ML18143B446

Proposed Tech Specs Reflecting Reorganization of Nuclear Operations Dept & Changes to LER Sys

ML18143B446
Person / Time
Site:	Surry
Issue date:	08/30/1985
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18143B445	List: ML18143B444 ML18143B446
References
NUDOCS 8509100204
Download: ML18143B446 (43)

Text

ATTACHMENT 2 CONSOLIDATED SET OF PROPOSED TECHNICAL SPECIFICATION CHANGES Incorporates changes associated with VEPCO's letters of November 2 and 30, 1984; April 12 and 17, 1985, and this letter.

8509100204 850830 PDR ADDCK 05000280 p PDR

e List of Proposed Revised Pages Page 1.0-5 Figure 6.1-1 Page 3.l-15a Figure 6.1-2 Page 3 .1-24 Page 3.7-21 Page 6.2-1 Page 3 .11-2 Page 6.3-1 Page 3 .11-3 Page 6.4-2 Page 3.11-4 Page 6.4-3 Page 3.11-5 Page 6.4-4 Page 3 .11-6 Page 6.4-5 Page 3 .11-7 Page 6.5-1 Page 3.12-7 Page 6.6-1 Page 4.9-15 Page 6.6-2 Page 4.19-8 Page 6.6-4 Page 4.19-10 Page 6.6-5 Page 6. 6-10 Table 4.19-2 Page 6.6-12 Page 6.6-15 Page 6 .1-1 Page 6.6-16 Page 6.1-2 Page 6.6-17 Page 6.1-6 Page 6.1-7 Page 6.1-8 Page 6 .1-11 Page 6.1-12 Page 6 .1-15

e TS 1.0-5 for operational activities provided that they are under administra-tive 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.

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

TS 3.l-15a

2. The specific activity of the reactor coolant shall be limited to :::; 1.0 µCi/cc DOSE EQUIVALENT-131 whenever the reactor is critical or the average temperature is greater than 500°F.

3. The requirements of D-2 above may be modified to allow the specific activity of the reactor coolant >1.0 µCi/cc DOSE EQUIVALENT I-131 but less than 10. 0 µCi/cc DOSE EQUIVALENT I-131. Following shutdown, the unit may be restarted and/or operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided that operation under these circumstances shall not exceed 10 percent of the unit's total yearly operating time. With the specific activity of the reactor coolant >1.0 µCi/cc DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding 10.0 µCi/cc DOSE EQUIVALENT I-131, the reactor shall be shut down and cooled to 500°F or less within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after detection. With the total cumulative operating time at a primary coolant specific activity> 1.0 µCi/cc DOSE EQUIVALENT I-131 exceeding 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> in any consecutive 6 month period, prepare and submit a Special Report to the NRC, Regional Administrator, Region II, within 30 days indicating the number of hours above this limit.

4. If the specific activity of the reactor coolant exceeds 1.0 µCi/cc DOSE EQUIVALENT I-131 or 100/E µCi/cc, a report shall be prepared and submitted to the Commission pursuant to Specification 6.2. This report shall contain the results of the specific activity analysis together with the following information:

a. Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded,

b. Fuel burnup by core region,

c. Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded,

e 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.1.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.

Basis 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 INSTRUMENT TOTAL NO. MINIMUM CHANNELS OF CHANNELS OPERABLE

1. Auxiliary Feedwater Flow Rate 1 per S/G 1 per S/G

2. Reactor Coolant System Subcooling Margin Monitor 2 1

3. PORV Position Indicator (Primary Detector) 1/valve 1/valve

4. PORV Position Indicator (Backup Detector) 1/valve 0

s. PORV Block Valve Position Indicator 1/valve 1/valve

6. Safety Valve Position Indicator (Primary Detector) 1/valve 1/valve e

7. Safety Valve Position Indicator (Backup Detector) 1/valve 0

8. Reactor Vessel Coolant Level Monitor 2 1

9. Containment Pressure 2 1

10. Containment Water Level (Narrow Range) 2 1

11. Containment Water Level (Wide Range) 2 1

12. Contaiment High Range Radiation Monitor 2 1 (Note 1, band c only)

13. Process Vent High Range Effluent Monitor 2 2 (Note 1, a, b, and c)

14. Ventilation Vent High Range Effluent Monitor 2 2 (Note 1, a, b, and c)

15. Main Steam High Range Radiation Monitors 3 3 (Note 1, a, b, and c)

(Units 1 and 2)

16. Aux. Feed Pump Steam Turbine Exhaust Radiation 1 1 (Note 1, a, b, and c)

Monitor 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 6.2 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 operable.

e TS 3.11-2

c. The surveillance requirements for liquid effluents are given in Table 4.9-1.

d. The reporting requirements of section 6.2 are not applicable.

