DISTRIBUTION AFTER ISSUANCE OF OPERATING LICENSE 195 …

DISTRIBUTION AFTER ISSUANCE OF OPERATING LICENSE

NRC.*oRM 195 U.S. NUCLEAR REGULATORY COM 0 R

(2-76)

-NfRC DISTRIBUTION roR PART 50 DOCKET MATERIAL

TO- FROM: Duke Power Co. DATE OF DOCUMENT

EMr., Edson G. Case Charlotte, N. C. 28242 11/21/77 William 0. Parker DATE RECEIVED

11/23/77

mETTER ONOTORIZED PROP INPUT FORM NUMBER OF COPIES RECEIVED

XORIGINAL /ZUNCLASSIFIED OcoPv

ESCRIPTION Furnishing response to NRC'S ENCLOSURE

Itr dtd 10/11/77 concerning minimum

qualifications of the individual perfoiming the function of Radiation Protection Manager

i.e., the Station Health Physicist, at Oconne Nuc. StAtion...Advising personnE appointed to the position of Station HE&lth Physicist are and will continue to be cualified

as specified in ANSI N18.1-1971...

3 p

PLANT NAME: OCONNE UNITS 1, 2, & 3

jcm 11/25/77

SAFETY FOR ACTION/ NFORMATION BRANCH CHIEF: (7)

INTERNAL DISTRIBUTION

T.COLLINS__ _ _ ___ _ _ ____ _ _

EXTERNAL DISTRIBUTION ____________ CONTROL NUMBER

CS16 CYSSENT CATEG .

kYP___________

DUKE POWER COMPAN A POWER BUILDING

422 SOUTH CHURCH STREET, CHARLOTTE, N. C. 28242

WILLIAM 0. PARKER,JR.

VICE PRESIDENT TELEPHONE: AREA 704 STEAM PRODUCTION 373-4083

November 21, 1977

Mr. Edson G. Case, Acting Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Attention: Mr. A. Schwencer, Chief Operating Reactors Branch #1

Reference: Oconee Nuclear Station Docket Nos. 50-269, -270, -287

Dear Mr. Case:

In your letter of October 11, 1977, you provided additional comments on the minimum qualifications of the individual performing the function of Radiation Protection Manager (RPM), i.e., the Station Health Physicist, at Oconee Nuclear Station. Our position remains that the qualifications established in ANSI N18.1-1971 are appropriate minimum requirements for this position as stated in our May 13, 1977 response to your initial correspondence dated March 9, 1977.

Your letter advised that Regulatory Guide 1.8 does not require the RPM to have a Bachelor's Degree. However, it does state that the RPM must have nine years of training and experience. This experience level is on the same level as that required for the position of station Manager and Technical Services Superintendent. These positions require ten and eight years of training and experience,respectively, and represent the two levels of supervision above the RPM. The fact that the station Manager and the Technical Services Superintendent have greater responsibilities than the RPM is evident, therefore, it is expected that the minimum requirements of the RPM should not be greater than these two positions.

In addition, the position of Station Health Physicist at Oconee Nuclear Station is not identical to the position of RPM.as described in Regulatory Guide 1.8. In Duke's unique situation, the Station.Health Physicist and the Station Health Physics organization are supported by a General Office Health Physics staff called the System Health Physics Unit. Therefore, the Station Health Physicist at Oconee is not the sole person who fulfills the role of RPM as defined by Regulatory Guide 1.8. The System Health

7732 0063

Mr. Edson G. Case Page 2 November 21, 1977

Physics Unit establishes the health physics program for Oconee Nuclear Station, provides technical direction for conducting the programs, establishes the environmental radioactivity monitoring program and the emergency plan; audits the efficacy of these programs and modifies them as required and coordinates a centralized Radiological Laboratory which provides personnel dosimetry, instrument calibration and environmental monitoring.

As described in our May 13, 1977 letter, the System Health Physics Unit represents over 60 man-years of direct power reactor health physics experience. The staff of the System Health Physics Unit consists presently of 12 people, nine of whom are professionals in the Health Physics field, including four with Master Degrees and four with Bachelors Degrees.

