B13181, Proposed Tech Spec Sections 3/4.4,3/4.6 & 3/4.7 Re RCS
| ML20244B084 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 06/30/1989 |
| From: | CONNECTICUT YANKEE ATOMIC POWER CO. |
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| Shared Package | |
| ML20244B073 | List: |
| References | |
| B13181, NUDOCS 8906120270 | |
| Download: ML20244B084 (215) | |
Text
{{#Wiki_filter:,. _ - -_ _ _ , - Docket No. 50-213 B13181 l 1 l i l 1
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l Attachment 1 ) i Haddam Neck Plant { 1 Proposed Revised Technical Specifications ! l i I
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Section 3/4.4 REACTOR COOLANT SYSTEM
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.3/4.4 REACTOR COOLANT SYSTEM l- 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION I
l- LIMITING CONDITION FOR OPERATION 3.4.1.1 The following reactor coolant loops shall be in operation, with associated loop stop valves OPERABLE: l a. All reactor coolant loops in operation with the reactor above 65% 1, of RATED THERMAL POWER, or
- b. At least three coolant loops in operation
- with the reactor less than or equal to 65% of RATED THERMAL POWER.
APPLICABILITY: MODES I and 2. ACTION: i With less than the above required reactor coolant loops in operation or the l associated loop stop valves not OPERABLE be in at least HOT STANDBY within 6 I hours. SURVEILLANCE RE0VIREMENT l 4.4.1.1.1 At least once per 12 hours the above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant, and that power is available to the loop stop valves. 4.4.1.1.2 At least once per 18 months, cycle the loop stop valves through one complete cycle of full travel. l
- The loop out of service may be idled (cold leg stop valve closed) or isolated (cold and hit leg stop valves closed).
HADDAM NECK 3/4 4-1
REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 The following number of Reactor Coolant Loops listed below shall be OPERABLE end in operation.*
- a. At least three reactor coolant loops shall be OPERABLE and at least two reactor coolant loops shall be in operation if the reactor trip breakers are closed and the control rod drive lift coils energized, or
- b. At least two reactor coolant loops shall be OPERABLE and at least one reactor coolant loop shall be in operation if the reactor trip breakers are open or the control rod drive lift coils are de-energized.
The reactor coolant loops are;
- a. Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,
- b. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,
- c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump, and
- d. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump.
APPLICABILITY: MODE 3. ACTION:
- a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
i
- b. With only one reactor coolant loop in operation and the reactor trip system breakers closed and the control rod drive lift coils energized, within I hour either open the reactor trip system breakers or de-energize the control rod drive lift coils.
- c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor
- Coolant System, immediately open or verify open the Reactor Trip System breakers, and initiate corrective action to return the required loop to operation. l l
- All reactor coolant pumps may be de-energized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
l HADDAM NECK 3/4 4-2 I u J
REACTOR COOLANT SYSTEM HOT STANDBY SURVEILLANCE RE0VIREMENTS 4.4.1.2.1 At'least once per 7 days the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE. 4.4.1.2.2 At least once pcr 12 hours the required steam generators.shall be determined OPERABLE. 4.4.1.2.3 At least once per 12 hours the required reactor coolant loops shall be verified in operation and circulating reactor coolant. 4.4.1.2.4' At least once per 12 hours, if required, verify that the reactor trip system breakers are open or the control rod drive lift coils are de-energized. v HADDAM NECK 3/4 4-3
REACTOR COOLANT SYSTEli HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 The following number of heat removal loops listed below shall be OPERABLE
- and in operation:
- a. At least three reactor coolant loops shall be OPERABLE and at least two reactor coolant loops shall be in operation if the reactor trip breakers are closed and the control rod drive lift coils energized, or
- b. At least two heat removal loops (RCS or RHR) shall be OPERABLE and at least one heat removal loop (RCS or RHR) shall be in operation if the reactor trip breakers are open or the control rod drive lift coils are de-energized.
The heat removal loops are;
- a. Reactor Coolant Loop 1 and its associated steam generatar and reactor coolant pump,**
- b. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,**
- c. - Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,**
- d. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump,**
- e. RHR LOOP A, and
- f. RHR LOOP B.
APPLICABILITY: MODE 4. All reactor coolant pumps and RHR pumps may be deenergized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core autlet temperature is maintainod at least 10*f below saturation temperature.
** A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperstwes less than or equal to 315'F unless the secondary water temperature of each steam generator is less than 20*F above each of the Reactor Coolant System cold leg temperatures.
i HADDAM NECK 3/4 4-4
REACTOR COOLANT-SYSTEM HOT SHUTDOWN'> LIMITING CONDITION FOR OPERATION fcontinued) ACTION:
- a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status and open or verify open the Reactor Trip system breakers within I hour. If the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours.
- b. With no loop in operation, suspend all operations involving a reduction in. boron concentration of the Reactor Coolant System,
. immediately open or verify open the Reactor Trip System breakers and initiate corrective action to return the required loop to operation.
SURVEILLANCE'RE0VIREMENTS
'4.4.1.3.1 ' At least once per 7 days the required reactor coolant pump (s),
if not'in operation, shall be determined OPERABLE. 4.4.1.3.2 _At least once per 12 hours the required steam generator (s) shall be~ determined OPERABLE by verifying secondary side narrow range water level to be greater than or equal to 25%. 4.4.1.3.3 At least once per 12 hours one reactor coolant or RHR LOOP shall be verified in operation and circulating reactor coolant. 4.4.1.3.4 At least once per 12 hours, if required, verify that the reactor trip system breakers are open or the control rod drive lift coils are de-energized. h HADDAM NECK 3/4 4-5
REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one RHR LOOP shall be OPERABLE and in operation,* the Reactor Trip System breakers shall be open or the control rod drive lift coils shall be de-energized, and either:
- a. One additional RHR LOOP shall be OPERABLE, or The secondary side narrow range water level of at least two b.
unisolated steam generators shall be greater than 25%. APPLICABILITY: MODE 5 with reactor coolant loops filled.** ACTION: i
- a. With one of the RHR LOOPS inoperable and with less than .the required steam generator level, immediately initiate corrective l action to' return the inoperable RHR LOOP to OPERABLE status or l restore the required steam generator water level as soon as '
possible.
- b. With no RHR 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 RHR ! LOOP to operation.
- c. With the Reactor Trip System breakers closed and the control rod I drive lift coils energized, within I hour either open the Reactor )
Trip System breakers or de-energize the control rod drive lift ' coils. I
- The RHR pump may be deenergized for up to I hour provided: (1) no j operations are permitted that would cause dilution of the Reactor l Coolar.t System boron concentration, and (2) core outlet temperature is !
maintained at least 10*F below saturation temperature.
** A teactor coolant pump in an unisolated locp shall not be started with !
one or more of the Reactor Coolant System cold leg temperatures less ! than nr Equal to 315"F unless the secondary water temperature of each l steam geatrator is less than 20*F above each of the Reactor Coolant ! System cold leg temperatures. ! l i j l i HADDAM NECK 3/4 4-6 _ _. _ _ _ - _ _ . s
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P REACTOR COOLANT SYSTEM COLD SHUTOOWN - LOOPS FILLED SURVEILLANCE REQUIREMENTS (continued) 4.4.1.4.1.1 At' least once per 12 hours the secondary side water level of at least two' steam generators when required shall be determined to be within limits. 4.4.1.4.1.2 At.least once per.12 hours, an RHR LOOP shall be determined to be in operation and circulating reactor coolant. 4.4.1.4.1.3' The' RHR LOOP not in operation but required shall be determined OPERABLE- at-least once per 7 days by verifying breaker alignments and
- indicated power availability..
4.4.1.4.1.4 At least once per 12 hours verify that either the Reactor Trip System breakers are open or the control rod drive lift coils are' de-energized. HADDAM NECK 3/4 4-7
REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two RHR LOOPS shall be OPERABLE
- with at least one RHR LOOP in operation ** and either the Reactor Trip System breakers shall be open or the control rod drive lift coils shall be de-energized.
APPLICABILITY: MODE 5 with reactor coolant loops not filled. ACTION:
- a. With less than the above required RHR LOOPS OPERABLE, immediately initiate corrective action to return the required RHR LOOPS to OPERABLE status as soon as possible.
- b. With no RHR 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 RHR
-LOOP to operation.
- c. With the Reactor Trip System breakers closed and the control rod drive lift coils ~ energized, within I hour either open the Reactor Trip System breakers or de-energize the control rod drive lift coils.
SURVEILLANCE RE0VIREMENTS 4.4.1.4.2.1 At least once per 12 hours, a RHR LOOP shall be determined to be in operation and circulating reactor coolant. 4.4.1.4.2.2 At least once per 7 days the RHR LOOP not in operation shall be determined OPERABLE by verifying breaker alignments and indicated power availability. 4.4.1.4.2.3 At least once per 12 hours verify that either the Reactor Trip System breakers are open or the control rod drive lift coils are j de-energized. 1 One RHR loop may be inoperable for up to 2 hours for surveillance t testing provided the other RHR loop is OPERABLE and in operation. j
- The RHR pump may be de-energized for up to I hour provided: (1) no ccerations are permitted that would cause dilution of the Reactor Coolant Systen boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature, j l
HADDAM NECK 3/4 4-8
REACTOR COOLANT SYSTEM ISOLATED LOOP l LIMITING CONDITION FOR OPERATION 1 I 3.4.1.5 The RCS loop stop valves of an isolated locp* shall be shut and either:
- a. The power removed from the valve operators, or
- b. The boron concentration of the isolated loop shall be maintained I greater than or equal to the boron concentration of the operating loops.
APPLICABILITY: MODES 1, 2 ACTION: With the requirements of the above specification not satisfied, either:
- a. Remove power from the valve operators within one hour, or l b. Increase the boron concentration of the isolated loop to within the limits within 4 hours, or i
- c. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE RE0VIREMENTS 4.4.1.5.1 At least once per 12 hours, if required, verify that power is removed from the valve operators. 4.4.1.5.2 At least once per 12 hours, if required, verify that the boron concentration of an isolated loop is greater than or equal to the boron concentration of the operating loops. l
- A loop is considered to be isolated when the hot and cold leg stop valves are both closed.
HADDAM NECK 3/4 4-9
REACTOR COOLANT SYSTEM-
- ISOLATED LOOP LIMITING' CONDITION FOR OPERATION 3.4.I'.6 The RCS loop stop valves of an isolated loop
- shall be. shut and either:
J. a. The power removed from the valve operators, or b.. The boron concentration of the isolated loop shall be maintained greater than or equal to the boron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification
' 3. 9.1. .
APPLICABILITY: MODES 3, 4, 5, and 6 ACTION:. With the requirements of the above specification not satisfied, either: a.. Remove power from the valve operators within one hour, or
- b. Increase the boron concentration of the isolated loop to within the 1imits within 4 hours, or
~
- c. Be in a least HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE RE0VIREMENTS 4.4.1.6.1 At least once per 12 hours, if required, verify that power is removed from the' valve operators. 4.4.1.6.2 At least once per 12 hours, if required, verify that the boron concentration of an isolated loop is greater than or equal .to the boron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification 3.9.1. p
- A loop is considered to be isolated when the hot and cold leg stop valves are both closed.
l HADDAM NECK 3/4 4-10
REACTOR COOLANT SYSTEM ISOLATED LOOP STARTUP: LIMITING CONDITION FOR OPERATION 3.4.I'.7 A reactor coolant loop shall remain isolated until:
- a. The temperature at the cold leg of the isolated loop is within-20*F of,the highest cold leg temperature of the operating loop (s) ,
- b. .The boron concentration of the isolated loop is greater than or equal to the boron concentration required to meet SHUTDOWN MARGIN requirements of specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification 3.9.1.
APPLICABILITY: MODES 3, 4, 5 and 6. ACTION:
"a.' 'With the requirements of the above specification not satisfied, do not open the isolated loop stop valves,
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE RE0UIREMENTS 4.4.1.7.1 The isolated loop cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve. 4.4.1.7.2 At least once per refueling outage the stop valve / temperature interlock shall be determined operable by verifying that the cold leg stop valve does not.open. if the cold leg temperature in the loop is more than 20*F cooler than the highest temperature of the remaining operating loops. 4.4.1.7.3 Within 30 minutes prior to opening the loop stop valves, the isolated loop shall be determined to have a baron concentration greater than or equal to the boron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2. or 3.1.1.3 or the refueling boron concentration of Specification 3.9.1. 4.4.1.7.4 At least once per refueling outage the reactor coolant pump, loop stop and bypass valve interlock operability shall be demonstrated. 1
- An operating loop (s) may be a Reactor Coolant loop (s) or a Residual Heat Removai loop (s).
t I
** The above requirements do not apply when a hot leg stop valve is opened I to create an idled loop.
l l' HADDAM NECK 3/4 4-11
REACTOR COOLANT SYSTEM l IDLED LOOP LIMITING CONDITION FOR OPERATION p l 3.4.1.8 The cold leg loop stop valve of an idled loop
- shall be shut and either:
- a. The power removed from the valve operator, or
- b. The boron concentration of the idled loop shall be maintained greater than or equal to the boron concentration of the operating loops.
f i APPLICABILITY: MODES 1, and 2.
' ACTION:
With the requirements of the above specification not satisfied, either:
- a. Remove power from the valve operator within one hour,
- b. . Increase the boron concentration of the idled loop to within the limits within 4 hours, or
- c. Be in at least HOT STANDBY within the next 6 hours and in COLD-SHUTDOWN within the following 30 hours.
SURVEILLANCE RE0VIREMENTS 4.4.1.8.1 At least once per 12 hours, if required, verify that power is removed from the valve operator. 4.4.1.8.2 At least once per 12 hours, if required, verify that the boron concentration of an idled loop is greater than or equal to the boron concentration of the operating loops.
- A loop is considered to be idled when the hot leg stop valve is open and the cold leg stop valve is closed.
4 HADDAM NECK 3/4 4-12 _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ - _ _ __ _ _ _ _ _ . i
l
. REACTOR COOLANTESYSTEM' F
IDLED LOOP LIMITING CONDITION FOR OPERATION I 3. 4.1. 9 . The cold. leg stop valve of an idled loop shall be shut.and either:
- a. The power removed from the valve operator, or
- b. The boron concentration of the idled loop shall be maintained greater than or equal to the boron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification
3.9.1. APPLICABILITY
MODES 3, 4, 5 and 6 ACTION: With' the' requirements of the above specification not satisfied, either:
- a. Remove power from the valve operator within one hour,
- b. Increase the boron concentration of the idled loop to within the limits within 4 hours, or
- c. Be in at least HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.4.1.9.1 At least once per 12 hours, if required, verify that power is removed from the valve operators. 4.4.1.9.2 At least once per 12 hours, if required, verify that the boron concentration of an idled loop is greater than or equal to the boron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification 3.9.1.
- A loop is considered to be idled when the hot leg stop valve is open and the cold leg stop valve is closed.
HADDAM NECK 3/4 4-13
l l I : REACTOR COOLANT SYSTEM IDLED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.10 A reactor coolant loop shall remain idled until:
- a. The temperature at the cold leg of the idled loop. is within 20*F of the highest cold leg temperature of the operating loops,
- b. The-boron concentration of the idled loop is greater than or equal .
to the boron concentration of the operating loops,
- c. The reactor is no greater than 60% RATED THERMAL POWER.
APPLICABILITY: MODES I and 2.
-ACTION:
- a. With the requirements of the above specification not satisfied, do not open .the idled loop cold leg stop valve.
- b. .The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.4.1.10.1 The idled loop cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of.the operating loops within 30 minutes prior to opening the idled loop cold leg stop valve. 4.4.1.10.2 Within 30 minutes prior to opening the idled loop cold leg stop valve, the reactor shall be determined to be less than 60% RATED THERMAL POWER. 4.4.1.10.3 Within 30 minutes prior to opening the idled loop cold leg stop valve, the idled loop shall be determined to have a boron concentration greater than or equal to the boron concentration of the operating loops. HADDAM NECK 3/4 4-14
REACTOR COOLANT SYSTEM IDLED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4.1.11 A reactor coolant loop shall remain idled until:
- a. The temperature at the cold leg of the idled loop is within 20*F of the highest cold leg temperature of the operating loop (s),*
- b. The boron concentration of the idled loops is greater than or equal to the boron concentration required to meet' the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1,.3 or the refueling boron concentration of Specification 3.9.1.
