ML13310B540

Amend 77 to License DPR-13,changing Tech Specs to Provide Functional Redundancy in DHR Methods During All Modes of Plant Operation

ML13310B540
Person / Time
Site:	San Onofre
Issue date:	07/06/1984
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:
Shared Package
ML13310B539	List: ML13310B538 ML13310B540 ML13310B541
References
DPR-13-A-077, TAC-42131 NUDOCS 8407090234
Download: ML13310B540 (17)

Text

4 REG,

UNITED STATES NUCLEAR REGULATORY COMMISSION o

WASHINGTON, D. C. 20555 SOUTHERN CALIFORNIA EDISON COMPANY AND SAN DIEGO GAS AND ELECTRIC COMPANY DOCKET NO. 50-206 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 77 License No. DPR-13

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Southern California Edison Company and San Diego Gas and Electric Company (the licensees) dated December 5, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

-2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Provisional Operating License No.

DPR-13 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 77, are hereby incorporated in the license. -Southern California Edison Company shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective 90 days from the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION r ~. Den is M..rutc field, Chief Ope ating R ct rs Branch #5 sion of Li nsing

Attachment:

Changes to the Technical Specifications Date of Issuance: July 6, 1984

9 ATTACHMENT TO LICENSE AMENDMENT NO. 77 PROVISIONAL OPERATING LICENSE NO. DPR-13 DOCKET NO. 50-206 Revise Appendix A Technical Specifications and Bases by removing the following pages and by inserting the enclosed pages. The revised pages contain the captioned amendment number and marginal lines indicating the area of change.

Remove Pages Insert Pages la la 10 10 10a 10b 10c 10d 11 11 38 38 38a 38a 38b 38b 38c 38d 43c 43d

~

la -9 Residual Heat Removal (RHR) Train A train of components that includes: one RHR pump aligned with one RHR heat exchanger; one component cooling water pump aligned with the same RHR heat exchanger and with one component cooling water heat exchanger; and one salt water pump aligned with the same component cooling water heat exchanger.

CDezational Moe -

mode An Operational Hooe (i.e., mode) shall correspono to any one inclusive comoination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

TABLE 1.2 OPERATIONAL HOES REACTIVITY

% RATED AVERAGE C01.ANT MCCE CONOITION, K ff THERAL POWER*

TEmPERATURE

1.

POWER OPERATION

>0.99 5%

>350 0F

2.

STARTUP ZC.99

< 5%

>3500F

3.

HOT STANDBY

<0.99 0

850 A. HOT SHUTDOWN 50.95 0

3500F> Tavg> 200o0

5.

COLD SHLTC wN 1.0.95 0

.<2000F

6.

REFLELINC.*

.<0.95 0

14 CF Exclucing decay heat.

• Rezctor vessel h.ead unbolted or removed and fuel in the vessel.

Amendent No 56 77

-10 3.1.2 OPERATIONAL COMPONENTS AoDlicability: Applies to the operating status of the reactor coolant system equipment and related equipment. For the applicable surveillance requirements, see Table 4.1.2.

Objective:

To identify those conditions of the reactor coolant system necessary to ensure safe reactor operation.

Soecification: A. At least one pressurizer safety valve shall be operable or open when the reactor head is on the vessel, except for hydrostatic tests.

B. The reactor shall not be made critical or maintained critical unless both pressurizer safety valves are operable.

C. During Modes 1 and 2, all three reactor coolant loops and their associated steam generators and reactor coolant pumps shall be in operation. With less than the above required coolant loops in operation, be in at least Hot Standby within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, except as modified by Specification 0 below.

D. The limitations of Specification C may be suspended during Modes 1 and 2 as follows:

1. Operation may be conducted with 0, 1, 2, or 3 reactor coolant pumps operating during low power physics testing at less.than-5% of full power.

2. Whenever reactor power is less than 10% of full power, operation with one or two reactor coolant pumps operating shall be limited to less than 24 consecutive hours.

E. During Mode 3, the following specifications shall apply:

1. At least two of the reactor coolant loops listed below shall be operable:

a. Reactor coolant loop A and its associated steam generator and reactor coolant pump.

b. Reactor coolant loop B and its associated steam generator and reactor coolant pump.

Amendment No.",

7 7

-lOa

c. Reactor coolant loop C and its associated steam generator and reactor coolant pump.

2. At least one of the above reactor coolant loops shall be in operation.*

3. With less than the above required reactor coolant loops operable,- restore the required loops to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.

With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immnediately initiate corrective action to return the required reactor coolant loop to operation.