It is felt that the Station Health Physicist qualified to ANSI N18.1-1971, supported by the System Health Physics Unit provides a more comprehensive radiation protection program than a single individual qualified to Regulatory Guide 1.8. Our position is further supported by a December, 1976 proposed revision to ANSI N18.1 which states under Section 4.4, Radiation Protection, "At the time of initial core loading or appointment to the position, whichever is later, the responsible person shall have a minimum of four years experience in applied radiation protection. A minimum of two of the four years shall be in radiation protection work at an operating nuclear power plant. In-addition, he shall have a minimum of two years of related technical training or have satisfactorily completed. a comprehensive specialty program in radiation protection. In addition, the owner organization shall have either in the person described above or another person and either on site or off site an individual possessing the following qualifications. A Bachelor's Degree or the equivalent in a science or engineering subject, including formal training in radiation protection. He should have at least five years of professional experience in applied radiation protection. (A post graduate degree may be considered equivalent to one year of professional experience and a doctor's degree may be considered equivalent to two years of professional experience where course work related to radiation protection is involved). At least three years of this professional experience should be in applied radiation protection work at an operating nuclear power plant."

Since the Station Health Physicist (RPM) is supported by two levels of management and a central Health Physics organization, and the requirements of ANSI N18.1-1971 are minimum requirements, it is considered that the proposed qualifications of Regulatory Guide 1.8 are not appropriate for Duke Power Company. In addition, it is considered that the program for qualification of the Oconee Station Health Physicist will assure that personnel assigned to this position are fully capable of performing the required duties.

Mr. Edson G. Case Page 3 November 21, 1977

It is, therefore, concluded that personnel appointed to the position of Station Health Physicist are and will continue to be qualified as specified in ANSI N18.1-1971.

Very truly yours,

William 0. Parker, Jr.

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Duke Power Cimipany ATTh: Mr. Uiliavm . Parker, Jr.

VIce President Star Production

Post Office Bex 2178 422 South Church Street Charlotte, North Carolina 2242

Gantiem~en

On or ebout November 8, 1.977, you should have receIved ' Sulleti 0o. 77-05 on Electrical Connector Assemblies frwom the Regional Office of the 1C Office of inspection and Tforcement and a suppleeant to tJis bulletin o- or -about Hoveter 15, 1977, (coples enclosed). This Pulletin reQuests that, within 30 days, you provide Inforation on whether yer Gcorec 1, 4 and 3 factlities utilize electrical connector assermbles within containiment that could be subject to an accident. envIronment and are rmquired to be opertble during an accident; and, if4they do, to iiprOie CocMentatlo of the v quaificton testing performed for these connectors.

Due to the safety significance of this matter, the staff also conducted a telOhone survey of all operating plants to obtain preliminary inforrati on which facilities utilze such connectors and the adequacy of gualification testing of any sucheconnectors. -As you know, this survey determined that such connectors are used In your 3conee 1, 2 and 3 facilitis. Frther, as a res.Ilt of meetings held Ith your representatives daring the week.of Iovember 7, 1977, we deterined or' a preibinary basis, that the environmental testing of the electrical connectors used in your facilities arnd the documentation thereof, apnear to be satIsfactory. Confirmation of this is to be provided by your resoonse to the IC tulletin and the supplement thereto.

DATE*I *I ________ _______ _

NRC FORM 318 (9-76) NRCM 0240 us e. eENErovnPiF iri Sm8FCESI3 . 20 68S

. uke Power Cpany - 2 NOV 18 1977

If, however, in responding to the Bull Dtia, you are not able to provide complete documentotio of the required qualification tosting of all applicable coonectors, the Bulletin requires that you submit plas and programs toward qualifying existing equipment or plans for replacing mnqualified connectors with qual i fied equipment.

This is to advise you that if you are not able to provide the necessary documentation, then, pursuant to 10 CFR 50.54(f)" at the same time that you respond to the Bulletin, you are requested to submait to the Office of.Nuclear Reactor Regulation, Nuclear Regul'atory Commission, W'as h-igto.,Jt'. C. 555, Justification- that continued operation of the facilities fIr whatever period it will take to environrentfally qualify the connectors in accorlance with NRC regulations would not create an undue risk to the health and safety ofjthe pulic

Please advise us if you have any questions relative to thisatter

Sincerely,

dszorG. Case, Acting Director Office of Nuclear Reactor Pegulation

Encl osures: DSRBTO As stated DISTRIBI N

Docket cc W/enclosures; NRC PDR

See nLocal PDR .

