ML18152A574

Proposed Tech Specs Reflecting New LER Sys,Per 10CFR50.72 & 50.73

ML18152A574
Person / Time
Site:	Surry
Issue date:	11/30/1984
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18152A573	List: ML18152A572 ML18152A574
References
NUDOCS 8412120013
Download: ML18152A574 (21)

Text

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.e ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGE

~riR12120013.. 841130-p ADOCK 050002SO PDR

I.

TS 1. 0-5 for operational activities provided that they are under adminis-trative control and are capable of being closed immediately if required.

2.

Blind flanges are installed where required.

3.

The equipment access hatch is properly closed and sealed.

4.

At least one door in the personnel air lock is properly closed and sealed.

5.

All automatic containment isolation valves are operable or are locked closed under administrative control.

6.

The uncontrolled containment leakage satisfied Specification 4. 4.

Reportable Event A Reportable Event shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

Basis TS 3.1-24

b.

With both PORV's inoperable, depressurize the RCS within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> unless Specification 3.1.G.l.b.(4) is in effect.

When the RCS has been depressurized, open one PORV or establish the conditions listed below.

Maintain the RCS depressurized until both PORV's have been restored to operable status.

(1)

A maximum pressurizer narrow range level of 33%.

(2)

The series RHR inlet valves open and their re-spective breakers locked open or an alternate letdown path operable.

(3)_

Limit charging flow to < 150 gpm.

(4)

Safety Injection accumulator discharge valves closed and their respective breakers locked open.

c.

When the conditions noted in 3. 1. G. 2. b. ( 1) through 3.1.G.2.b. (4) above are required to be established, their implementation shall be verified at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3.

In the event that the Reactor Coolant System Overpressure Mitigating System is used to mitigate a RCS pressure transient, a Special Report shall be prepared and submit-ted to the Commission pursuant to Specification 6.6 within 30 days.

The report shall describe the circum-stances initiating the transient, the effect of the mitigating system or the administrative controls on the transient and any corrective actions necessary to prevent recurrence.

The operability of two PORV' s or the RCS vented through an opened PORV ensures that the Reactor Vessel will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when the Reactor Coolant average temperature is~ 350°F and the Reactor Vessel Head is bolted.

When the Reactor Coolant average temperature is > 350°F, overpressure protection is provided by a bubble in the pressurizer and/or pressurizer safety valves.

A single PORV has adequate relieving

TABLE 3.7-6 ACCIDENT MONITORING INSTRUMENTATION

1.

2.

3.

4.

5.

6.

7.

8.

9.

INSTRUMENT Auxiliary Feedwater Flow Rate Reactor Coolant System Subcooling Margin Monitor PORV Position Indicator (Primary Detector)

PORV Position Indicator (Backup Detector)

PORV Block Valve Position Indicator Safety Valve Position Indicator (Primary Detector)

Safety Valve Position Indicator (Backup Detector)

Reactor Vessel Coolant Level Monitor Containment Pressure

10.

Containment Water Level (Narrow Range)

11.

Containment Water Level (Wide Range)

TOTAL NO.

OF CHANNELS 1 per S/G 2

1/valve 1/valve 1/valve 1/valve I/valve MINIMUM CHANNELS OPERABLE 1 per S/G 1

1/valve 0

1/valve 1/valve 0

1 1

1 1

12.

Contaiment High Range Radiation Monitor 2

2 2

2 2

2 2

3 1 (Note 1, band c only)

13.

Process Vent High Range Effluent Monitor

14.

Ventilation Vent High Range Effluent Monitor

15.

Main Steam High Range Radiation Monitors (Units 1 and 2) 2 2

3 (Note 1, (Note 1, (Note 1, a, b, and c) a, b, and c) a, b, and c)

16.

Aux. Feed Pump Steam Turbine Exhaust Radiation Monitor 1

1 (Note 1, a, b, and c)

Note 1: With the number of operable channels less than.required by the Minimum Channels Operable requirements

a. Initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

b. Either restore the inoperable channel to operable status within 7 days of the event, or

c. Prepare and submit a Special Report to the commission pursuant to specification 6.6 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to ~perable.