F. During Mode 4, the following specifications shall apply:

1. At least two of the reactor coolant loops/residual heat removal (RHR) trains listed below shall be Operable:

a. Reactor coolant loop A and its associated steam generator and reactor coolant pump.

b. Reactor coolant loop B and its associated steam generator and reactor coolant pump.

c. Reactor coolant loop C and its associated steam generator and.reactor coolant pump.

d. Residual heat removal (RHR) pump G-14A and one associated RHR train.

e. Residual heat removal (.RHR) pump G-14B and one associated RHR tr-ain.

2. At least one of the above loops/trains shall be in operation.*

• All reactor coolant pumps may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 40oF below saturation temperature.

• • All reactor coolant pumps and residual heat removal pumps may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 400F below saturation temperature.

Amendment No-. 77

10b

3. With less than the above required loops/trains operable immediately initiate corrective action to return the required loops/trains to operable status as soon as possible; if the remaining operable loop/train is an RHR train, be in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4. With no loop or-train in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immeaiately initiate corrective action to return one required loop or train to operation.

6. During Mode 5 with reactor coolant loops filled, the following specifications shall apply:

1. At least one residual heat removal (RHR) train shall be operable and in operation*, and either

a. One additional RHR train shall be operable,** or

b. The secondary side water level of at least two steam generators shall be greater than or equal to 256 inches of narrow range on cold calibrated scale.

2. With less than the above required loops/trains operable, or with less than the requi.red steam generator level, immediately initiate corrective action to return the required loops/trains to operable status or to restore the required level as soon as possible.

3. With no RHR train in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system ad immediately initiate corrective action to return the required RHR train to operation.

H. During Mode 5 with reactor coolant loops not filled, the following specifications shall apply:

• The RHR pump may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 400F below saturation temperature.

• One RHR train may be inoperable for up to-2 hours for surveillance testing, provided the other RHR train is operable and in operation.

Amendment No. 77

10c

1. Two residual heat removal (RHR) trains shall be operable* and at least one RHR train shall be in operation**

2. With less than the above required RHR trains operable, immediately initiate corrective action to return the required RHR trains to Operable status as soon as possible.

3. With no RHR train in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system-and immediately initiate corrective action to return the required RHR train to operation.

Basis:

One pressurizer safety valve is sufficient to prevent over-pressurizing when the -reactor is subcritical, since its relieving capacity is greater than that required by the sum of the available heat sources, i.e., residual heat, pump energy and pressurizer heaters.

Prior to reducing boron concentration by dilution with make up water either a reactor coolant pump or a residual heat removal pump is specified to be in operation in order to provide effective mixing. During boron injection, the operation of a pump, although desirable, is not essential.

The boron is injected into an inlet leg of the reactor coolant loop.

Thermal circulation which exists whenever there is residual heat in the core and the reactor coolant system is filled and vented, will cause the boron to flow to the core.

Lack of further mixing cannot result in areas of reduced boron concentration within the core. Prior to criticality the two pressurizer safety relief valves are specified in service in order to conform to the system relief capabilities.(l)

• One RHR train may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR train is operable and in operation.

• • The RHR pump may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 400F below saturation temperature.

Amendment No. 77

• 1) ltd The plant is designed to have all three reactor coolant loops operational during normal power operation (Modes 1 and 2).

Under these conditions, the DNB ratio will not drop below 1.30 after a loss of flow with a reactor trip.( 2)(3) With one reactor coolant loop not in operation, this specification requires that the plant be in at least Hot Standby within one hour. However, except-ion is taken whenever reactor power is less than 10% of full power. Heat transfer analyses show that reactor heat equivalent to 8% of full power can be removed with natural circulation only; hence, for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the specified upper limit of 10% of. full power with 1 or 2 reactor coolant pumps operating provides a substantial safety factor.

In modes other than Modes 1 and 2, functional redundancy in the core heat removal methods (not necessarily system redundancy) i :specified to satisfy single failure considerations.

Functional redundancy, as applied to the San Onofre Unit 1 power plant, includes use of diverse heat removal methods.

Furthermore, single failure considerations apply only to active components.

In Mode 3, a single reactor coolant loop provides sufficient capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In Mode 4 and Mode 5-(reactor coolant loops filled), a single reactor coolant loop or RHR train provides sufficient capability for removing decay heat; but single failure

• considerations require that at least two methods (either RCS loop or RHR train) be OPERABLE.

In Mode 5 (reactor coolant loops not filled), a single RHR train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of any of the steam generators as a heat removing component, require that at least two RHR trains be OPERABLE.