APPLICABILITY: MODES 3, 4, 5 and 6 ACTION: With the requirements of the above specification not satisfied, do not open the idled-loop cold leg stop valve. SURVEILLANCE 4.4.1.11.1 The idled loop. cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of the operating loop (s) within 30 minutes prior to opening the idled loop cold leg stop valve. 4.4.1.11.2 Within 30 minutes prior to opening the idled loop cold leg stop
. valve, the idled loop shall be determined to have a boron concentration greater than or equal to the boron concentration required to meet the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 or 3.1.1.3 or the refueling boron concentration of Specification 3.9.1. If an idled loop is being started within 30 minutes after a reactor trip, this surveillance requirement may be waived if the cold leg loop stop valve is closed for less than 15 minutes.
4.4.1.11.3 At least once per refueling outage the stop valve / temperature interlock shall be determined operable by verifying that the cold leg stop valve does not open if the cold leg temperature in the loop is more than 20*F cooler than the highest temperature of the remaining operating loops. 4.4.1.11.4 At least once per refueling outage the reactor coolant pump, loop stop and bypass valve interlock operability shall be demonstrated. An operating loop (s) may be a Reactor Coolant loop (s) or a Residual Heat Remaval loop (s). HADDAM NECK 3/4 4-15
l l REACTOR COOLANT SYSTEM ! 3/4.4.2 SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum ~of one pressurizer Code safety valve shall be OPERABLE with the lift setting associated with the OPERABLE Code safety valve within 1% of its design Setpoint*. APPLICABILITY: MODE 4 except when Specification 3.4.9.3 is applicable. l l ACTION: I i
~With no pressurizer Code safety valve OPERABLE, immediately suspend all !
operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode. SURVEILLANCE RE0VIREMENTS l 4.4.2.1 In addition to the requirements of specification 4.0.5,.each l
. pressurizer Code safety valve shall be demonstrated OPERABLE by checking its setpoint each refueling. i 1
l l
- The lift setting pressure shall correspond to ambient conditions of the I valve at nominal operating temperature and pressure.
1 HADDAM NECK 3/4 4-16
REACTOR COOLANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with respective lift settings of 2485 psig 1%, 2535 psig 1% and 2585 psig 1% in accordance with their respective nameplates.* APPLICABILITY: MODES 1, 2, and 3. ACTION: With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.4.2.2 In addition to the requirements of Specification 4.0.5, each pressurizer Code safety valve shall be demonstrated OPERABLE by checking its setpoint each refueling.
- The lift setting pressure shall correspond to embient conditions of the valve at nominal operating temperature and pressure.
HADDAM NECK 3/4 4-17
REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with:
- a. Water level within 5% of the programmed level of Figure 3.4-1 during periods when THERMAL POWER is maintained constant,* and
- b. At least two groups of pressurizer heaters capable of being powered from an emergency power source and each having a capacity ,
of at least 150 kW. ' APPLICABILITY:' MODES 1, 2, and 3. ACTION:
- a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in H0T SHUTDOWN within the following 6 hours.
- b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor Trip System breakers open within 1 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE RE0VIREMENTS 4.4.3.1 At least once per 12 hours the pressurizer water level shall be determined to be within its limit. 4.4.3.2 At least once per 92 days the capacity of each of the above required groups of pressurizer heaters shall be verified. 4.4.3.3 The emergency power supply of the pressurizer heaters shall be demonstrated operable at least once per 18 months by manually transferring 1 power from the normal to the emergency power supply and energizing the l heaters.
- During periods when THERMAL PGWER is being changed the pressurizer water level may be outside the 5% band for periods not to exceed I hour.
l HADDAM NECK 3/4 4-18 l
I~ FIGURE 3.4-1 PRESSURIZER PROGRAMMED WATER LEVEL HADDAM NECK 3/4 4-19
Tw . f
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-- .. ~ ... . - - . . . . . ..- . . . .
So , - . . .
~
40 .- . .
..M .
1 . .. s ..
-w .. _ .. . .. .
-g
- a. ,,
.m 30< - --- -.
. . - ~ * -- . - - - - . .
j - 25 - W g' 20 - - -- - - - - - - - - 8 _ g.- .. . .. . . un _ . m - w m . . . . . C. .
- 10 , - - .. .
44 0 - 500_ . _ _.. . ,520 . 535 550
. 562 SYS Tave, OF . __ .
FIGURE 3.4-1
- PRESSURIZER PROGRAMMED WATER LEVEL I
HADDAM NECK -v i shv-fi ___.____.___.m ---- - - - -
REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR.0PERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. The Setpoint for the PORVs and their block valves shall be greater than or equal to 2325 psig and less than or equal to 2350 psig. The emergency control air supply shall have a minimum pressure of 118 psig. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one or more PORV(s) inoperable, because of excessive seat leakage, within I hour either restore the PORV(s) to OPERABLE status or close/ verify closed the associated block valve (s);
otherwise, be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one PORV inoperable due to causes other than excessive seat
-leakage or low emergency control air supply pressure, within I hour either restore the PORV to OPERABLE status or close/ verify closed the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With both PORVs inoperable due to causes other than excessive seat leakage or low emergency control air supply pressure, within I hour either restore each of the PORVs to OPERABLE status or close/ verify closed their associated block valves and remove power from the block valves and be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
- d. With the emergency control air supply pressure for the PORVs less than 118 psig restore the emergency control air supply to OPERABLE status within 6 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by: I a. Performance of a CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST, and l
- b. Operating the valve through one complete cycle of full travel. I HADDAM NECK 3/4 4 20 l
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS 4.4.4.2 Each PORV block valve shall be demonstrated OPERABLE at least once per 18 months by performance of an ANALOG CHANNEL OPERATIONAL TEST. 4.4.4.3 Each PORY block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4. 4.4.4.4 The control air supply for the PORVs shall be demonstrated OPERABLE at least once per 18 months by verifying that the control air supply does not drop more than 0.3 psi in one hour when isolated from Containment Air Supply System. 4.4.4.5 The air supply for the PORVs shall be verified by local pressure indication once every 92 days. 4.4.4.6- The emergency air and power supply for the PORVs and block valves shall be-demonstrated OPERABLE at least once per 18 months by: l 1. Manually transferring power and air, and control power from the normal to the emergency power and air supplies, and;
- 2. Operating the valves through a complete cycle of full travel.
l l l 1
)
i HADDAM NECK 3/4 4-21
REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator in a nonisolated reactor coolant loop shall. be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. l ACTION: With one or more steam generators inoperable in unisolated reactor coolant loops, restore the inoperable steam generator (s) to OPERABLE status prior to increasing Tavg above 200'F. SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. 4.4.5.1 Steam Generator Samole Selection and Inspection - Each steam generator shall be determined OPERABLE during refueling by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
- a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;
- b. The first sample of tubes selected for each inservice inspection of each steam generhter shal? include:
I 1 HADDAM NECK 3/4 4-22 1
l REACTOR COOLANT SYSTEM ' SURVEILLANCE REQUIREMENTS (Continued) i
- 1) All nonplugged tubes that previously had detectable wall i penetrations (greater than or equal to 20%) above the tubes rolled region and nonplugged tubes with an imperfection in the region one-half inch below the uppermost one inch of sound roll,
- 2) Tubes in those areas where experience has indicated potential problems, and
- 3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
If any tube does not permit the passage of a 0.460 inch probe, this tube shall be plugged.
- c. -The tubes-selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and
- 2) The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories: Cateaory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. i C-2 One or more tubes, but net raore than 1% of th9 total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are i degraded tubes. C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defeci.tve. NOTE: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be I included in the above percentage calculations. This does not l apply within the tube-to-tubesheet roll region. 1 HADDAM NECK 3/4 4-23
REACTOR COOLANT SYSTEM 1 SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frequencies - -The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
- a. Inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months-after the previous inspection. If two consecutive inspections result in all inspection results falling into the C-1 category, the inspection interval may be extended to a maximum of once per 40 months;
- b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals I fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and
- c. Additional, unscheduled inservice inspections shall be performed-on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent I to any of the following conditions:
1 1) Primary-to-secondary tubes leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or
- 2) A seismic occurrence greater than an Operating Basis Earthquake, or
- 3) A loss-of-coolant accident which causes significant d wiations in temperature or pressure which affect the integrity of the steam generator tubes, or requires actuation of the engineered safeguards.
- 4) A main steam line or feedwater line break.
i HADDAM NECK 3/4 4-24
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.4 Acieriance Criteria I
- a. As used in this specification:
1)- -Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings l or specifications. Eddy-current testing indications below , 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections;
- 2) Degradation means a service-induced cracking, wastage, pitting, wear or general corrosion occurring on either inside or outside of a tube;
- 3) Decraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation above the tubes rolled region. Also a tube with
. an . imperfection in the region one-half inch below the .
uppermost one inch of sound roll is considered a degraded tube;
- 4) % Degradation means the percentage of the tube wall thickness affected or removed by degradation;
- 5) Defect means an imperfection of such severity that it exceeds the plugging limit or an imperfection which is located within the uppermost one inch of the tube's sound roll. A tube or sleeve containing a defect is defective;
- 6) Pluaaina limit means the imperfection depth at or beyond which the tube or sleeve shall be repaired or removed from service. The tube plugging limit shall be equal to 50% of the nominal tube wall thickness for tubes.*
For the rolled region (bottom four inches of the tube) the following criteria apply: a) Any tube containing an imperfection within the uppermost' one inch of sound roll shall be repaired or removed from service. b) Any imperfection is acceptable, below the uppermost one inch of sound roll.
- The plugging limit for sleeves will be determined prior to the first refueling outage following sleeve installation.
HADDAM NECK 3/4 4-25
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS'(Continued)
- 7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specification 4.4.5.3c., above;
- 8) Tube 'Insoection means an inspection of the steam generator tube from the hot leg entry point completely around the U-bend to the top support of the cold leg; or an inspection from the point of entry (hot leg or cold leg) completely around the U-bend to the opposite end.
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or sleeve
- all tubes exceeding the plugging limit u defined in Specification 4.4.5.4.a.6 and plug all defective sleeves) required by Table 4.4-2.
4.4.5.5 Reports
- a. Following the completion of each inservice inspection of steam generator tubes, a Special Report documenting the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days pursuant to Specification 6.9.2;
- b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 90 days following the completion of the inspection. This Special Report shall include:
) 1) Number and extent of tubes inspected,
- 2) location and percent of wall-thickness penetration for each indication of an imperfection, and
- 3) Identification of tubes plugged or sleeved.
- c. Results of steam generator tube inspections which fall into
} Category C-3 shall be reported in a Special 9.eport to the Commission pursuant to Specification 6.9.2 prinr to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
* ~T3e sleeving shall be performed in accordance with the Connecticut Yankee Steam Generator Sleeving Report. Revision 1, transmitted by letter dated .lanuary 7, 1986.
)
l l I HADDAM NECK 3/4 4-26 i
TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection None No. of Steam Generators per Unit Four o First Inservice Inspection (I) All Second & Subsequent Inservice Inspections One(2) TABLE NOTATIONS l (1) The first inservice inspection was performed in 1975.
'(2) The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 12% of the tubes in.that steam generator if the results of the first or previous inspections indicate that all-steam generators are performing in like manner. Note that under some circumstances, the operating conditions in one or more steam generators may be found to-be more severe than those in other steam generators.
R Under such circumstances the sample sequence shall be modified to 1 inspect the most severe conditions.- l 1 l l~ 1. I 1
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l ! l l l i 1 l L HADDAM NECK 3/4 4-27
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REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 -The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:
- a. The Containment Atmosphere Gaseous Radioactivity Monitoring System, and
- b. The Volume Control Tank Level (Narrow Range) Monitoring System and the Containment Main Sump Level Monitoring System.
APPLICABILITY: MODES 1, 2, 3, and 4.
. ACTION:
- a. -With the Containment Atmosphere Gaseous Radioactivity Monitoring System Inoperable, operation may continue provided grab samples of the containment atmosphere are obtained and analyzed for gross noble gas activity at least once per 24 hours. Restore the inoperable monitor to OPERABLE status within 7 days or, in lieu of any report required by Specification 6.9.2, prepare and submit a special report to the Commission pursuant to Specification 6.9.3 within 30 days outlining actions taken, cause of inoperability and plans for restoring the monitor to OPERABLE status.
- b. With the Volume Control Tank Level Monitoring System inoperable, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With the Containment Main Sump Level (Narrow Range) Monitoring System inoperable, restore the inoperable level monitoring system to l
OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- d. With both the Containment Main Sump Level (Narrow Range) Monitoring System and the Containment Atmosphere Gaseous Radioactivity Monitoring System inoperable, operation may continue provfded grab l samples of the containment atmosphere are ob! tined and analyzed for gross noble gas activity at least onco per 4 hours. Rettore the Containment Main Sump Level (Narrow Range) Monitoring System to l OPERABLE STATUS within 72 hours froin time of its initial loss or be {
in at least HOT STANDBV within the next 6 hours and in COLD SHUTDOWN 4 within the following 30 hours. 1 HADDAM NECK 3/4 4-29 l
i REACTOR COOLANT SYSTEM LEAKAGE DETECTION SYSTEMS SURVEILLANCE RE0VIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
- a. Containment Atmosphere Gaseous Radioactivity Monitoring System-performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and
- b. Containment Main Sump Level (Narrow Range) Monitoring System-performance of a CHANNEL CHECK at least once per 24 hours and a CHANNEL CALIBRATION
- at least once per 18 months.
, c. Volume Control- Tank Level Monitoring System-performance of CHANNEL CHECK
- at least once per 24 hours, CHANNEL CALIBRATION at least once per 18 months and ANALOG CHANNEL OPERATIONAL TEST at least once every 90 days.
l l l
- Following any seismic event greater than OBE (one half the Safe Shutdown
' Earthquake) the indicated calibration, or check shall be performed prior to declaring such affected instruments operable. 1^ HADDAM NECK 3/4 4-30 l L - . - - - - - - - - _ - - _ _ _ _ -
REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY 1 U MITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified_in Table 3.4-1. APPLICABILITY: At all times. ACTION: MODES 1, 2, 3, and 4:
- a. With any one or more chemistry parameters in excess of its Steady-State Limit but within its Transient Limit, restore the parameter to within its Steady-State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and
- b. ~ With any.one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours a.nd in COLD SHUTDOWN within the following 30 hours.
At All Other Times: With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4. SURVEILLANCE RE0VIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3. HADDAM NECK 3/4 4-33 !
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+ ,- REACTOR COOLANT SYSTEM CHEMISTRY LIMITS'
" l STEADY-STATE = TRANSIENT E8RAMETER LIMIT LIMIT
't' Dissolved Oxygen
- 1ess than or 11ess than or -
equal to ' equal to. 0.10 ppe: -1.00 ppm s-Chloride ; less than or :less than or equal to' equal to 0.15_ ppa ' 1.50 ppm
' Fluoride .less than or less than or equal to equal' to
> 0.15 ppm 1.50 ppm"
*- Limit not applicable with T avg less than or, equal to'250*F.
L HADDAM NECK- 3/4 4-34 j .. e
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l} :- l ls' E., ' TABLE 4.4-3 BEACTOR COOLANT SYSTEM l' CHEMISTRY LIMITS SURVEILLANCE REQUIREMENT 5' SAMPLE AND PARAMETER ANALYSIS FREQUENCY B Dissolved Oxygen *- At least once per 72 hours l; q= Chloride At-least once per 72 hours Fluoride' At least once per 72 hours
- Not required with T avg le'ss than or equal to 250*F.
HADDAM NECK 3/4 4-35
REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specif'c activity of the reactor coolant shall-be limited to:
- a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and
- b. Less than or equal to 68/E microcuries per gram of gross radioactivity.
APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1, 2, and 3*:
- a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or for more than 800 hours in any consecutive 12-month period, or exceeding the limit line shown on Figure 3.4-2, be in at least HOT STANDBY with T avg less than 500*F within 6 hours; and
- b. With the specific activity of the reactor coolant greater than 68/E microcuries per gram of gross radioactivity, be in at least HOT STANDBY with T avg less than 500*F within 6 hours.
