ML20202J835

Proposed Tech Specs,Incorporating Reporting Requirements of 10CFR50.72 & 50.73,per Generic Ltr 83-43

ML20202J835
Person / Time
Site:	Fort Calhoun
Issue date:	07/08/1986
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20202J806	List: ML20202J801 ML20202J835
References
GL-83-43, TAC-53219, NUDOCS 8607170154
Download: ML20202J835 (17)

Text

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6' TABLE OF CONTENTS (Continued)

Pace 4.3 Nuclear Steam Supply System (NSSS) 4-3 i

4.3.1 Reactor Coolant System 4-3 4.3.2 Reactor Core and Control 4-3 4.3.3 Emergency Core Cooling 4-3 i

j 4.4 Fuel Storage 4-4 4.4.1 New Fuel Storage 4-4 4.4.2 Spent Fuel Storage 4-4 4.5 Seismic Design for Class I Systems 4-5 ij 5.0 ADMINISTRATIVE CONTROLS 5-1 i

i 5.1 Responsibility 5-1 i

5.2 Organization 5-1 5.3 Facility Staff Qualifications 5-la

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5.4 Training 5-3 ll 5.5 Review and Audit 5-3 i i 5.5.1 Plant Review Committee (PRC) 5-3 ll 5.5.2 Safety Audit and Review Committee (SARC) 5-5 5.5.3 Fire Protection Inspection 5-8a lf 5.6 Reportable Event Action 5-9 l

5.7 Safety Limit Violation 5-9 i

5.8 Procedures 5-9 5.9 Reporting Requirements 5-10 t

5.9.1 Routine Reports 5-10 l

5.9.2 Reportable Events 5-12 l

5.9.3 Special Reports 5-15 5.9.4 Unique Reporting Requirements 5-15 5.10 Records Retention 5-18

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5.11 Radiation Protection Program 5-19

<t 5.12 Deleted 5.13 Secondary Water Chemistry 5-20 5.14 Systems Integrity 5-21 5.15 Post Accident Radiological Sampling and Monitoring.

5-21 i

l 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS 6-1 i

t 6.1 Limits on Reactor Coolant Pump operation 6-1 i

6.2 Use of a Spent Fuel Shipping Cask 6-1 1

6.3 Auxiliary Feedwater Automatic Initiation Setpoint..

6-1 l

6.4 Operation With Less Than 75% of Incore Detector Strings Operable 6-1 c

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i iii Amendment No.JJ,J/,f),

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3.0 SURVEILLANCE RE0VIREl4ENTS t

3.3 Iteactor Coolant System, steam Generator Tubes, and Other Components Subject to ASNE XI foiler & Pressure Vessel Code Inspection and Testing Surveillance (Continued)

Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top l

support of the cold leg.

i (ii) The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all t

tu'es containing through-wall cracks) required by l

Table 3-8.

c.

Reporting Requirements (1) Following each in-service inspection of steam j

generator tubes, the number of tubes plugged in each steam generator shall be reported to the Comission within 30 days.

(ii) The complete results of the steam generator tube inservice inspection shall be reported to the i

Commission within six (6) months following com-l pletion of the inspection. This report shall include:

1.

flumber and extent of tubes inspected.

f' 2.

Location and percent of wall thickness penetration for each imperfection.

3.

Identification of tubes plugged, t

(iii) Results of steam generator tube inspections which

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fall into Category C-3 require prompt notification i

of the Commission and shall be reported pursuant

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to Section 5.6 of the Technical Specifications i

prior to resumption of plant operation. The j

written followup of this report shall provide a description of investigations conductec to deter-1 mine cause of the tube degradation and corrective measures taken to prevent recurrence.

i (3) Surveillance of Reactor Coolant System Pressure Isolation Valves i

a.

Periodic leakage testing

• on each valve listed in Table 2-9 l

shall be accomplished prior to entering the power operation l

To satisfy ALARA reoutrements, Icakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with 1

approved procedures and supported by computations showing that the method ji is capable of demonstrating valve compliance with the leakage criteria.

Ll Amendment flo. 4, Order dated 4/20/81 3-25

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STEAM GENERATOR TUBE INSPECTION 2

E 1st SAMPtE IriSPECTION I

2ND SAMPLE INSPECT 10t!

3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required 2

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A cinimum of C-1 None N/A N/A N/A N/A 3

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• 300 tubes l

per S.G.

C-2 Plug defective tebes C-1

!!one N/A N/A and inspect addi-Plug defective tubes C-1 Hone l

tional 600 tubes in C-2 and inspect additional C-2 Plug defective tubes this S.G.

