ML15223A732

Proposed Tech Specs 2.1-2,2.1-3,2.1-7,2.3-5,3.2-1,3.2-2, 3.3-1,3.3-2,3.3-3,3.3-4,3.3-5,3.3-6,3.5-9,3.5-10,3.5-15, 3.5-15a & b,3.5-18,3.5-18a,b,c,d & e,3.5-21,3.5-21a & B, 3.5-24,3.5-24a & b,3.8-2 & 3.8-3 Re Core Protection

ML15223A732
Person / Time
Site:	Oconee
Issue date:	05/29/1981
From:	DUKE POWER CO.
To:
Shared Package
ML15223A731	List: ML15223A730 ML15223A732 ML19352B163
References
NUDOCS 8106030243
Download: ML15223A732 (33)

Text

ATTACHMENT 2 DUKE POWER COMPANY OCONEE NUCLEAR STATION Proposed Technical Specification Revision Oconee 1, Cycle 7 Pages 2.1-2 3.5-15b 2.1-3 3.5-18 2.1-7 3.5-18a 2.3-5 3.5-18b 3.2-1 3.5-18c 3.2-2 3.5-18d 3.3-1 3.5-18e 3.3-2 3.5-21 3.3-3 3.5-21a 3.3-4 3.5-21b 3.3-5 3.5-24 3.3-6 3.5-24a 3.5-9 3.5-24b 3.5-10 3.8-2 3.5-15 3.8-3 3.5-15a II0 as1zo s3o

can be related to DNB through the use of the BAW-2 correlation (1). The BAW-2 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat. flux distributions. The local DNB ratio (DNBR), definedas the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. A DNBR of 1.30 corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating.conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1A represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power (112 percent).when four reactor coolant pumps are operating (minimum reactor coolant flow is 106.5 percent of 131.3 x 106 lbs/hr.). This curve is based on the combination of nuclear power peaking factors, with potential effects of fuel densification and rod bowing, which result in a more conservative DNBR than any other shape that exists during normal operation.

The curves of Figure 2.1-2A are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and rod.bowing:

1. The 1.30 DNBR limit produced by the combination of the radial peak, axial peak and position of the axial-peak that yields no less than a 1.30 DNBR.

2. The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 20.05 kw/ft for Unit .1.

Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the reactor power imbalance produced by the power peaking.

The specified flow rates for Curves 1, 2, and 3 of Figure 2.1-2A correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.

The curve of Figure 2.1-1A is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3A.

The magnitude of the rod bow penalty applied to each fuel cycle is equal to or greater than the necessary burnup independent DNBR rod bow penalty for the ap plicable cycle minus a credit of 1% for the flow area reduction factor used in the hot channel analysis (3).

All plant operating limits are presently based on an original method of cal culating rod bow penalties that are more conservative than those that would be obtained with new approved procedures (3). For Cycle 7 operation, this sub rogation results in a 10% DNBR margin, which is partially used to offset the reduction in DNBR due to fuel rod bowing.

2.1-2

The maximum thermal power for three-pump operation is 87.18 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.08 =

80.68 percent power plus the maximum calibration and instrument error. The maximum thermal power for other coolant pump conditions are produced in a similar manner.

For Figure 2.1-3A, a pressure-temperature point :above and to the left of the curve would result in a DNBR greater than 1.30.

References (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, March, 1970.

(2) Oconee 1, Cycle 4 - Reload Report - BAW-1447, March, 1977.

(3) Oconee 1, Cycle 7 - Reload Report - BAW-1660, March, 1981 2.1-3

THERMAL POWER- EVEL, %

120

(-28, 112)

ACCEPTABLE

(-45, 100) OPERATION- 100 (+45, 100) 28,87. 2)2

.(( a ~(28,87..2)

Z.~ACCEP TABL E 3&4 PUMP

-4574)OPERATION +4574

(-28,59.4) (28,59.4)

(-45,48) OPERATION

(+45,48

-- 40 20

-60 -40 -20 0 20 40 60 Reactor Power Imnalance, %

CURVE RC FLOW (GPM) 1 374,880 21 280,035 2 180,0 CORE PROTECTION SAFETY LIMITS UNIT 1 OCONEE NUCLEAR STATION FIGURE 2.1 - 2A 2.1-7

THERMAL POWER LEVEL,,%

120 (X,Y)

-15,1s) --- 1107, 807)

0. 90 4 PUMP M2 1.O OPERATIN 100 (36,8) 90 36, ) (-15,8so.7) 1(7,so.7)M

(-

3&4 PUMP OPERATION

-- 70

(-36,62) (36,62)

(17, 52.9)

(-15,-52.9) 50 2, 3 &

PUMP 30 IOPERATION . 40

-30

-20 II II

-10

-40 -20 0 20 40 Reactor Power Iinalance, %

PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UNIT I OCONEE NUCLEAR STATION FIGURE 2.3 - 2A 2.3-8

3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS Applicability Applies to the high pressure injection and the chemical addition systems.