MODES 1, 2, 3, 4, and 5: With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or treater than 68/E microcuries per gram of gross radioactivity, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits. SURVEILLANCE RE0VIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4. With I avg greater than or equal to 500*F. HADDAM NECK 3/4 4-36
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-OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY GREATER ~THAN 1 microcurie / gram' DOSE. EQUIVALENT I-131 LHADDAMNECK. 3/4.4-37
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l REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM
-3 LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-3, 3.4-4 and 3.4-5 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of 50*F in any one hour period with T less than or equal to 200*F and. a maximum heatup rate of 60*F ina My one hour period with T,y, greater than 200*F and less than or equal to 550*F, and
- b. A maximum cooldown rate of 100*F in any one hour with T greater than or equal to 200*F and a maximum cooldown rate of 36VP in any one hour with T ave greater than or equal to 70*F but less than L 200*F, and 1
l c. A maximum temperature change of less than or equal to 10*F in any 1-hour period during inservice hydrostatic and leak testing operations. APPLICABILITY: At all times. ACTION: With any of the above limits exceeded;
- a. -Restore the temperature and/or pressure to within the limit within 30 minutes;
- b. Perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; and
- c. Determine that the Reactor Coolant System remains acceptable for continued operation.
Otherwise, be in at least HOT STANDBY within the next 6 hours and reduce the RCS Tavg and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours. HADDAM NECK 3/4 4-39
SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as - required by 10 CFR Part 50, Appendix H, in accordance with the schedule in , Table 4.4-5. The results of these examinations shall be used-to update Figures 3.4-3, 3.4-4 and 3.4-5. l l l HADDAM NECK 3/4 4-40
FIGURE 3.4-3 CONNECTICUT YANKEE LIMIT CURVE FOR HYDR 0 STATIC AND LEAK TESTING APPLICABLE FOR 22.0 EFFECTIVE FULL POWER YEARS. (TERROR - 10*F, PERROR = 60 PSIG) HADDAM NECK 3/4 4-41
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(TERROR _=10*F,P$RROR.-60PSIG) lt 1 HADDAM NECK 3/4 4 43
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I REACTOR COOLANT SYSTEM ~ PRESSURIZER l s L LIMITING CONDITION FOR OPERATION 3.4.9.2 Th'e pressurizer temperature shall be limited to: L A maximum heatup of 100*F in any 1-hour period,
- a. .
- b. A maximum cooldown of 200*F in any 1-hour period,
- c. A maximum Reactor Coolant System (RCS) temperature difference of E 200*F, and-
- d. Greater than or equal to 70*F whenever pressurizer pressure is greater than 500 psig.
l APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits:
- a. Restore the temperature to within the limits within 30 minutes,.
- b. Perform an engineering evaluation to determine the' effects of the out-of-limit condition on the structural integrity of the j pressurizer, and
- c. Determine that the pressurizer remains acceptable for continued operation.
Otherwise , be in at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours. l
' SURVEILLANCE RE0VIREMENTS 4.4.9.2.1 The pressurizer temperatures for heatup and cooldown shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.
4.4.9.2.2 The spray water temperature differential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation. 4.4.9.2.3 The RCS temperature differential shall be determined to be within the limit at least once per 12 hours. 4.4.9.2.4 The pressurizer pressure shall be determined to be less than 500 psig at least once per hour whenever the pressurizer is not drained and the pressurizer temperature is less than 70*F. HADDAM NECK 3/4 4-45
REACTOR COOLANT SYSTEM LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following Low Temperature Overpressure Protection (LTOP) Systems shall be OPERABLE:
- a. Two LTOP spring-loaded relief valves (SLRV) with a lift setting of 380 ( 3%) psig with their respective motor-operated isolation valves in the open position, or
- b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 7 square inches (3 inches nominal inside diameter).
APPLICABILITY: MODE 4 when the temperature of any RCS cold leg of an unisolated loop is less than or equal to 315'F. MODE 5 and MODE 6 with the reactor vessel head on. The above overpressure protection system shall be placed in service prior to placing the RHR system into service and shall remain in service unless the requirements of Specification 3.4.9.3.b are satisfied. ACTION:
- a. With one LTOP SLRV. inoperable, restore the inoperable LTOP SLRV to OPERABLE status within 7 days, or depressurize and vent the RCS through at least a 7 square inch vent within 8 hours.
- b. With both LTOP SLRVs inoperable, depressurize and vent the RCS through at least a 7. square inch vent within 8 hours.
- c. In the event either the LTOP SLRVs or the RCS vent (s) of Specification 3.4.9.3b are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 i days. The report shall describe the circumstances initiating the transient, the effect of the LTOP SLRVs or the above RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.
- d. The provisions of Specification 3.0.4 are not applicable.
HADDAM NECK 3/4 4-46
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each LTOP SLRV shall be demonstrated OPERABLE by verifying the SLRV isolation valve is open at least once per 72 hours when the L10P SLRV is required to be OPERABLE. 4.4.9.3.2 The RCS vent (s) of Specification 3.4.9.3b shall be verified to be open at'least once per 12 hours when the vent (s) 4re being used for overpressure protection, except when the vent pathway is provided with a valve that is locked, sealed or otherwise secured in the open position.
,4.4.9.3.3- Once per COLD SHUTDOWN, the LTOP System Isolation Valve Interlocks and Alarms'shall be demonstrated OPERABLE by the performance of. an. ANALOG CHANNEL OPERATIONAL TEST.
4.4.9.3.4 Once per refueling, the LTOP System shall be demonstrated OPERAULE by the performance of a CHANNEL CALIBRATION and an ANALOG CHANNEL OPERATIONAL TEST. t HADDAM NECK 3/4 4-47
u : REACTOR COOLANT SYSTEM l- 3/4.4.10 STRUCTURAL INTEGRITY-1 LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.10. APPLICABILITY: All MODES. ACTION:
- a. With the structural integrity of any ASME Code Class I component (s).not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDTT considerations.
- b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.
l c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.
- d. -The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.4.10 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. HADDAM NECK 3/4 4-48
[. L REACTOR COOLANT SYSTEM 3/4.4.11 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION L 3.4.11 At least one Reactor Coolant System (RCS) vent path consisting'of at least two valves in series capable of being powered from 125-volt D.C. buses shall be OPERABLE
- and closed at each of the following locations:
- a. Reactor Vessel Head,-and
- b. Pressurizer Steam Space.
APPLICABILITY: MODES 1, 2, 3, and 4. l ACTION:
- a. With the pressurizer vent path inoperable, STARTUP and/oi l POWER OPERATION may continue provided that: 1) the-inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path and 11) one power operated relief valve (PORV) and its associated block valve is OPERABLE; otherwise, restore either the inoperable vent path or one PORV and its associated block valve to OPERABLE status within 30 days, or submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the path to OPERABLE status.
- b. With the reactor vessel head vent path inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next i 10 days outlining the cause of the malfunction and the plan :
l for restoring the path to OPERABLE status, l
- Power to the valves may be removed.
HADDAM NECK 3/4 4-49 f N___ -_ _ __
.__7__ SURVEILLANCE REQUIREMENTS 4.4.11 Each RCS vent path shall be demonstrated OPERABLE at least once per-18 months by:
- a. Verifying all manual isolation valves. in each vent' path are
,, locked in the open position,
- b. Cycling each valve in the vent path through at least one complete cycle of full- travel from the control room during COLD SHUTDOWN or REFUELING, and
- c. . Verifying flow through the RCS vent paths during venting during COLD SHUTDOWN or REFUELING.
1 1 HADDAM NECK- 3/4 4-50
3/4.4 REACTOR COOLANT SYSTEM , . BASES l E4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above that point which provides 95% confidence at a 95% probability level that DNB has not occurred during all normal operations and anticipated transients. In MODES I and 2 with one reactor coolant loop not in operation this specification requires that the plant remain below 65% power. With less than the required reactor coolant loops in operation, the plant shall be in at least H0T STANDBY within 6 hours. The loop isolation valves are required to be OPERABLE in the operating loops in order to terminate the primary to secondary leak path in the event of a steam generator tube rupture. In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident. Single failure considerations require that three loops be OPERABLE. A single reactor coolant loop provides sufficient heat removal capability for decay heat if a bank withdrawal accident can be prevented (i.e., by opening the reactor trip system breakers or de-energizing the control rod drive lift coils). Single failure considerations require that two loops be OPERABLE. l l In MODE 4, two reactor coolant loops provide sufficient heat removal capability ~for removing decay heat even in the event of a bank withdrawal accident. Single failure considerations require that three loops be OPERABLE. A single reactor coolant or RHR loop provides sufficient heat removal capability for decay heat if a bank withdrawal accident can be prevented, i.e., by opening the reactor trip system breakers or de-energizing the control rod drive lift coils. Single failure l considerations require that two loops be OPERABLE. In MODE 5 with reactor coolant loops filled, a single RHR loop provides sufficient hcat removal capability for removing decay heat. A bank withdrawal accident is prevented by openir.g the reactor trip system breakers or de-energizing the control rod drive lift coils. Single failure considerations require that at least two RHR loops be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat. A bank withdrawal accident is prevented by opening the reactor trip system breakers or de-energizing the control rod drive lift coils. Single failure considerations and the unavailability of the steam generators as a heat removing component require that at least two RHR loops be OPERABLE. The operation of one Reactor Coolant Pump (RCP) 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 with baron reduction will, therefore, be within the capability of operator recognition and control. HADDAM NECK B3/4 4-1
3/4.4 REACTOR COOLANT SYSTEM BASES' 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION (continued) The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 315'F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 20*F above each of the RCS cold leg temperatures. The requirement to maintain the boron concentration of an isolated / idled loop greater than or equal to the boron concentration of the operating loops or the baron concentration required to meet SHUTDOWN MARGIN requirements I ensures that no unacceptable reactivity addition to the core could occur during startup of an isolated / idled loop. Verification of the boron concentration in an isolated / idled loop prior to opening the stop valves provides a reassurance of the adequacy of the boron concentration in the l isolated / idled loop. Startup of an isolated / idled loop could inject cool water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by prohibiting isolated / idled loop startup until its i temperature is within 20*F of the operating loops. 3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. The required relieving capacity of each safety valve is 240,000 lbs. per hour at 2,485 psig as acsumed in the safety analysis. Each safety valve is conservatively designed to relieve 293,300 lbs. per hour of saturated steam at 2485 psig. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures. During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. The combined relief capacity of all of these valves is greater than the ; maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. HADDAM NECK B3/4 4-2 1
REACTOR COOLANT SYSTEM BASES 3/4.4.3 PRESSURIZER The limit on the water level in the pressurizer assures that the parameer is maintained within that assumed in the safety analyses. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the l capability of the plant to control Reactor Coolant System pressure and l establish natural circulation. l 3/4.4.4 RELIEF VALVES l Operation of the power-operated relief valves (PORVs) minimizes the l undesirable opening of the spring-loaded pressurizer Code safety valves and provide an alternate means of core cooling. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a PORV become inoperable. One of two redundant PORV relief trains must be OPERABLE to adequately cool the core in the event that the steam generators are not available to remove core decay heat. When in automatic mode, all PORVs and block valves open automatically on high pressurizer pressure. The PORVs and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on: (a) a modification of Regulatory Guide 1.83, Revision 1 and
'(b) Previous Eddy Current Examination results. Inservice inspection of steams generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any ttbe degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits imposed by plant chemistry guidelines which minimize corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may occur. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary system imposed in Specification 3.4.6.2, Operational Leakage. Operating plants have demonstrated that primary to secondary leakage specified in Specification 3.4.6.2 can readily be detected by radiation monitoring of steam generator blowdown. Despite chemistry controls, several forms of steam generator corrosion have been found throughout the industry. One of the most prevalent forms of HADDAM NECK B3/4 4-3
REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) corrosion is wastage, which involves a general thinning of the tube wall over a large area. Wastage is easily detectable during inservice inspection, so any flaws of this form will be detected during the scheduled refueling outage inspections. Plugging or sleeving is required for any tubes with defects that have penetrated through 50% or more of the tube-wall thickness. Since wastage is a very slow corrosion process, unplugged tubes or sleeves with less than 50% through-wall defects do not pose a safety problem. Of course, these tubes are inspected each outage to measure their defect size, and they are plugged or repaired if necessary. Another form of corrosion is denting, which is caused by the rapid i production of iron oxide within the tube / support crevice region. As this I oxide is produced, it fills the gap between the tube / support structure subsequently pushing with sufficient force on the tube to cause a dent. This dent causes stresses which can lead to stress corrosion cracking of the tube. Those tubes which do not permit the passage of a 0.460 inch diameter probe are plugged. Stress corrosion cracking is caused by a combination of stress in the tube i with or without coincident adverse chemistry. Once a stress corrosion crack l begins to form, it may propagate rapidly. If this occurs, a through-wall l crack may develop during operation and cause primary-to-secondary leakage. To limit the extent of tube leakage during operation, the l primary-to-secondary leakage has a limit of 150 gallons per day per steam generator as defined in Section 3.4.6.2c. Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 150 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown or air ejector exhaust. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or sleeved as defined in Specification 4.4.5.4.a.6. Pitting is another form of corrosion. Pits are small, pinhole-type defects which are caused by chemical impurities adhering to the tube. Pits can corrode the tubes faster than wastage, but because of their size, they have very little effect on tube integrity. Pits can be detected during inservice inspections, and are plugged or repaired if the defect size exceeds 50%. Also, leakage caused by pitting ouring operation would be monitored and measured as discussed above. Testing has shown that pits will cause tube leaks before affecting tube integrity. If a defect should develop in service, it would be found during scheduled inservice steam generator tube examinations. Plugging or sleeving will be required for all tubes with imperfections equal to or exceeding the plugging limit as defined in Specification 4.4.5.4.a.6. Tubes containing sleeves with imperfections exceeding the plugging limit will be plugged. Steam generator tube inspections of operating plants have demonstrated the HADDAM NECK B3/4 4-4
f REACTOR COOLANT SYSTEM BASES capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness and to axially locate the degradation in the rolled region with an accuracy of 10.1 inch. The plugging or sleeving limit defined in Specification 4.4.5.4.a.6 is based l on the requirements set forth in Regulatory Guide 1.121. These requirements are also the bases for demonstrating that a tube imperfection is acceptable regardless of its depth provided it is located below one inch of sound roll. ! An imperfection within this 1" region due to manufacturing or repair l activities is acceptable and considered insignificant. An indication which developed in service within this 1" region may be considered significant. Whenever the results of any steam generator tubing inservice inspection fall l into Category C-3, these results will be reported to the Commission in a Special Report pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. Such cases will be considered by the , Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current I inspection, and revision of the Technical Specifications, if necessary. 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. This technical specification ensuras a reliable means of detecting unidentified leakage in the reactor coolant system which potentially could be due to a circumferential through-wall flaw in primary system piping. The required instrument sensitivity is 1 gallon per minute in 4 hours as stated in Condition (2) of Generic Letter 84-04. Because the Volume Control Tank Level Monitoring System and the Containment Main Sump Level (Narrow Range) Monitoring System are not seismically qualified, surveillance requirements for a seismic event greater than one half the Safe Shutdown Earthquake (SSE) are imposed. The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. i HADDAM NECK B3/4 4-5
REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be protr 4i: placed in COLD SHUTDOWN. Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage. The total steam generator tube leakage limit of 0.4 gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The 0.4 gpm limit is consistent with the assumptions used in the analysis of these accidents. The 150 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event.of a main steam line rupture or under LOCA conditions. The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems. The limitation on the combined leakage from the RHR System and the charging system provides an indication of impending seal failures. The radiological consequences associated with the combined leakage is acceptable. This leakage will be considered as a portion of the IDENTIFIED LEAKAGE. The specified allowable leakage from the listed ECCS valves is sufficiently low to ensure early detection of possible in-series check valve failure. It is apparent that when pressure isolat4on is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure. The Surveillance Requirements for these ECCS valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from these ECCS valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. HADDAM NECK B3/4 4-6
I I REACTOR COOLANT SYSTEM 1 BASES 4 , 3/4.4.7 CHEMISTRY 1 The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor ! Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature l dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without ! having a significant effect on the structural integrity of the Reactor l Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits. The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action. 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specifi: activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident. The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 1 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-2, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1 microcurie / gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-2 must be restricted to no more than 800 hours per year (approximately 10% of the unit's yearly operating time) since the activity levels allowed by Figure 3.4-2 increase the 2-hour thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture. HADDAM NECK B3/4 4-7
REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radiciodine level is to be determined every 4 hours. If the gross specific activity level and radiciodine level in the reactor coolant ! were at their limits, the radioiodine contribution would be approximately 1%. In a release of reactor coolant with a typical mixture of radioactivity, the actual radioiodine contribution would probably be about 20%. Reducing T to less than 500*F prevents the release of activity should a steamgeneNEortuberupturesincethesaturationpressureofthereactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G:
- 1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-3, 3.4-4 and 3.4-5 for the service period specified thereon:
- a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and
- b. Figures 3.4-3, 3.4-4 and 3.4-5 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
- 2. These limit lines shall be calculated periodically using methods provided below,
- 3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F, HADDAM NECK B3/4 4-8 g
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
- 4. The pressurizer heatup and cooldown rates shall not exceed 100*F/hr and 200*F/hr,respectively.