EDO Reading NRR Reading VStello KRGol ler ASchwencer DNeighbors SMSheppard OELD OI&E(3) DEisenhut TB Abernathy JRBuchanan ACRS(16)

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NRC FORM 318 (9-76) NRCH 0240 US s. GOVERNMENT PRINTING OPPICEs 1976 - 828-82

1 _ .9 UNITED STATES

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

November 18, 1977

Docket Nos. 50-269 50-270

and 50-287

Duke Power Company ATTN: Mr. William 0. Parker, Jr.

Vice President Steam Production

Post Office Box 2178 422 South Church Street Charlotte, North Carolina 28242

Gentlemen:

On or about November 8, 1977, you should have received IE Bulletin No. 77-05 on Electrical Connector Assemblies from the Reqional Office of the NRC Office of Insoection and Enforcement and a supplement to this Bulletin on or about November 15, 1977, (copies enclosed). This Bulletin requests that, within 30 days, you provide information on whether your Oconee 1, 2 and 3 facilities utilize electrical connector assemblies within containment that could be subject to an accident environment and are reQuired to be operable during an accident; and, if they do, to provide documentation of the environmental qualification testing performed for these connectors.

Due to the safety significance of this matter, the staff also conducted a telephone survey of all operating plants to obtain preliminary information on which facilities utilize such connectors and the adequacy of qualification testing of any such connectors. As you know, this survey determined that such connectors are used in your Oconee 1, 2 and 3 facilities. Further, as a result of meetings held with your representatives during the week of November 7, 1977, we datermined on a preliminary basis, that the environmental.testing of the electricnl connectors used in your facilities and the documentation thereof, appear to be satisfactory.. Confirmation of this is to be provided by your response to the IE Bulletin and the suppleme'if thereto.

Duke Power Company - 2

If, however, in responding to the Bulletin, you are not able. to provide complete documentation of the required qualification testing of all applicable connectors, the Bulletin requires that you submit plans and programs toward qualifying existing equipment or plans for replacing unqualified connectors with qualified equipment.

This is to advise you that if you are not able to provide the necessary documentation, then, pursuant to 10 CFR 50.54(f), at the same time that you respond to the Bulletin, you are requested to submit to the Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, Washington, D. C. 20555, justification that continued operation of the facilities for whatever period it will take to environmentally qualify the connectors in accordance with NRC-regulations would not create an undue risk to the health and safety of the public.

Please advise us if you have any questions relative to this matter.

Sincerely,

Edson G. Case, Acting Director . Office of Nuclear Reactor negulation

Enclosures: As stated

cc w/enclosures: See nexit page

Duke Power Company - 3 - November 18, 1977

cc: Mr. William L. Porter Duke Power Company P. 0. Box 2178 422 South Church Street Charlotte, North Carolina 28242

J. Michael McGarry, III, Esquire DeBevoise & Liberman 700 Shoreham Building 806-15th Street, NW., Washington, D.C. 20005

Oconee Public Library 201 South Spring Street Walhalla, South Carolina 29691

NUCLEAR REGULATORY COMMISSIO0 OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D. C. 20555

November 8, 1977

IE Bulletin 77-05

ELECTRICAL CONNECTOR ASSEMBLIES

Description of Circumstances

Recent tests conductedby the Sandia Laboratories of electrical connector/cable assemblies in a simulated post-LOCA containment environment (LWR) demonstrated that the assemblies failed to perform in an acceptable mapner. The connectors are the pin and socket type, with metal shell and screw couplings. The specific test specimens w-.ere manufactured by Bendix, ITT Cannon and Gulton Industries using combinations of Anaconda and ITT Surprenant cables. Details of the specific connector/cable combinations, test conditions, test results and other pertinent information are described in the Attachment.

