-2048 U. S. Nuclear Regulatory Commission Att: Document Control Desk Washinytun, MC 20555 Gentlemen:

Subject Oyster Creek Nucl1-.-ar (;enerat ing Station (OCNGS)

Docket 50-219 SEP Topic 1ll-7B, Drywell Shield Wall Integrity Our letter dated November 19, 1993 transmitted ABB Impell Corporation Report No.

1037-00196-01, Rev.0 which provides calculated stresses in the concrete and reinforcing bars in the drywell shield wall above elevation 95 ft.

The results indicate that stresses are well below allowables taking into consideration the existing (cracked) condition of the

'iield wall.

During refueling outages, the ioŽactor cavity is flooded and the inside surface of the drywell shield wall is exposed to some water due to leakage past the steel plate covering the cavity surface.

This water could enter the pre-existing cracks in the concrete wall and wet the surface of the steel reinforcing.

However, during normal operation, very little moisture is present in the vicinity of the drywe~l shield wall due to the relatively high temperatures and the fact that the cavity is not flooded.

In our recent phone conversatie-concerning the subject matter, the NRC staff reque:;ted GPU Nuclear to establish a crack monitoring program for the drywell

.concrete shield wall to provide confirmatory information regarding shield wall conditio,,s. During the phone conversation, GPU Nuclear Informed the NRC staff.

that. structural systems engineer is assigned to the Oyster Creek site.

Ihe systems engineer is responsible for ensuring that the structures at Oyster Creek are monitored and evaluated.

The NRC vxpressed a desire that a formal program to monitor cracks in the drywe)l sl*nld wall be established.

9404290153 940419 P

PDR 2.j i,)

i -

I) -...

GPU Nuclear Corporation is a subsididary of Guon.,l Fublic UIlilih*s Co(.oiporalion

C321-94 2048 Page 2 GPU Nuclear agrees to perform the periodic inspections of the drywel I shield wall as requested by the NRC staff.

Therefore, GPU Nuclear is developing a program to ensure monitoring of con( ret e c)nd it ions during each refuel ing outage and a f ormal u idel ine for performing the monitoring (e.g. visual inspec t. i on-. for c rac..k growth and/or sta i n i ng of the coritcrete).

The program and gouidel ioe wI 1

II in pkice prior to refuel ing outage 15R.

Sincerely, R. W. Keaten I) i *i'c:t. r'r, fec:hn ical I FuncLion I t c.c A(!

n i rtart.or

• • q ion 1 N.

(..

16 1 iden I I,ii pvc f. or Oy't. r (.. ree(k N16. Pro ji, t Maniii.or

Nuclear GPU Nuclear Corporation One Upper Pond Road Parsippany, New Jersey 07054 201 316-7000 TELEX 136-482 Writers Direct Dial Number.

September 15. 1995 C321-95-2235 5000-95-0088 U. S. Nuclear Regulatory Commission Ali: Document Control Desk Washington, DC 20555 Gentlemen:

Subject:

Oystcr Creek Nuclear Generating Station (OCNGS)

Docket No. 50-219 Facility Operating License No. DPR-16 Drywell Corrosion Monitoring Program

References:

(1)

GPU Nuclear Letter C321-92-2163, "Oyster Creek Drywell Containmeht," May 26, 1992.

(2)

NRC Letter dated June 30, 1992. "Oyster Creek Drywell Containment."

(3)

GPU Nuclear Letter C321-93-2100, "Oyster Creek Drywell Inspection," March 25, 1993.

In compliance with Item (3) of References I and 2, and Reference 3, GiU Nuclear has (1) assessed the condition of the drywell based upon inspections performed at Oyster Creek during the 15R Outage and is (2) submitting an extended drywell inspection plan for the remaining life of th2: plant. GPU Nuclear remains committed, as stated in Reference 1, to' continue taking drywell thickness measurements for the life of the plant.

Through the I5R Outage, GPU Nuclear's drywell containment vessel thickness monitoring program, Item (2) of References I and 2, consisted of ultrasonic thickness (UT) measurements taken at the sandbed region and upper elevations (cylinder, sphere) of the drywell during refueling outages and other outages of opportunity.

Assessment of the most recent UT data taken during the 15R Outage has determined that there is no evidence of ongoing corrosion in the upper elevations of the drywell and that corrosion has been arrested in the sandbed region of the drywell which was cleaned of sand and rust and coated during the 14R Outage (December 1992). The autached table summarizes the 15R Outage UT inspection results for both the sandbed region and upper

. 0

-400..