- 5. System preservice hydrotests and inservice leak and hydrotests shall be ,
performed at pressures in accordance with the requirements of ASME ! Boiler and Pressure Vessel Code, Section XI. The fracture toughness properties of the ferritic materials in the reactor l vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-73, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves", April 1975. Heatup and cooldown limit curves are calculated using'the most limiting value of the nil-ductility reference temperature, RT at the end of 22 effective full power years (EFPY) of service life. YRI,22EFPYservicelife: period is chosen such that.the limiting RT at the 1/4T location in the core region is greater than the RT ftNSTlimiting unirradiated material. TheselectionofsuchalimitingRYDT assures that all components in the Reactor Coolant System will be oper$9dd conservatively in accordance with applicable Code requirements. The reactor vessel materials have been tested to determine their initial ; RT . Reactor operation and resultant fast neutron (E greater than 1 MeV) 1 irYSliationcancauseanincreaseintheRT Therefore, an adjusted reference temperature, based upon the fluen$0T.and chemical content of the i material in question, can be predicted. The heatup and cooldown limit , curves of Figures 3.4-3, 3.4-4, and 3.4.5 include predicted adjustments for ; this shift-in RT NDT* ! 4 i i l HADDAM NECK B3/4 4-9
REACTOR COOLANT SYSTEM BASES I PRESSURE / TEMPERATURE LIMITS (Continued) i Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50. The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semielliptical. surface defect. with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capability of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most
-limiting value of the nil-ductility reference temperature, RT is used.
and this includes the radiation-induced shift, delta RT chfe,spondingto theendoftheperiodforwhichheatupandcooldowncurNI, regenerated. The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that'the total stress intensity factor, K , for the combined thermal and pressure stresses at any time during heatup of cooldown cannot be greater than the reference stress intensity factor, K for the metal temperature at that time. K is obtained from the rhe,rencefracturetoughnesscurve,definedinkhpendixGtotheASMECode. The K IR is given by the equation: KIR = 26.78 + 1.223 exp (0.0145(T-RTNDT + 160)) (1) Where: K is the reference stress intensity factor as a function of the metal temhrature T and the metal nil-ductility reference temperature RT N Thus, the governing equation for the heatup-cooldown analysis is defined g. Appendix G of the ASME Code as follows: CKyg + kit less than or equal to K IR (2) HADDAM NECK B3/4 4-10
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) Where: KIM = the stress intensity factor caused by membrane (pressure) stress, kit = the stress intensity factor caused by the thermal gradients, KIR = constant provided by the Code as a function of temperature relative to the RT NDT f the material, C = 2.0 for level A and B service limits, and C - 1.5 for inservice hydrostatic and leak test operations. At any time during the heatup or cooldown transient, K is determined by the metal temperature at the tip of the postulated flak the appropriate
, and the reference fracture toughness curve. The thermal value forres stresses RT@Iing from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, K for the reference flaw is computed. FromEquation(2)thepressurestrelk, intensity factors are obtained and, from these, the allowable pressures are calculated.
C00LDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside ! of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown
~
rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that at any given reactor ccolant temperature, the delta T developed during cooldown results in a higher value of K at the 1/4T location for finite cooldown rates than for steady-state opdhation. Furthermore, if conditions exist such that the increase in K exceeds K the calculated allowable pressure during cooldown will be grd$ter than tU, steady-state value. HADDAM NECK B3/4 4-11 l . _ _ _ _ _ _ -
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period. HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the insioe of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K for the 1/4T crack during heatup is lower than the K for the 1/4T cradk during steady-state conditions at the same coolant Temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K 's for steady-state andfiniteheatupratesdonotoffseteachotherandthdR pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steaG-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup i produce stresses which are tensile in nature and thus tend to reinforce any l pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis. i HADDAM NECK B3/4 4-12 ,
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEMS The OPERABILITY of two spring-loaded relief valves (SLRVs) or an RCS vent opening of greater than 7 square inches ensures that the RCS will be protected from presure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 315'F. Either SLRV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 20*F above the RCS cold leg temperatures, or (2) the start of a charging pump (centrifugal) and its injection into a water-solid RCS. The Maximum Allowed SLRV Setpoint for the Low Temperature Overpressure Protection System (OPS) is derived by analysis which models the performance of the OPS assuming various mass input and heat input transients. Operation with a SLRV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the SLRV Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all but one charging pump (centrifugal or HADDAM NECK B3/4 4-13
REACTOR COOLANT SYSTEM BASES LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEMS (Continued) metering) while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 20*F above RCS cold leg temperature. The Maximum Allowed SLRV Setpoint for the OPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5. 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(1). 3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the RCS that could inhibit natural circulation core cooling. The OPERABILITY of at least one RCS vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function. The valve redundancy of the RCS vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path. The function, capability and testing requirements of the RCS vents are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. HADDAM NECK B3/4 4-14
Section 1.0 DEFINITIONS _ _ _ - - _ _ - _ _ _ _ _ _ - . _ _ _ _ _ - - . _ _ _ _ . _ - - - _ - - _ _ - _ - - _ _ . - - - _ - _ - _ - -- __ _ . _ - _ - - - . . - _ _ - - - - - _ _ _ _ _ _ _ _ - - . - - - - _ - _ - - - - _ - - _ - -_ _ - . - a
DEFINITIONS CONTAINMENT INTEGRITY 1.6 CONTAINMENT INTEGRITY shall exist when:
- a. All penetrations required to be closed during accident conditions are either:
- 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or
~2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as noted below:
Note 1) Normally-closed isolation valves SS-SOV-150A, SS-S0V-150B, SS-S0V-150C, SS-S0V-1500, SS-S0V-151A, SS-S0V-151B, SS-S0V-151C, and SS-S0V-151D which fail closed on loss of power and are capable of being closed within 60 seconds of a containment isolation actuation signal (CIAS) by an operator utilizing normal control switches and normal position indication within the main control room may be opened for periodic testing. Note 2) Normally-closed manual isolation valves SI-V-863A, B, C, and D, SA-V-413, CC-V-884, and SS-V-999A may be opened for periodic surveillance and containment boundary (vent .and drain) manual valves may be opened for diagnostic checks to ensure Technical Specification limits or to ensure system operability are maintained. While these valves are open, a locally stationed operator will be in direct communication with the main control room. This ensures the valves are capable of being closed within 60 seconds of a CIAS.
- b. The equipment hatch is closed and sealed,
- c. The air lock is in compliance with the requirements of Specification 3.6.1.3,
- d. The containment leakage rates are within the limits of J Specification 3.6.1.2, and l
- e. The sealing mechanism associated with each penetration (e.g.,
welds, bellows, or 0-rings) is OPERABLE. CONTROLLED LEAKAGE 1.7 CONTROLLED LEAKAGE shall be that seal water flow returned from the reactor coolant pump number 2 seals. I i HADDAM NECK 1-2
DEFINITIONS CORE ALTERATION 1.8 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe position. DOSE E0VIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of U. S. Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluent for the Purpose of Evaluating Compliance with 10CFR50, Appendix I," Revision 1, October 1977. HADDAM NECK 1-3
Section 3/4.6 CONTAINMENT SYSTEMS
i 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within I hour or be in at least HOT STANDBY.within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE0VIREMENTS , 4.6.1.1. Primary CONTAINMENT INTEGRITY shall be demonstrated: k a. At. least once per 31 days by verifying that all penetrations
- not capable of being closed by OPERABLE containment automatic isolation valves'and required to be closed during accident
-conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as noted in DEFINITION, 1.6 - CON 1AINMENT INTEGRITY. 1
- b. By verifying that the containment air lock is in compliance with the requirements of Specification 3.6.1.3, and
- c. After each closing of each penetration subject to Type B testing, except the containment air lock, if opened following a Type A or B test, by leak rate testing the seals with gas at a pressure not less ;
.than Pa, 39.6 psig, and verifying that when the measured leakage !
rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rate is less than 0.60 La. 1 I Except valves, blind flanges, and deactivated automatic valves which l are located inside the containment or PAB pipe trench and are locked, l sealed, or otherwise secured in the closed position. These i penetrations shall be verified closed during each COLD SHUTDOWN in ! excess of 48 hours except that such verification need not be performed 1 more often than once per 92 days. I ll HADDAM NECK 3/4 6-1
CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:
- a. An overall integrated leakage rate of less than or equal to La, 0.18 % by weight of the containment air per.24 hours at Pa, 39.6 psig, or
- b. A combined leakage rate of less than 0.60 La for all penetrations
- and valves subject to Type B and C tests, when pressurized to Pa.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With either the measured overall integrated containment leakage rate exceeding 0.75 La or the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 La, restore the overall integrated leakage rate to less than 0.75 La and the combined leakage rate for all penetrations subject to Type B and C tests to less than 0.6g La prior to increasing the Reactor Coolant System temperature above 200 F. SURVEILLANCE RE0VIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule, and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972:
- a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 10 month intervals during shutdown at a pressure not less than Pa, 39.6 psig during each 10-year service period. The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection;
- Penetrations P11, P63, P69 and P80 may be tested at a pressure substantially greater than Pa. This represents an exemption to Appendix J, Paragraph III.C.2.(a).
HADDAM NECK 3/4 6-2 i
SURVEILLANCE RE0VIREMENTS (Continued)
- b. If any periodic Type A test fails to meet 0.75 La, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 La, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 La at which time the above test schedule may be resumed or a corrective action plan may be prepared and submitted to the NRC that provides an acceptable alternative contingent on NRC approval.
- c. The accuracy of each Type A test shall be verified using the relationship:
(LTM +ol - 0.25 L,) 1 L cI (LTM + lo + 0.25 L,) where: L TM is the percent measured containment leakage per 24 hours at pressure Pt ' L o is the percent superimposed leakage, L c is the percent leakage obtained from the supplemental test result, and L a is replaced with Ltfor reduced pressure tests.
- d. Type B and C tests shall be conducted at intervals no greater than 24 months and at a pressure not less than Pa, 39.6 psig, using halogen gas detection, soap bubble, pressure decay, or other methods of equivalent sensitivity, except for tests involving:
- 1) Air locks, and
- 2) Purge supply and exhaust isolation valves with resilient material seals.
i e. Air locks shall be tested and demonstrated OPERABLE by the l requirements of Specification 4.6.1.3; Purge supply and exhaust isolation valves with resilient material f. seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.9.9;
- g. The provisions of Specification 4.0.2 are not applicable.
l HADDAM NECK 3/4 6-3
p -CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS
. LIMITING CONDITION FOR OPERATION 3.6.1.3 The containment air lock shall be OPERABLE with:
- a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air-lock door shall be closed, and
- b. An overall air lock leakage rate of less than or equal to 0.05 La at Pa, 39.6 psig.
APPLICABILITY: MODES 1, 2, 3, AND 4. ACTION:
- a. With one containment air lock door inoperable:
- 1. Maintain at least one OPERABLE air lock door closed and either restore the. inoperable air lock door to OPERABLE status within 24 hours or lock the OPERABLE air lock door closed;
- 2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days;
- 3. Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and
- 4. The provisions of Specification 3.0.4 are not applicable.
- b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door-closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- HADDAM NECK 3/4 6-4
CONTAINMENT SYSTEMS
. SURVEILLANCE REQUIREMENTS 4.'6.1.3 The containment. air lock shall be' demonstrated OPERABLE:
- a. Within 72-hours following each closing, except when the. air lock.
- is being used for multiple entries, then-at least once~ per 72
' hours, by. verifying no greater than '.01 L from pressure decay or other equivalent method when the volume bltween' door seals is
, pressurized to greater than or equal to 10 psig for at least.15 minutes;i '
- b. By conducting overall air lock leakage tests at' not less than P,,
39.6 psig, and verifying.the overall air lock leakage rate is within its limit:
- 1) At least once per 6 months,* and.
~2)~ Prior to establishing CONTAINMENT INTEGRITY l
The provisions of Specification 4.0.2 are not applicable. HADDAM NECK 3/4 6-5
' CONTAINMENT SYSTEMS.
INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between
-2 and +2.8 psig.
APPLICABILITY: MODES 1, 2, 3,-and 4. ACTION: With the containment internal pressure outside of the limits above, restore
.the internal pressure to within the limits within I hour or be in at least HOT-STANDBY within the next 6 hours and in COLD SHilTDOWN within the ci following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours. l-HADDAM NECK 3/4 6-6
h.:,- [i CONTAINMENT SYSTEMS AIR TEMPERATURE l . LIMITING CONDITION FOR OPERATION 3.631.5 Primary containment average air temperature shall not exceed 140 F.
-APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION: With the containment average air temperature greater than.140 F, reduce the average air temperature to within the limit within 8 hours, or be in at least HOT-STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE0VIREMENTS
'4.6.1.5 The primary containment average. air temperature shall be the weighted average of the temperatures of all OPERABLE temperature detectors and shall be determined at least once per 24 hours.
HADDAM NECK 3/4 6-7
CONTAINMENT SYSTEMS i CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in { - Specification 4.6.1.6.1. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE0VIREMENTS 4.6.1.6.1 Containment Vessel Surfaces. The structural integrity of the exposed accessible interior and exterior surfaces of the containment vessel, including the liner plate, shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of these surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradation. 4.6.1.6.2 Reoorts. Any abnormal degradation of the containment vessel structure detected during the above required inspections shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2 within 15 days. This report shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken. i HADDAM NECK 3/4 6-8
I U CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM l
}
LIMITING CONDITION FOR OPERATION
-3.6.1.7- The 42-inch containment purge supply;and exhaust isolation valves and.the 8-inch bypass valve shall be OPERABLE and locked closed.*
APPLICABILITY: MODES 1, 2,_3, and 4. l ACTION: With a 42-inch containment purge supply and/or exhaust isolation valve (s) or- I an:8-inch bypass valve open or not locked closed, close and/or lock close
- that valve'or isolate the penetration (s) within I hour; otherwise, be in at' :
.least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the !
.following 30 hours. l SURVEILLANCE REQUIREMENTS q
, 4.6.1.7 : The 42-inch containment purge supply and exhaust isolation valves ! and: the 8-inch bypass. valve shall be verified to be locked closed at least
~
once per 31 days. j l
~!
l 1 Containment purge capability may be rendered' inoperable when the-- , reactor is critical.by placing a blank flange on the 42-inch purge air l exhaust penetration inside the reactor. containment for a period of 7 l days. If the blank flange cannot be removed within 7 days, then the {
. reactor shall be shut' down within 24 hours. '
l HADDAM NECK 3/4 6-9 =- - -__ _ _ _ _ _ - - _
m _ - _ - - _ 1 CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves shall be OPERABLE with isolation times less than or equal to the required isolation times. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one or more of the isolation valve (s) inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:
- a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or l
- b. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or I
- c. Isolate each affected penetration within 4 hours by use of at least i one closed manual valve or blind flange, or
- d. Be in at least HOT STANDBY within the next 6 hours and f:. COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE RE0VIREMENTS 4.6.3.1 Each isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work i is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time, as required. l l HADDAM NECK 3/4 6-12 _ _ -- ]
I CONTAINMENT !YSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by verifying that on Safety Injection Actuation test signal, each automatic isolation valve actuates to its isolation position. 4.6.3.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5. HADDAM NECK 3/4 6-13
L 3/4.6 CONTAINMENT SYSTEMS i BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions. 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakaoe rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, Pa. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 La, as applicable, during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests. The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50. 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air lock is required to meet the restrictions on CONTAINMENT INTEGRITY and containment
. leak rate. Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 7.5 psid and (2) the containment peak pressure does not exceed the design pressure of 40 psig during LOCA conditions. The maximum peak pressure expected to be obtained from a LOCA event is 39.6 psig. The limit of 2.8 psig for initial positive containment pressure will ensure that the containment design pressure of 40 psig will not be exceeded and is consistent with the safety analyses. HADDAM NECK B3/4 6-1
4 (
' CONTAINMENT SYSTEMS BASES - 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the containment safety analysis for a LOCA and a main steam line break inside the containment. Measurements shall be taken from all OPERABLE temperature detectors to determine the average air temperature.