Objective To provide for adequate boration under all operating conditions to assure ability to bring the reactor to a .cold shutdown condition.

Specification The reactor shall not be critical unless the following conditions are met:

3.2.1 Two high pressure injection pumps per unit are operable except as specified in 3.3.

3.2.2 One source perunit of concentrated soluble boric acid in addition to the borated water storage tank is available and operable.

This source will be the concentrated boric acid storage tank contain ing at least the equivalent of 1020 ft3 of 8700 ppm boron as boric acid solution with. a temperature at least 10aF above the crystalliza tion temperature. System piping and valves necessary to establish a flow path from the tank to the high pressure injection system shall be operable and shall have the same temperature requirement as the concentrated boric acid storage tank.. At least one channel of heat tracing capable of meeting the above temperature requirement shall be in operation. One associated boric acid pump shall be operable.

If the concentrated boric acid storage tank with its associated flow path is unavailable, but the borated water storage tank-is available and operable, the concentrated boric acid storage tank shall be re stored to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the reactor shall be placed in a hot shutdown condition and be borated to a shutdown margin equivalent to 1% Ak/k at 200'F within the next twelve hours; if the concentrated boric acid storage tank has not been restored to opera-.

bility within the next 7 days the reactor shall be placed in a cold shutdown condition within an. additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

If the concentrated boric acid storage tank is available but the borated water storage tank is neither available nor operable, the borated water storage tank shall be restored to operability within one hour or the reactor shall be placed in a hot shutdown condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in a cold shutdown condition within an addition al 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.2-1

Bases The high pressure injection system and chemical addition system provide con trol of the reactor coolant system boron concentration.(1) This is normally accomplished by using any of the three high pressure injection pumps in series with a boric acid pump associated with'either the boric acid mix tank or the concentrated boric acid storage tank. An alternate method of boration will be the use of the high pressure injection pumps taking suction directly from the borated water storage tank.(2)

The quantity of boric acid in storage in the concentrated boric acid storage tank or the borated water storage tank is sufficient-to borate the reactor coolant system to a 1% Ak/k subcritical margin at cold conditions (700 F) with the maximum worth stuck rod and no credit for xenon at the'worst time 'incore life. The current cycles for each unit, Oconee 1, Cycle 7, Oconee 2, Cycle 5, and Oconee 3, Cycle 6 were analyzed with the most limiting case selected as the basis for all three units. Since only the present cycles were analyzed, the specifications will be re-evaluated with each reload. A minimum of 1020 ft3 of 8,700 ppm boric acid in the concentrated boric acid storage tank, or a minimum of 350,000 gallons of 1835 ppm boric acid 'in the borated water storage tank (3) will satisfy the requirements. The volume requirements include a 10%

margin.and, in addition, allow for a deviation of 10 EFPD in the cycle length.

The specification assures that two supplies are available whenever the reactor is critical so that a single failure will not prevent boration to a cold con dition. The required amount of boric acid can be added in several ways. Using only one 10 gpm boric acid pump taking suction from the concentrated boric acid storage tank would .require approximately 12.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to inject the required.

boron. An alternate method of addition is to inject boric~ acid from the borated water storage tank using the makeup pumps.- The required boric acid can be injected in less than six hours using only one of the makeup Pumps.

The concentration of boron in the concentrated boric acid storage tank may be higher than the concentration which would crystallize at ambient conditions.

For this reason, and to assure a flow of boric acid is available when needed, these tanks and their associated piping will be kept at least 100F above the crystallization temperature for the concentration present. The boric acid concentration of 8,700 ppm in the concentrated boric acid storage tank cor responds to a crystallization temperature of 770 F and therefore a temperature requirement of 87oF. Once in the high pressure injection system, the concen trate is sufficiently well mixed and diluted so that normal system temperatures assure boric acid solubility.

REFERENCES (1) FSAR, Section 9.1; 9.2 (2) FSAR, Figure 6.2 (3) Technical Specification 3.3 3.2-2

3.3 EMERGENCY CORE COOLING, REACTOR BUILDING COOLING, REACTOR BUILDING SPRAY, AND LOW PRESSURE SERVICE WATER SYSTEMS Applicability Applies to the emergency core cooling, reactor building cooling, reactor building spray, and low pressure service water systems.

Objective To define the conditions necessary to assure immediate availability of the emergency core cooling, reactor building cooling, reactor building spray and low pressure service water systems.

