ML20212H636

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Proposed Tech Specs Requiring Two Reactor Coolant Loops to Be in Operation During Mode 3
ML20212H636
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 02/27/1987
From:
TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20212H634 List:
References
NUDOCS 8703060172
Download: ML20212H636 (7)
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i REACTOR COOLANT SYSTEM etf.[gui-M WO roebr toolo n locys 'm opeptic^ vAen M ST M yk,/Qe;er 3 c4 Spees,, breaker.t ace Clcssi stol ai" Me. reacho( coo \ct s {co p g c c @ n; 6eA %e.

LIMITING CONDITION FOR' OPERATION 3eador Go Q+w br(edec Cre cM.n '

, C , E' Reactor Coolant Loop 'A and its associated steam generetor and reactor coolant pump, 2' Reactor coolant Loop B and its associated steam generator b* and reactor coolant pump, o

J: Reactor Coolant Loop C aM its associated steam generator and reactor coolant pump, gl,N' Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.

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APPLICA8ILITY: MODE 3 ACTION- N **in*ne beaver.s * *tos

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a. With less than the above required reactor coolant loops OPERA 8LE, i 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 o, / in HOT SHUTDOWN within the next 12. hours.

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c. ,tr. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and isenediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIRENENTS l

4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERA 8LE once per 7 days by verifying correct breaker alignments and indicated power availability.

8 4.4.1.2.2 The required steam generators shall be determined OPERABLE by y verifying secondary side water level to be greater than or equal to 21 percent at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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oo 4.4.1.2.3 At--S=t 'rhe cqdcc4 fS er,e-Reactor Coolant loopvshall be verified to be in gn operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. g nhc Do "All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided CfQ (1) no operations are permitted that would cause dilution of the reactor o coolant system boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperatur ,

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SEQUOYAH - UNIT 1 3/4 4-la Amendment No. 12 k'

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LINITING CONDITION FOR'

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3.4.1.2 g At least bo of the reactor coolant loops listed below shall be OPERABLE / 4 0, X Reactor Coolant Loop A and its associated steam generator ~

and Reactor Coolant pump, 21 Reactor Coolant Loop B and its associated steam generator g,

and Reactor Coolant pump, a y Reactor

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Coolant Loop C and its associated steam generator and Reactor Coolant pump, s't.,A Reactor Coolant Loop D and its associated steam generator and Reactor Coolant pump.

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APPLICABILITY: MODE 3 and Me % 6cTr;p W+h bop h Gerd s ACTION: 43%on1 one f"CbC.C*C 4re%r3 m -ne c69 cosmw, w,%y,{ w op %

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With less $$a7the itboNNrEqhiN[Ne' actor 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 i

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 />.

c,-tr With no Reactor Coolant loop 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 Reactor Coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required Reactor Coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power-availability.

4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 21 percent at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

p ce V 4.4.1.2.3 A g;;;yeired_ :-- Reactor Coolant loop'shall be verified to.be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"All Reactor Coolant pumps may be de-energized for up to I hour provided coolant system boron concentration, and (2) core outlet te maintained at least 10*F below saturation temperature. '

l SEQUOYAH - UNIT 2 3/4 4-2

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In MDOE 3, two reactor coolant loops provide sufficicnt heat removal capability for removing core decay heat even in the event of a bank withdrawal accJdent; however, a single reactor coolant loop provides sufficient heat

- removal capacity if a bank withdrawal accident can be prevented, i.e., by '

opening the Reactor Trip System breakers. Single failure considerations require that two loops be OPERA 8LE at all times. ,

3/4.4 REACTOR COOLANT SYSTEM BASES 4

3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in opera-

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tion, and maintain DN8R above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least H0T STAN08Y within 1 hcur.

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In MODE 4, a single reactor coolant loop or residual heat removal (RHR) R16 loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least, two loops be OPERA 8LE.

Thus, if the reactor coolant loops are not OPERA 8LE, this specification '

.s requires two RHR loops to be OPERA 8LE. 116

.( ' In MODE 5, single failure considerations require that two RHR loops be OPERA 8LE.

The operation of one Reactor Coolant Pump or one RHR 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 eith baron reduction will, therefore, be

. within the capability of operator recognition and control.

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l 3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS from being Each safety valve is designed pressurized above its Safety Limit of 2735 psig.

to relieve 420,000 lbs per hour of saturated steam at the valve set point.

The relief capacity of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown. In the event that no MAR 251982 8 3/4 4-1 Amendment No. 12 SEQUOYAH - UNIT 1

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In M00E 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers. Single failure considerations ,

require that two loops be OPERA 8LE at all times.

3/4.4 REACTOR COOLANT SYSTEM l

BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation, and maintain DN8R above 1.30.during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in

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operation this specification requires that the plant be in at 1 east HOT STAN08Y within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.  :

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In MODE 4, a single reactor coolant loop or residual heat removal (RHR)

, loop provides sufficient heat removal capability for re. moving decay heat; but single failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPENABLE, this specification requires two RHR loops to be OPERA 8LE.

