ML13326A663

Notice of Issuance & Availability of Amend 58 to License DPR-13

ML13326A663
Person / Time
Site:	San Onofre
Issue date:	11/06/1981
From:	Wambach T Office of Nuclear Reactor Regulation
To:
Shared Package
ML13326A660	List: ML13326A659 ML13326A661 ML13326A662 ML13326A663
References
TAC-44128, NUDOCS 8111300188
Download: ML13326A663 (25)

Text

7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-206 SOUTHERN CALIFORNIA EDISON COMPANY AND SAN DIEGO GAS & ELECTRIC COMPANY NOTICE OF ISSUANCE OF AMENDMENT TO PROVISIONAL OPERATING LICENSE The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No.

58 to Provisional Operating License No.

DPR-13, issued to Southern California Edison Company and San Diego Gas and Electric Company (the licensees), which revised the Technical Specifications for operation.

of the San Onofre Nuclear Generating. Station Unit No. 1 (the facility) located in San Diego County, California.

The amendment is effective 30 days from its date of issuance.

The amendment approves changes to the Appendix A Technical Specifi cations which -incorporate certain of the TMI-2 Lessons Learned Category'"A" requirements. These requirements concern (1) Emergency Power Supply Requirements, (2) Valve Position Indication', (3) Instrumentation for Inadequate Core Cooling, (4) Containment Isolation, (5) Shift Technical Advisor, and (6) Containment Sphere Hydrogen Detection and Control.

Note that these Technical Specifications are to be implemented within 30 days.

The application for amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate 8111300188 811106 PDR ADOCK.05000206 P

PDR-,

7590-01 findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter-I, which at 'set forth in the license amendment.

Prior public notice of this amendent was hot required since the amendment does not involve a significant hazards consideration.

The Commission has determined that the issuance of this amendment will not result in any significant-environmental impact that pursuant to 10 CFR §51.5(d)(4),

an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared-in connection with issuance of-this amendment.

For further details with respect to this action, see (1) the application for amendment dated May-7, 1981, (2) Amendment No.

58 to License..No. DPR-13, and (3) the Commission's related Safety Evaluation.

All of these items are available for public inspection at the Commission's-Public Document Room, 1717 H Street, N. W., Washington, D. C., and at the Mission Viejo Branch Library, 24851 Chrisanta Drive, Mission Viejo, California. A single copy of items.(2) and (3) may be obtained upon request addressed to the U. S.

Nuclear Regulatory Commission, Washington, D. C.

20555, Attention:

Director, Division of Licensing.

Dated at Bethesda, Maryland this 6th day of November, 1981.

FOR THE NUCLEAR REGULATORY COMMISSION Thomas V. Wambach, Acting Chief Operating Reactors Branch #5 Division of Licensing

TABLE 3.5.7-1 AUXILIARY FEEDWATER INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION a) Stm. Gen. Water Level-Low

i.

Start Motor Driven Pumps 3

2 2

1, 2, 3, 4 F

ii. Start Turbine-Driven Pump 3

2 2

1, 2, 3, 4 F

ACTION F - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required CHANNEL TEST provided the inoperable channel is placed in the tripped condition withIn 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION G - With more than one channel inoperable, an operator shall assume continuous surveillance and actuate manual initiation of auxiliary feedwater, if necessary. Restore the system to no more than one channel inoperable within 7 days, or be in HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

C+

TABLE 3.5.7-2 AUXILIARY FEEDWATER INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES a> Steam Generator Water Level-Low 2:5% of narrow range 2 0% of narrow range instrument span each instrument span each steam generator steam generator

-34 3.6 Containment Systems 3.6.1 Containment Sphere

>Applicability:

Applies to the operating status of the containment sphere.

Objective:

To ensure containment integrity.

Specification:

A. Leakage The reactor coolant system temperature shall not be increased above 200*F if the containment leakage exceeds the maximum acceptable values specified in Surveillance Standard 4.3.

B. Access to Containment

.(1)

Containment integrity shall not be violated unless the reactor coolant system is below 500 psig and a.

shutdown margin greater than 1%Zl.k/k with all rods inserted is maintained for the most reactive temperature.