2. Dose

a. The dose or dose commitment to the maximum exposed member of the public from radioactive materials in liquid efflu-ents released, from each reactor unit, to unrestricted areas shall be limited:

(i) During any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to the critical organ, and (ii) During and calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to the critical organ.

b. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

e e TS 3.11-3

3. Liquid Radwaste Treatment

a. The Liquid Radwaste Treatment System shall be used to reduce the redioactive materials in liquid waste prior to their discharge when the projected dose due to liquid effluent releases to unrestricted areas (see figure 5.1-1) when averaged over 31 days would exceed 0.06 mrem to the total body or 0.2 mrem to the total body or 0.2 mrem to the critical organ.

b. With radioactive liquid waste being discharged with-our treatment and in excess of the above limits, prepare and submit to the Commission within 30 days pursuant to Specification 6.2 a Special Report that includes the following information:

(i) Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or sub-system, and the reason for the inoperability, (ii) Action(s) taken to restore the inoperable equip-ment to operable status, and (iii) Summary description of action(s) taken to pre-vent a recurrence.

TS 3.11-4 B. Gaseous Effluents

1. Dose Rate

a. The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the site boundary (see figure 5.1-1) shall be limited to the following:

(i) For noble gases: less than or equal to 500 mrems/yr.

to the total body and less than or equal to 3000 mrems/yr. to the skin, and (ii) For iodine-131, for tritium, and for all radio-nuclides in particulate form with half lives greater that 8 days: less than or equal to 1500 mrems/yr. to the critical organ.

b. With the dose rate(s) exceeding the above limits, without delay restore the release rate to within the above limit (s).

c. The reporting requirements of section 6.2 are not applicable.

2. Dose-Noble Gases

a. The air dose due to noble gases released in gaseous effluents, from each reactor unit, from the site to areas at and beyond the site boundary (see figure 5.1-1) shall be limited to the following:

(i) During any calendar quarter: less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation and,

. TS 3.11-5 (ii) During any calendar year: less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

b. With the calculated air dose from radioactive noble gases in** gaseous effluents exceeding any of the above limits, prepare and pursuant .to submit to the Specification Commission within 6.2, a Special 30 Report days, that I

identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

3. Dose-I-131, Tritium, and Radionuclides in Particulate Form

a. The dose to the maximum exposed member of the public from all I-131, from tritium, and from all radionuclides in particulate form with half-lives greater that 8 days in gaseous effluents .released, from each reactor unit, from the site to areas at and beyond the site boundary (see figure 5.1-1) shall be limited to the following:

(i) During any calendar quarter: less than or equal to 7.5 mrems to the critical organ, and (ii) During any calendar year: less than or equal to 15 mrems to the critical organ.

b.

• TS 3.11-6 With the calculated dose from the release of I-131, tri-tium, and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to commission within 30 days, pursuant to Specification 6.2, a Special Report that identifies the cause(s) the for I

exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

4. Gaseous Radwaste Treatment

a. The appropriate portions of the Gaseous Radwaste Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases, from each reactor unit, from the site to areas at and beyond the site boundary (see Figure 5. 1-1) would exceed O. 2 mrad for gamma radiation and 0.4 mrad for beta radiation when averaged over 31 days.

b. The Ventilation Exhaust Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases, from each reactor unit, from the site to areas at and be):ond the site boundary (see Figure 5 .1-1) would exceed O. 3 mrem to the critical organ when averaged over 31 days.

c.

• TS 3.11-7 With gaseous waste* being discharged without treatment and in I

excess of the above limits,prepare and submit to the Commission within 30 days, pursuant to Specification 6.2, a Spe~ial Report that includes the following information:

(i) Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equip-ment or sub-systems, and the reason for the inoperability, (ii) Action(s) taken to restore the inoperable equipment to operable status, and (iii) Summary description of action(s) taken to prevent a recur-rence.