Specification 3.3.1 High Pressure Injection (HPI) System

a. Prior to initiating maintenance on any component of the HPI system, the redundant component shall be tested to assure operability.

b. When the reactor coolant system (RCS), with fuel in the core, is in a condition with temperature above 350 0 F and reactor power less than 60% FP:

(1) Two independent trains, each comprised of an HPI pump and a flow path capable of taking suction from the borated water storage tank and discharging into the reactor coolant system automatically upon Engineered. Safeguards Protective System (ESPS) actuation (HPI segment) shall be operable.

(2) Test or maintenance shall be allowed on any component of the HPI system provided one train of the HPI system is operable.

If the UPI system is not restored to meet the requirements of Specification 3.3.1.b(1) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

If the requirements of Specification 3.3.1.b(1) are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the reactor shall be placed in a condition with RCS temperature below 350 0 F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. For all Units, when reactor power is greater than 60% FP:

(1) In addition to the requirements of Specification 3.3.1.b(l) above, the remaining HPI pump and valves 3HP-409 and 3HP-410 shall be operable and valves HP-99 and HP-100 shall be open.

(2) Tests or maintenance shall be allowed on any component of the HPI system, provided two trains of HPI system are operable.

If the inoperable component is not restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, reactor power shall be reduced below 60% FP within an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3.3-1

3.3.2 Low Pressure Injection (LPI) System

a. Prior to initiating maintenance on any component of the LPI system, the redundant component shall be tested to assure operability.

b. When the RCS, with fuel in the core, is in a condition with pressure equal to or greater than 350 psig or temperature equal to or greater than 250 0 F:

(1) Two independent LPI trains, each comprised of an LPI pump and a flowpath capable of taking suction from the borated water storage tank and discharging into the RCS automatically upon ESPS actuation (LPI segment), together with two LPI coolers and two reactor building emergency sump isolation valves (manual or remote-manual) shall be operable.

(2) Tests or maintenance shall be allowed on any component of the LPI system provided the redundant train of the LPI system is operable. If the LPI system is not restored to meet the re quirements of Specification 3.3.2.b(1) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the. requirements of Specification 3.3.2.b(1) are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 250'F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

333. Core Flood Tank (CFT) System When the RCS is in a condition with pressure above 800 psig both CFT's shall be operable with the electrically operated discharge valves open and breakers locked open and tagged; a minimum level of 13 t .44 feet (1040 t 30 ft.3) and one level instrument channel per CFT; a minimum concentration of borated water in each CFT of 1835 ppm boron; and pressure at 600+/- 25 psig with one pressure instrument channel per CFT.

3.3.4. Borated Water Storage Tank (BWST)

When the .RCS, with fuel in the core, is in a condition with pressure equal to or greater than 350 psig or temperature equal to or greater than 250 0'F

a. The BWST shall have operable two level instrument channels.

(1) Tests or maintenance shall be allowed on one channel of BWST level instrumentation provided the other channel is operable.

(2) If the BWST level instrumentation is not restored to meet the requirements of Specification 3.3.4..a above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the requirements of Specification 3.3.4.a are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 250 0 F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I 3.3-2

b. The BWST shall contain a minimum level of 46 feet of. water having a minimum concentration of 1835 ppm boron at a minimum temperature of 40 F. The manual valve, LP-28, on.the discharge line shall be locked open. If these requirements are not met, the BWST shall be considered unavailable and action initiated in accordance with Specification 3.2.

3.3.5 Reactor Building Cooling (RBC) System

a. -Prior to initiating maintenance on any component of the RBC system, the redundant component shall be tested to assure operability.

b. When the RCS, with fuel in the core, is in a condition with pressure equal to or greater.than 350 psig or temperature equal to or greaterthan 250 F and subcritical:

(1) Two independent RBC trains, each comprised of an RBC fan, associated cooling unit, and associated ESF valves.shall be operable.

(2) Tests or maintenance shall be allowed on any component of the RBC system provided on train of the RBC and one train of the RBS are operable.. If the RBC system is not restored to meet the requirements of Specification 3.3.5.b(l) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in'a.condition with RCS pressure below 350 psig and RCS temperature below 250 F with in an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. When the reactor is critical:

(1) In addition to the requirements of Specifications 3.3.5.b(l) above, the remaining RBC fan, associated cooling unit, and associated ESF.valves shall be operable.

(2 Tests or maintenance shall be allowed on one RB train under either of the following conditions:

(a) One RBC train may be out of.service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(b) One RBC.train may be out of service for 7 days provided both RBS trains are operable.