(2) Containment integrity shall not be violated when the reactor coolant system is open to the containment atmosphere unless a shutdown margin greater than 5%

k/k is maintained with all control rods inserted.

(3) Positive reactivity changes shall not be made by rod drive motion or boron dilution whenever the containment integrity is not intact.

C. Internal Pressure The reactor shall not be made-critical, nor.be allowed to' remain critical, if the containment sphere internal pressure exceeds 0.4 psig, or the internal vacuum exceeds 2.0 psig.

Basis:

The bases for the shutdown margins and 500 psig pressure are as follows:

k/k Event Basis for Adeauacy 1% (Below 500 Violation of Safety injection system. dis psig)

Containment armed; no credible automatic or operator action could cause return to criticality.

5%

Open reactor Provides adequate margin so coolant system that maintenance activities can be carried out with the reactor head removed.

Amendment No.

58

TABLE 3.6.2-1 POWER OPERATED OR AUTOMATIC CONTAINMENT ISOLATION VALVE

SUMMARY

DESCRIPTION INSIDE SPHERE ALIGNMENT*

OUTSIDE SPHERE ALIGNMENT*

.1. Sphere Sump Discharge CV-102 ISV-108)

B CV-103 SV-10)

A RCS Ur Tk Discharge CV-104 SV-110 B

CV-105 SV-111)

A

3. RCS Dr Tk Vent CV-106 SV-112 B

CV-107 (SV-113 A

4. N2 to RCS Drain Tank and PRT CV-536 A

6-535 a

5. ORMS 1211/1.212 Sphere CV-147 (SV-1212-1)

B SV-1212-9 A

Sample Supply

6. ORMS 1211/1212 Sphere CV-146 (SV-1212-6)

B SV-1212-8 A

Sample Return

1. A Stm. Gen. Stm. Sample None SV-119 A

8..BStm. Gen. Stm. Sample.

None SV-120 A

9. C Stm. Gen. Stm. Sample None SV-121 A

10. A Stm. Gen. Blowdown Sample None SV-123 A

11. 8 Stm. Gen. Blowdown Sample None SV-122 A

12 C Stm. Gen. Blowdown Sample None SV-124 A

13. Service Water to Sphere CV-537 A

CV-115(SV-1Z6)

B

14. Service Air to Sphere Check Valve SV-25 A

15. '1.

Loop C Vent SV-702B A

SV-702A B

16. S1 Loop B Vent SV-7020 A

SV-702C B

17. PRT Gas Sample CV-948**

A CV-949 SV-9491 B

18. RC Loop Sample (CV-955. CV-956, CV-962)**

A CV-957 SV-957

19. Pressurizer Sample (CV-951, CV-953)**

A CV-992 SV-992

20. phere Purge Air Supply POV-9 SV-29)

A

21. Sphere Purge Air Outlet POV-10 SV-30)

A

22. Sphere Equalizing/Sphere Vent CV-116 (SY-27)

B CV-10 SY48)

A Inst. Air Vent CV-40 (SV-19)

B

23. Primary Makeup to.Press CV-533 A

CV-34 B

Rif. Tk

24. Cont. Cooling Out CV515*

A

25. eont. CoolingIn CV.516**

B

26. N2 Supply to PORV CV-532**

B Check Valve

27. Letdown CV-525**

A CV-526" B

28. Seal Water Return CV-527*

A CV.528**

B

29. Hydrogen Monitoring System SV-3004 B

SV-2004 A

Logic Nest C, Train A is aligned to power train F; Logic Nest D. Train B is aligned to power train G.

• ' These valves do not receive an automatic containment isolation signal.

They are operated by remote manual switch (RMS).

-44b 4.1.3 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION (The applicable Technical Specifications remain under review and will be issued at a later date.)

-44c 4.1.4 CONTAINMENT ISOLATION INSTRUMENTATION Applicability:

Applies to instrumentation which actuates the containment sphere isolation.valves, containment sphere purge and exhaust valves, and containment sphere instrumentation vent header val ves.

Objective:

To ensure'reliability of the containment sphere isolation provisions.

Specification:

A. Each instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL TEST operations for the MODES and at the frequencies shown in Table 4.1.4-1.

B. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

Basis:

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

References:

(1) NRC letter dated July 2, 1980,.from D. G. Eisenhut to all pressurized water reactor licensees.

TABLE 4.1.4-1 CONTAINMENT ISOLAMON INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES' IN WHICH CHANNEL CHANNEL CHANNEL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED Containment Isolation (Valves listed in Table 3.6.2-1) a)

Manual N.A.

N.A.

M(1) 1, 2, 3, 4 b) Containment N.A.

R M(2) 1, 2, 3 Pressure-High c) Sequencer Subchannels N.A.

N.A.

M 1, 2, 3, 4 d) Safety Injection

1) Containment Pressure-High N.A.

R M(2) 1, 2, 3

2) Pressurizer Pressure-Low N.A.

R M

1, 2, 3, 4 Purge and Exhaust Isolation (POV-9, POV-10, CV-10, TTO, CV-116) a) Manual N.A.

N.A.

M(1) 1, 2, 3, 4 b).Containment Radioactivity-High S

R M

1, 2, 3, 4

44e TABLE 4.1.4-1 (Continued)

TABLE NOTATION (1) Manual actuation switches shall be tested at least once per 18 months during shutdown.

All other circuitry associated with manual safeguards actuation shall-receive aGCHANNEL TEST at least once per 31 days.

(2) The CHANNEL TEST shall Include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

44f 4.1.5 Accident Monitoring Instrumentation Ap p3ca bii ty,:

Applies to the accident monitoring instruments shown in Table 4.1.5-1 for MODES 1, 2 and 3.

Objectives; To ensure the reliability of the accident monitoring instrumentation.

Specification:

A. Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and.CHANNL CALIBRATION operations at the frequencies shown in Table 4.1.5-1.

Basis:

The surveillance requirements specified for these systems ensure that the overall functional capability is maintained comparable to the.-original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

References:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

TABLE 4.1.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION Pressurizer Water Level M

R Auxiliary Feedwater Flow Indication*

M R

Reactor Coolant System Subcooling Margin Monitor M

R PORV Position Indicator M

R PORV Block Valve Position Indicator M

R Safety Valve Position Indicator M

R 4 See footnote of Table 3.5.6-1.

44h

4.

1.6 Pressurizer Relief Val ves Applicability: Applies to the power operated relief valves (PORVs) and their associated block valves for MODES 1, 2 and 3.

Objective:

To ensure the reliability of the PORVs and block valves.

Specification:

A. Each.PORV shall be demonstrated OPERABLE:

1. At least once per 31 days by performance of a CHANNEL TEST, which may include valve operation, and

2.

At least once per 18 months by performance of a CHANNEL CALIBRATION.

B. Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve:through one complete cycle of full travel.

C. The backup nitrogen supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months, by transferring motive power from the normal air supply to the nitrogen supply and operating the valves through a complete cycle of full travel.

Basis:

The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The air supply for both the relief valves and the block valves is capable of being supplied from a backup passive nitrogen source to ensure the ability to seal this possible RCS leakage path.

References:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

44i 4.1.7 Pressurizer

'Appliedbility: Applies to pressurizer heaters and pressurizer water level-for MODES 1, 2 and 3.

Objective:

To ensure proper pressurizer water volume and to ensure the capability to energize the pressurizer heaters from the emergency diesel generator.

Specification: A. The pressurizer water level shall be determined to be between 5% and 70% at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal supply to the emergency diesel generator and energizing the heaters.

Basis:

The requirement that the pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency diesel generator provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.

References:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

0e

- 44j 4.1.8 Auxiliary Feedwater Instrumentaton..

Applicability:

Applies to the instruments shown in Table 4.1.8-1.

Objective:

To ensure reliability of automatic initiation of the auxiliary feedwater pumps.

Specification:

A. -Each.instrmentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL TEST operations for the MODES and at the frequencies shown in Table 4.1.8-1.

Basis:

The surveillance-requirements specified for this instrumentation ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

References:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

TABLE 4.1.8-1 AUXILIARY FEEDWATER INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES IN WHICH CHANNEL CHANNEL CHANNEL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED a)

Steam Generator Water Level-Low S

R M

1, 2, 3, 4

0.1

- 4.3 Containment Systems 4.3.1 Containment Testing Applicability:

Applies to containment leakage.