5. Explosive Gas Mixture

a. The concentration of hydrogen or oxygen in the waste gas holdup system shall be limited to less than or equal to 4% by volume.

b. With the concentration of hydrogen or oxygen in the waste gas holdup system exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

6. Gas Storage Tanks

a. The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to i4,600 curies of noble gases (considered as Xe-133).

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 1% 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 TS 4.9-15 Eis the counting efficiency (as counts per disintegration),

Vis the sample size (in units of mass or volume),

2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),

is the radioactive decay constant for the particular radionuclide, and t for environmental samples is the elapsed time between sample collection (or end of the sample collection period) and time of counting Typical values of E, V, Y, and t should be used in the calculation.

It should be recognized that the LLD is defined as an 2- priori (before the fact) limit representing the capability of a measurement system and not as !. posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contri-buting factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.6.b.2.

' .. I W,;-

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. Nwnber 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.

TS 4.19-10 withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readpy 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 results 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

• r STEAM GENERATOR TUBE INSPECTION 1st SAMPLE INSPECTION 2nd SAMPLE INSPECTION 3rd SAMPLE INSPECTION Sample Size Result Action Required Result ~ction Required Result Action Required A minimum of C-1 None NIA NIA NIA NIA 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 addit- tubes 4S tubes in this S_.G.

C-3 Perform action for C-3 result of first sample C-3 Perform action for C-3 result of

. NIA NIA first sample C-3 Inspect all tubes in All other None NIA 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 NIA 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

TS 6.1-1 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization, Safety and Operation Review Specification A. The Station Manager shall be responsible for the overall operation of the facility. In his absence, the Assistant Station Manager (Operations and Maintenance) shall be responsible for the safe operation of the facility. During the absence of both, the Station Manager will delegate in writing the succession to this responsibility.

1. The off-site organization for facility management and technical support shall be as shown on TS Figure 6.1-1.

B. The Station organization shall conform to the chart as shown on TS Figure 6.1-2.

1. Each member of the facility staff shall meet or exceed the minimum qualifications of ANS 3 .1 (12/79 Draft)

• for comparable positions, and the supplemental requirements specified in the March 28, 1980 NRC letter to all licensees, except for the Superintendent - Health Physics who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

• Exceptions to this requirement are specified in VEPCO' s QA Topical Report, VEP-1, "Quality Assurance Program, Operations Phase."

e TS 6.1-2

2. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and response and analysis of the plant for transients and accidents.

3. The Station Manager is responsible for ensuring that re-training and replacement training programs for the facility staff are maintained and that such programs meet or exceed the requirements and recommendations of Section 5. 5 of ANSI (12/79 Draft)* and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in the March 28, 1980 NRG letter to all licensees, and shall include familiarization with relevant industry operational experience identified by the SEC Staff.

4. Each on-duty shift shall be composed of at least the minimum shift crew composition for each unit as shown in Table 6 .1-1.

5. A health physics technician shall be on site when fuel is in the reactor.

6. All core alterations shall be observed and directly super-vised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

• Exceptions to this requirement are specified in VEPCO's QA Topical Report, VEP-1, "Quality Assurance Program, Operations Phase."

TS 6.1-6 C. Organization units to provide a continuing review of the opera-tional and safety aspects of the nuclear facility shall be constituted and have the authority and responsibilities outlined below:

1. Station Nuclear Safety and Operating Committee (SNSOC)

a. Function The SNSOC shall function to advise the Station Manager on all matters related to nuclear safety.

b. Composition The SNSOC shall be composed of the:

Chairman Assistant Station Manager, Nuclear Safety and Licensing Vice Chairman Assistant Station Manager, Operations and Maintenance Member Superintendent - Operations Member Superintendent - Maintenance Member Superintendent - Technical Services Member Superintendent - Health Physics

c. Alternates All alternate members shall be appointed in writing however, no more than two alternates shall partici-pate as voting members in SNSOC activities at any one time.

d. Meeting Frequency The SNSOC shall meet at least once per calendar month and as convened by the SNSOC Chairman or his desig-nated alternate.

e. Quorum A quorum of the SNSOC shall consist of the Chairman or Vice Chairman and two members including alternates.