1200 tubes in this Perform action for S.G.

C-3 C-3 result of first l

-L sample l

Perform action for C-3 C-3 result of first N/A N/A l

sample 2l C-3 Inspect all tubes in The this S.G., plug de-second fective tubes and S.G.

None N/A N/A inspect 600 tub s is C-1 9

in other S.G.

The Perfnrm action for N/A f/A second C-2 result of second Prompt notification S.G.

sample to NRC pursuant to is C-2 Tech Spec. 5.6 The Inspect all tubes in l

second the second S.G. and S.G.

plug defective tubes.

is C-3 Prompt notification N/A N/A to NRC pursuant to Tech Spec'5.6 l

N/A = Not Applicable

d.

3.0 SURVEILLANCE REQUIREMENTS 3.3 Reactor Coolant System. Steam Generator Tubes, and Other Compon'ents Subject to ASME XI Boiler & Pressure Vessel Code Inspection and Testing Surveillance (Continued) i Basis Undetected prolonged leakage of borated reactor coolant onto carbon. steel l

sets up an unusual corrosion mechanism.

Detection of this leakage at an j

early stage can dest be accommodated directly after an outage and before startup.

The inspection program specified in Specification 3.3(1) places major emphasis on the areas of highest stress concentration as determined by general design evaluation and experience with similar systems.' The inspections will rely on non-destructive analysis methods utiliz,ing up-

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to-date analyzing equipment and trained personnel.

Volumetric i'nspection of the reactor pressure vessel is to be performed completely from the I

outside diameter. The testing techniques and acceptance criteria of l

Section XI of the ASME B&PV Code will be utilized, except where specific relief is granted by the Commission.

The surveillance requirements for inspection of the stehm generator tubes ensure that the structural integrity of this portion of the RCS will be l

maintained.

The program for in-service -inspection of the steam generator l

tubes is based on a modification of Regulatory Guide 1.83, Revision 1, dated July 1975.

In-service inspection of steam generator tubing is i

i essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive j

degradation due to design, manufacturing errors, or in-service conditions that lead to corrosion.

In-service inspection of steam generator tubing also provides a means of I

characterizing the nature and cause of any tube degradation so that i

corrective measures can be taken.

Wastage-type defects are unlikely with proper chemistry treat. ment of the secondary coolant.

However, even if a defect should develop in service, it l will be found during scheduled in-service steam generator tube examina-tions.

Plugging will be recuired for all tubes with imperfections exceed-ing the plugging limit of 40% of the tube nominal wall thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing in-service inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Section S.6 of the Technical' Specifications prior to l the resumption of plant operation.

Such cases will be considered by the 3

Commission on a case-by-case basis and may result in a requirement for

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analysis, laboratory examinations, tests, additional cddy-current inspec-tien, and revision of the Technical Specifications, if necessary.

References i

l (1) USAR, Section 4.5.3 Amendrrent No. /$, Order dated 4/20/81 3-29 L

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.5.O ADMINISTRATIVE CONTROLS 1

5.5 Review-and Audit (continued) 5.5.1 Plant Review Committee (PRC) (continued) l Resoonsibilities

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5.5.1.6 The Plant Review Committee shall be responsible for:

1 a.

Review of 1) all procedures required by Specification 5.8 and changes thereto, 2) any other proposed j!

procedures or changes therto as determined by the i

Manager - Fort Calhoun Station to affect nuclear safety.

l' b.

Review of all. proposed tests and experiments that affect nuclear safety.

]l c.

Review of all proposed changes to the Technical Specifications.

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d.

Review of all proposed changes or modifications to plant jj systems or equipment that affect nuclear safety.

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Investigation of all violations of the Technical Specifications and shall prepara and forward a report covering evaluation and and recommendations to prevent recurrence to the Division Manger - Nuclear Production 1

and to the Chariman of the Safety Audit and Review Committee.

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f.

Review of facility operations to detect potential safety hazards.

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!j' g.

Performance of special reviews and investigations and reports thereon as requested by the Chairman of the j;

Safety Audit and Review Committee.

1 i h.

Review of the Site Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the Safety Audit and Review Committee.

i.

Review of the Site Emergency Plan and implementing l<

procedures and shall submit recommended changes to the il Chairman of the Safety Audit.and Raview Committee.

i j.

Review of all Reportable Events.

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5.5.1.7 The Plant Review Committee shall:

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I a.