Objective:

To verify that leakage from the containment sphere is maintained within specified values.

Specifications:

I.

Integrated Leakage Rate Tests, Type A A. Test Pressure Peak pressure tests are conducted at a test pressure greater than or equal to 49.4 psig, and reduced pressure-tests are conducted at a test pressure greater than or equal to 24.7 psig.

B. Acceptance Criteria For the peak pressure test program the containment sphere leakage rate measured is less than 0.090 wt%/24 hours of the initial content of the containment air at the calculated peak pressure of 49.4 psig. For the reduced pressure test program to be conducted at 24.7 psig, the measured leakage rate shall' be less than 0.064 wt%/24 hours of the initial content of the containment atmosphere at the calculated peak pressure of 49.4 psig.

The accuracy of each Type A test is verified by a supplemental test which (1) confirms the accuracy of the Type A test by verifying that the difference between supplemental and Type A test data is within 25% of 0.12 wt%/24 hours.for the peak pressure test or 0.085 wt%/24 hours for the reduced pressure test, and (2).requires the quantity of air injected into the containment during' the supplemental test to be equivalent to at least 75 percent of the total allowable leakage rate at 49.4 psig.

C. Frequency (1) An integrated leak rate test shall be performed as follows:

(a) Within 24 months from the date of initial criticality.

(b) Within 26 months from the date of the test in "a" above.

(c) Within 39 months from-the date of the test in "b" above.

(d) Within every 39 months from the date of the previous test The intervals specified in a b, c, and d may be va.ried within an allowake-of plus-' montois and minus-8 months to coincide with planned shutdown.

In the event it is determined during any one test that the containment leakage rate does not meet the acceptability limit specified in "8" above, the condition shall be corrected, aretest made, and' the testing frequency shall revert Amendment-No.

4.-

back to item "a" of the above schedule.

-48 II. Penetration Testing The combined leakage rate of all penetrations and all containment isolation valves subject to leakage rate tests shall not exceed 0.072 wt%/24 hours off the initial content of the containment atmosphere at the calculated peak pressure of 49.4 psig.

A. Types D, E, and Electrical Penetrations (1). Tests Leakage tests of types D, E, and electrical penetrations through the containment sphere shall be performed at an initial-pressure-(beginning of test) of 49.4 psig.

(2) Frequency For Type D penetration, testing shall be accomplished during shutdown when the reactor is depressurized if the test has not been performed within the previous 6 months but in no case at intervals greater than 2 years.

Type E and all electrical penetrations shall be tested at a frequency of at least every 6 months.

B. Personnel Air Locks

1) Test Leakage tests of personnel air locks shall be performed at an initial pressure (beginning of test) of approximately 10 psig.

(2)

Frequency During operation, personnel air locks shall be tested at a frequency of at least every 6 months.

Isol ation Val ve Testing All isolation valves shall be tested for leak rate characteristics. -Isolation valves normally operating-with pressure less than 50 psig shall be tested at an initial pressure (beginning of test) of 49.4 psig.

Amendment 24, 58

-51a 4.3.2 CONTAINMENT ISOLATION VALVES Application:

Applies to the containment isolation valves listed in Table 3.6.2-1 for MODES 1, 2, 3 and 4.

Objective:

To ensure reliability of containment isolation valves.

Specification: A. The isolation valves specified in Table 3.6.2-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power-circuit by performance of a cycling test.

B. Each isolation valve specified in Table 3.6.2-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:

1. Verifying that on containment isolation test signal, each automatic isolation valve actuates to its isolation position.

2.-Verifying that on a containment radiation-high test signal, each-purge supply and-purge outlet automatic valve actuates to its isolation position.

Basis:

The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

References:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

51b 4.3.3 Hydrogen Monitors and Hydrogen Recombiners Application:

Applies to containment sphere hydrogen monitors and hydrogen recombiners-for MOUES 1 and 2.

Objective:

To. ensure reliability of the equipment and systems required for-the detection and control of hydrogen gas.