(c) If the inoperable RBC train is not restored to meet the requirements of Specification 3.3.5.c(l) within the time permitted by Specification 3.3.5.c(2)(a) or (b), the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the requirements of Specification 3.3.5.c(l) are not met within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350.psig and RCS temperature below 2500F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.3-3

3.3.6 Reactor Building Spray (RBS) System

a. Prior to initiating maintenance on any component of the RBS system, the redundant component shall be tested to assure operability.

b. When the RCS, with fuel in the core, is in a condition with pressure equal to or greater than 350 psig or temperature equal to or greater than 2500F and subcritical:

(1) One RBS train, comprised of an RBS pump and a flowpath capable of taking suction from the LPI system and discharging through the spray nozzle header automatically upon ESPS actuation (RBS segment) shall be operable.

(2) Tests or maintenance shall be allowed on any component of the RBS system under the following conditions:

(a) One RBS train may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided two RBC train are operable.

(b) If the inoperable RBS train is not restored to meet the requirements of Specification 3.3.6.b(l) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a condition with the RCS pressure below 350 psig and RCS temperature below 2500?

within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. When the reactor is critical:

(1) In addition to the requirements of Specifications 3.3.6.b(l) above, the other RBS train comprised of an RBS pump and a flowpath capable of taking suction of the 121 system and discharging through the spray nozzle header automatically upon ESPS actuation (RBS segment) shall be operable.

(2) Tests or maintenance shall be allowed on one RBS train under either of the following conditions:

(a) One RBS train may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(b) One RBS train may be out of service for 7 days provided all three RBC trains are operable.

(c) If the inoperable RBS train is not restored to meet the requirements of Specification 3.3.6.c(l) above within the time permitted by Specification 3.3.5.c(2)(a) or (b),

the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the requirements of Specification 3.3.6.c(l) are not met within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the. reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temp erature below-2500 within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.3-4

3.3.7 Low Pressure Service Water (LPSW)

a. Prior to initiating maintenance on any component of the LPSW system, the redundant component shall be tested to assure operability.

b. When the RCS, with fuel in the core, is in a condition with pres sure equal to or 0 greater than 350 psig or temperature equal to or greater than 250 F (1) Two LPSW pumps for the shared Unit 1, 2 LPSW system and two LPSW pumps for the Unit 3 LPSW system shall be operable with valves LPSW-108, 2LPSW-108, and 3LPSW-108 locked open.

(2) Tests or maintenance shall be allowed on any component of the LPSW system provided the redundant train of the LPSW system is operable. If the LPSW system is not restored to meet the requirements of Specification 3.3.7.b(l) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the requirements of Specification 3.3.7.b(l) are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following. hot .shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig 0

and RCS temperature below 250 within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Bases Specification 3.3 assures.that, for whatever.condition the reactor coolant system is in, adequate engineered safety feature equipment are operable.

For operation up to 60% FP, two high pressure injection pumps are specified.

Also, two low pressure injection pumps and .both core flood tanks are required.

In the event that the need for emergency core cooling.should occur, func tioning of one high pressure injection pump, one low pressure injection pump, and both core flood tanks will protect the core, and in the event of a main coolant loop severence, limit the peak clad temperature to less than 2,2000 and the metal-water reaction to that representing less than 1 percent of the clad. (1) Both core flooding tanks are required as a single core flood tank has insufficient inventory to reflood the core.

The requirement to have three HPI pumps and two HPI flowpaths operable during power operation above 60% FP is based on considerations of potential small breaks at the reactor coolant pump discharge piping for which two HPI trains (two pumps and two flow paths) are required to assure adequate core cooling.(2)

The analysis of these breaks indicates that for operation at or below 60% FP only a single train of the HPI system is needed to provide the necessary core cooling.

The borated water storage tanks are used for two purposes:

(a). As a supply of borated water for accident conditions.

(b) As a supply of borated water for flooding the fuel transfer canal during refueling operation.(3) 3.3-5

Three-hundred and fifty thousand (350,000) gallons of borated water ( a level of 46 feet in the BWST) are required to supply emergency core cooling and reactor building spray in the event of a loss-of-core cooling accident. This amount fulfills requirements for emergency core cooling. The borated water storage tank capacity of 388,000 gallons is based on refueling volume require ments. Heaters maintain the borated water supply at a temperature to prevent freezing. The boron concentration is set at the amount of boron required to maintain .the core 1 percent subcritical at 700F without any control rods in the core. The minimum value specified in the tanks is 1835 ppm boron.

It2 has been shown for the worst design basis loss-of-coolant accident ( a 14.1 ft hot leg break).that the.Reactor Building design pressure will not be exceeded with one spray and two coolers operable. (4) Therefore, a maintenance period of seven days is acceptable for one Reactor Building cooling fan and its associated.cooling unit provided two Reactor Building spray systems are operable for seven.days or one Reactor Building spray system provided all three Reactor Building cooling units are operable.