1.0

)0 P. 0 oy GPU Nuclear Corporation is a subsidiary of Genpral Publi, (Itilille's;.oronratinn

C321-95-2235 Page 2 surfaces indicates that, after 21 months of service, the coating is performing satisfactory with no signs of deterioration such as blisters, flakes, discoloration, etc.

GPU Nuclear's extended inspection plan for the Oyster Creek drywell containment vessel covers both the upper elevations of the drywell and the coated sandbed region.

For the upper elevations of the drywell, this program will perform UT measurements during the 16R Outage (currently scheduled to begin September, 1996) and, as a minimum, again during every other refueling outage (18R, 20R, etc.). The UT measurement locations will be the nine areas identified as most severely corroded. Assessment of the most recent UT data taken during the 15R Outage has determined (and will be reconfirmed by the 16R inspections) that there is no evidence of ongoing corrosion in the tipper elevations of the drycll. After each inspection, a technical assessment of the drywell condition will be made. any appropriate corrective action will be taken, and any necessary additional inspections would be scheduled to ensure that drywell integrity is maintained for the remaining life of the plant.

For the sandbed region of the drywell, this program will perform visual inspection of the external epoxy coating during the 16R Outage and, as a minimum, again during the 18R Outage (year 2000). The epoxy coating has an estimated life of 8-10 years which makes the current projected end of life between December, 2000 and December, 2002. Coating inspection shall be by direct (physical) and/or remote methods on a sample basis. Bascd upon these inspections, a technical assessment of the coating condition will be made, any appropriate corrective action will be taken, and the need for additional (post 18R) inspections will be determined to ensure that drywell integrity is maintained for the remaining life of the plant. In addition, while not technically required based upon the performance of the epoxy coating, UT thickness measurements will be taken one more time in the sandbed region during the 16R Outage, to the same extent as the 15R Outage inspections.

In compliance with Reference 3. GPLI Nuclear remains committed to inform the NRC prior to implementing any changes to this drywell inspection program.

Very truly yours,

,+

R. W. Keaten Vice President and Director Technical Functions Attachment RTZ/plp c:

Administrator, Region I Senior Resident Inspector Oyster Creek NRC Project Manager

TABLE 1 ACCEPTABLE MEAN DRYWELL THICKNESSES 15R OUTAGE INSPECTION DRYWELL THICKNESSES UT CODE LOCATION NOMINAL MEASURED REQUIRED MARGIN MINIMUMS (1)

Sandbed Region

1. 154" 0.806"

.736" (2)

.070" (3)

Sphere (el. 50' - 2")

0.770" 0.733" 0.541" 0.192" Sphere (el. 51' - 10")

0.722" 0.61W" 0.518" 0.177" Sphere (c1. 60' - I1")

0.722" 0.709" 0.518"

0. 191" Cylinder (el. 87' - 5")

0.640" 0.613" 0.452" 0.161" (I)

Thinnest Location as measured during the (2)

Controlled by buckling.

(3)

Corrosion arrested (sandbed region coated 15R outage, September, 1994.

in 14R outage).

oU, UNITED STATES I

oNUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20566-401 November 1, 1995 Mr. John J. Barton Vice President and Director GPU Nuclear Corporation Oyster Creek Nuclear Generating Station P.O. Box 388 Forked River, NJ 08731

SUBJECT:

CHANGES IN THE OYSTER CREEK DRYWELL MONITORING PROGRAM (TAC NO. M93658)

Dear Mr. Barton:

In a letter dated September 15, 1995, GPU Nuclear Corporation (GPUN) stated that they assessed the condition of the drywell based upon inspections performed at Oyster Creek during the 15R refueling outage (15R) and submitted an extended drywell inspection plan for the remaining life of the plant.

GPUN also stated that they remain comitted, as stated in their letter of May 26, 1992, to continue taking drywell thickness measurements for the life of the plant.

The staff has reviewed the information provided by GPUN and tvncludes that changes in the drywell corrosion monitoring program as planned by GPUN is acceptable if GPUN commits to additional inspection within approximately 3 months after discovery of water leakage from the pools above the reactor cavity.

Our safety evaluation is enclosed.

Within 30 days of the date of this letter, we request that you provide your intent to perform additional inspection within approximately 3 months after discovery of water leakage.

This requirement affects nine or fewer respondents and, therefore, is not subject to the Office of Management and Budget review under P.L.

No.96-511.