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the. facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 39.6 psig in the event of a LOCA. A visual inspection in conjunction with the Type A leakage tests is sufficient to demonstrate this capability. 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 42-inch containment purge supply and exhaust isolation valves and the 8-inch bypass valve are required to be closed and locked closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves locked closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the Containment Purge System. To provide assurance that.these containment valves cannot be inadvertently opened, the valves are locked closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to lock the valve closed. Containment post. accident hydrogen venting can be accomplished by two methods. One uses the containment air particulate monitoring system and the other uses the containment purge exhaust system. These methods are not required in any short time frame after an accident;.it is expected that months may elapse. In any event, if the systems are not operable because of maintenance reasons, they can be made operable. System operability can be readily obtained provided access into the containment is not required. HADDAM NECK B3/4 6-2
CONTAINMENT SYSTEMS BASES Containment purge is utilized as a back-up means of venting hydrogen from the containment following a loss-of-coolant accident. The containment air particulate monitoring system provides the primary means of purging because it provides adequate purge flow to prevent an explosive mixture build-up while allowing fine control of the release of radioactivity during purges. When necessary to effect repairs to the containment purge or purge bypass isolation valves, a blank flange must be applied to the 42" purge air exhaust penetration inside the reactor containment so that the containment remains leak tight. This renders the purge system inoperable for a finite time. Seven days is considered a reasonable length of time for repair parts to be received, installed and the system retested for leak tightness and returned to service. 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. UFSAR Table 7.3-1 lists all containment isolation valves. The addition or deletion of any containment isolation valve shall be made in accordance with Section 50.59 of 10CFR50 and approved by the Plant Operations Review Committee. i HADDAM NECK B3/4 6-3 __-____-_____-__----_a
r-L:.- {. l l Section 3/4.7 PLANT SYSTEMS _ _ _ _ _ . _ ._ ._ __ -- l
)
3/4.7 PLANT SYSTEMS l 3/4.7.1' TURBINE CYCLE SAFETY VALVES 1 LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator of an unisolated reactor coolant loop shall be OPERABLE with lift settings as specified in Table 3.7-1. APPLICABILITY: MODES 1, 2, and.3. ACTION: I a. With four reactor coolant loops and associated steam generators in operation and one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided that within 4 hours the inoperable valve is restored to OPERABLE
' status; otherwise, be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With three reactor coolant loops and associated steam generators in operation and one or more main steam line Code safety valves associated with an operating loop inoperable, operation in MODES
.1, L and 3 may proceed provided that within 4 hours the ir, tcrable valve is restored to OPERABLE status; otherwise, be in HOT. STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS ,. 4.7.1.1 In addition to the requirements of Specification 4.0.5, each main steam line code safety valve associated with each steam generator shall be demonstrated OPERABLE by checking its setpoint each refueling. HADDAM NECK 3/4 7-1
.-}.
.y r
TABLE 3.7-1 STEAM LINE SAFETY VALVES PER LOOE f VALVE NUMBER J,JFT SETTING d 1R* VALVE SIZE a.- MS SV.ll, 21, 31,'41 985 psig 6Q8
- b. ~ MS SV 12, 22, 32, 42 1015 psig 6Q8 i c. MS SV 13, 23, 33, 43 1025 psig 6Q8 U d. MS SV 14, 24, 34, 44 1034 psig -608 4
The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. HADDAM NECK 3/4 7-2
PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 Two independent steam turbine-driven auxiliary feedwater pumps capable of being powered from an OPERABLE steam supply system anJ associated flow paths shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one steam turbine driven auxiliary feedwater pump inoperable, restore the inoperable auxiliary feedwater pump to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours,
- b. With two steam turbine-driven auxiliary feedwater pumps inoperable; restore at least one of the inoperable auxiliary feedwater pumps to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. Restore both the auxiliary feedwater pumps to OPERABLE status within 72 hours from time of initial loss of the first auxiliary feedwater pump or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the next 6 hours.
SURVEILLANCE RE0VIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE at least once per 31 days on a STAGGERED TEST BASIS by:
- a. Verifying that each steam turbine-driven pump develops a discharge pressure of greater than or equal to 800 psig at a steam supply pressure of greater than or equal to 300 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3;
- b. Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position, and 4.7.1.2.2 At least once per 18 months, the Auxiliary Feedwater System shall be demonstrated OPERABLE by:
- a. Verifying the cepability of each pump to attain rated flow of 450 gpm at 1050 psig;
- b. Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Auxiliary Feedwater Actuation signal; and
- c. Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation signal.
HADDAM NECK 3/4 7-3
1
' PLANT SYSTEMS AUXILIARY FEEDWATER SUPPLY LIMITING CONDITION FOR OPERATION 3.7.1.3 The demineralized water storage tank (DWST) shall be OPERABLE with a minimum contained volume of 50,000 gallons of water and the primary water storage tank (PWST) shall be OPERABLE with a minimum contained volume of-80,000 gallons .of water.
APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With the DWST inoperable, restore the DWST to OPERABLE status within 4 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With the PWST inoperable, within'4 hours:
- 1. Restore the PWST to OPERABLE status, or
- 2. Provide an equivalent supply from an alternate source, or
- 3. Be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The DWST and PWST shall be demonstrated OPERABLE at least once per 12 hours by verifying the contained water volume is within its limits. 4.7.1.3.2 The Recycle Primary Water Storage Tank-(RPWST) shall be demonstrated OPERABLE at least once per 12 hours by verifying the contained water volume is equivalent to the PWST requirements when the RPWST is the alternate water source for the PWST.
'e HADDAM NECK 3/4 7-4
'l; PLANT SYSTEMS SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microcurie per gram DOSE EQUIVALENT I-131.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the specific activity of the Secondary Coolant System greater than 0.1 microcurie per gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD' SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the Secondary Coolant System shall be determined to be within the limit by performance of the sampling and. analysis program of Table 4.7-1. 1 l HADDAM NECK 3/4 7-S o
TABLE 4.7:1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS , AND ANALYSIS FREQUENCY 1., Gross Radioactivity At least once per 72 hours. Determination
- 2. Isotopic Analysis'for DOSE a) Once per 31 days, whenever the gross EQUIVALENT 1-131 radioactivity determination Concentration indicates concentrations greater than 10% of the allowable limit for.
radioiodines. b) Once per 6 months, whenever the gross radioactivity determination indicates concentrations less than or equal to 10% of the allowable limit for radioiodines. HADDAM NECK 3/4 7-6
1 PLANT SYSTEMS MAIN STEAM LINE TRIP VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line trip valve (MSLTV) shall be OPERABLE. APPLICABILITY: -MODES 1, 2, and 3. ACTION: MODE 1: With one MSLTV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours or the respective reactor coolant loop is removed from service and its associated main steam non-return valve is closed; otherwise, be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
' MODES 2 and 3: With one MSLTV inoperable, subsequent operation in MODE 2 or 3 may proceed provided the MSLTV is maintained closed or the respective reactor coolant loop is removed from service and its associated main steam non-return valve is closed; otherwise, be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.5 Each MSLTV shall be demonstrated OPERABLE by:
- a. ' Verifying a partial stroke when tested pursuant to Specification 4.0.5.
- b. -Verifying simultaneous full closure of all four MSLTVs within 10 seconds
- during each COLD SHUTDOWN unless performed in the previous 3 months.
- With the assistance of the air operator.
HADDAM NECK 3/4 7-7
i PLANT SYSTEMS-1
. 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2- The pressure and temperature of each steam generator shall bo limited such that:'
l- a. The_ temperature of the reactor coolant in a steam generator shall be greater than 70*F when the pressure of the reactor coolant in '< the steam generator is greater than 500 psig.
- b. - _ The temperature of a steam generator vessel shall-be greater than-70'F when' the pressure of the secondary coolant in the steam generator is greater than 200 psig,
- c. The maximum heatup or cooldown rate is less than or equal to 100*F in any 1-hour period,
- d. The temperature difference across the tube sheet shall not exceed 100*F.
-APPLICABILITY: At- all times.
ACTION: With the requirements of the above specification not satisfied:
- a. Restore the steam generator pressure and/or temperature to within the limits within 30 minutes, and
- b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator. Determine that the steam generator remains acceptable
-for continued operation prior to increasing its temperature above 200*F.
SURVEILLANCE REQUIREMENTS 4.7.2.1 The pressure in each side of the steam generator shall be determined to be within the limit at least once per hour when the temperature of either the reactor or secondary coolant is less than or ecual to 70*F. 4.7.2.2 The heatup and cooldown rates of each steam generator shall be determined at least once per 30 minutes by monitoring the Reactor Coolant System temperature during heatup or cooldown operations. HADDAM NECK 3/4 7-8
L-i
- PLANT SYSTEMS 3/4.7.3 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two service water headers servicing safety-related equipment shall be OPERABLE.
' APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION: With only one. service water header OPERABLE, restore at least two headers to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.3 - At least two service water. headers shall be demonstrated 0PERABLE:
- a. At least once per 31 days by verifying that each safety related valve (manual, power-operated or automatic) in the service water system that is not locked, sealed, or otherwise secured in position, is in.its correct position; and
- b. At least once per 18 months during shutdown, by verifying that:
- 1) Each safety related automatic valve that is required to actuate, goes to its correct position on a Loss-of-Normal Power or Safety Injection Test Signal..
- 2) Each service water system pump starts automatically on a Loss-of-Normal Power Test Signal.
HADDAM NECK 3/4 7-9
i PLANT SYSTEMS 3/4.7.4 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.4 All safety related snubbers shall be maintained in an OPERABLE condition. APPLICABILITY: MODES 1, 2, 3, and 4. (MODES 5 and 6 for snubbers. located on systems required OPERABLE in those MODES.) ACTION: With one or more snubbers inoperable, within 72 hours replace or restore the 4 inoperable snubber (s) to OPERABLE status and perform an engineering ' evaluation for Specification 4.7.4.C on the supported component or declare the supported system inoperable and follow the appropriate ACTION statement for that system. SURVEILLANCE RE0VIREMENTS 1 ! 4.7.4 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program,
- a. Visual Inspection Schedule All snubbers shall be visually inspected in accordance with the following schedule:
Number of Snubbers Found Inoperable During Inspection Next Required Or Durina Inspection Interval Inspection Interval *# 0 18 months i 25% 1 12 months 25% 2 6 months i 25% 3, 4 124 days i 25% 5,6,7 62 days 25% 8 or more 31 days i 25% The snubbers may be categorized into two groups: those accessible, and l those inaccessible during reactor operation. Each group may be inspected independently in accordance with the above schedule. The inspection interval shall not be lengthened more than one step at a time. i # The provisions of Specification 4.0.2 are not applicable. 1 HADDAM NECK 3/4 7-10 j
i
! PLANT SYSTEMS l
3/4.7.4 SNUBBERS SURVEILLANCE RE0VIREMENTS (Continued)
- b. Visual Inspection Acceptance Criteria Visual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILITY, and, (2) attachments to the foundation or supporting structure are secure.
The inspection will also identify indications of excessive leakage from hydraulic snubber reservoirs or connections. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per Specifications 4.7.4.d or 4.7.4.e, as applicable.
- c. Functional Tests At least once per 18 months during shutdown, a representative sample of at least 10% of the total of each type of snubber in use in the plant shall be functionally tested either in place or in a bench test. For each snubber that does not meet the functional test acceptance criteria of Specification 4.7.4.d or 4.7.4.e, an additional 5% of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested.
The representative sample selected for the functional test sample plan shall be randomly selected from the snubbers of each type and I reviewed before beginning the testing. The review shall ensure, as far as practicable, that they are representative of the various I configurations, operating environments, range of size, and capacity of snubbers of each type. Consideration shall be given to possible effects such as vibration in the vicinity of equipment such as pumps, turbines, valves, etc. or fluid forces from relief valve discharge in the selection of the representative sample. Functional test samples shall be rotated such that different snubbers are tested at each test interval. In addition to the regular sample, snubbers which failed the previous functional test shall be retested during the next test period. If a spare snubber has been installed in place of a failed snubber, then both the failed snubber (if it is repaired and installed in another position) and the spare snubber shall be retested. Test results of these snubbers may not be included for the re-sampling. HADDAM NECK 3/4 7-11
L I PLANT SYSTEMS 3/4.7.4 SNUBBERS r SURVEILLANCE RE0VIREMENTS (Continued) i i
.If any snubber selected for functional testing either fails to ;
lockup or fails to move, i.e., frozen in place, the cause will be ! evaluated and if caused by manufacturer or design deficiency, all
. . snubbers of the same design subject.to the same defect shall be i functionally tested. This testing requirement shall be independent of the requirements stated above for snubbers not i meeting the functional test acceptance criteria. ;
For the snubber (s) found inoperable, an engineering evaluation shall be performed on the components which are supported by the snubber (s). 'The purpose of this engineering evaluation shall be 4 to determine if the components supported by the snubber (s) were ' adversely affected by the inoperability of the snubber (s) in order to ensure that the supported component remains capable of meeting the designed service requirement.
- d. Hydraulic Snubbers Functional Test Acceptance Criteria i The hydraulic snubber functional test shall verify that: i
- 1. Activation (restraining action) is achieved within the ;
specified range of velocity or acceleration in both tension and compression; l
- 2. Snubber bleed, or release rate, where required, is within the specified range in compression or tension. For snubbers specifically required to not displace under continuous load, the ability of .the snubber to withstand load without displacement shall be verified.
- e. Mechanical Snubbers Functional Test Acceptance Criteria The mechanical snubber functional test shall verify that:
- 1. The force that initiates free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.
- 2. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.
- 3. Snubber release rate, where required, is within the specified range in compression or tension. For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified.
HADDAM NECK 3/4 7-12 l
PLANT SYSTEMS 3/4.7.4 SNUBBERS SURVEILLANCE RE0VIREMENTS (Continued)
- f. Snubber Service Life Monitorina A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained as required by Specification 6.10.
Concurrent with the first inservice visual inspection and at least once per 18 months thereafter, the installation and maintenance records for each safety related snubber shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to tl'.e next scheduled snubber' service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement or reconditioning shall be indicated in the records. l l i l HADDAM NECK 3/4 7-13 _- _ _ _ _ _ _ _ __a
y PLw41' SYSTEMS 3/4.7.5 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.5 Each sealed source containing radioactive material either in excess of 100 microCuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcurie of removable contamination. APPLICABILITY: At all times. ACTION:
- a. Each sealed source with removable contamination in excess of the above limits shall be immediately withdrawn from use and either:
- 1. Decontaminate and repair the sealed source, or
- 2. Dispose of the sealed source in accordance with Commission Regulations.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.7.5.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:
- a. The licensee, or
- b. Other persons specifically authorized by the Commission or an Agreement State.
The test method shall have a detection sensitivity of at least 0.005 microcurie per test sample. 4.7.5.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below.
- a. Sources in use - At least once per 6 months for all sealed snuces containing radioactive materials:
- 1) With a half-life greater than 30 days (excluding Hydrogen 3), and
- 2) In any form other than gas.
HADDAM NECK 3/4 7-14
PLANT SLSls..i . 3/4.7.5 SEALED SOURCE CONTAMINATION
; SURVEILLANCE REQUIREMENTS (Continued)
{
- b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or. transfer to another licensee unless tested within the previous 6 months. Sealed
- sources and fission detectors transferred without a certificate
. indicating the last test date shall be tested prior to being placed into use; and
- c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.
4.7.5.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removal contamination. HADDAM NECK 3/4 7-15
p PLANT SYSTEMS 3/4.7.6 FIRE SUPPRESSION SYSTEMS FIRE WATER SUPPLY / DISTRIBUTION SYSTEM 1 LIMITING CONDITION FOR OPERATION ! 3.7.6.1 The Fire Water. Supply / Distribution System shall'be OPERABLE with: L
- a. At least two high pressure fire suppression pumps, each with a !
capacity of 2500 gpm (at Rated Pressure), automatic start capability with their discharge aligned to the fire suppression header, and i l- b. An OPERABLE flow path capable of taking suction from the Connecticut River and transferring the water through distribution piping with OPERABLE sectionalizing control valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or Spray System riser required to be OPERABLE per Specifications 3.7.6.2, 3.7.6.5, and 3.7.6.6. APPLICABILITY: At all times. ACTION:
- a. With one pump inoperable, restore the inoperable pump to OPERABLE . !
status within 7 days or provide an alternate method of supplying fire water.