Three low pressure service water pumps serve Oconee Units 1 and 2 and two low pressure service water pumps serve Oconee Unit 3. There is a manual cross connection on the supply headers for Units 1, 2, and 3. One low pressure service water pump per unit is required for normal operation. The normal oper ating requirements are greater than the emergency requirements following a loss-of-coolant accident.

Prior to. initiating maintenance on any of the components, the redundant component (s) shall be tested to assure operability. Operability shall be based on the results of testing as required by Technical Specification 4.5.

The maintenance period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable if the operability of equipment redundant to that removed from service is demonstrated immedi ately prior toremoval. The basis of acceptability is a likelihood of fail ure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following such demonstration.

REFERENCES (1) ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Babcock &

Wilcox, 'Lynchburg, Virginia, June 1975.

(2) Duke Power Company to NRC letter, July 14, 1978, "Proposed Modifications of High Pressure Injection System".

(3) FSAR, Section 9.5.2 (4) FSAR, Supplement 13 3.3-6

f. If the maximum positive quadrant power tilt exceeds the Maximum Limit of Table 3.5-1, the reactor shall be shut down within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the ther mal power and the Nuclear Overpower Trip Setpoints allowable for the reattor coolant pump combination are restricted by a reduc tion of 2% of thermal power for each 1% tilt for the maximum tilt observed prior to shutdown.

g. Quadrant power tilt shall be monitored on a minimum frequency of once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 15% full power.

3.5.2.5 Control Rod Positions

a. Technical Specification 3.1.3.5 does not prohibit the exercising of individual safety.rods as required by Table 4.1-2 or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.

b. Except for physics tests, operating rod group overlap shall be 25% +/- 5% between two sequential groups. If this limit is ex ceeded, corrective measures shall be.taken immediately to achieve an acceptable overlap. Acceptable overlap shall be attained within two hours or the reactor shall be placed in a hot shutdown condition within an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. Position limits are specified for regulating and axial.power shap ing control rods. Except for physics tests or exercising control rods, the regulating control rod insertion/withdrawal limits are, specified on figures 3.5.2-1A1, 3.5.2-1A2, and 3.5.2-1A3 (Unit 1);

3.5.2-1B1, and.3.5.2-1B2 (Unit-2); 3.5.2-II, 3.5.2-1C2 and

• 3.5.2-1C3 (Unit 3) for four pump operation, and on figures 3.5.2-2A1, 3.5.2-2A2, and 3.5.2-2A3 for three pump operation and 3.5.2-2A4, 3.5.2-2A5, and 3.5.2-2A6 for two pump operation (Unit 1); 3.5.2-2B1, and 3.5.2-2B2 (Unit 2); and 3.5.2-2CI, 3.5.2-2C2 and 3.5.2-2C3 (Unit 3) for two or three pump operation.

Also, excepting physics tests or exercising control rods, the axial power shaping control rod insertion/withdrawal limits are specified on figures 3.5.2-4A1, and 3.5.2-4A2 (Unit 1); 3.5.2-4BI1 and 3.5.2-4B2, (Unit 2); 3.5.2-4G1, 3.5.2-4C2,.and 3.5.2-4C3 (Unit 3).

If the control rod position limits are exceeded, corrective mea sures shall be taken immediately to achieve an acceptable control rod position. An acceptable control.rod.position shall then be attained within two hours. -The minimum shutdown margin required by Specification 3.5.2.1 shall be maintained at all times.

3.5-9

3.5.2.6 Xenon Reactivity Except for physics tests, reactor power shall not be increased above the power level-cutoff shown in Figures 3.5.2-lAl, 3.5.2-lA2, and 3.5.2-1A3 for Unit 1; Figures 3.5.2-1B1, and 3.5.2-1B2, for Unit 2; and Figures 3.5.2-1i1, 3.5.2-1C2, and 3.5.2-1C3 for Unit.3 unless one of the following conditions is satisfied:.

1. Xenon reactivity did not deviate more than 10 percent from the equilibrium value for operation at steady state power.

2. Xenon reactivity deviated more than 10 percent but is now within 10 percent of the equilibrium value for operation at steady state rated power and has passed its final maximum or minimum peak during its approach to its equilibrium value for operation at the power level cutoff'.

3. Except for xenon free startup (when 2. applies), the reactor has operated within a range of 87 to 92 percent of rated thermal power for a period exceeding. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

3.5.2.7 Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power.

Except for physics tests, imbalance shall be maintained within the envelope defined by Figures 3.5.2-3Al, 3.5.2-3A2, 3.5.3-3B1, 3.5.2-3C1, 3.5.2-3C2, and 3.5.2-3C3. If the imbalance is not within the envelope -defined by these figures, corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within two hours, reactor power shall be reduced until imbalance limits are met.