Sincerely, Arexander W. Dromerick, Senior Project Manrager Project Directorate 1-3 Division of Reactor Projects -

1/11 Office of Nuclear Reactor Regulation

Enclosure:

Safety Eval iation cc w/enclh See next page

AEG 4:10 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 2056-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION DRYWELL MONITORING PROGRAM GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO.

50-219 GPU Nuclear Corporation (GPUN),

the Oyster Creek Nuclear Generating Station licensee, previously, in a letter dated May 26, 1992, committed to conduct ultrasonic thickness (UT) measurements of the drywell at refueling outages (RO) and at other outages of opportunity.

The areas to be monitored are the upper elevations and the sandbed regions of the drywell where corrosion had been detected.

During the 14th RO (December 1992) the sandbed region of the drywell was cleaned of sand and rust, and coated.

During the 15th RO the licensee made UT measurements at the sandbed region and at the upper elevations (cylinder and sphere) of the drywell.

In a letter dated September 15, 1995, GPUN stated that they assessed the results of the inspection and determined:

(1) there is no evidence of ongoing corrosion in the upper elevations and (2) the corrosion of the sandbed region has been

rrested.

On the basis of this finding the licensee has proposed to reduce their inspection program as follows:

1.

For the upper RO (September inspection, a be performed.

elevations, UT measurements will be made during the 16th 1996) and during every second RO, thereafter.

After each determination will be made i.t additional inspection is to

2.

For the sandbed region visual inspection of the coating as well as UT measurement of the shell will be made during the 16th RO.

The coating will be inspected again during the 18th RO (.year 2000).

Based on the results of inspection of the coating, determinations will be made for additional inspections.

The licensee has provided a table of UT measurement results from inspection.

This table shows the locations of the measurements, as-constructed thickness, the minimum as measured thickness, the required thickness and the corrosion margin available.

the 15th RO the nominal ASME Code On the basis of the information provided, the staff finds the proposed change to the licensee's previous inspection commitment to be reasonable and acceptable.

However, since water leaking from the pools above the reactor cavity has been the source of corrosion, the licensee should make a commitment to the effect that an additional inspection of the,irywell will be performed about 3 months after the discovery of any water leaKage.

Principal Contributor:

C. P. Tan Date: November 1, 1995 9511060071 951101 PDR ADOCK 05000219 P

PDR

~

p'~ATTACIMENT

~~

EVALUATION REPORT

~

~

oyster Crook Nuclear Generatin Statlion Docket Io, 5O-.

Engiinelindg a iGolllSn

',CJ5 hfl" CP Nc e

Statiohn liconsee, previo ully

{(Ma 1992)oi¢ tted to. conduct ul traonIthiC kfle(iUT) measuremet of" he drywellat, rlfuel Ing outagecS(0), and at-other outgles ioopportunit.

-Th"'area to. be monitored.are t upper, elevations and the sandbed regions oW 1'drywllwtItecorrosion l had been dotcteod. During the4th RO (Dec.mber..

199*2) the sIndbed ve ion.ofrthe dryullS was

,claned.,ofsand and.rust and..d

Ine.

l thelth RO' theI ic'ns made UTmeasurments at-the sandb ed

.c oatidi.;

Ou,

.,t.

h.d 1.. h hgion n

upperd el evations(clinder. nd sphere) of the dr.el Te,

.Aoý 4 licensee Assessed the results of.the Inspection and determined: (1 there Is no evidence of ongoing corrosion i'n. the upper, eleovations and (2) the corrosion of the sandbed region has been arrested.y-On the basis of this finding the license* has proposed to educe their*lnspectioo program as follows:

'ii 1, For the upperelit@vtIonsf

, UT measurements will be made during the 16th RO (September, 1996) and during every second RO, there after.

, After each Inspection,,-a determlnation will be made if additional i nspectlon, isto.lbo performed.

2.'

For the sandbed region visual'inspettiofl of the coating as well as UT measurement of, the shell will be made during theo16th RO. The coating will be-Inspected again durtng the 18th RO (year 2000).o' Based on the ults:of-nspection.of.thoec~ati nizdet"I nitions wil1.be made for

&;2dditional'inspections. aigAtriaifswl emd o The licentsee has pro'vided a tabl of1 UT measurement results from the 15th RO inspection.

s table shovsthe locations of the masurements, the nominal as-constructed thickness,., the minimum* as measured thickness, the ASHE Coderequird thickness end the corrosion margin available.,

!On the bai*s o

• theInformation provided' the staff finds the proposed change to thelicaensee!s previouslnsipection commitment to be reasonable and acceptabl a;Howeve*i wat

""leaking from the pools above the reactor,

, cav tj ;has been the source of.onrrosion, thelic-nsee should make a commitment the oeffect.thatiin*additional o

i, nspection of the drywol 1 will be performed A.bout three" mnths after. the*discovery olf.n-y water, leakagoe.'c,:.