While electrical connectors of the type tested are not normally used in applications that are required to survive LOCA conditions, it is not possible in the absence of specific information to conclude that such applications do not exist. Further, it is unknown whether other manufacturers have supplied similar assemblies, whether such assemblies have been properly qualified !for the intended service or whether these types: of assemblies are utilized in applications that must continue to operate reliably in a LOCA environment.

Action To Be Taken By Licensees and Permit Holders:

For all power reactor facilities with an operating license or a construction permit:

1. Determine whether your facility utilizes or plans to utilize electical connector assemblies of the type. tested by Sandia Laboratori. , or any other similar type, in 'systems that are located inside contaiument, are subject to a LOCA environment\and are required to be operable during a-LOCA.

2. If any such applications are identified, review the adequacy of qualification testing for the assemblies and submit the documentation for NRC review.

1 of 2

IE Bulletin 77-05 November 8, 197

3. If evidence is not available to support a conclusion of adequacy, submit your plans and programs toward qualifying existing equipment or your plans for replacing unqualified assemblies with qualified equipment.

4. Provide your response in writing within 30 days for facilities with an operating license and within 60 days for facilities with a construction permit. Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the U. S. Nuclear Regulatory Comission, Office of Inspection and Enforcement, Division of Reactor Construction Inspection, Washington, D. C. 20555.

Approved by GAO, B180225 (R0072); clearance expires 7-31-80. Approval was given under a blanket clearance specifically for identified generic problems.

Attachment: Trip Report by W. R. Rutherford

Electrical Connector Assemblies

2 of 2

IE Bulletin-77-05 November 8, 1977

TRIP REPORT by

W. R. Rutherford


On September 1, 1977 a meeting was held in Alburquerque, New Mexico to

investigate the electrical connector assembly malfunctions or failures

that occurred during tests under LOCA conditions performed by Sandia

Laboratories. The following is a description of the equipment, test

scope and results of these tests.

Equipment

The test assemblies of particular interest consisted of three types of

connectors: Bendix, ITT Cannon, and Gulton installed on two types of

cables; Anaconda and ITT Surprenant.

1. Bendix Connector: A 3 conductor/No. 12 AWG with crimp pin conductors, anodized aluminum shell, silicone rubber insert, rigid back plane

potting, pliable over-potting.

2. ITT Cannon Connector; A 3 conductor/No. 12 AWG with crimp pin conductors, anodized aluminum shell, silicone rubber insert, anodized

aluminum back shell, rubber packing boot, mechanical retaining clamp.

3. Gulton Connector: A 3 conductor/No. 12 AWG with crimp pin conductor,

stainless steel shell, hard fiber insert, pin back sealed with RTV 112,

stainless shell, back plane poured with Sylgard potting, mechanical

clamp termination.

4. Anaconda Cable: A 3 conductor/No. 12 AWG, tinned copper conductor,

30 mil ethylene propylene rubber insulation 15 mil Hypalon jacket,

cable asbestos tape, 60 mil Hypalon jacket, rated 600 volts, cable

diameter 0.55".

5. ITT Surprenant Cable: A 3 conductor/No. 12 AWG, tinned copper conductor,

30 mil Exane II insulation, silicone glass tape, 65 mil Exane jacket,

rated 600 volts, cable diameter 0.455".

Attachment A Page 1 of 3


Test Scope

The three tests performed by Sandia were composed of two sequential and one simultaneous exposure to LOCA environments. In each case the equipment was exposed to radiation and thermal aging prior to operating under the simulated LOCA conditions. Figures 1 and 2 describe the test profiles for sequential and simultaneous tests respectively (Sandia tests were designed to study synergistic effects). Each of the tests satisfy the intent of IEEE 323-1974. The assemblies were electrically loaded to 20 amperes and 600 volts at the start of the tests. Insulation resistance and capacitance measurements were recorded during the tests to indicate damage.