Specification: A. Each hydrdgen monitor shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases containing:

1. Two volume percent hydrogen, balance nitrogen.

2. Six volume percent hydrogen, balance nitrogen.

B.

Each hydrogen recombiner system shall be demonstrated OPERABLE at least once per 6 months by verifying that the minimum-heater sheath temperature increases to greater than or equal to 7000F within 90 minutes. Upon reaching 7000F, increase the power setting to maximum power for 2 minutes and verify that the power meter reads greater than or equal to 60 Kw.

C. Each hydrogen recombiner system shall be demonstrated OPERABLE at least once per 18 months by:

1. Performing a CHANNEL CALIBRATION of all recombiner instrumentation and control circuits.

2. Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure (i.e., loose wiring or structural connections, deposits or foreign materials, etc.),,and

3. Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the test in Specification B above. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.

Basis:

The OPERABILITY of-the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit (or the purge system) is capable of controlling the expected hydrogen generation associated with radiolytic decomposition of water and corrosion of metals within containment.

(Cumulative operation of the purge system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31-day period is sufficient to reduce the buildup of

51c moisture on the adsorbers and HEPA filters). These hydrogen control systems are consistent with the recommendations of Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a LOCA," March, 1971.

References:

(1) Regulatory Guide 1.7, "Control of Combustible Gas

-Concentrations in Containment Following a LOCA,"

March, 1971.

-71 TABLE 6.2.2.2 MINIMUM SHIFT CREW COMPOSITION#

LICENSE CATEGORY APPLICABLE MODES QUALIFICATIONS 1, 2, 3 & 4 5 & 6 SRO 1

1*

RO 2

1 Non-Licensed Auxiliary 1

1 Operator Shift Technical Advisor 1

None Required

• Does not include the licensed Senior. Reactor Operator or Senior Reactor Operator limited to Fuel Handling, supervising CORE OPERATIONS.

1. Shift crew composition may be one less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accomodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition' to within the minimum requirements of Table 6.2.2.2. This provision does not permit any shift crew position to'be unmanned upon shift change due to an oncoming shift crewman being late or absent.

Amendment No. V{, 58

• 72a

-6.3 Unit Staff Qualifications 6.3.1 Each member of the unit, staff shall meet.or exceed the-minimum qualifications of.ANSI N18.1-1971 for comparable positions, except for (1) the Health Physics Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) the Shift Technical Advisor who 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.

6.4 Training 6.4.1 A retraining and replacement training program for the faciity staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations, of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements of Section 27 of the National Fire Protection Association Code - 1976."

Amendment No. ;

, 54, 8

-88

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

g. Records of Quality Assurance activities as required by the QA Manual.

h. Records of reviews performed for changes made tb procedures or equipmnt or reviews or tests and experiments pursuant to 10 CFR 50.59.

i.

Records of meetings of the OSRC and the NARC.

j.

Records for.Environmental Qualification which are covered under the provisions of paragraph 6.12.

6.10.3 The following records shall be retained for two years:

a. Records of facility radiation and contamination surveys.

b. Records of training of facility personnel.

6.11 RADIATION PROTECTION PROGRAM 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.

6.12 ENVIRONMENTAL QUALIFICATIONS A. By no later than June 20, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of:

Division of Operating Reactors "Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 "Interim Staff Position on Environmental Qualification of Safety Related Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License DPR-13 dated October 24, 1980.

B By no later than December 1, 1980, completeand auditible records.

must be available and maintained at a central location which describe the environmental qualification method used'for all safety-related electrical'equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588.

Thereafter, such records should be updated and maintained current as equipment is replaced,.further tested, or otherwise further qualified.

Amendment No. 2-,,

58

-89 6.13 SYSTEMS INTEGRITY The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

This pro-gram shall include the following:

1.

Provisions establisbing preventative maintenance and periodic visual inspection requirements, and

2. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

6.14 IODINE MONITORING The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1.

Training of personnel,

2.

Procedures for monitoring, and

3.

Provisions f6r maintenance of sampling and analysis equipment.

6.15 BACKUP METHOD FOR DETERMINING SUBCOOLING MARGIN The licensee shall implement a program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This pro gram shall include the following:

1. Training of personnel, and

2. Procedures for monitoring.

Amendment No.

58