Recommend in writing to the Manager - Fort Calhoun Station approval or disapproval of items considered

'e under 5.5.1.6(a) through (d) above.-

5-4 Amendment No. 9, Jp, l

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5.0 ADMINISTRATIVE CONTROLS 5.5 Review and Audit (continued) 5.5.1 Elant Review Committee (PRC) (continued) 5.5.1.7 b.

Render determinations in writing with regard to whether or not each item considered under 5.5.1.6(a) through (e) above i

constitutes an unreviewed safety question.

I c.

Provide immediate written notification to the Division Manager -

Nuclear Production and the Safety Audit and Review Committee of 4

disagreement between the Plant Review Comittee and the Manager -

l Fort Calhoun Station; however, the Manager - Fort Calhoun Station shall have responsibility for resolution of such disagreements pursuant to 5.1.1 above.

Records 5.5.1.8 The Plant Review Comidttee shall maintain written minutes of each meeting and copies shall be provided to the Division Manager - Nuclear Production and Chairman of the Safety Audit and Review Comittee.

5.5.2 Safety Audit and Review Comittee (SARC)

Function 5.5.2.1 The Safety Audit and Review Comittee shall function to provide the independent review and audit of designated activities in the areas of:

i a.

nuclear power plant operation b.

nuclear engineering c.

chemistry and radiochemistry l

d.

metallurgy i

e.

instrumentation and control f.

radiological safety g.

mechanical and electrical engineering i

h.

quality assurance Composition j4'!

5.5.22 Th Safety Audit and Review Comittee shall be composed of:

i Chairman:

Division Manager - Quality Assurance and Regulatory Affairs-Member:

Vice President - Nuclear Production, Production 1i Operations, Fuels, and QA & RA.

Member:

Vice President - Engineering and General Services Member:

Divisica Manager - Engineering Member:

Division Manager - Nuclear Production

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Member:

OPPD Operations, Engineering, and Technical Support staff Member:

Qualified Non-District Affiliated Consultants as Required

j and as Determined by SARC Chairman I

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5-5 Amendment No. J, JJ,

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5.0 ADMINISTRATIVE CONTROLS 5.5 Review and Audit (cantinuCd) 5.5.2 Safety Audit and Review Committee (SARC) (continued) 5.5.2.7 c.

Proposed test or experiments which involve an unreviewed f

safety question as defined in section 50.59, 10 CFR.

I d.

Proposed changes in Technical Specifications or l

licenses.

I e.

Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, i

or of internal procedure or instructions having nuclear i

safety significance.

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f.

Significant operating abnormalities or deviations from i

normal and expected performance of plant equipment that affect nuclear safety.

f g.

All Reportable Eyents.

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h.

Any indication of an unanticipated deficiency in some aspect of design or operation of safety related struc-tures, systems, or components.

i' i.

Reports and meeting minutes of the Plant Review j

Committee.

i The Chairman of the Safety Audit and Review Committee (SARC) i!

may designate subgroups, special working committees, or au-i1 dit teams as he deems necessary in order to carry out the responsibilities of the SARC.

These subgroups, committees, l

or audit teams will perform the SARC responsibilities and report on their activities for review at the next regularly j

scheduled SARC meeting following any groups's action.

Audit 5.5.2.8 Audits of facility activities shall be performed under the cognizance of the Safety Audit and Review Committee.

These audits shall encompass:

a.

The conformance of facility operation to all provisions contained within the Technical Specifications and applicable license conditions at least once per year.

b.

The performance, training and qualifications of the entire facility staff at least once per year.

c.

The results of all actions taken to correct deficiencies occuring in facility equipment, structures, systems or method of operation that affect nuclear safety at least i

once per six months.

d.

The performance of all activities required by the Qual-ity Assurance Program to meet the criteria of Appendix B,

10 CFR 50, at least once per two years.

l 5-7 Amendment No. 60

S.

5.0 ADMINISTRATIVE CONTROLS 5.6 ReDortable Event Action 5.6.1 The following actions shall be taken in the event of a REPORTABLE EVENT:

a.

The Commission shall be notified pursuant to the requirements of 10 CFR 50.72, if applicable.

b.

Each Reportable Event shall be reviewed by the Plant Review Committee and submitted to the Chairman of the Safety Audit and Review Committee and the Division Manger - Nuclear Production.

c.

Submit reports of Reportable Events pursuant to the requirements of Specification 5.9.2.

5.7 Safety Limit Violation 5.7.1 The following actiong shall be taken in the event a Safety Limit is violated:

a.