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e TS 3.12-7

a.

The hot channel factors shall be determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the power level adjusted to meet the require-ment of Specification 3.12.B.1, or

b.

If the hot channel factors are not determined within two hours, the power level and high neutron flux trip setpoint shall be reduced from rated power 2% for each percent of quadrant tilt.

c.

If the quadrant to average power tilt exceeds+/- 10%, the power level and high neutron flux trip setpoint will be reduced from rated power 2% for each percent of quadrant tilt.

7.

If, except for physics and rod exercise testing, after a further period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the power tilt in Specification 3.12.B.5 above is not corrected to less than 2%:

a.

If design hot channel factors for rated power are not exceeded, an evaluation as to the cause of the discrepancy shall be made and a special report issued to the Nuclear Regulatory Commission.

b.

If the design hot channel factors for rated power are exceeded and the power is > 10%, the Nuclear Regulatory Commission shall be notified and the Nuclear Overpower, Nuclear Overpower ~T, and Overtemperature ~T trips shall be reduced I% for each percent the hot channel factor exceeds the*rated power design values.

c.

If the hot channel factors are not determined, the Nuclear Regulatory Commission shall be notified and the Overpower

e e

TS 4;19-8 F.

Reports

a.

Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.

b.

The complete results of the steam generator tube inservice inspection shall be reported on an annual basis for the period in which the inspection was completed.

This report shall include:

1.

Number and extent of tubes inspected.

2.

Location and percent of wall-thickness penetration for each indication of an imperfection.

3.

Identification of tubes plugged.

c.

Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported by special report prior to resumption of plant operation.

The report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

e TS 4.19-10 withstand the loads imposed during normal operation and by postulated accidents.

Operating plants have demons.trated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be de-tected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant.

However, even if a defect of similar type should develop inservice, it will be found during scheduled inservice steam generator tube examination.

Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.19.E.a, if 40% of the tube nominal wall thickness.

Steam generator tube inspections of operating plants have demonstrated the capability of reliably detecting degradation that has penetrated 20%

of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these r.esults will be reported to the Commission by special report prior to resumption of plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations tests, additional eddy current inspection, and revision of the Technical Speci-fication, if necessary.

TABLE 4.19-2 STEAM GENERATOR TUBE INSPECTION 1st SAMPLE INSPECTION 2nd SAMPLE INSPECTION 3rd SAMPLE INSPECTION Sample Size Result Action Required Result iAction Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A S Tubes per S.G.

C-2 Plug defective tubes C-1 None N/A N/A and inspect additional C-2 Plug defective C-1 None 2S tubes in this S.G.

tubes and C-2 Plug defective inspect a'ddi t-tubes 4S tubes in this S.G.

C-3 Perform action for C-3 result of first sample C-3 Perform action for N/A N/A C-3 result of first sample C-3 Inspect all tubes in All other None N/A N/A this S.G., plug defec-S.G.s are tive tubes & inspect C-1 2S tubes in each other S.G.

Some S.G.s Perform action for N/A N/A Special Report C-2 but no C-2 result of additional second sample S.G. are C-3 Additional Inspect all tubes S.G. is C-3 in each S.G. and plug defective N/A N/A tubes Special Report Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection I

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TS 6.1-7

f.

Responsibilities The SNSOC shall be responsible for:

(1)

Review of a) all proposed normal, abnormal, and emergency operating procedures and all proposed maintenance procedures and changes

thereto, b) any other proposed procedures or changes thereto as determined by the Station Manager which affect nuclear safety.

(2)

Review of all proposed test and experiment pro-cedures that affect nuclear safety.

(3)

Review of proposed changes to Technical Specifi-cations.

(4)

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

(5)

Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Manager-Nuclear Operations and Maintenance, and to the Director-Safety Evaluation and Control.

(6)

Review of all Reportable Events and special reports submitted to the NRC.