3.5.2.8 The control rod'drive patch panels shall be locked at all times with limited access to be authorized by the manager or his designated alternate..

3.5.2.9 The operational limit-curves of Technical Specifications 3.5.2.5.c and 3.5.2.7 are valid for a nominal design cycle length, as defined in the Safety Evaluation Report for the appropriate unit and cycle.

Operation beyond the nominal design cycle length. is permitted provided that an evaluation is.performed to verify that the operational limit curves are valid for extended operation.' If the operational limit curves are not valid for the extended period of the operation, appropriate limits will be established and the Technical Specification curves will be modified as required.

3.5-10

(144. 5 102) (284, 102) (300, 102) 100 (281,92)

POWER LEVEL 80 (271,80) . CUTOFF OPERATION RESTRICTED 100% FP L..

60 SHUTDOWN MARGIN LIMIT (88, 50) (200, 50.

40 OPERATION NOT ALLOWED OPERATI ON ACCEPTABLE 20

• (37, 15). (90 15)

(0, 11. 8)

(0. 2. 5)1 0

0 100 200 300 GR 5 I Rod Index (W tndrawn) 0 75 100 0 25 100 0 25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 0 to 50 (+10, -0) EFPD UNIT 1 OCONEE NU.CLEAR STATiON FIGURE 3.5.2 1Al

(144.5, 102 . (277.5,102) . (300,102) 100 (214.5,92)

POWER LEVEL (264.5,80) . CUTOFF 100% FP OPERATION RESTRICTED SHUTDOWN MARGIN t3 60 LIMIT (88,50) (200,50)

OPERATION OPERATION ACCEPTABLE NOT ALLOWED 40 20 (37, 15)(915 (0,11. 8)

(0.2.5)0 0 100 200 300 GR 5 1 Rod 1ndex (witlidrawn)

GR 5 0GR 6 25 I1I G p 105lo 0 25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 50 (+10, -0) to 200 + 10 EFPD UNIT 1 OCONEE NUCLEAR STATOON FIGURE 3.5.2 - 1A2

(220, 102) 274.5,102) O300, 100 (271,92)

POWER LEVEL 80OPERATION NOT ALLOWED (264.580) CUTOFF 100% FP 80 (6.,0)

OPERATION RESTRICTED 60 7M (160,50) (20050)

SHUTDOWN MARGIN 40 LIMIT OPERATION ACCEPTABLE C3 ul Q a- 20 (82, 15) (0 15 (0,8.5)

(0,2.5 0 100 300 GR 5 Rod Index (witnarawn) 0 15 10o GR II 0 25 5R111 0 0 25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION AFTER 200 + 10 EFPD UNIT 1 OCONEE NUCLEAR STATION FIGURE 3.5.2 - 1A3

100 OPERATION NOT ALLOWED 80

a. 1451 cu OPERATION RESTRICTED 60 SHUTDOWN MARGIN (200, 50)

L IMIT 40 (88,38)

OPERATION ACCEPTABLE 20 (37_1 75) (9O, 15)

(0,9.35)

(0,2.5 0 100 200 300 GR 5 II I Rod Index (wi tndrawn) 0 GR 6 1PO 0 25 GR7 5 PO 0 25 100 ROD POSITION LIMITS FOR THREE PUMP OPERATION FROM 0 to 50 (+10, -0) EFPD UNIT 1 OCONEE NUCLEAR STAT50N FIGURE 3.5.2 - 2Al

@100 80 a 144C 5377 (258, 77) *(300, 77)

OPERATION NOT ALLOWED OPERATION RESTRICTED 60 SHUTDOWN MARGIN LIMIT (20050) 40 (88,38) OPERATION ACCEPTABLE 02 20 (37 11.75)

(9015)

(0,9. 35)

(0.2.5) 0 100 200 300 G

CR 5 0 Rod nex (witndrawn) 0 75 100 O 25 75 100 G7I o 25 1oo ROD POSITION LIMITS FOR THREE PUMP OPERATION FROM 50 (+10, -0) to 200 +.10 EFPD UNIT 1 OCONEE NUCLEAR STATION FIGURE 3.5.2 - 2A2

100 80(220,17 (258777)

OPERATION NOT ALLOWED.

!B RESTD 60 SHUTOWN MARGIN (200,50)

LIMIT 2 40 (160,38) OPERATION ACCEPTABLE CD(133,29)

Lu

~20 OPERATION RESTRICTED (82, 11.75)

(0,6.875) (54,10)

(0,2. 5) 0 100 200 300 IRon naex (witndrawn)

GR 5 0 5 GRR 5 I 100 GR6 25 75 100 GR I I 0 25 100 ROD POSITION LIMITS FOR THREE PUMP OPERATION AFTER 200 + 10 EFPD UNIT 1 OCONEE NUCLEAR STATION FIGURE 3.5.2 - 2A3

@10o 0 80 0) 0L) 60 OPERATION NOT ALLOWED (300,52) i=

ca(144.5, 52).