- b. With the Fire Water Supply / Distribution System otherwise inoperable, establish an hourly Fire Watch Patrol for the affected areas, and provide a backup Fire Water Supply / Distribution System within 24 hours or the plant shall be placed in at least Hot Standby within the following 6 hours and in Cold Shutdown within the following 24' hours.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
HADDAM NECK 3/4 7-16
..4
- S i, s
.. (
PLANT SYSTEMS 4 ,
. SURVEILLANCE REQUIREMENTS (4.**,1.1 The Fire Water Supply / Distribution System shall be de onstrated
'O .LE:-
_a. ;At least once per 31 day's on a STAGGERED TEST-BASIS by starting each pump and operating it with flow. :The diesel pump shall run-for 30 minutes and the electric pump for 5 minutes.
- b. 'At least once per 31 days by verifying that each valve (manual, .
power-operated or automatic)'in' the flow path that is not locked,
, sealed, or otherwise. secured is in its correct position,
- c. ' At'least once per 12 months by cycling each testable valve in the flow path.through at least one complete cycle of. full travel,
- d. At'least once per 18 months by performing a. system. functional test which includes simulated automatic actuation of the system throughout us operating sequence, and
- 1) Verifying that each pump develops at least 2500 gpm at a system. head of greater than or equal to.100 psig at Rated Speed, 2). Cycling each valve'in the flow path that is not testable during plant operation through at least one complete cycle.of full travel,'and r
Verifying that. each high pressure pump. starts automati: ally
^
- 3)-
and sequentially to maintain the Fire Water Supply / Distribution System pressure greater than or. equal to 75 psig,
- e. At least' once per 3 years by performing a flow test of the Distribution System to assure System capabilities have not i degraded.
! 4.7.6.1.2 The fire pump diesel engine shall be demonstrated OPERABLE: 1
- a. At least once per 31 days by verifying:
- 1) The fuel storage tank contains at least 130 gallons of fuel, and
- 2) The diesel starts from ambient conditions and operates for at least 30 minutes with flow.
1 1.
'HADDAM NECK 3/4 7-17
b
. PLANT SYSTEMS
-SURVEILLANCE REQUIREMENTS (CONTINUED)
At least' once per 92 days by verifying that a sample of diesel b. fuel from the fuel storage tank, obtained in accordance with ASTM-- 0270-1975, is within the acceptable limits specified in. Table 1 of ASTM D975-1977 when checked for viscosity, water and sediment.
- c. At least once per 18 months, during shutdown, by:
- 1) . Subjecting the diesel to an inspection in.accordance with
. procedures prepared in conjunction with its manufacturer's recommendations for the class of service, and
- 2) Verifying the diesel starts from ambient conditions on the auto-start signal and operates for greater than or equal to 30 minutes while loaded with the fire pump.
1 4.7.6.1.3 The fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1) The electrolyte level of each battery is above the plates, and
- 2) The individual overall battery voltage is greater than or equal to 24 volts.
- b. At least once per 92 days by verifying that the specific gravity-is appropriate for continued service of the battery, and
- c. At least once per 18 months by verifying that:
- 1) The batteries, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, and 1
- 2) The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anticorrosion material.
I HADDAM NECK 3/4 7-18
PLANT SYSTEMS SPRAY AND/OR SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION 1 3.7.6.2 . The following Spray and/or Sprinkler Systems shall be OPERABLE:.
- a. Sprinkler Systems
- 1) Maintenance shop annex (old warehouse),
- 2) Turbine building, El. 21'6",
- 3) Turbine building mezzanine,
~4) Turbine building beneath the turbine floor and under the high
. pressure turbine,
- 5) Turbine building between column lines C and D and column numbers 8 and 12 under the 59' 6"' elevation *
- 6) Lube Oil Storage room, 7). Cable Spreading area,*
- 8) Primary Auxiliary building,*
- 9) Service l Building,
- 10) Chemistry Lab,
- 11) Machine Shop,
- 12) ' Intake Structure (diesel fuel storage room)
- 13) Diesel Fire Pump.
- 14) Service Water Pumps Water Curtain *, and-
- 15) Cable Spreading Hallway Water Curtain *
- b. Deluge Spray Systems
- 1) High pressure turbine, (manual),
2)- Hydrogen seal oil unit, (manual),
- 3) Turbine lube oil reservoir, (manual) and 4). Turbine Building Crane Well (automatic).
c.- Preaction Systems 1)- Turbine building mezzanine under generator,
- 2) Diesel Generator Room'A, and
- 3) Diesel Generator Room B.
APPLICABILITY:'Whenever systems, structures, components, or equipment protected by the spray and/or sprinkler systems are' required to be OPERABLE. ACTION:
- a. With one or more of the above required Spray and/or Sprinkler q l Systems inoperable, within.1 hour establish a continuous fire ;
i watch with backup fire suppression equipment for those areas in l which redundant systems or components could be damaged or present i an exposure fire hazard to other components (these areas are 1 asterisked (*)); for other areas, establish an hourly fire watch patrol.
- b. The provisions of Specification 3.0.3 and 3.0.4 are not a applicable.
- Areas where redundant systems or components could be damaged.
HADDAM NECK 3/4 7-19
l j PLANT SYSTEMS SURVEILLANCE RE0VIREMENTS 4.7.6.2 Each of the above required Spray and/or Sprinkler Systems shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power-operated, er automatic) in the flow path, that is not locked, sealed or otherwise secured in position, is in its correct position,
- b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel,
- c. At least once per 18 months:
- 1) By performing a system functional test which includes simulated automatic actuation of the system, and:
a) Verifying that the system alarm check / main actuation valves in the flow path actuate to their correct positions on a simulated test signal, and b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.
- 2) By visual inspection of the spray headers to verify their integrity, and
- 3) By visual inspection of each nozzle to verify that there is no blockage and that each spray area is not obstructed.
- d. At least once per 3 years by performing an air and/or water flow test through each open head system and verifying the supply piping and nozzles are unobstructed.
HADDAM NECK 3/4 7-20
m ' PLANT SYSTEMS CO2 SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.6.3 The following High Pressure CO, Systems shall be OPERABLE with the minimum number of storage bottles shown Bach having a net weight of not less than 90% charge weight:
- a. Cable Vault - 18 bottles, and
- b. Primary Auxiliary Building Charcoal Filters - 8 bottles.
APPLICABILITY: Whenever systems, structures, components, or equipment protected by the High Pressure C02 Systems are required to be OPERABLE. ACTION: a.. With one or more of the above required High Pressure C0 Systems inoperable, within I hour establish a continuous fire w$tch with backup fire suppression equipment.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.6.3 Each of the above required High Pressure CO2 Systems shall be demonstrated OPERABLE:
- a. At least once per 6 months by verifying C0 storage bottle weight
'tobeatleast90%offullchargeweight,$nd
- b. At least once per 18 months by:
- 1) Verifying the system alarms and associated interlocks actuate manually and automatically upon receipt of a simulated test signal, and
- 2) Performance of an air flow test through system headers and nozzles to assure no blockage.
HADDAM NECK 3/4 7-21
I 1 g.
' PLANT SYSTEMS'.
HALON SYSTEMS LIMITING CONDITIONS FOR OPERATION 3.7.6-4 The following Halon 1301 Systems shall be OPERABLE with the minimum number of storage containers shown each having a net weight of not less than 95% and a pressure of.not less than 324 psig:
- a. Switchgear Room - 8 containers,
- b. Control Room - 3 containers.
APPLICABILITY: Whenever systems, structures, components, or equipment protected by the Halon System are required to be OPERABLE. ACTION:
- a. With one or more of the above required Halon 1301 Systems inoperable, within 1 hour establish a continuous fire watch with backup fire' suppression equipment.
b.. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.7.6.4 Each of the above required Halon 1301 Systems shall be demonstrated OPERABLE:
- a. At least once per 6 months by verifying each Halon storage container net weight to be at least 95% of full charge weight and with a pressure of not less than 324 psig (adjusted for temperature),and
- b. At least once per 18 months by:
- 1) Verifying the system valves and associated interlocks actuate manually and automatically upon receipt of a simulated test signal, and
- 2) Performance of an air flow test through headers and nozzles to assure no blockage.
HADDAM NECK 3/4 7-22
-PLANT SYSTEMS FIRE HOSE STATIONS ~
LIMITING CONDITION FOR OPERATION 3.7.6.5 'The fire hose stations given in Table 3.7-4 shall be OPERABLE. h APPLICABILITY:-Whenever systems, structures, components, or equipment in the areas protected by the Fire Hose Stations are required to be OPERABLE. ACTION:
- a. With one or more of the fire hose stations listed in Table 3.7-4 inoperable,
- 1) Within one hour provide a hose of equivalent capacity, or
- 2) Establish a continuous fire watch for the area of the affected hose station if the inoperable hose provides primary suppression capability and provide a hose of equivalent capacity within 24 hours,-or
- 3) Establish an hourly fire vatch patrol if the affected hose provides backup suppression capability and provide a hose of equivalent capacity within 24 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.7.6.5 ~ Each of the fire hose stations given in Table 3.7-4 shall be demonstrated OPERABLE:
- a. At least once per 31 days by visual inspection of the fire hose stations to assure all required equipment is at the station,
- b. At least once per 12 months, by:
- 1) Removing the hose for inspection, conducting a hose hydrostatic test at a pressure of at least 50 psig above maximum system pressure and re-racking, and
- 2) Inspecting all gaskets and replacing any degraded gaskets in the couplings.
- c. At least once per 3 years, by opening each hose station valve to verify valve OPERABILITY and no flow blockage.
HADDAM NECK 3/4 7-23
TABLE 3.7-4 FIRE HOSE STATIONS ' 1
.i FIRE WATER H0SE STATION 1 DESIGNATION AND LOCATION ELEVATION PROTECTED AREA )
i FHC - 3, Service Bldg, Corridor 21'6 Diesel- general Rooms-Locker / -j Shower: Area & H. P. Support area. FHC --4, Service Bldg. Corridor 21'6" Cable Spreading Area-Primary _ i Auxiliary Building - Cable-Vault FHC - 5, I&C Corridor 59'6" Corridor to CAS
'FHC - 6,. Service Bldg, Corridor 21'6" Cable Spreading Area-Primary Auxiliary Building FHC - 7, Auxiliary Building 21'6" Primary Auxiliary Bldg.
FHC - 8, Auxiliary Building-35'6" Primary Auxiliary Bldg. FHC - 9, Screenwell Bldg. 21'6" Service water pumps FHC - 10, Switchgear Room 41'6" Switchgear FHR - 1, Turbine Building, North 21'6" Air Compressors FHR - 2, Turbine Building, North 21'6" 011 Storage Room FHR - 3, Turbine Building, East 21'6" Under main turbine generator FHR - 4, Turbine Building, East 21'6" Feed pumps FHR - 5, Turbine Building, South 21'6" Under main turbine generator FHR'- 6, Turbine Building, South 21'6" Under main turbine generator FHR - 7, Turbine Building, West 21'6" Lube oil reservoir FHR - 8, Turbine Building, West 21'6" seal oil unit FHR - 9 Turbine Building, East 37'5" H, Feedwater control FHR - 10, Turbine Building 37'6" Feed heater #1 FHR - 11, Turbine Building, East 37'6" Switchgear Room FHR - 12, Turbine Building, West 37'6" Hp vapor extractors FHR - 13, Turbine Building, North 59'6" HTgh pressure turbine FHR - 14, Turbine Building, East 59'6" Control Room FHR - 15, Turbine Building, East 59'6" Low pressure turbines FHR - 16, Turbine Building, South 59'6" Generator / exciter FHR - 17, Turbine Building, South 59'6" Generator / exciter FHR - 18, Turbine Building, West 59'6" Low pressure turbines FHR - 19, Turbine Building, West 59'6" High pressure turbine FHR - 21, Maint. Shop Annex, North 21'6" Elect. cables FHR - 22, Maint. Shop Annex, South 21'6" Elect cables FHR - 23, Maint. Shop, East 21'6 " Elect. cables FHR - 24, Spent Fuel building, East 21'6" Fuel Storage Building FH Control Room 59'6" Main Control Board FH Switchgear Room 41'6" Switchgear FH Auxiliary Building at FHC #7 21'6" Auxiliary Building FH Auxiliary Building at FHC #8 35'6" Auxiliary Building FH Service Building at FHC #4 21'6" Cable Vault , Notes: FHR - Fire Hose Reel FH = Fire Hose FHC - Fire Hose Cabinet HADDAM NECK 3/4 7-24
PLANT SYSTEMS
. YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES LIMITING CONDITION FOR OPERATION 3.7.6.6 The yard fire hydrants and associated hydrant hose houses given~in Table 3.7-5 shall be OPERABLE.
APPLICABILITY:'Whenever systems, structures, components, or equipment in the areas protected by the Yard Fire Hydrants are required to be OPERABLE. ACTION:
- a. With one of more of the yard fire hydrants-(and associated hydrant hose houses) given in Table 3.7-5 inoperable, 1)- Within one hour connect sufficient additional lengths of 2% inch diameter hose on an OPERABLE hydrant to provide service to the unprotected area (s) if the inoperable hydrant or hydrant hose house is the primary means of fire suppression for any safety-related areas, or
- 2) Establish an hourly fire watch patrol in the safety-related area (s) protected by the inoperable hydrant or hydrant hose house if fire detection is available in the safety-related area (s) and provide the additional hose within 24 hours, or
- 3) Establish a continuous fire watch in the safety-related area (s) protected by the inoperable hydrant or hydrant hose house if fire detection is not available in the area and provide the additional hose within 24 hours, or
- 4) Provide additional hose within 24 hours if the inoperable hydrant or hydrant hose house is the backup means of fire suppression for any safety-related area (s).
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.7.6.6 Each of the yard fire hydrants and associated hydrant hose houses given in Table 3.7-5 shall be demonstrated OPERABLE:
- a. At least once per 31 days by visual inspection of the hydrant hose house to assure all required equipment is at the hose house, and that the hydrant is accessible and unobstructed,
- b. At least once per 6 months (once during March, April or May and {
once during September, October or November), visually inspecting ) each yard fire hydrant and verifying that the hydrant barrel is dry and that the hydrant is not damaged, and 4 HADDAM NECK 3/4 7-25
h p SURVEILLANCE REQUIREMENTS (Continued)-
- c. At least;once per 12 months ~, by:
- 1) Conducting a hose hydrostatic test (of the hose in the hydrant' hose house) at a pressure at least 50 psig above the 1, maximum system pressure,
- 2) Inspecting all the gaskets and replacing any degraded gaskets :
- in the couplings, and
- 3) Opening each valve and flushing each hydrant to verify valve OPERABILITY and no blockage.
HADDAM NECK 3/4 7-26
----me- :
v: - , ! j n' TABLE 3.7-5 O / p YARD FIRE HYDRANTS L LOCATION FIRE HYDRANT NUMBER !( Fire hydrantisouth of screen house * #1 FP-V-164 Fire' hydrant south of turbine building- #2 FP-V-165' i. Fire h'ydrant by . fuel . storage ~ tank * #3 FP-V-208 Fire hydrant south of radwaste storage building #4 FR-V-206 Fire' hydrant by recycle primary water storage tank * #5 FP-V-205
,' Fire hydrant'by north of tank farm * #6 FP-V-201
~ Fire hydrant. north of switchgear building #7 FP-V-200 Fire hydrant north of screen house #8 FP-V-196 Fire hydrant north west of diesel building #9 FP-V-425
- These fire hydrants have associated fire houses, HADDAM NECK 3/4 7-27
PLANT SYSID$ 3/4.7.7 FIRE RATED ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.7.7 All fire rated assemblies (walls, floor / ceiling, cable tray enclosures, and other fire barriers) separating safety-related fire areas or separating portions of redundant systems required OPERABLE for safe shutdown within a fire area and all sealing devices in fire rated assembly i penetrations (fire doors, fire dampers, cable, piping, and ventilation duct penetration seals) shall be OPERABLE. APPLICABILITY: At all times unless otherwise determined that the separation of safety-related fire areas or separating portions of redundant systems important to safe shutdown within a fire area is not required based on the MODE of operation. ACTION:
- a. With one or more of the above required fire rated assemblies and/or penetration sealing devices inoperable, within I hour:
- 1. Determine that the fire areas / zones on both sides of the affected fire rated assembly and/or penetration sealing device are monitored by either an OPERABLE fire detection or automatic suppression system at the fire barrier and establish a fire watch patrol that inspects both areas at least once per hour, n
- 2. Establish a continuous fire watch on at least one side of the affected fire rated assembly and/or penetration seal, E
- 3. Temporarily repair the inoperable fire rated assembly and/or sealing device with a qualified temporary assembly and/or seal and classify it as temporary.