(7)

Review of facility operations to detect poten-tial nuclear safety hazards.

(8)

Performance of special reviews, investigations or analyses and report thereon as requested by the Chairman of the Station Nuclear Safety and Operating Committee.

e TS 6.1-11 (3)

Changes in the Technical Specifications or license amend-ments relating to nuclear safety prior to implementation except in those cases where the change is identical to a previously reviewed proposed change.

(4)

Violations and Reportable Events such as:

(a)

Violations of applicable codes, regulations, order, Technical Specifications, license requirements or internal procedures or instructions having safety significance; (b)

Significant operating abnormalities or deviations from normal or expected performance of station safety-related structures,

systems, or components; and (c)

All Reportable Events.

Review of events covered under this paragraph shall include the results of any investigations made and the recommen-dations resulting from such investigations to prevent or reduce the probability of recurrence of the event.

(5)

The Quality Assurance audit program at least once per 12 months and audit reports.

e 6.2 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE EVENT IN STATION OPERATION Specification A.

The following actions shall be taken for Reportable Events:

TS 6.2-1

1.

A report shall be submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

2.

Each Reportable Event shall be reviewed by the SNSOC and sub-mitted to the Director - Safety Evaluation and Control and the Vice President-Nuclear Operations.

TS 6.5-1 6.5 STATION OPERATING RECORDS Specification A.

Records and logs relative to the following items shall be retained for 5 years, unless a longer period is required by applicable regula-tions.

1.

Records of normal plant operation, including power levels and periods of operation at each power level.

2.

Records of principle maintenance activities, including inspec-tion repair, substitution, or replacement of principle items of equipment pertaining to nuclear safety.

3.

Record of all Reportable Events.

4.

Record of periodic checks, inspections, and calibrations per-formed to verify that surveillance requirements are being met.

5.

Records of any special reactor test or experiments pursuant to IO CFR 50.59.

6.

Records of changes made in the Operating Procedures pursuant to 10 CFR 5 0. 5 9.

7.

Records of shipment of radioactive material.

8.

Records of leakage testing of miscellaneous radioactive source test results, in units or microcuires, for leak tests performed pursuant to Technical Specification 4.16.

TS 6.6-1 6.6 STATION REPORTING REQUIREMENTS In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforcement unless otherwise noted.

A.

Routine Reports

1.

Startup Report A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any addditional specific details required in license conditions based on other commitments shall be included in this report.

Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following r

t

TS 6.6-2 resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earli-est.

If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commerical power operations),

supplementary reports shall be submitted at least every 3 months until all three events have been completed.

2.

Annual Operating Report1/

Deleted

e TS 6.6-4 (1)

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures> 100 mrem/yr and their associated man rem exposure acccording to work and job functions, 2/e.g.,

operations and surveillance, in-service reactor inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, measurements.

Small exposures totaling or film badge

< 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.

3.

Monthly Operating Report Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the Reactor Coolant System PORV's or safety valves, shall be submitted on a

monthly basis to the Director, Office of Management and Program

Analysis, U.S.

Nuclear Regulatory Commission, Washington, D.C.

20555, with a

copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.

TS 6.6-5 Pages 6.6-5 through 6.6-9 have been deleted.

TS 6. 6-10 B.

Unique Reporting Requirements

1.

Inservice Inspection Evaluation Special summary technical report shall be submitted to the Director of Reactor Licensing, Office of Nuclear Reactor Regu-lation, NRC, Washington, D.C.

20555, after 5 years of oper-ation.

This report shall include an evaluation of the results of the inservice inspection program and will be reviewed in light of the technology available at that time.

2.

Annual Radiological Environmental Operating Report. 1 Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The initial

• • report shall be submitted prior to May 1 of the year following inital criticality.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate),

and previous environmental surveillance reports, and an assess-ment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification 3.11.D.2.a.

TS 6.6-15

3.

Containment Leak Rate Test Each containment integrated leak rate test shall be the subject of a summary technical report.