The unit shall be placed in at least HOT SHUTDOWN within one hour.

Ij b.

The safety Limit violation shall be reported to the Division Manager - Nuclear Production and to the Chair-4 man of the Safety Audit and Review Committee (SARC) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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c.

A Safety Limit Violation Report shall be prepared.

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report shall be reviewed by the Plant Review Committee.

i This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence.

d.

The Safety Limit Violation Report shall be submitted to the Chairman of the Safety Audit and Review Committee and the Division Manager - Nuclear Production within 14 days of the violation.

5.8 Procedures j

l 5.8.1 Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed j

the minimum requirements of sections 5.1, and 5.3 of ANSI N18.7-1972 and Appendix A of USNRC Regulatory Guide 1.33 except as provided in 5.8.2 and 5.8.3 below.

I 5.8.2 Each procedure and administrative policy of 5.8.1 above, and changes thereto, shall be reviewed by the Plant Review Com-mittee and approved by the Manager - Fort Calhoun Station prior to implementation and periodically as set forth in

.'j each document.

i 5.8.3 Temporary changes to procedures 5.8.1 above may be made provided:

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5-9 Amendment No. p, JJ, JJ, 84

5.0 ADMINISTRATIVE CONTROLS 5.9 Reportina Reouirements (Continued) 5.9.1 Routine Reports (Continued) work and job functions, E e.g., reactor operations and surveil-lance, inservice inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling outages.

The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

c.

Monthly Ooeratino Report - Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Direc-tor, Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC, 20555, with a copy to the appropriate Regional Office, no leter than the fifteenth of each month following the calendar month covered by the report. This monthly report shall also include a statement regarding any challenges or failures to the pressurizer power operated relief valves or safety valves occurring during the subject month.

5.9.2 Reportable Events A Licensee Event Report (LER) shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC, 20555, with a copy to Region IV of the NRC, within 30 days after discovery of any event meeting the requirements of 10 CFR 50.73.

YThis tabuk$ tion supplements the requirements of Paragraph 20.407 of CFR Part 20.

5-12 Amendment No. 9, 24, 35, 85

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l 5.0 ADMINISTRATIVE CONTROLS 5.10 Record Retention 5.10.1 The following records shall be retained for at least five years:

I a.

Records and logs of facility operation covering time

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interval at each power level.

4 b.

Records and logs of principal maintenance activities, l

inspections, repair and replacement of principal items j

of equipment related to nuclear safety.

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c.

Licensee Events Reports (LER).

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d.

Records of surveillance activities, inspections and cali-brations required by these Technical Specifications.

e.

Records of reactor tests and experiments.

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f.

Records of changes made to Operating Procedures.

g.

Records of radioactive shipments.

h.

Records of annual physical inventory of all source material of record.

l 5.10.2 The following records shall be retained for the duration of the Facility Operating License:

Records of drawing changes reflecting facility design a.

modifications made to systems and equipment described in the Final Safety Analysis Report.

i b.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

c.

Records of facility radiation and contamination surveys.

d.

Records of radiation exposure for all individuals entering radiation control areas.

e.

Records of gaseous and liquid radioactive material released to the environs, f.

Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.

g.

Records of training and qualification for current members of the plant staff.

5-18 Amendment No. 75

l 5.0 ADMINISTRATIVE CONTROLS 5.10 Record Retention (Continued) 5.10.2 Continued h.

Records of in-service inspections performed pursuant to these Technical Specifications.

1.

Records of Quality Assurance activities required by the QA Manual.

j. Records of reviews performed for changes made to procedures or equipment or reviews of tests, and experiments pursuant to 10 CFR 50.59.

k.

Records of meetings of the Plant Review Committee and the Safety i

Audit and Review Committee.

1.

Records of Environmental Qualification of Electric Equipment pursuant to 10 CFR 50.49.

m.

Records of the service '.ives of all hydraulic and mechanical snubbers listed on Table 2-6(a) and (b) including the date at which the service life commences and associated installation and maintenance records.

n.

Records of analyses required by the Radiological Environmental Monitoring Program.

5.10.3 A complete record of the analysis employed in the selection of any fuel assembly to be placed in Region 2 of the spent fuel racks will be 4

retained as long as that bundle remains in Region 2 (reference Technical Specifications 2.8(12) and S.8.4).