The equipment assemblies with respect to the sequextial and simultaneous tests performed were as follows:

1. Sequential Tests (Two)

.Gulton Connector/ITT Cable 1 Asstmbly Gulton Connector/Anaconda Cable 1 Assembly Bendix Connector/ITT Cable 2 Assumblies ITT Connector/ITT Cable I Assembly

2. Simultaneous Test (One)

ITT Connector/ITT Cable 1 Asstembly Bendix Connector/ITT Cable 1 Assenbly Bendix Connector/Anaconda Cable 2 Assemblies

Test Results

Both ITT Cannon connector assemblies and both Gultan connector assemblies showed almost immediate damage in terms of insulation resistance and capacitance as the 70 psig steam was applied.

The ITT Cannon connector assembly failures appeared to be back plane boot seal leakage failures. The assembly construction did not contain potting compound (by design) to protect the pin backs. Therefore, boot failure leads directly to connector failure.

In the case of the Gulton Assemblies, failures were attributed to both the mating surface interface and the back plane seal. The design uses a rigid insert around the mating pins and the 0-ring seals are



bypassed by an alignment key slot. This design may lead to leaks due to non-uniform confinement of the 0-ring which could cause arcing between pins. Neutron radiography revealed inadequate amounts of potting compound (voids) and cracking of potting compound. These conditions could account for back plane failures. Neutron radiography performed on untested connectors revealed similar conditions, i.e., voids and cracking, thus indicating an apparent quality control problem at Gulton's facility. Other.problems detected were identified as:

1. The shrink tube used over the pin cable interface was split lengthwise and had pulled away.

2. The potting material showed virtually no adhesion to, or sealing between, the.cable jacket, insulation, and the connector shell.

3. The mechanical clamp had been secured so tightly that it cut the cable jacket.

The Bendix connector assembly was the only type to survive an entire test cycle. One Bendix/Anaconda assembly malfunctioned after about eight days into the 10 psig profile and the Bendix/ITT assembly experienced decreasing resistance and increasing capacitance through the simultaneous tests until both readings were off scale at the end of the 10 psig profile. A second Bendix/Anaconda assembly survived the sinultaneous tests. During the sequential tests only Bendix and ITT Cannon assemblies were involved and both assemblies failed. The failures of these assemblies would be difficult to define as either connector or cable failures. The ITT cable exhibited a shrinking characteristic which could tave provided a leak path through the sealing medium of the connector.



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LISTING OF IE BULLETINS ISSUED IN 1977

Bulletin Subject Date Issued Issued To No.

77-04 Calculational Error 11/4/77 All PWR Power Affecting the Design Reactor Facilities Performance of a System with an Operating for Controlling pH of License (OL) or Containment Sump Water Contruction Permit (CP) Following a LOCA

77-03 On-Line Testing of 9/12/77 All W Power the W Solid State Reactor Facilities Protection System with an Operating

License (OL) or Construction Permit (CP)

77-02 Potential Failure 9/12/77 All Holders of Mechanism in Certain Operating Licenses W AR Relays with (OL) or Construc

Relays with Latch tion Permits (CP) Attachments

77-01 Pneumatic Time 4/29/77 All Holders of Delay Set Point Operating Licenses Drift (OL) or Construc

tion Permits (CP)

Enclosure 2 Page 1 of 1

NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D. C. 20555

November 14, 1977

IE Bulletin 77-05A


Description of Circumstances

This Bulletin is a supplement to IE Bulletin 77-05, issued November 8, 197-, and the two documents should be considered together..

Bulletin 77-05 described the failure, under test conditions, of a general type of electrical connector. The tests were intended to simulate postLOCA conditions, therefore the action requested in the Bulletin focussed on the qualification of electrical connectors for use inside containment.

Since issuance of Bulletin 77-05, our attention has been called to circumstances which indicate that the scope of the action requested should be expanded and therefore the response to Bulletin 77-05 should reflect the expanded scope.

Electrical connectors should be qualified to perform their intended function after having been subjected to accident conditions if they are contained ina system whose function is to mitigate that accident.

.The location of the connectors that must be qualified is not limited to those inside containment.

Action To Be Taken By Licensees and Permit Holders:

1. Actions requested by Bulletin 77-05 should be expanded to include all connectors in safety systems which are required to function to mitigate an accident where the accident itself could adversely affect the ability of the system to perform its safety function. The examination is not to be limited to only LOCA's nor to areas only

,within containment.