I In MODE 5 single failure considerations require that two RHR loops be OPERA 8LE.

i The operation of one Reacto' Coolant r Pump or one RHR pump provides adequate I

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 baron reduction will, therefore, be within the capability of operator recognition and control.

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, ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 (TVA SQN TS 82)

Justification to require two reactor coolant loops to be in operation during Mode 3.

DESCRIPTIbN OF CHANGE This change will revise the Reactor Coolant System (RCS) specifications while in Mode 3 to require two reactor coolant loops to be in operation when the reactor trip system breakers are closed. The bases will also be revised to show the time when~two reactor coolant loops are required to provide the necessary heat removal capability.

REASON FOR CHANCE In June'of 1984, the Westinghouse Safety Review Committee determined that a potential unreviewed safety question exists as the result of the identification of an inconsistency in the assumptions between the accident analysis in the Final Safety Analysis Report (FSAR) and the technical specifications. This issue concerns the number of operating reactor coolant pumps when the plant is between residual heat removal (RHR) operation and hot zero power (HZP). This stage of operation is known as Mode 3 in the Westinghouse Standard Technical Specifications.

j The accident analyses in the FSAR which are performed at HZP conditions are intended to bound the colder conditions of Mode 3 between HZP and RHR

! operation. Of the accident analyses presented in the FSAR, three are performed at HZP; steamli'ne rupture, RCCA ejection, and uncontrolled bank i

withdrawal from subcritical. However, the only accident requiring reanalysis due to the technical specification inconsistency is the bank withdrawal from suberitical event. The FSAR analysis of the bank withdrawal from suberitical event assumes that all reactor coolant pumps are operating.when the, plant is at HZP and not operating in a special test. The results of the Westinghouse reanalysis show that two reactor coolant pumps in operation are adequate to mee.t RCS design limits. Thus, the proposed change to the technical specifications will require two reactor coolant loops to be in operation during Mode 3.

l Justification for Change Westinghouse has analyzed the bank withdrawal from suberitical event assuming two pumps in operation at HZP. This-accident was analyzed using the current Westinghouse analytical methods to demonstrate that the departure from nucleate boiling ratio (DNBR) remains above the limit value, which is the acceptance criterion for Condition II events. Using the current methods, the uncontrolled RCCA bank withdrawal from

.suberitical accident is performed in three stages: first, an average core nuclear power transient calculation, then an average core heat flux calculation, and finally the DNBR calculation. The TWINKLE computer code is used to calculate the nuclear power transient, the FACTRAN code to calculate, the heat flux, and the THINC code to calculate the DNBR. ,

(These codes are already referenced in the FSAR.) These methods have '

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been previously approved by NRC in the review of the accident analyses on individual plant dockets.

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In general, the assumptions listed in FSAR section 15.2.1.2 for the RCCA bank withdrawal from suberitical accident apply to the reanalysis. Additional assumptions used in the reanalysis are:

1. A conservative value for the moderator temperature coefficient was used in the analysis to yield the maximum peak heat flux.
2. The most limiting axial and radial power shapes, associated with having the two highest combined worth sequential control banks in their high worth position, were assumed in the DNB analysis.
3. Two reactor coolant pumps were assumed to be in operation.

The results of the reanalysis show that the minimum DNBR remains above the limiting value at all times during the transient. Therefore, the reactor core and reactor coolant system will not be adversely affected, and no cladding damage and no release of fission products to the RCS is predicted in the event of an RCCA withdrawal from suberitical accident with two reactor coolant pumps operating.

Sequoyah placed administrative controls in place on August 10, 1984 to require two reactor coolant loops to be in operation or one reactor coolant loop if the control rods are on the bottom and the control rod drive system is tagged to prevent rod withdrawal. This proposed change adds this requirement'to the technical specifications.

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-/* , ENCLOSURE 3

! PROPOSED. TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 (TVA SQN TS 82)

Determination of no significant hazards considerations for proposed changes which will require two reactor coolant loops to be in operation during Mode 3.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No, the reactor coolant pumps (RCPs) play two different roles at hot zero power (HZP). During the startup role, all four RCPs are put in operation to provide additional heat generation desired for startup. In the cooldown role, the number of RCPs in operation varies according to present conditions. The proposed change will require at least two pumps to be running, which provides sufficient heat removal capability for removing core decay heat.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

No, this change corrects an inconsistency between the FSAR and the' technical specifications. .The current analysis for a bank withdrawal from suberitical event assumes all four RCPs are running. The results of the Westinghouse ~ reanalysis for only two RCPs running show that the departure from nucleate boiling ratio (DNBR) remains abo'e v the limit value, which is the ' acceptance criterion for this event. Thus, no ,

cladding damage and no release of fission products to the reactor coolant system is predicted as a ' result of DNB.

3. Does the proposed amendment involve a significant reduction in a margin of safety? ,

No, of the three accidents analyzed at HZP, the bank withdrawal from 4

suberitical is the only event that requires more than one RCP to be running. The change in the technical specifications from one pump required running to two pumps running will not have any significant impact on the margin of safety.

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