All temporary fire rated assemblies and/or sealing devices shall be permanently repaired within 30 days, or implement ACTION 1 or 2 above.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.7.7.1 At least once per 18 months, the above required fire rated assemblies and penetration sealing devices shall be verified OPERABLE by performing a visual inspection of:
- a. The exposed surfaces of each fire rated assembly, HADDAM NECK 3/4 7-28
PLANT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued)
- b. At least 10% of the fire penetration pipe and cable assembly penetrations. This 10% shall include proportionally equal numbers :
of fire penetration pipe and cable assemblies containing each type j of penetration seal material. If a type of penetration seal ' material does not satisfy test acceptance criteria, an additional sample of such fire protection pipe and cable assembly h penetrations as was initially inspected shall be visually inspected. This examination process shall be continued until a. i sample of the particular seal material type equal to the number l initially inspected satisfies testing acceptance criteria. l Samples shall be selected such that all fire penetration pipe and I cable assembly penetrations will be inspected every 15 years. 4.7.7.2 At least once per 18 months each fire damper shall be verified ! OPERABLE by performing a TRIP ACTUATING DEVICE OPERATIONAL TEST. 4.7.7.3 Each of the required fire doors shall be verified OPERABLE by inspecting the automatic hold-open, release and closing mechanista and latches at least once per 6 months, and by verifying:
- a. That each locked closed fire door is closed at least once per 7 days,
- b. That each unlocked fire door with electrical supervision is closed at least once per 7 days.
- c. That each unlocked fire door without electrical supervision is closed at least once per 24 hours.
HADDAM NECK 3/4 7-29
PLANT SYSTEMS 3/4.7.8 FLAMMABLE LIOUIDS CONTROL LIMITING CONDITION FOR OPERATION 3.7.8 Flammable liquids in volumes greater than 1 pint shall be restricted from the control room except under the following conditions: l a. Written permission is obtained from the Supervising Control Operator or Shift Supervisor,
- b. The flammable liquid is contained in a suitable
- container not to exceed one quart in volume, and
- c. A dedicated fire watch is assigned to the activity.
APPLICABILITY: At all times. ACTION:
- a. Upon determination that flammable liquids in volumes greater than 1 pint are contained in the control room and do not meet the conditions stated above, the flammable liquid shall be removed immediately from the control room.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.7.8 None.
- A suitable container is non-spillable and has a flame arrestor in the nozzle.
HADDAM NECK 3/4 7-30
PLANT SYSTEMS 3/4.7.9 FEEDWATER ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.9 The~ feedwater isolation valves specified in Table 3.7-6 shall be OPERABLE with isolation times as shown in Table 3.7-6. APPLICABILITY: MODES 1, 2, 3 and 4 ACTION: With one or more' feedwater isolation valve (s) specified in' Table 3.7-6 ' inoperable:
- a. Restore the inoperable valve (s) to OPERABLE status within 72 hours; or
- b. Be in at'least H0T STANDBY within the next 6 hours and in COLD-SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.9.1 The isolation valves specified in Table 3.7-6 shall be demonstrated OPERABLE prior to returning the valve to service after. maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time, as required. 4.7.9.2 Each feedwater isolation valve specified in Table 3.7-6 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by verifying that on containment isolation actuation signal (CIAS), each feedwater isolation valve actuates to its isolation position. 4.7.9.3 The isolation time of each valve of Table 3.7-6 shall be determined to be within its limit when tested pursuant to Specification 4.0.5. HADDAM NECK 3/4 7-31
J s TABLE 3.7-6 FEEDWATER ISOLATION VALVES MAXIMUM ISOLATION VALVE FUNCTION JJE FW-MOV-11 Feedwater Isol. Valve ~ 70 Seconds FW-MOV-12 Feedwater Isol. Valve 70 Seconds 1 FW-MOV Feedwater Isol. Valve 70 Seconds l- FW-MOV-14 Feedwater Isol. Valve 70 Seconds 1 i l 1 I' HADDAM NECK 3/4 7-32 i
PLANT SYSTEMS i 3/4.7.10 EXTERNAL FLOOD PROTECTION LIMITING CONDITION FOR OPERATION 3.7.10 Flood protection shall be initiated in accordance with plant l procedures for all safety-related systems, components, and structures when the water level of the Connecticut River reaches 12 feet mean sea level, USGS datum, at CY, or upon receiving flood warning from CONVEX or the United - States National Weather Service, Northeast River Forecast Center. ! l l APPLICABILITY: At all times. ACTION:
- a. With the water level at 12 feet above mean sea level and when l forecasts from either CONVEX or the United States National Weather Service, Northeast River Forecast Center predict a level of 19 feet mean sea level, initiate the flood protection measures outlined in applicable plant procedures.
i b. With the water level at 16 feet above mean sea level and when forecasts from either CONVEX or the United States National Weather Service, Northeast River Forecast Center predict a level of 19 feet mean sea level, initiate and complete within eight hours the flood protection measures outlined in applicable plant procedures.
- c. With the water level at 19.0 feet above mean sea level, USGS datum, initiate an orderly shutdown to MODE 3, if applicable, considering plant operating status.
SURVEILLANCE RE0VIREMENTS 4.7.10 The water level at CY shall be determined to be within the limits by:
- a. Measurement at least once per 24 hours when the water level is forecasted by either CONVEX or the National Weather Service, Northeast River Forecast Center to be above elevation 12 feet mean sea level, USGS datum, and
- b. Measurement at least once per two hours when the water level is equal to or above elevation 11 feet mean sea level, USGS datum.
l HADDAM NECK 3/4 7-33
I PLANT SYSTEMS 3/4.7.11 PRIMARY AUXILIARY BUILDING AIR CLEANUP SYSTEM LIMITING CONDITION FOR OPERATION 3.7.11 The Primary Auxiliary Building Air Cleanup System shall be OPERABLE and in operation. 6 APPLICABILITY: MODES 1, 2, 3, and 4. Also, during operations involving movement of fuel assemblies or control rods within the containment. ACTION:
- a. With the Primary Auxiliary Building Air Cleanup System inoperable during MODES 1, 2, 3 and 4 restore the inoperable system to OPERABLE status within 7 days or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With the Primary Auxiliary Building Air Cleanup System inoperable or not in operation during operations involving movement of fuel assemblies or control rods within the containment, suspend all operations involving movement of fuel assemblies or control rods within the containment.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.7.11 The Primary Auxiliary Building Air Cleanup System shall be demonstrated OPERABLE and in operation:
- a. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following major painting, fire, or chemical release in any ventilation zone communicating with the system by:
- 1) Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and t'ne system flow rate is greater than 20,000 cfm;
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 10% at test corditions of 860F, 95% relative humidity, atmospheric pressure, and 40 feet / min face velocity in accordance with ASTM D3803; and HADDAM NECK 3/4 7-34
PLANT SYSTENS SURVEILLANCE REQUIREMENTS (Continued)
- 3) Verifying a system flow rate of greater than 20,000 cfm during system operation when tested in accordance with ANSI N510-1975.
- 4) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 1.6 inches Water Gauge while. operating the system at a flow rate of 20,000 cfm i10%.
- b. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1975 for a DOP test aerosol while operating the system at a flow rate of greater than 20,000 cfm; and
- c. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1975 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of greater than 20,000 cfm.
- d. After every 720 hours of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a. of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 10% at test conditions of 860F, 95% relative humidity, atmospheric pressure and 40 feet / min. face velocity in accordance with ASTM D3803.
- e. At least once per 18 months, verify a system flowrate of greater than 20,000 cfm during system operation aligned for Containment Purge.
HADDAM NECK 3/4 7-35
3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCE. 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary Coolant System pressure will be limited to below 110%,(1100 psia), of its design pressure of 1000 psia during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser). The specified valve lift settings and relieving capacities are in accordance with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition. The design total relieving capacity for all valves on all of the steam lines is 9,504,000 lbs/hr which is 120% of the total secondary steam flow of 7,872,000 lbs/hr at 100% RATED THERMAL POWER. 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of offsite power. Each steam turbine-driven auxiliary feedwater pump has a capacity sufficient to ensure adequate delivery of feedwater flow to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F within the Residual Heat Removal System operating range. With one auxiliary feedwater pump inoperable, the safest mode of operation is HOT SHUTDOWN with the decay heat removal function capable of being provided by the RHR System. With two steam turbine-driven feedwater pumps inoperable, at least one pump must be restored to OPERABLE with 24 hours from the time that the second pump is declared inoperable, or be in HOT STANDBY within the next six hours and in HOT SHUTDOWN with the following six hours. In addition, both the pumps must be restored to OPERABLE within 72 hours from the time of initial loss of the first pump or be in HOT STANDBY in the next six hours and HOT SHUTDOWN within the following six hours. 3/4.7.1.3 AUXILIARY FEEDWATER SUPPLY The OPERABILITY of the demineralized water storage tank (DWST) and primary water storage tank (PWST) with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 16 hours with steam discharge to the atmosphere concurrent with total loss-of-offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. HADDAM NECK B3/4 7-1
PLANT SYSTEMS l l BASES AUXILIARY FEEDWATER SUPPLY (Continued) In addition, the auxiliary feedwater system can be initiated manually. In this case, feedwater is available from the DWST by gravity feed to the auxiliary feedwater pump. The specified 50,000 gallons of water in the DWST is adequate for decay heat removal for a period of at least 2 hours. Within this period, decay heat removal demands are reduced to approximately 150 gpm. Makeup water is available during this period from the PWST which contains a minimum volume of 80,000 gallons. The PWST transfer pumps can l transfer 200 gpm from the PWST to the DWST. An alternate supply can be I previded from the 100,000 gallons Recycled Primary Water Storage Tank. ! 3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident 0.4 gpm reactor-to-secondary tube leak in the steam gener5 tor of the affected steam line. These values are consistent with the assumptions used in the safety analyses. 3/4.7.1.5 MAIN STEAM LINE TRIP VALVES The OPERABILITY of the main steam line trip valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam line trip valves within the closure times of the Surveillance Requirements are consistent with the assumptions used in the safety analyses. 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-inducedstressesinthesteamgeneratorsdonotexceedthgmaximum allowable fracture toughness stress limits. Thg limitations of 70 F and 200 psig are based on a steam generator RTNDT of 10 F and are sufffcient to prevent brittle fracture. The heatup and cooldown rate of 100 F/hr for the steam generators are specified to ensure that stresses in these vessels are maintained within acceptable design limits. . 3/4.7.3 SERVICE WATER SYSTEM The OPERABILITY of the Service Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analysis. A service water header is comprised of the two service water pumps associated with each diesel generator and the safety-related piping and components. HADDAM NECK B3/4 7-2
PLANT SYSTEMS BASES 3/4.7.4 SNUBBERS All snubbers are required to be OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained during and following a seismic or other event initiating
' dynamic loads. Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure, or failure of the system on which they are installed, would have no adverse effect on any safety-related system.
The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule. When the cause.of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection or are similarly located or exposed to the same envircamental conditions, such as temperature, radiation, and vibration. When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversely affected by the inoperability of the snubber. The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system. To provide assurance of snubber functional reliability, a representative sample of the installed snubbers will be functionally tested during plant shutdowns at 18-month intervals. These tests will include stroking of snubbers to verify freedom of movement over the full stroke, restraining characteristics, and drag force (if applicable). Ten percent (10%) of the total of each type of snubber represents an adequate sampling for these tests. Observed failures on these samples require testing of additional units. Hydraulic snubbers and mechanical snubbers may each be treated as a different entity for the above surveillance program. HADDAM NECK B3/4 7-3 l
PLANT SYSTEMS BASES The service life is evaluated via manufacturer input and information with consideration of snubber service conditions and associated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, high temperature area, location of potential fluid transient loading, etc.) The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of age and operating conditions. (Due to implementation of the snubber service life monitoring program after several years of plant operation, the historical records to date may be incomplete. ) Snubber service life records will provide a statistical basis for future consideration of snubber service life. 3/4.7.5 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism. 3/4.7.6 FIRE SUPPRESSION SYSTEMS The OPERABILITY of the Fire Suppression Systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety-related equipment is located and to ensure any fires occurring outside safety related areas do not affect safety related areas. The Fire Suppression System consists of the Fire Water Supply / Distribution System, Spray And/0r Sprinkler Systems, C07 Systems, Halon Systems, Fire Hose Stations, and Yard Fire Hydrants and associated Hydrant Hose Houses. The Fire Suppression Systems are adequate to minimize potential damage to safety-related equipment and are a major element in the facility Fire Protection Program. HADDAM NECK B3/4 7-4
PLANT SYSTQi1 l BASES In the event that portions of the Fire Suppression Systems are inoperable, I alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression. The Surveillance Requirements provide assurance that the minimum OPERABILITY requirements of the Fire Suppression Systems are met. In the event the Fire Suppression Water System becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant. 3/4.7.7 FIRE RATED ASSEMBLIES The functional integrity of the fire rated assemblies and barrier penetrations ensures that fires will be confined (for a time of at least equal to the minimum time rating of the associated fire barrier) from spreading to adjacent portions of the facility. These design features minimize the possibility of a single fire rapidly involving several areas of the facility prior to detection and extinguishing of the fire. The fire barrier penetrations are a passive element in the facility Fire Protection Program and are subject to periodic inspections. Fire barrier penetrations, including cable penetration barriers, fire doors and dampers are considered functional when the visually observed condition is the same as the as-designed condition. During periods of time when a barrier is not functional, alternate measures are taken to prevent the possible spread of fire. These measures include verifying the operability of fire detection or suppression systems on both sides of the affected barrier and establishing a fire watch patrol, or posting a continuous fire watch in the vicinity of the affected barrier, or installation of a temporary fire stop pending restoration of the permanent seal. 3/4.7.8 FLAMMABLE LIOVIDS CONTROL The control of flammable liquids in the control room substantially reduces any fire loadings in the control room. Specification 3/4.7.8 also satisfies a NRC condition of an Appendix R control room exemption issued November 14, 1984. HADDAM NECK B3/4 7-5
PLANT SYSTEMS BASES 3/4.7.9 FEEDWATER ISOLATION VALVES The accident analysis for a main steam line break assumes that the main feedwater isolation valves will close on a containment isolation actuation signal (CIAS). Also, the closure of these valves based on a CIAS is credited in determining the Pressure / Temperature limits for the purpose of environmental qualification. The feedwater isolation valves act as a backup to the feedwater regulation valves in the event a feedwater regulation valve fails open during a Main Steam Line Break. 3/4.7.10 EXTERNAL FLOOD PROTECTION The thresholds regarding flood protection ensure that facility protective actions will be taken (and the orderly shutdown of the plant to MODE 3 will be made) in the event of flood conditions. The estimated Connecticut River probable maximum flood (PMF) level, including wave effects (i.e., still water level), is 39.5 feet mean sea level. Normal flood control measures provide protection to safety-related equipment to El. 30 feet mean sea level . Normal flood protection to this elevation is based on a low probability of exceedance and structural capacity limitations. Based on the one to two day rise period of the PMF, alternative means of providing decay heat removal for flooding events up to the PMF is provided in A0P 3.2-24. 3/4.7.11 PRIMARY AUXILIARY BUILDING AIR CLEANUP SYSTEM PAB ventilation is accomplished using two supply units, two roof ventilation exhaust units, the ventilation and purge subsystem and the ducting required to connect the units in a common ventilation system for the entire building. Both of the supply units are provided with filtration and steam heating capabilities. The ventilation and purge exhaust subsystem includes prefilters, high efficiency particulate and charcoal filters and two large capacity exhaust fans. The radiological consequences analyses for loss-of-coolant accidents assume Primary Auxiliary Building efficiencies which are ensured by this Technical Specification. Also, in consideration of a fuel handling accident inside containment, (i.e., when the containment is being purged) the purge discharge would be directed through the Primary Auxiliary Building charcoal filters. Credit is again taken for these filters in reducing the radiological consequences. HADDAM NECK B3/4 7-6
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Docket'No; 50-213' B13181 ; y [p. .