2. Responses should be provided within 30 or 60 days, as appropriate, of the date of this supplement.

Approved by GAO, B180225 (R0072); clearance expires 7-31-80. Approval was given under a blanket clearance specifically for identified generic problems.

NOV 1 7 1977 0RW0 Q SMSheppard

Docket No 5-6 50-270

S- and 50-287

NRC Public Document Room Local Public Document Room

By letter dated October 18, 1977, Duke Power Company provided

a report concerning the seismic investigations performed at the

Jocassee Dam. Because of its bulky.nature, a copy of this report

will not be placed in the PDR's. It can, however, be made available

for inspection and copying upon advadce notice.

Don Neighbors, Project Manager Operating Reactors Branch #1 Division of Operating Reactors

OFFICE* ORB #1

SURNAME~ DNeighbors~njf

DATE / /77 NRC FORM 318 (9-76) NRCM 0240 U) S. GOVERMEiN1j PRINTING OFFICE: 1976 - 426.62A

NOv 1 0 1977

Docket 50-270

and 50-287

Duke Power Company AT1N4 Mr. William 0. Parker, .>

Vice .President - Steam Production 422 South Church Street P. 0. Box 2178 Charlotte, North Carolina 28242

Gentlemen:

We have completed a preliminary review of the overpressure protection system for Oconee 1, 2 and 3. . e have found that the system currently installed does not meet all the .criteriaestablished by the NRC.

Your system maintains a gas blanket or bubble in the pressurizer at all times' in conjunction with a single.. low pressure. setpoint relief valve. This design has certain advantages over other concepts because failure of an operator to manually enable the low pressure setpoint of the relief valve does not totally defeat protection against a pressure transient. We have concluded that,your system adequately accommodates all postulated overpressure transients with the exception of an inadvertent initiation of safety injection by the high pressure injection (HPI) pump.

Based on your analyses, we have-identified high pressure injection as.the limiting-mass addition overpressure transient. Operation of the HI pump, which is capable of delivering flow against full system operating pressure, is required whenever a reactor coolant pump is in operation. Since the discharge of the HPI pump is isolated from the reactor coolant system by a single injection valve, a single error or equipment failure could open the injection valve and initiate a.pressure increase inside the reactor coolant system pressure boundary (RCSPB). If failure of the single low setpoint power operated relief valve is then assumed as the single failure following initiation of the event, your analysis shows that operator action would be required within five minutes to maintain the RCSPB pressure below Appendfx G limits. This-is, otin accordance with NRC criteria which does not allow credit for operator intervention for ten minutes.

OFFICEs

_NRC FORM 318 (9-76) NRM 0240 . - u* s:. GOVERN4MENT PRINTINGE OPPICE: 1976 - 626424

Duke Power Company 2 NOV 1 -1

Since your system does not fullysatisfy our established criteria, we request that you propose system modifications that will provide overpressure protection in full conformance with NRC criteria, and that you provide a value-fimpact assessmaent on schedule and cost to make all necessary hardware changes.

It is also our positiO that to- assu re proper lignmnt of the overpressure protection system during plant cooldown, an enabling alarm must be provided which monitors the system enabling switch and the position of the isolation valve-upstream of the power operated relief valve (PORV).

Your previous submittals do not provide adequate electrical circuit and logic diagrams of the overpressure protection system to permit a thorough review. Please provide the following:

1. RCS overpressure protection.system diagram.

2. Logic diagram.

3. Control circuitry diagram..

4. Instrument loop diagram.

5& ,Annunciator system schematic.

6. Overpressure protection Control display and layout.

In addition to the items discussed above, we have identified several concerns related to PORV maintenance and HPI testing for the currently proposed system. If the reliefvalve requires maintenance the upstream isolation valve would need to be closed thereby removing the single relief valve from service. Therefore we request that you propose Technical Specifications which.stipulate that when the reactor vessel temperature is below the minimum value, for which the vessel can be fully pressurized the PORV may be removed from service for a short period of time only if: (1) chargingpumps are out of service and all HPI injection valves are closed and power removed, or (2) the vessel head is removed. Regarding HPI testing, we request that the HPI valve be allowed to be cycled only if all HPI pumps are out of service, or vessel temperature is above the minimum value for which the vessel can be fully pressurized, or the reactor vessel head Is removed.