The operation of one reactor coolant pump or one residual heat removal pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control (4).

Amendment No. 77

S-11

References:

(1) Final Engineering Report and Safety Analysis, Sections 9 and 10 (2) Final Engineering Report and Safety Analysis, Paragraph 10.2 (3) Supplement No. 1 to Final Engineering Report and Safety Analysis, Section 3, Question 9 (4) NRC letter dated June 11, 1980 from 0. G. Eisenhut to all operating pressurized water reactors.

Amendment No.

77

-38 3.8 Fuel Loadino and Refueling Applicability: Applies to fuel handling and refueling operations.

For the applicable surveillance requirements, see Table 4.1.2.

Obiective:

To prevent incidents during fuel handling operations that could affect public health and safety.

Specification: A. During refueling operations (Mode 6):

1. Radiation levels in the containment and spent fuel building shall be monitored.

2. Core..subcritical neutron flux shall be continuously monitored during the entire refueling period by not less than two neutron monitors, each with continuous visual indication and one with continuous audible indication.

3. For water levels in the refueling pool, greater than elevation 40 feet, 3 inches (See 7. below for reference elevation), the following specifications shall apply:

a. At least one of the following methods of decay heat removal shall be in operation and circulating reactor coolant at a flow rate of > 400 gpm:

(1) One RHR train.

(2) One refueling water pump taking suction from the refueling pooT through the recirculation heat exchanger (with supporting heat removal systems operating), and discharging via the safety injection system piping to one reactor coolant loop cold leg.

b. With less than one method of decay heat removal in operation, except as provided in c. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the reactor coolant system.

Immediately initiate corrective action to return the required decay heat removal method to operating status as soon as possible. In addition, within four hours, close all containment penetrations that provide direct access from the containment atmosphere to the outside atmosphere.

c. The decay heat removal capability may be removed from operation for up to one hour per eight hour period.

OCT 3 1 19gnendment No. 25.

t rS7

38a

4. Whenever the water level in the refueling pool is less than elevation 40 feet 3 inches [see (7) below for reference elevatTon], the following specifications shall.apply:

a. Two of the following methods of decay heat removal shall be:operable, and at least one shall be in operation circulating reactor coolant:

(1) Residual heat removal (RHR) pump G-14A with

.one RHR train.

(2) Residual heat removal (RHR) pump 6-14B with a second RHR train.

(3) One refueling water pump taking suction from the refueling pool through the recirculation heat exchanger (with supporting heat removal systems operating), and discharging via the safety injection system piping to one reactor coolant loop cold leg.

b. With less than the required methods of decay heat removal operable, immediately initiate corrective action to return the required methods of decay heat removal to operable status as soon as possible or to establish water level in the refueling pool of at least 40 feet 3 inches [see (7) below for reference elevation].

c. With none of the required methods of decay heat removal in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system. Immediately initiate corrective action to return the required decay heat removal method to operation. In addition, within four hours, close all containment penetrations that provide direct access from the containment atmosphere to the outside atmosphere.

5. During reactor vessel head removal and while loading and unloading fuel from the reactor, the more restrictive of the following reactivity conditions shall be mot:

a. A shutdown margin greater than 5% &k/k,

b. A boron concentration greater than or equal to 2,000 ppm.

6. The reactor shall be subcritical for at least 148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br /> prior to movement of irradiated fuel in the reactor pressure vessel.

Amendment No.

3?77 772

(1)(1 38b

7. Borated water to insure the shutdown margin as specified in Item A.(5) above shall be maintained to an elevation not less than 40 feet 3 inches in the refueling pool during movement of fuel assemblies and RCC's. Reference elevation is sea level, mean lower low water.

8. If any of the specified limiting conditions for refueling Is not met, refueling of the reactor shall cease, work shall be initiated to correct the violated conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be carried out.

B. With fuel assemblies in the spent fuel storage pool:

1. Loads in excess of 1,500 pounds shall be prohibited from travel over fuel assemblies in the storage pool.

2. Borated water to insure the shutdown margin as specified in Item A(5) above shall be maintained to an elevation not less than 40 feet 3 inches in the spent fuel storage pool.

Reference elevation is sea level, mean lower low water.

3. With the requirement of B(2) above not satisfied suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limits within four hours.

Basis:

During refueling the reactor refueling cavity is filled with approximately 240,000 gallons of borated water whose concentration is sufficient to maintain the reactor subcritical by greater than 5% Ak/k or to a boron concentration greater than or equal to 2,000 ppm, whichever is more restrictive. Operation of one method of decay heat removal is provided to assure continuous mixing flow of refueling water through the reactor vessel during the refueling period.(1)

Borated water injection capability is provided as per Specification 3.2 Part A in the unlikely event there is any need during the refueling period.