2100, 50)

SHT0 PERATION 40 MARGIN LIMIT RESTRICTED (88, 26) 00 OPERATION ACCEPTABLE 20 (90, 15)

(0.6.9)

(0,2.5)I 0 100 200 300 GR 5 IRod Index (witndrawn) 04 7 1 R 0 25 75 100 GR 7 0 25 100 ROD POSITION LIMITS FOR TWO PUMP OPERATION FROM 0 to 50 (+10, -0) EFPD UNIT 1 k, 0 a OCONEE NUCLEAR STATION FIGURE 3.5.2 - 2A4

100 cu 80 0

OPERATION 60 NOT ALLOWED 4.-o (144. 5, 52) (204, 52) (300, 52)

(200, 50)

SHUTDOWN PERAT ION CD40 MAGNRESTRICTED c LIMIT OPERATION cu(88, 26).

C3 ACCEPTABLE

~ 200 (90,15)

(0.6.9)

(0,2. 5) 0 100 200 300 Rod Index (wItntrawn) 0 75 100 GR 6 0 25 GR 7 5 90 0 25 100 ROD POSITION LIMITS FOR TWO PUMP OPERATION FROM 50 (+10, -0) to 200 + 10 EFPD UNIT 1 OC ONEE NUCLEAR STATION FIGURE 3.5.2 - 2A5

@00 80 C3

• 1o OPERATION NOT ALLOWED 60 (220, 52)

(300, 52)

SHUTDOWN MARGIN

- 40 LIMIT Ln (160, 26) OPERATION 20 ACCEPTABLE a- 20 -OPERATION RESTRICTED (285 (0.5 25)

(0, 2.5) I 0 0 00 200 300 Index (wi tndrawn) 0Rod GR5 0 75 100 J GR 6 0 25 5 100 0 25 100 ROD POSITION LIMITS FOR TWO PUMP OPERATION AFTER 200 + 10 EFPD UNIT 1 OCONEE NUCLEAR STATION FIGURE 3.5.2 - 2A6

OPERATION RESTRICTED

( 14, 102) 100 (17.5, 102) 15,92) (17. 5,92)

(-25,80) 80 (20,80)

OPERATION 60 ACCEPTABLE 40 3c 20

-30 -20 -10 0 10 20 Axial Power Imnal ance (%)

POWER IMBALANCE LIMITS FOR OPERATION FROM 0 to 50 (+10, -0) EFPD UNIT I 00KEPOWE OCONEE NUCLEAR STATION FIGURE 3.5.2 - 3A1 3.5-21

OPERAT ION RESTRI CTED (75 102) (17.5, 102) 9,92) * (17.5,92)

30. 5 80 c80 (20,80)

OPERATION 60 ACCEPTABLE 40 20

-30 -20 -10 0 10 20 Axial Power Imoalance (%)

POWER IMBALANCE LIMITS FOR OPERATION FROM 50 (+10, -0) to 200 + 10 EFPD UNIT 1 DEPWE OCONEE NUCLEAR STATION FIGURE 3.5.2 - 3A2 3.5-21a

OPERATION RESTRICTED

(-23.5, 102) 5 15 02) 100

(-24.5,92) . . (17,92)

(-31,80) *8 3 80 c80 - 20, 80) 2.O 60 OPERATION ACCEPTABLE 40 20

-30 -20 -10 0 0 20 Axial Power Imnalance(%

POWER LMBALANCE LIMITS FOR OPERATION AFTER 200 + 10 EFPD UNIT 1 OCONEE NUCLEAR STATION FIGURE 3.5.2 -3A3 3.5-21b

(8 5.102) (35, 102) l00

.5,92) (35. 92) OPERATION RESTRI CTEO 80 (0,80)

S 60 OPERATION ACCEPTABLE 20 0

0 20 40 60 80 100 APSR Position (Percent Witnorawn)

APSR POSITION LIMITS FOR OPERATION FROM 0 to 50 (+10, -0) EFPD UNIT 1 OCONEE NUCLEAR STATION FIGURE 3.5.2 - 4Al 3.5-24

(8.5,102) (32. 5,102) 100 OPERATION RESTRICTED 0o (080 (41,80) 60 -6

-- OPERATION ( 00-50)

ACCEPTABLE

- 40 20 0'

0 20 40 80 80 100 APSR Position (%witnarawn)