'.,-.,o

a CPUl Nuclear Corporation One Upper Pcrid Road

~

N uclear Parsippany, 14ew Jersey 07054 201-316-7000 TELEX 136-482 Writers Direct Dial Number.

December 15, 1995 5000-95-098 C321-95-2360 U. S. Nuclear Regulatory Commission Att: DocumentControl Desk Washington, DC 20555 Gentlemen:

Subject:

Oyster Creek Nuclear Generating Station (OCNGS)

Docket No. 50-219 Facility Operating License No. DPR-16 Drywell Corrosion Monitoring Program

References:

(1)

NRC Letter dated November 1, 1995, "Changes in the Oyster Creek Drywell Monitoring Program."

(2)

GPU Nuclear Letter C321-95-2235, "Drywell Corrosion Monitoring Program," September 15, 1995.

Reference I requested GPU Nuclear to make a commitment, as part of the proposed extended Oyster Creek Drywell Monitoring Program (Reference 2), to perform "...additional inspection within approximately 3 months after discovery of water leakage from pools above the reactor cavity." Subsequent discussion with the NRC Staff provided clarification that this request was made to address contingency actions should water leakage be discovered during power operation between scheduled drywell inspections. The requirement was not meant to apply to minor leakage associated vith normal refueling activities.

Accordingly, GPU Nuclear proposes to commit to take the following actions should water leakage not associated with normal refueling outage activities be discovered during power operation.

(1) The Oyster Creek NRC Resident Inspector will be notified of thc discovery of leakage.

(2) The source of leakage will be investigated and appropriate corrective actions taken.

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S (3) An evaluation of the impact of the leakage on drywell structural integrity will he performed to ensure sufficient structural margin is maintained for operation to the next scheduled drywell inspection.

(4) In the unexpected event that the evaluation of the impact of the leakage on drywell structural integrity does not ensure sufficient structural margin will be maintained for operation to the next scheduled outa,,e, an additional drywell inspection wViii be performed within approximately 3 months after discovery of water leakage.

If you have any questions or comments on this submittal, please contact Mr. Ron Zak.

Corporate Regulatory Affairs at (201) 316-7035.

Very truly yours.

R. W. Keaten Vice President and Director Technical Functions c:

Administrator, Region I Senior Resident Inspector Oyster Creek NRC Project Manager

Mr. Michael B. Roche Vice President and Director GPU Nuclear Corporation Oyster Creek Nuclear Generating Station P.O. Box 388 Forked River, NJ 08731 SUB.]ECT:

CHANGES IN THE DRYWELL CORROSION MONITORING PROGRAM (TAC N40.

M92688)

Dear M-. Roche:

In a letter dated November 1, 1995, NRC Informed GPU Nuclear Corporation (GPUN) that the changes to the previously committed Drywell Corrosion Monitoring Program as delineated in GPUN's letter dated September 15,

1995, are acceptable.

However, GPUN is required to make a commitment to perform additional inspections of the drywell 3 months after the discovery of any water leakage.

GPUN felt such a requirement is too broad to be cost effective.

In a letter dated December 15, 1995, GPUN clarified its commitment and an understanding between the NRC staff and GPUN has been reached.

The requirement is to address water leakage discovered during power operation between scheduled drywell inspections.

The requirement was not meant to apply to minor leakage associated with normal refueling activities where minor leakage is defined as less than 12 GPM (gallons per minute).

GPUN indicated that prior to each refueling outage, a refueling cavity and equipment pool inspection and leak assessment plan is put in place and the plan has been found to be successful In prior outages.

For leakages not associated with refueling activities, GPUN will investigate the source of leakage, take corrective actions, evaluate the impact of the leakage and, if necessary, perform an additional drywell inspection about 3 months after the discovery of the water leakage.

Based on the additional information provided by GPUN, the staff finds GPUN's commitment to perform the inspections acceptable.

Sincerely, Pr iinal1 sion hI 9602220207 960215 PDR ADOCK 05000219 P

P I)R Docket No.

50-219 cc:

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