The requirement that at least one method of decay heat removal be in operation: ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140OF as required during refueling, and (ii) sufficient cooling circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

Amendment No.,3 6.?S 7 7 0

77

38c The requirement to have two methods of decay heat removal operable when the refueling pool water level is less than elevation 40 feet 3 inches ensures that a single failure of an operating component will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling; thus, in the event of a failure of an operating component, adequate time is available to initiate alternate means to-cool the core.(2)

In addition to the above safeguards, interlocks are utilized during refueling to.ensure safe handling.( 3) These include:

(1) An interlock on the lifting hoist to prevent lifting of more than one fuel assembly at any one time.

(2) The spent fuel transfer mechanism can accommodate only.

one fuel assembly at a time.

The restriction on movement of loads in excess of 1,500 pounds (i.e., the nominal weight of a fuel assembly, RCC, and associated handling tool) over fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (ii) any possible distortion of 'fuel in the storage racks will not result in a critical array. This assumption is consistent with activity release assumed in the accident analysis.

Requiring a minimum water elevation of 40 feet 3 inches in the refueling pool, and similarly in the spent fuel storage pool, ensures that (1) at least 23 feet of water would be available to remove 99% of the iodine gap activity assumed to be released in the event of a dropped and damaged fuel assembly, and (ii) there will be at least twelve feet of water above the top of the fuel rods of a withdrawn fuel assembly so as to limit dose rates at the top of the water in accordance with Section 4.2.6 of the facility FSA. Reference elevation is sea level, mean lower low water.

Finally, detailed written procedures are provided, and are carried out under close supervision by licensed personnel.

The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel assures that sufficient time has elapsed to allow the radioactive decay of short-lived fission products.

Amendment No. 6 7 7i

38d

References:

(1) Supplement No. 1 to Final Engineering Report and Safety Analysis, Section 5, Questions 8 and 9 (2) NRC letter dated June 11, 1980 from D. G. Elsenhut to all operating pressurized water reactors.

(3) Final Safety Analysis, Paragraph 2.9 Amendment No.. 77

-43c Check Freqjency

15.

Reactor Coolant

a. Per Technical Specifications

a. Once per 12 Loops/Residual 3.1.2.C and 3.1.2.D, in Mode 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Heat Removal and Mode 2 verify that Loops all required reactor coolant loops are tn operation and circulating reactor coolant.

b. Per Technical Specification 3.1.2.E, in Mode 3 verify

1. At least two required reactor 1. Once per 7 coolant pumps are operable days with correct breaker align ments and indicated power availability.

2. The steam generators

2. Once per 12 associated with the two hours required reactor coolant pumps are operable with secondary side water level

> 256 inches of narrow range on cold calibrated scale.

3. At least one reactor coolant

3. Once per 12 loop is in operation and hours circulating reactor coolant.

c. Per Technical Specification 3.1.2.F, in Mode 4 verify

1. At least two required (RC or

1. Once per 7 RHR) pumps are operable with days correct breaker alignments and indicated power availability.

2. The requtred steam

2. Once per generators are operable 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with secondary side water level > 256 inches of narrow range on cold calibrated scale.

3. At least one reactor coolant

3. Once per loop/RHR train is in operation 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and circulating reactor coolant.

d. Per Technical Specifications 3.1.2.G and 3.1.2.H, in Mode 5 verify, as applicable:

Amendment No. 77

43d Check Frequency

1. At least one RHR train is in 1. Once per operation and circulating 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> reactor coolant.

2. When required, one additional 2. Once per 7 RHR train is operable with days correct pump breaker alignments and indicated power availability.

3. When required, the secondary

3.

Once per 12 side water level of at least hours two steam generators is

> 256 inches of narrow range on cold calibrated scale.

e. Per Technical Specification

e. Once per 12 3.8.A.3, in Mode 6, with water hours level in refueling pool greater than elevation 40 feet 3 inches, verify that at least one method of decay heat removal is in operation and circulating reactor coolant at a flow rate of at least 400 gpm.

f. Per Technical Specification 3.8.A.4, in Mode 6, with water level in refueling pool less than elevation 40 feet 3 inches, verify

1. At least one decay heat

1. Once per 12 removal method is in hours operation and circulating reactor coolant.

2. One additional decay heat

.2. Once per 7 removal method is operable days with correct pump breaker alignments and indicated power availability.

Amendment No. 77