Upon completion of the initial containment leak rate test specified by proposed Appendix J to 10 CFR 50, a special report shall, if that Appendix is adopted as an effective rule, be submitted to the Director, Division of Reactor Licensing, USNRC, Washington, D.C.

20555, and other containment leak rate tests specified by Appendix J that fail to meet the acceptance criteria of the appendix, shall be the subject of special summary technical reports pursuant to Sec-tion V.B of Appendix J:

a.

"Report of Test Results - The initial Type A tests shall be subject of a summary technical report submitted to the Commission approximately 3 months after the conduct of the test.

This report shall include a schematic arrangement of the leakage rate measurement system, the instrumenta-tion used, the supplemental test method, and the test program selected as applicable to the initial test, and all subsequent periodic tests.

The report shall contain an analysis and interpretation of the leakage rate test data to the extent necessary to demonstrate the accepta-bility of the containment's leakage rate in meeting the acceptance criteria."

"For periodic tests, leakage rate results of Type A, B, and C tests ~hat meet the acceptance criteria of Sections III.A.7, III.B.3, respectively, shall be reported in the licensee's periodic operating report.

Leakage test re-sults of Type A, B, and C tests that fail to meet the acceptance criteria of Sections III.A.7, III.B.3, and III. C. 3, respectively, shall be reported in a separate summary report that includes an

e TS 6.6-16 analysis and interpretation of the test data, the least squares fit analysis of the test data, the instrument error analysis, and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the acceptance criteria.

Results and analyses of the supplementai verification test em-ployed to demonstrate the validity of the leakage rate test measurements shall also be included."

4.

Initial Containment Structural Test A special summary technical report shall be submitted to the Director, Division of Operating Reactors, USNRC, Washington, D. C.

20555, within 3 months after completion of the test.

This report will include a summary of the measurements of deflections, strains, crack width, crack patterns observed, as

• well as comparisons with predicted values of acceptance cri-teria.

C.

TS 6.6-17 Special Reports In the event that the Reactor Vessel Overpressure Mitigating System is used to mitigate a RCS pressure transient, submit a Special Report to the Commission within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs or the administrative controls on the transient and any corrective action necessary to prevent recurrence.

FOOTNOTES

1.

A single submittal may be made for a multiple unit station.

The submittal should combine those sections that are common to all units at the station.

2.

This tabulation supplements the requirements of §20.407 of 10 CFR Part 20.

i

ATTACHMENT 2 DISCUSSION OF PROPOSED TECHNICAL SPECIFICATION CHANGE Generic Letter No. 83-43 requested all licensees to revise their Technical Specifications to comply with 10CFRSO. 72 and 50. 73, 10CFRSO. 72 has been revised to indicate the immediate notification requirements for operating nuclear power reactors.

10CFRS0.73 is new and provides for a revised Licensee Event Report (LER) System.

The following changes to the Surry 1 and 2 Technical Specification should be made to comply with the new rules:

1.

Throughout the Technical Specifications, revise the term "Reportable Occurence" to become "Reportable Event",

2.

The definition of "Reportable Event" shall read, "a Reportable Event shall be any of those conditions specified in Section 50.73 to 10CFR Part SO.",

3, Delete Technical Specifications 6.6.2, 6.6.2.a and 6.6.2.b,

4.

Throughout the Technical Specifications, delete the references to Technical Specifications 6.6.2, 6.6.2.a and 6.6.2.b,

5.

Insert where applicable, the reference to Section 50.73 to 10CFR Part SO,

6.

Renumber Technical Specification 6.6.a to follow the outline format of the other Surry Technical Specifications,

7.

Throughout the Technical Specification revise the references to Technical Specification 6.6, In addition, minor editorial and typographical errors are corrected.

These proposed changes to the Surry 1 and 2 Technical Specifications do not pose a significant hazards consideration and are administrative in nature. These changes have been requested by the Nuclear Regulatory Commission to reflect the new requirements of 10CFRS0.72 and 50.73.