5.11 Radiation Protection Proaram Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

5-19 Sfpg//ggggg/J9//g/ES Amendment No. JJ, SJ, 93

DISCUSSION, JUSTIFICATION, AND SIGNIFICANT HAZARDS CONSIDERATIONS This application proposes to amend the Fort Calhoun Station Technical Speci-fications to provide conformance with 10 CFR 50.72 and 10 CFR 50.73. The changes are proposed in response to Generic letter 83-43, " Reporting Re-quirements of 10 CFR 50, Sections 50.72 and 50.73, and Standard Technical Specifications." These changes are in a format similar to the recommendations of Generic Letter 83-43.

Included are changes to the Table of Contents and to Sections 3.0 and 5.0 of the Fort Calhoun Station Technical Specifications to ensure continuity of reporting requirements.

The Table of Contents has been change to reflect the section title changes for sections 5.6 and 5.9.2.

Section 5.6 is changed to " Reportable Event Action" from Reportable Occurrence Action" and section 5.9.2 is changed to " Reportable Events" from Reportable Occurrences". These changes are consistent with the recommendations of Generic Letter 83-43.

Section 3.3 Specification (2)e(iii) has been changed to reference Section 5.6 for reportability as opposed to referencing Section 5.9.2.

The same change has been made in Table 3-8, Section C-3 and the Basis for Specification 3.3(2)e(iii). These changes will ensure consistency of reporting requirement references within the Technical Specifications.

Section 5.5.1.6, Plant Review Committee Responsibilities, has been changed by addition of item J., which requires review of all Reportable Events. This change is consistent with the recommendations of Generic Letter 83-43.

Section 5.5.2.7, Safety Audit and Review Committee (SARC) review responsi-bilities, has been changed by rewording item g, in accordance with the recommendations of Generic Letter 83-43. The SARC will review all reportable events rather than only those events requiring NRC notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Changes to Section 5.6 of the Technical Specifications are proposed to ensure compliance with the reporting requirements of 10 CFR 50.72 and reflect termin-ology changes consistent with the noted section of 10 CFR 50. The change to Technical Specification 5.6.la provides for notification of the Commission pursuant to 10 CFR 50.72, when applicable. Specification 5.6.lb is changed to delete reference to 24-hour notifications and for consistency with terminology used elsewhere in the Technical Specifications. A new specification, 5.6.lc, is added to reference Specification 5.9.2, regarding Licensee Event Reports.

Specifications in Section 5.7 of the Fort Calhoun Station Technical Specifica-tion are being changed to provide specific instructions to be taken in the event of a safety limit violation. The proposed changes will provide for placement of the plant in Hot Shutdown within one (1) hour.

Hot Shutdown is equivalent to the CE Standard Technical Specification Hot Standby. Therefore, this change is consistent with the CE-STS. The requirement to make a 14-day written report to the NRC has been deleted.

10 CFR 50.36(c)(7) no longer re-quires this report.

ATTACHMENT B

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Section 5.9.lc, Monthly Operating Report, is changed to read "... Director, Office of Inspection and Enforcement,..." from ".. 0ffice of Management and Program Analysis,...".

This change is consistent with to whom the monthly reports are currently submitted.

In addition, the words "to arrive" have been deleted from the statement regarding submittal of the report. OPPD cannot reasonably assure arrival of a document by any date. This change is also consistent with Standard Technical Specifications.

Specifications in Section 5.9.2 of these Technical Specifications have been changed to reference 10 CFR 50.73. The conditions under this section, which formerly described when an LER was required, have been removed. They are superseded by the requirements of the regulations.

Section 5.10.lc has been changed to " Licensee Event Reports (LER)" from

" Reportable Occurrence Reports". This change is consistent with recommenda-tions of Generic Letter 83-43.

The Section 5.10.2 which appeared on page 5-18 has been renumbered as 5.10.3 and has been moved to its appropriate place. The current specification con-tained two sections, both numbered 5.10.2.

Sioriificant Hazards Condisderations 1.

This proposed amendment will not involve an increase in the probability or consequence of an accident previously evaluated. The proposed changes concerning Licensee Event and Significant Event reporting are administra-tive in nature and do not affect the surveillance or operability of any system which functions to prevent or mitigate the consequences of a pre-viously analyzed accident.

2.

The proposed amendment will not create the possibility of a new or dif-ferent accident than previously evaluated. The proposed changes to the reporting requirements are administrative in% ature. The proposed changes do not affect the design or surveillance and operability requirements of the plant systems.

3.

This proposed amendment does not involve a reduction in a margin of safety.

The proposed changes are administrative in nature and do not affect any de-sign, surveillance, or operability requirements which would have an effect on safety margins.

ATTACHMENT B l

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