ML16138A584

Amends 155,155 & 152 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Amending Tech Specs to Support Operation of Unit 3 at Full Power During Cycle 10 & to Provide More Conservative Curve for Unit 1 Power Imbalance Envelope

ML16138A584
Person / Time
Site:	Oconee
Issue date:	03/19/1987
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Duke Power Co
Shared Package
ML16138A585	List: ML16138A584 ML16138A586
References
DPR-38-A-155, DPR-47-A-155, DPR-55-A-152 NUDOCS 8703300187
Download: ML16138A584 (16)

Text

0C UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

155 License No. DPR-38

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

A. The application for amendment by Duke Power Company (the licensee) dated December 12, 1986, as revised on January 29, 1987 and supplemented on February 11, 1987, 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 1; 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;

0. 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.3 of Facility Operating License No. DPR-38 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No.

155, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

8703300187 870319 PDR-ADOCK 05000269.

P PDR

3. This license amendment is effective as of the date of its issuance.

FOP THE NUCLEAR REGUIATORY COMMISSION

/oh F. Stol z, Director, fPW Proiect Directorate 46 li-ision of PWR Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 19, 1987

o0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLFAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 155 License No. DPR-47

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

A. The application for amendment by Duke Power Company (the licensee) dated December 12, 1986, as revised on January 29, 1987, and supplemented on February 11, 1987, 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. 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 Facility Operating License No. DPR-47 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 155, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance.

FOP THE NUCLEAR REGULATOPY COMMISSION h -f. Stolz, Director,"'

Project Directorate #6 Division of PWR Licensing-B

Attachment:

Changes to the Technical Specifications Date of Issuance: March 19, 1987

o UNITED STATES o

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 152 License No. DPR-55

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

A. The application for amendment by Duke Power Company (the licensee) dated December 12, 1986 as revised on January 29, 1987, and supplemented on February 29, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and requlations 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; add 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. 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 Facility Operating License No. DPR-55 is hereby amended to.read as follows:

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 152, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

-2

3. This license amendment is effectivP as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Jo

f.

Stolz, Directb r Project Directo te #6

-fivision of PWP licensing-B

Attachment:

Changes to the Technical Specifications Date of Issuance: March 19, 1987

ATTACHMFNT TO LICENSE AMENDMENTS AMENDMENT NO.

155 TO DPR-38 AMENDMENT NO.

155 TO DPR-47 AMENDMENT NO.

152 TO DPR-55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment numbers and contain vertical lines indicating the area of change.

Remove Pages Insert Pages 3.2-2 3.2-2 3.3-3 3.3-3 3.3-6 3.3-6 3.5-8 3.5-8 3.5-10 3.5-10 3.5-12 3.5-12 3.5-24 3.5-24 3.5-26 3.5-26 3.8-3 3.8-3

Bases The high pressure injection system and chemical addition system provide control of the reactor coolant system boron concentration.(l)

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 (70'F) with the maximum worth stuck rod and no credit for xenon at the worst time in core life. The current cycles for each unit 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 ftl of 11,000 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 condition. 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 1s 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 11,000 ppm in the concentrated boric acid storage tank corresponds to a crystallization temperature of 886F and therefore a temperature requirement of 98 0F.

Once in the high pressure injection system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures assure boric acid solubility.

REFERENCES (1) FSAR, Sections 9.3.1, and 9.3.2 (2) FSAR, Figure 6.0.2 (3) Technical Specification 3.3

• 2010 ppm boron for Unit 3, Cycle 10 only.

Amendment No. 155, 155, 152 3.2-2

b.

The BWST shall contain a minimum level of 46 fee-of water having a minimum concentration of 1835**ppm boron at a minimum temperature of 500'.

The manual valve, L?-28, on the discharge line shall be locked open.

If these requirement:3 are not met, the BWST shall be considered unavailablc 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 greater than 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 one 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(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 condition with RCS pressure below 350 psig and RCS temperature below 250aF 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(1) above, the remaining RBC fan, associated cooling unit, and associated ESF valves shall be operable.

(2) Tests or maintenance shall be allowed on one RBC 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 RBC trains are operable.*

(c) If the inoperable RBC train is not restored to meet the requirements of Specification 3.3.5.c(1). 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(1) 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 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 />.

• For the "3A" RBC train, aone-time extension of inoperability is granted in order to allow for.repair, provided both RBS trains are operable and that the "3A" RBC train is returned to service no later thai 11:59 p.m., April 20, 1985.

2010 ppm boron for Unit 3, Cycle 10 only.

Amendment No. 155,.155, 152 3.3-3

Three-hundred and fifty thousand (350,000) gallons of borated wa-tt!r a level of 46 feet in the BWST) are required to supply emergency core coo.Ling 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 requirements. Heaters maintain the borated water supply at a temperature above 500? to lessen the potential for thermal shock of the reactor vesuel during high pressure injection system operation. The boron concentration is set at the amount of boron required to maintain the core 1 percent subcritical at 70aF without any control rods in the core. The minimum value specified in the tanks is 1835* ppm boron.

It has been shown for the worst design basis loss-of-coolant accident (a 14.1 ft2 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 oper able 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 I and 2 and two low pressure service water pumps serve Oconee Unit 3. There is a manual cross connection on the supply headers for Unit 1, 2, and 3. One low pressure service water pump per unit is required for normal operation. The normal operating requirements are greater than the emergency requirements following a loss-of-coolant accident.

Prior to initiating maintenance on any of the components, the redundant compo nent(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 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> -prior to removal.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period prior to removal is adequate to permit efficient scheduling of manpower and equipment testing while ensuring that the testing is performed directly prior to removal.