APSR POSITION LIMITS FOR OPERATION FROM 50 (+10, -0) to 200 + 10 EFPD UNIT 1 OCONEE NUCLEAR STATION FIGURE 3.5.2 -4A2 3.5-24a

(8.5, 102) (34.5, 102) 100

.(8. 5,92) *(36,92) OPERATION RESTRICTED 8o (0,80) (42. 5,80) 60 OPERATION F (00T50)

ACCEPTABLE 0 4 0) 20 0 20 40 80 80 - 100 APSR Position (% witnerawn)

A.PSR POSITION LIM4ITS FOR OPERATION AFTER 200 + 10 EFPD UNIT I OCONEE UKEPOWER NUCLEAR STATION FIGURE 3.5.2 - 4A3 3.5-24b

3.8.9 If any of the above specified limiting conditions for fuel loading and refueling are not met, movement of fuel into the reactor core shall cease; action shall be initiated to correct the conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be made.

3.8.10 The reactor building purge system, including the radiation monitor, RIA-45, which initiates purge isolation, shall be tested and verified to be operable immediately prior to refueling operations.

3.8.11 Irradiated fuel shall not be moved from the reactor until the unit has been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

3.8.12 Two trains of spent fuel pool ventilation.shall be operable with the following exceptions:

a. With one train of spent fuel pool ventilation inoperable, fuel movement within the storage pool or crane operation with loads over the storage pool may proceed provided the operable spent fuel pool ventilation train is in operation and discharging through the Reactor Building purge filters.

b. With no spent fuel pool ventilation filter operable, suspend all operations involving movement of fuel within the storage pool or crane operations with loads over the storage pool until at least one train of spent fuel pool ventilation is restored to operable status.

3.8.13 a. Prior to spent fuel cask movement in the Unit 1 and 2 spent fuel pool, spent fuel stored in the first 36 rows of the pool closest to the spent fuel cask handling area shall be decayed a minimum of 55 days.

b. Prior to spent fuel cask movement in the Unit 3 spent fuel pool, spent fuel stored in the first 20 rows of the pool closest to the spent fuel cask handling area shall be decayed a minimum of 43 days..

3.8.14 No suspended loads of more than 3000 lb m shall be transported over spent fuel stored in either spent fuel pool.

3.8.15 a. N2*guel which has an.enrichment greater than 3.5'weight percent U (46 grams of U2 35per axial centimeter of fuel assembly) will be stored in the spent fuel pool for Units 1 and 2.

b. N ;uel which has a 3 nrichment greater than 4.3 weight percent U (57 grams of U per axial centimeter of fuel assembly)

Bases Detailed written procedures will be available for use by refueling personnel.

These procedures, the above specifications, and the design of the fuel hand ling equipment as described in Section 9.7 of the FSAR incorporating built-in interlocks and safety features, provide assurance that no incident could occur 3.8-2

during the refueling operations that would result in a hazard to public health and safety. If no change is being made in core geometry, one flux monitor is sufficient. This permits maintenance on the instrumentation.

Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The low pressure injection pump is used to maintain a uniform boron concentration. (1) The shutdown margin indicated in Specification 3.8.4 will keep the core subcritical, even with all control rods withdrawn from the core. (2) The boron concentration will be maintained above 1835 ppm. Although this concentration is sufficient to maintain the core k (0.99 if all the control rods were removed from the core, only a few con e ff =

trol rods will be removed at any one time during fuel shuffling and replace ment. The k with all rods in the core and with refueling boron concen eff tration is approximately 0.9. Specification 3.8.5 allows the control room operator to inform the reactor building personnel of any impending unsafe condition detected from the main control board indicators during fuel movement.

The specification requiring testing of the Reactor Building purge isolation is to verify that these components will function as required should a fuel hand ling accident occur which resulted in the release of significant fission products.

Specification 3.8.11 is required, as the safety analysis for the fuel handling accident was based on the assumption that the reactor had been shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.(3)

The off-site doses for the fuel handling accident are within the guidelines of 10CFR100; however, to further reduce the doses resulting from this acci dent, it is required that the spent fuel pool ventilation system be operable whenever the possibility of a fuel handling accident could exist.

Specification 3.8.13 is required as the safety analysis for a postulated cask handling accident was based on the assumptions that spent fuel stored as indicated has decayed for the amount of time specified for each spent fuel pool.

Specification 3.8.14 is required to prohibit transport of loads greater than a fuel assembly with a control rod and the associated fuel handling tool(s).

REFERENCES (1) FSAR, Section 9.7 (2) FSAR, Section 14.2.2.1 (3) FSAR, Section 14.2.2.1.2 II 3.8-3

ATTACMENT 3 B&W Reload Report Oconee 1 Cycle 7