OFFICE.. t.. sURNAME.

OA.E* I..... NRC FORM 318 (9-76) NRCM 0240 uS e.-.OVERNMENT PRINTING OFFICRI 1976 - 626*624

Duke Power Company ' - 3 . NOV 1 0 1977

We will require that your TechnicalSpecifications identify the system enabling temperature and the PORV setpoint. In addition, you should propose specifications related to system testing.

These maintenance and testing restrictions .should be examined to assure compatibility with present Technical Specification requirements regarding the operability and periodic testing of ECC and emergency boration systems. Also, since the impact of the proposed Technical Specifications will be considered by.us in determintng the acceptability of the proposed overpressure-mitigating system, you should provide a thorough evaluation of the effect of these.maintenance and testing requirements on the susceptibility of the reactor coolant system to a pressure transient, - .

You should provide the above requested information and the proposed Technical .Specifications within 45 days of receipt of this letter. For the interim period prior toInstallation of a system which meets all the NRC criteria, you should continue to use. the overprotection mieasures of your installed system. We realize that although your -oerpressure mitigating system does.not meet the NRC criteria, It will provide adequate protection for.theAnderim period.

Sincerely,

A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactor$

cc: See next page

DISTRIBUTION: Dockets DEisenhut NRC PDR TBAbernathy Local PDR JRBuchanan ORB#1 Rdg ACRS (16) VStello File KRGoller OELD. OI&E (3) ASchwencer DNeighbors SSheppard

OFIE ORB ~ ORB#~

DATE/77 11 1 / NRC FORM 318 (9-76) NRCM 0240 u3 i. GOVERNMENT PRINTING OFFICEZ 1976- 026-624

Duke Power Company -.4 - November 10, 1977

cc: Mr. William L. Porter Duke Power Company P. 0. Box 2178 422 South Church Street Charlotte, North Carolina 28242


Oconee Public Library 201 South Spring Street Walhalla, South Carolina 29691

NOV 1 0 1977

Docket No. 50-269


Vice President - Steam Production 422 South Church Street P. 0 13ox 2178 Charlotte, North Carolina 28242.,.

Gentlemen:

It has recently come to our attention that one of the reporting requirements in the Oconee Nuclear Station Technical Specifications is not consistent with this reporting requirement as it is specified in the Technical Specifications in force at other facilities nor is it in conformance.with NRC intentions. in this regard.

Specifically, for events requiring prompt notification to the NRC, the Oconee Technical Specificationsas presently worded would permit this prompt notification to be by a.choice of either telephone, telegraph, mailgram, or facimile transmission. .A.It is our intent that the prompt reporting of such events should be done promptly both orally by telephone and in writing by telega0h, mailgram or facimile transmission.

The lack of a requirement to promptly document such reportable occurrences creates what we consider to be an unacceptable inconsistency in reporting requirements and record keeping and could hamper the NRC's ability to rapidly and accurately disseminate notification of such events.

We have enclosed a page from another facility's Technical Specifications which provides you an example of thewording you should use in your Technical Specifications.

0 FF ICE 9

NRC FORM 318 (9-76) NRCM 0240~ *u. s. GOVERNMENT PRINTING OPPICE: 1976 - 626-624

iDuke Power Company -NOV 1 0.1977

Accordingly, within 15 days of receipt of this letter,,you should request an amendment to the Oconee Technical Specifications to change the reporting requirements on prompt notiflaation with written followup to be consistent withthe staffs lntentf as shown in the enclosure.

Sincerely,

A. Schwencer, Chief Operating Reactors branch #1 Division of Operating Reactors

Enclosure: Sample Technical Specification

cc w/enclosure: See next page

DISTRIBUTION: Docket NRC PDR Local PDR ORB#l Rdg VStello KRGoller ASchwencer DNeighbors SSheppard OELD OI&E(3) DEisenhut TBAbernathy JRBuchanan ACRS (16) File

OFFICE) ORB#1

SURAM~~ DNei hors ar ASchwen e SU R N A M E ........ ... ................ ................................... ....................... ........................... .......................