The basis of accept ability is the low likelihood of failure within a clearly defined 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following redundant componepat testing.

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 9.3.3.2 (4)

FSAR, Section 15.14.5

• 2010 ppm boron for Unit 3, Cycle 10 only.

3.3-6

1.

Either the quadrant power tilt shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to within its Steady State Limit or,

2.

The reactor thermal power shall be reduced below 100% full power by 2% thermal power for each 1% of quadrant power tilt in excess of the Steady State Limit, and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be reduced within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by 2% thermal power for each 1% tilt in excess of the Steady State Limit.

If less than four reactor coolant pumps are in operation, the allowable thermal power for the reactor coolant pump combination shall be reduced by 2% for each 1% excess tilt.

c.

Quadrant power tilt shall be reduced within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to within its Steady State Limit or,

1.

The reactor thermal power shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable power for the re actor coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be reduced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combination.

d.

If the quadrant power tilt exceeds the Transient Limit but is less than the Maximum Limit of Table 3.5-1 and if there is a simultaneous indication of a misaligned control rod then:

1. Reactor thermal power shall be reduced within 30 minutes at least 2% for each 1% of the quadrant power tilt in excess of the Steady State Limit.

2.

Either quadrant power tilt shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to within its Transient Limit or,

3.

The reactor thermal power shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable power for the reactor coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be reduced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combination.

e.

If the quadrant power tilt exceeds the Transient Limit but is less than the Maximum Limit of Table 3.5-1, due to causes other than simultaneous indication of a misaligned control rod then:

1. Reactor thermal power shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable power for the 'reactor coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combina tion.

Amendment No. 155, 155, 152 3.5-8

3.5.2.6 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 3.5.2-10 (Unit 1).

If the imbalance 3.5.2-11 (Unit 2) 3.5.2-12 (Unit 3) 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.

3.5.2.7 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.

3.5.2.8 The operational limit curves of Technical Specifications 3.5.2.5.c and 3.5.2.6 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 pro vided 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 Specifica tion curves will be modified as required.

Amendment No.

155, -155,-152 351

The rod position limits are based on the most limiting of the following three criteria:

ECCS power peaking, shutdown margin, and potential ejected rod worth.

Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out position (1).

The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.65% Lk/k at rated power.

These values have been shown to be safe by the safety analysis (2,3,4,5) of hypothetical rod ejection accident. A maximum single inserted control rod worth of 1.0%

k/k is allowed by the rod position limits at hot zero power. A single inserted control rod worth of 1.0% 6k/k at beginning-of-life, hot zero power would result in a lower transient peak thermal power, and, therefore, less severe environmental consequences than a 0.65%Ak/k ejected rod worth at rated power.

Control rod groups are withdrawn in sequence beginning with Group 1. Groups 5, 6, and 7 are overlapped 25 percent.

The normal position at power is for Group 7 to be partially inserted.

The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established to prevent the linear heat rate peaking increase associated with a positive quadrant power tilt during normal power operation from exceeding:

7.50% for Unit 1. The limits shown in Specification 3.5.2.4 7.50% for Unit 2, 7.50% for Unit 3 are measurement system independent. The actual operating limits, with the appropriate allowance for'observability and instrumentation errors, for each measurement system are defined in the station operating procedures.

The quadrant tilt and axial imbalance monitoring in Specification 3.5.2.4 and 3.5.2.6, respectively, normally will be performed in the process computer.

The two-hour frequency for monitoring these quantities will provide adequate surveillance when the computer is out of service.

Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours without specification violation. Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.

Operating restrictions resulting from transient xenon power peaking are implicitly included in the limits of Section 3.5.2.5 (control rod positions) and 3.5.2.6 (reactor power imbalance).

Since these limits are set during the cycle-specific maneuvering analysis to prevent excessive power peaking by transient xenon at all power levels, there is no need for any hold at a power level cutoff below 100% FP.

Amendment No. 155, 155, 152 3.5-12

REACTOR POWER,%FP

(-17.0,102.0) 1 (27.0,102.0) 100)

ACCEP ABLE

(-27.0,90.0)

OPER TION 80 60 RESTRICTED OPERATION RESTRICTED OPERATION 40 20

-100

-80

-60

-40

-20 0

20 40 60 80 100 IMBALANCE.%

OPERATIONAL POWER IMBALANCE ENVELOP FIRO 0-FPO :TO Enr "NIT 1 ma COEE NUCLEAR STATIAT Amendment No. 155, 155, 152 3.52 FIGURE 3.5.2-10

REACTOR POWER,%FP

(-17.0,102.0)

(27.0,102.0)

-100 ACCEPI ABLE

(-27.0,90.0)

OP OPER.\\TION

-80 60 RESTRICTED OPERATION RESTRICTED OPERATION 40 20

-100

-80

-60

-40

-20 0

20 40 60 80 100 IMBALANCE.%

OPERATIONAL POWER IMBALANCE ENVELOP

ROM 0 EFPD TO EOC UNIT 3 OCONEE NUCLEAR -STAT!0N Amendment No. 155, 155, 152 3.5-26 Figure 3.5.2-12

These procedures, the above specifications, and the design of the fuel handling equipment as described in Section 9.1.4 of the FSAR incorporating built-in interlocks and safety features, provide assurance that no incident could occur 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 purp 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 eff control 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 concentra eff tion is approximately 0.90. 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 houre.(3)

The off-site doses for the fuel handling accident are within the guidelines of 10 CFR 100; however, to further reduce the doses resulting from this accident, 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 9.1.4 (2) FSAR, Section 15.11.1 (3) FSAR, Section 15.11.2.1

• 2010 ppm boron for Unit 3, Cycle 10 only.

Amendment No. 155, 155, 152 3.8-3