• 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.

r-

3. Review of all proposed changes or modifications f

to plant systems or equipment that affect nuclear safety.

4. Review of proposed changes to Technical Specifi-cations and shall submit recommended changes to the Station Manager.

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 Vice President - Nuclear Operations and to the Director - Safety Evaluation and Control.

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

7. Review of facility operations to detect potential nuclear safety hazards.

8. Performance of special reviews, investigations or analyses and report thereon as requested by the Chairman of the SNSOC or Station Manager.

e TS 6.1-8

9. Review of the Plant Security Plan and implement-ing procedures and shall submit recommended l*

changes to the Station Manager.

10. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Station Manager.

11. Review of every unplanned onsite release of radioactive material to the environs exceeding the limits of Specification 3.11, including the preparation or reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President-Nuclear Operations and to the Director-Safety Evaluation and Control.

12. Review of changes to the Process Control Program and the Offsite Dose Calculation Manual.

g. Authority The SNSOC shall:

1. Provide written approval or disapproval of items considered under (1) through (3) above. SNSOC approval shall be certified in writing by an Assistant Station Manager. f

2. Render determinations in writing with regard to whether or not each item considered under 11 (1) 11 through 11 (5) 11 above constitutes an unreviewed safety question.

3. Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President - Nuclear Operations and the Director - Safety Evaluation and Control of disagreement between SNSOC and the Station Manager; however, the Station Manager shall have responsibility for resolution of such disagree-ments pursuant to 6.1.A above.

TS 6.1-11

3. Changes in the Technical Specifications or license amendments 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 devi-ations 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 recommendations 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 e TS 6.1-12

6. Any other matter involving safe operation of the nuclear power stations which is referred to the Director - Safety Evaluation and Control.

7. Reports and meeting minutes of the Station Nuclear Safety and Operating Committee.

f. Authority The Director - Safety Evaluation and Control shall report to and advise the Manager - Nuclear Programs f and Licensing, who shall advise the Vice President -

Nuclear Operations on those areas of responsibility specified in Section 6.1.C.2.d.

g. Records Records of SEC activities required by Specification 6.1.C.2.e shall be prepared and maintained in the 'SEC files and a summary shall be disseminated each calendar month as follows:

1. Vice President - Nuclear Operations

2. Nuclear Power Station Managers

3. Manager - Nuclear Operations Support

4. Manager - Nuclear Programs and Licensing

5. Executive Manager - Quality Assurance

6. Others that the Director - Safety Evaluation and Control may designate

TS 6.1-15

c. Records Records of the Quality Assurance Department audits shall be prepared and maintained in the department files. Audit reports shall be disseminated as indi-cated below:

1. Vice President - Nuclear Operations

2. Nuclear Power Station Manager

3. Manager - Nuclear Operations Support

4. Executive Manager - Quality Assurance

5. Manager - Nuclear Programs and Licensing

6. Director - Safety Evaluation and Control

7. Supervisor of area audited

8. Nuclear Power Station Manager-Quality Assurance

OFF-SITE ORGANIZATION FOR FACILITY MANAGEMENT AND TECHNICAL SUPPORT EXECUTIVE VHCE PRESIDENT POWER-C.0.0.

SENIOR VICE PRESIDENT POWER OPERATIONS e VICE PRESIDENT NUCLEAR OPERATIONS EXECUTIVE MANAGER MAINTENANCE & MANAGER PERFORMANCE QUALITY SERVICES I I ASSURANCE MANAGER NUCLEAR NUCLEAR MANAGER NUCLEAR STATION! STATION NUCLEAR MANAGER OPERATIONS MANAGER MANAGER PROGRAMS AND MANAGER POWER TRAINING SUPPORT SURRY NORTH ANNA LICENSING QUALITY I I SERVICES ASSURANCE.

I I DIRECTOR

,- L - - - .J I

NUCLEAR I I I I I TRAINING I DIRECTOR DIRECTOR DIRECTOR DIRECTOR .t,-3 I

DIR!ECTOR OPERATIONS AND SAFETY .