DATE~ 1 _1/ c ') l /77 1__o__/77 NRC FORM 318 (9-76) NRCM 0240 U3 S. GOVERNMENT PRINTING OFFICEU 1976 - 626-62A

00 4UNITED STATES / NUCLEAR REGULATORY COMMISSION

% WASHINGTON, D. C. 20555

November 10, 1977

Docket No. 50-269


Vice President - Steam Production 422 South Church Street P. 0. Box 2178 Charlotte, North Carolina 28242

Gentlemen:

It has recently come to our attention that one of the reporting requirements in the Oconee Nuclear Station Technical Specifications is not consistent with this reporting requirement as it is specified in the Technical Specifications in force at other facilities nor is it in conformance with NRC intentions in this regard.

Specifically, for events requiring prompt notification to the NRC, the Oconee Technical Specifications as presently worded would permit this prompt notification to be by a choice of either telephone, telegraph, mailgram, or facimile transmission. It is our intent that the prompt reporting of such events should be done promptly both orally by telephone and in writing by telegraph, mailgram or-facimile transmission.

The lack of a requirement to promptly document such reportable occurrences creates what we consider to be an unacceptable inconsistency in reporting requirement; and record keeping and could hamper the NRC's ability to rapidly and accurately disseminate notification of such events.

We have enclosed a page from another facility's Technical Specifications which provides you an example of the wording you should use in your Technical Specifications.

Duke Power Company 2 - November 10, 1977

Accordingly, within 15 days of receipt of this letter, you should request an amendment to the Oconee Technical Specifications to change the reporting requirements on prompt notification with written followup to be consistent with the staff's intent as shown in the enclosure.

Sincerely,

A. Schwencer, Chief Operating Reactors Branch #1. Division of Operating Reactors

Enclosure: Sample Technical Specification

cc w/enclosure: See next page

Duke Power Company - 3 - November 10, 1977

cc: Mr. William L. Porter Duke Power Company P., 0. Box 2178 422 South Church Street Charlotte, North Carolina 28242


Oconee Public Library 201 South Spring Street Walhalla, South-Carolina .29691

James P. O'Reilly, Director Nuclear Regulatory Commission, Region II

Office of Inspection and Enforcement 230 Peachtree Street, N.W., Suite 1217

Atlanta, Georgia 30303

a. Prompt Notification With Written Followup. The types of

events listed below shall he reported as expeditiously as

possible, but within 24 hours by telephone and confirmed by

telegraph, mailgram, or facsimile transmission to the Director of the appropriate Regional Office, or his designate no later than the first working day following the event, with a written followue report within two weeks. The writ

ten followun report shall include, as a minimum, a completed

copy of a licensee event report form. Information provided

on the licensee event report form shall be sunplemented, as needed, by additional narrative material to provide com

plete explanation of the circumstances surrounding the event.

(1) Failure of the reactor protection system or other

systems subject to limiting safety system settings to initiate the reouired protective function by the time a monitored paramater reaches the setpoint snecified as the limiting safety system setting in the technical specifications or failure to complete the required

protective function.

Note: Instrument drift discovered as a result of testing need not be reported under this item but may be reportable under items 6.9.2.a(5), 6.9.2.a(6), or 6.9.2.b(1) below.

(2) Operation of the unit or affected systems when any

parameter or operation subject to a limiting condi

tion is less conservative than the least conservative

aspect of the limiting condition for operation established in the technical specifications.

Note: If specified action is taken when a system is found

to be operating between the most conservative and

the least conservative aspects of a limiting condition for operation listed in the technical specifications, the limiting condition for operation is not considered to have been violated and need not be reported under this item, but it may be reportable under item 6.9.2.b(2)

(3) Abnormal degradation discovered in fuel cladding,

reactor. coolant. pressure boundary, or primaryv-contain

ment. . .

Note: Leakage of valve vacking or gaskets within the limits

for identified leakage set forth-in technical.specifications need not be reported under this item.

Amendment No. 5 Change No. 7

TS 6-16 12/18/75

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