Cl)

ADMINISTRATIVE EMERGENCY HEALTH PHYSICS MAINTENANCE EVALUATION "rj SUPERINTENDENT I SUPPORT SERVICES PLANNING AND CONTROL I-'*

(IQ NUCLEAR - _J °'

I-'

TRAINING I I-'

SURRY POWER STATION ORGANIZATION CHART VICE PRESl>ENl NUCLEAR OPERATIONS ,...

MANAGER EXECUTIVE A ..TENANCE ANll MANAGER QUALITY PERFORUANCE SERVICES ASSURANCE STATION ~ - - - - STATION

..... - - - - - - -~

NUCLEAR MANAGER MANAGER MANAGER SAFETY

• OPER.

POWER TRAINING ._EOMMfTTEE QUALITY SERVICES ASSURANCE I

DIRECTOR SUPERVISOR .

NUCLEAR QUALITY CONTAOL TRAIMING Q.A. ACTIVITl:8

---

1 AS818TANT ASSISTANT DIRECTOR OTA1'10N STATION MANAGER NUCLEAR MANAGER (0 a M)

(NS a L) SECURITY I

I I I SUPERINTENDEN1 8UPEAINTENDEN1 8UPERINTENDEN1 8UPERINTENDEN1 SUPER..TENDENT SUPER..TENDENT NUCLEAR TECHNICAL OPERATIONS MAINTENANCE PROJECTS HEALTH TRAINll'tG SERVICES RL PHYSICS SUPERVISOR SUPERVISOR SHFT SUPERVISOR LICENSING SUPV.-ENG. COORDINATOR SUPERVISOR ENGINEER.. GL. SUPERVISOR>- MECHANICAL,_ I- (SAFETY ENG. EMERGENCY ADMIN.

CHEMISTRY COORDINATOR (PLANNING) MAINTENANCE STAFF) PLANN ..G SERVICES 8L 1-:l ASST. SHIFT SHFT. SUPERVISOR BUSINESS en ELECTRICAL INSTRUMENT SUPERVISOR 1- I- I- TECHNCAL RECORDS L. L. SYSTEMS '"%j SUPERVISOR SUPEVISOR ADVISORS MANAGEMENT t-'*

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SUPERVl~OF SUPERVISOR

e TS 6.2-1 6.2 GENERAL NOTIFICATION AND REPORTING REQUIREMENTS Specification A. The following actions shall be taken for Reportable Events:

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. The Director-Safety Evaluation and Control- and Vice-President Nuclear Operations shall be notified of the results of this review.

B. Immediate notifications shall be made in accordance with Section 50.72 to 10 CFR Part 50.

.. TS 6.3-1 6.3 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED Specification A. The following actions shall be taken in the event a Safety Limit is violated:

1. The facility shall be placed in at least hot shutdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2. The Safety Limit violation shall be reported to the Commission, the . Vice President - Nuclear Operations, and the Director -

Safety Evaluation and Control within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SNSOC. 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.

4. The Safety Limit Violation Report shall be submitted to the Commission, the Director - Safety Evaluation and' Control, and the Vice President - Nuclear Operations within 14 days of the violation.

F e TS 6.4-2

1. The intent of 10 CFR 20.203(c)(2)(iii) shall be implemented by satisfying the following conditions:

a. The entrance to each radiation area in which the intensity of.,_. radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posted.

b. The entrance to each radiation area in which the intensity of radiation is equal to or greater than 1000 mrem/hr shall be provided with locked barricades to prevent unauthorized entry into these areas. Keys to these locked barricades shall be maintained under the administrative control of the Shift Supervisor on duty and/or Superintendent Health Physics.

c. All such accessible high radiation areas shall be surveyed by Health Physics personnel on a routine schedule, as determined by the Superintendent-Health Physics, to assure a safe and practical program.

d. Any individual entering a high radiation area shall have completed the indoctrination course designed to explain the hazards and safety requirements involved, or shall be escorted at all times by a person who has completed the course.

e. Any individual or group of individuals permitted to enter a high radiation area per 1. d above, shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.

TS 6.4-3

f. Entrance to areas with radiation levels in excess of 1 R/hr shall require the use of the "buddy system", whereby a minimum of two individuals maintain continuous visual and/or verbal communication with each other; or other mechanical and/ or electrical means to provide constant communication with the individual in the area shall be provided.

g. A Radiation Work Permit system shall be used to authorize and control any work performed in high radiation areas.

h. All buildings or structures, in or around which a high radiation area exists, shall be surrounded by a chain-link fence. The entrance gate shall be locked under admini-strative control, or continuously guarded to preclude unauthorized entry.

i. Stringent administrative procedures shall be implemented to assure adherence to the restriction placed on the entrance to a high radiation area and the radiation pro-tection program associated thereto.

2. Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. Process Control Program implementation.

b. Offsite Dose Calculation Manual implementation.

C. All procedures described in 6. 4 .A and 6. 4 .B, and changes thereto, shall be reviewed and approved by the Station Nuclear Safety and Operating Committee prior to implementation.

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e e TS 6.4-4 D. All procedures described in Specifications 6.4.A and 6.4.B shall be followed.

E. Temporary changes to procedures described in Specifications 6. 4 .A and 6.4.B which do not change the intent of the original procedure may be made, provided such changes are approved prior to implementa-tion by the persons designated below based on the type of procedure to be changed:

1. Administrative Cognizant Supervisor

2. Abnormal Shift Supervisor or Assistant Shift Supervisor

3. Annunciator Shift Supervisor or Assistant Shift Supervisor

4. Health Physics *Health Physicist

5. Emergency Shift Supervisor or Assistant Shift Supervisor

6. Maintenance *Cognizant Supervisor

7. Operating Shift Supervisor or Assistant Shift Supervisor

8. Periodic Test *Cognizant Supervisor

9. Start-up Test *Engineering Supervisor

10. Special Test *Engineering Supervisor

11. Quality Assurance Manager, Quality Assurance or Supervisor Quality Control

12. Chemistry *Chemist

• These procedures must have the approval of a licensed Senior Reactor Operator.

Such changes will be documented and subsequently reviewed and approved by the Station Nuclear Safety and Operating Committee within 14 days.

F. Temporary changes to procedures described in Specifications 6. 4 .A and 6.4.B which change the intent of the original procedures may be made, provided such changes are approved prior to implementation by the person designated below based on the type of the procedure to be changed.

e e TS 6.4-5

1. Administrative Station Manager

2. Abnormal Superintendent - Operations

3. Annunciator Superintendent - Operations

4. Health Physics Superintendent - Health Physics

5. Emergency Superintendent - Operations

6. Maintenance Mechanical Supervisor Electrical Supervisor Instrument Supervisor

7. Operating Superintendent - Operations

8. Periodic Test Engineering Supervisor

9. Start-up Test Engineering Supervisor

10. Special Test Engineering Supervisor

11. Quality Assurance Manager, Quality Assuiance or Supervisor

12. Chemistry Supervisor - Chemistry Such changes will be documented and subsequently reviewed and approved by the Station Nuclear Safety and Operating Committee.

G. In cases of emergency, operations personnel shall be authorized to depart from approved procedures where necessary to prevent injury to personnel or damage to the facility. Such changes shall be docu-mented, reviewed and approved by the Station Nuclear Safety and Operating Committee.

.) e 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, µnless 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 10 CFR 50.59.

6. Records of changes made in the Operating Procedures pursuant to 10 CFR 50.59.

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.

e 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 Administrator of the appropriate NRC Regional Office 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

.. e 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.

1

2. Annual Operating Report /

Deleted

. ,, . e e TS 6.6-4 (1) A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) re-ceiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and job f unctions, Z/ e.g.,

• reactor operations an d survei*11 ance, inservice inspection, routine maintenance, special main-tenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling 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.

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.

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

e - TS 6.6-10 B. Unique Reporting Requirements I

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.

1

2. Annual Radiological Environmental Operating Report.

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.

e TS 6.6-12 1

3. Semi-Annual Radioactive Effluent Release Report Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

The Radioactive Effluent Release Reports shall include a sunnnary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Tables 1, 2, and 3 of Appendix B thereof, The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an assessment of the radiation doses to the maximum exposed members of the public due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. Annual meteorological data shall be retained in a file on site and shall be made available to the NRC upon request . All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in the

TS 6.6-15

4. 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 that 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

• -.t ... V 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 supplemental verification test em-ployed to demonstrate the validity of the leakage rate test measurements shall also be included."

C.

Special Reports

- TS 6.6-17 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 K20.407 of 10 CFR Part 20.