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Haddam Neck Plant !
.. Description of Individual Proposed Changes.' j
- to the Technical Specifications and Discussion q on'the Significant Hazards Consideration i
3/4.4 -' Reactor Coolant System j 3/4.6 Containment. Systems ) 3/4.7 P1 ant Systems- -1 i i I a
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June 1989
" Docket No. 50-213 B13181' Haddam Neck Plant Technical Specification Section 3/4.4, Reactor Coolant System l.
L 1 June 1989
Attachment 2 Section 3/4.4 B13181/Page 1 Technical Specification Section 3/4.4. Reactor Coolant System The proposed revised Technical Specification (RTS) Section 3/4.4 has been prepared by converting the existing Technical Specification Section 3.3, Reactor Coolant System, Section 3.4, Combined Heatup, Cooldown and Pressure Limitations, Section 4.10, Inservice Inspection and Reactor Vessel Surveil-lance, Section 4.10.1, Inservice Inspection of Steam Generator Tubes. Table 4.2-1, Minimum Frequencies for Testing, Calibrating and/or Checking Instrument Channels, and Table 4.2-2, Minimum Equipment Check and Sampling Frequencies to a format consistent with the Westinghouse Standard Technical SpecificationsdHSTS). In addition, applicable Administrative Technical Specifications at the Haddam Neck Plant have also been included in the proposed RTS. The proposed changes are compared to the existing Technical Specifications and the H STS. A matrix summarizing this comparison is included in Attachment 3. Section 3/4.4.1.1. Start-up and Power Ooeration This section in the proposed RTS for the Haddam Neck Plant is the same as the existing Technical Specification Section 3.3.1.1. Th Technical Specification Section 3.3.1.1 was approved by the NRC.$2pxistingNo changes have occurred to this section other than the renumbering to achieve consistency with the H STS. The proposed Section is an enhancement to the existing Technical Specifications. It provides clear applicability, action and surveil-lance requirements modeled after the H STS. 3/4.4.1.2. Hot Standby This section in the proposed RTS for the Haddam Neck Plant is same as the existing Technical Specification Section 3.3.1.2. Th Technical Specification Section 3.3.1.2 was approved by the NRC.i3pxistingNo changes have occurred to this section other than the renumbering to achieve consistency with the the H STS. The proposed section is an enhancement to the existing Technical Specifications. It provides clear applicability, action and surveillance requirements modeled after the H STS. 3/4.4.1.3. Hot Shutdown This section in the proposed RTS for the Haddam Neck Plant is same as the existing Technical Specification Section 3.3.1.3 with the exception that the proposed RTS section no 1onger includes the definition of what an RHR loop consists of. This definition is a footnote in the existing Technical (1) Administrative Technical Specifications at the Haddam Neck Plant are administrative procedures that were implemented as an interim measure prior to converting the Technical Specifications to the H STS format. (2) F. M. Akstulewicz letter to E. J. Mroczka, " Technical Specifications for Cycle 15, Operation," dated November 12, 1987. (3) Ibid. l
Attachment 2 Section 3/4.4 B13181/Page 2 Specification (4)This definition is covered in the proposed RTS Section 1.0
" Definition," (Section 1.27) and is equivalent to the existing Technical Specification. This Specification Section 3.3.1.3 was approved by the NRC (5fxisting _TechnicalNo other changes have occurred to this se than the renumbering to achieve consistency with the W STS. The proposed section is an enhancement to the existing Technical Specifications. It provides clear applicability, action and surveillance requirements modeled after the H STS.
3/4.4.1.4.1. Cold Shutdown Looos Filled This section in the proposed RTS for the Haddam Neck Plant is the same as the existing Technical Specification 3.3.1.4g) The existing Technical Specifi-No changes have occurred to this cation 3.3.1.4.1 was approved by the NRC. section other than the renumbering to achieve consistency with the W STS. The proposed section is an enhancement to the existing Technical Specifications. It provides clear applicability, action and surveillance requirements modeled after the H STS. 3/4.4.1.4.2. Cold Shutdown loops Not Filled This section in the proposed RTS for the Haddam Neck Plant is the same as the existing Technical Specification Section 3.3.1.4.2. Technical Specification Section 3.3.1.4.2 was approved by the NRC.(gisting No changes have occurred to this section other than the renumbering to achieve consistency with the W STS. The proposed section is an enhancement to the existing Technical Specifications. It provides clear applicability, action, and surveillance requirements modeled after W STS. Significant Hazards Consideration - Sections 3/4.4.1.1. 3/4.4.1.2. 3/4.4.1.3. 3/4.4.1.4.1 and 3/4.4.1.4.2 In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because the changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The content of the RTS is the (4) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Amendment Request for Sections 1.0, 3/4.2, 3/4.9, 3/4.10, 5.0 and 6.0 of the revised Technical Specifications," dated October 26, 1988.
(5) F. M. Akstulewicz letter to E. J. Mroczka, " Technical Specifications for Cycle 15, Operation," dated November 12, 1987. (6) Ibid. I y (7) Ibid. L - - - - - - - - _ _ -
Attachment 2 Secttn 3/4.4 B13181/Page 3 same as the previously approved version of the existing Technical Specification Sections. No changes have occurred to this section other than the renumbering to achieve consistency with the H STS. Therefore, there is no increase in the probability or consequences of an accident previously analyzed.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. The proposed changes do not impact the operation of any component or system. The proposed changes do not introduce any new failures. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from those previously analyzed.
- 3. Involve a significant reduction in a margin of safety. Since the proposed changes do not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.
3/4.4.1.5. Isolated Loon Modes 1 and 2 This proposed RTS section provides a Limiting Condition for Operation (LCO) requirement for the loop stop valves of an isolated loop and boron cohcentra-tion requirements for Modes 1 and 2. These requirements are equivalent to those of the existing Technical Specificatg Section 3.3.1.5 for the asso-was submitted to the NRC. This ciated modes. A proposed amendment request amendment request was submitted to separate the original specification (which was applicable to Modes 1, 2, 3, 4, and 5) into two separate specifications; one for Modes 1 and 2 and another for Modes 3, 4, 5, and 6. No changes have occurred to this section other than renumbering to achieve consistency with the H STS. The proposed section is an enhancement to the existing Technical Specifications. It provides clear applicability, action, and surveillance requirements modeled after the H STS. 3/4.4.1.6. Isolated looo - Modes 3. 4. 5. and 6 This proposed RTS section provides an LC0 for the loop stop valves and boron concentration requirements of an isolated loop for Modes 3, 4, 5, and 6. These requirements are equivalent to those of the existing Technical Specifi-cation Section 3.3.1.5 for the associated modes with the following exceptions:
- 1) The proposed RTS LCO specifies boron concentration of the isolated loop to be maintained to meet shutdown margin requirements or the
- refueling boron concentration. The original specification required l the boron concentration of the isolated loop to be maintained greater than or equal to that of the operating loops.
- 2) The APPLICABILITY of the proposed RTS was expanded to include Mode 6, whereas the original specification was applicable to only Modes 3, 4, and 5.
l (8) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, Technical Specification Changes, Reactor Coolant System, dated January 10, 1989. l. l
l Attachment 2 1 Section 3/4.4 813181/Page 4 These changes were included in an amendment requestI9) which was submitted to the NRC on January 10, 1989. No changes have occurred to this section other than renumbering to achieve consistency with the H STS. The proposed section is an enhancement to the existing Technical Specifications. It provides clear applicability, action, and surveillance requirements modeled after the M.STS. 3/4.4.1.7. Isolated Loon Start-up This proposed RTS section provides the conditions for removing a reactor coolant loop from isolated status and is equivalent to the existing Technical Section 3.3.1.6 with the following exceptions:
- 1) The proposed RTS provides a footnote defining an operating loop.
- 2) The proposed RTS provides a footnote that exempts APPLICABILITY requirements when a hot leg stop valve is opened to create an idled loop.
- 3) The boron concentration and subcriticality requirements have been revised to maintain shutdown margin requirements.
- 4) Surveillance requirements have been modified to include loop stop valve temperature interlock verification per the existing Technical Specification Table 4.2-1 item 22.
- 5) APPLICABILITY was revised to include a statement that the provisions of Specification 3.0.4 do not apply.
- 6) An additional Surveillance requirement has been added (4.4.1.7.4) to demonstrate pump / valve interlock at least once per refueling outage.
These changes (except item 6 above) were included in the amendment request (10) which was submitted to the NRC on January 10, 1989. Item 6 has been added to the RTS Section to include the surveillance require-ment from the existing Technical Specification Table 4.2-1, item 23. This is necessary to ensure equivalence of the RTS with the existing Technical Speci-fications. The only other change is renumbering of the section per the H STS format. (9) Ibid. (10) Ibid.
Attachment 2-
~Section 3/4.4 B13181/Page 5
'3/4.4.1.8. Idled Loon - Modes 1 and 2 This proposed RTS Section provides an LC0 requirement for the cold leg loop
-stop valves.of an idled loop and boron concentration' requirements for Modes 1-and. 2. These requirements are equivalent to those of existing Technical Specifications Section 3.3.1.7 for the associated modes.
l These changes were included in an amendment requestIII) which was submitted to the NRC on' January - 10, 1989. No other changes have occurred to this section other than renumbering to achieve consistency with the M STS. 3/4.4.1.9. Idled loop - Modes 3. 4. 5. and 6 This proposed change provides an LC0 requirement for the cold leg loop stop valves of an idled . loop and boron concentration requirements for Modes 3, 4, 5, and 6. These requirements are equivalent to those of existing Technical Specification Section 3.3.1.7 for the associated modes with the following
- . exceptions
- 1) The propossd RTS LC0 specifies boron concentration of the isolated loop to- be' maintained to meet shutdown margin requirements or the refueling. boron concentration. The original specification required the boron concentration of the isolated loop to be maintained greater than or equal to that of the' operating loops.
- 2) _ The APPLICABILITY of the proposed RTS was expanded to include Mode 6, whereas the original specification was applicable to only Modes 3, 4, and 5.
These changes are consistent with an amendment request (12) which was submitted to the NRC on January 10, 1989. No other changes have occurred to this section other than renumbering to achieve consistency with the H STS. 3/4.4.1.10. Idled loop Start-uo The proposed RTS section provides conditions for removing a reactor coolant loop from -idled status for Modes 1 and 2 and is equivalent to the existing Technical Specification Section 3.3.1.8 for the associated modes with the following exceptions:
- 1) The proposed RTS deletes the requirement that the reactor be subcritical by more than 1000 pcm if more than one loop is idled.
This requirement is not applicable for Modes 1 or 2.
- 2) The APPLICABILITY of the proposed RTS section was revised to include the statement that the provisions of Specification 3.0.4 do not apply.
(11) Ibid. (12) Ibid.
L 1 Attachment 2 L ' Section 3/4.4 B13181/Page 6 The above changes were include g an-amendment request which was submitted to the NRC on January 10, 1989. No other changes have occurred to this section other than renumbering to achieve consistency with the M STS. 3/4.4.1.11. Idled looo Start-up The proposed RTS section provides conditions for removing a reactor coolant loop from idled status for Modes 3, 4, 5, and 6 and is equivalent to the existing Technical . Specification Section 3.3.1.8 for the associated modes with the following exceptions:
- 1) The proposed RTS section provides a footnote defining what an operating loop may be considered as.
- 2) The boron concentration and subcriticality requirements have been revised to maintain shutdown margin requirements.
- 3) The requirement that the reactor power be no greater than 60% has been removed. This is not applicable to Modes 3, 4, 5, or 6.
- 4) The proposed RTS surveillance requirement includes a waiver if the idled loop is being started within 30 minutes and the cold leg stop valve was closed for less than 15 minutes.
- 5) Additional surveillance requirements have been added to include stop valve temperature interlock verification per the existing Technical Specification Table 4.2-1 item 22, valve temperature interlocks, and item 23, pump / valve interlocks at least once per refueling outage.
These g es which (except items 4 and 5 above) were included in an amendment was submitted to the NRC on January 10, 1989. Item 4 was request added to allow reactor coolant pump restart post reactor trip (which requires closure of the cold leg stop valves) without having to perform the surveil-lance. Two of four reactor coolant pumps (RCP) coast down approximately one minute after turbine / reactor trip. In order to regain forced circulation, these RCPs can be repowered from offsite power. The procedure to restart these RCPs requires the cold leg loop isolation valve (LIV) to be isolated, the bypass valves opened, the RCP to be started, and then the cold leg LIV to be opened. During this evolution, the loop is technically an idled loop and the proposed RTS for idled loop start-up, including the surveillance requirements, need to be followed. During the immediate posttrip time frame (30 min.), forced RCS circulation is desirable. The slight variations in RCS temperature or boron concentration that could occur in the small time window when the cold leg LIV is closed will (13) Ibid. (14) Ibid.
Attachment 2
'Section 3/4.4 B13181/Page 7 not affect event consequences. The increase in core coolant flow will lessen event consequences. Thus, it is acceptable to waive the RCS temperature and boron concentration requirements in this particular situation.
Item 5 above has been added to incorporate the surveillance requirements from the existing Technical Specification Table 4.2-1 item 22 and 23. This is necessary to ensure that the RTS requirements are equivalent to the existing Technical Specifications. No other changes have occurred to this section other than renumbering to achieve consistency with the M STS. Significant Hazards Consideration - Sections 3/4.4.1.5. 3/4.4.1.6. 3/4.4.1.7. 3/4.4.1.8. 3/4.4.1.9. 3/4.4.1.10. and 3/4.4.1.11 In a letter dated January 10,1989,05) CYAPC0 submitted a proposed revision to the existing Technical Specification 3.3.1.5 through 3.3.1.8. As stated above, the proposed RTS sections are consistent with the changes proposed in that letter. In addition, additional changes were made to ensure total equivalence with the existing Technical Specifications. Therefore, the significant hazards consideration discussion included in the above referenced letter is also applicable to the proposed RTS. 3/4.4.2.1. Safety Valves Shutdown This proposed RTS section provides requirements for pressurizer Code safety valve operability for Mode 4 and is equivalent to the existing Technical Specification Section 3.3.2.1. The Surveillance Requirements in the proposed RTS section also include the additional testing requirements identified on Table 4.2-2 item 6 of the existing Technical Specifications. Incorporation of this testing in this section is necessary since Table 4.2-2 is being deleted by the RTS. This ensures total equivalence with the existing Technical Specifications. No other changes have occurred to this section other than renumbering to achieve consistency with the M STS. 3/4.4.2.2. Safety Valves Operatina This proposed RTS section provides requirements for pressurizer Code safety valve operability for MODES 1, 2, and 3 and is equivalent to existing Tech-nical Specification section 3.3.2.2. The Surveillance Requirements in the proposed STS section also include the additional testing requirements identi-fied on Table 4.2-2 item 6 of the existing Technical Specifications. Incorpo-ration of this testing in this section is necessary since Table 4.2-2 is being deleted by the RTS. This ensures total equivalence with the existing Tech-nical Specifications. No other changes have occurred to this section other than renumbering to achieve consistency with the H STS. l (15) Ibid.
Attachment 2 Section 3/4.4 B13181/Page 8 3/4.4.3. Pressurizer This proposed RTS section provides requirements for pressurizer operability and is equivalent to the existing Technical Specification Section 3.3.3 with the following exception: '
- 1) The RTS provides an additional surveillance requirement (sec- l tion 4.4.3.3). This is necessary to verify compliance with the l corresponding LCO (section 3.4.3.b). This surveillance requires that the pressurizer heaters be energized from emergency power once ;
every 18 months. This action would not place the plant in a mode , that would negatively impact safe plant operation. This is also consistent with the H STS. This represents an additional surveillance requirement and is more restrictive than the existing Technical Specifications. The only other change is the renumbering of the section per M STS format. 3/4.4.4. Relief Valves This proposed RTS section is equivalent to the existing Technical Specifica- ' tion Section 3.3.4.1 with the following exception:
- 1) The RTS requires the performance of an " ANALOG CHANNEL OPERATIONAL TEST" whereas the existing Technical Specification requires the i performance of a " CHANNEL FUNCTIONAL TEST." l As specified in the definition sections (the RTS 1.2 and the existing Tech-nical Specification 1.11) the intent of these tests are equivalent. This is considered only a nomenclature change. This proposed change also incorporates the surveillance requirements specified in the Administrative Technical
? specification 4.2 (Table 4.2-2 item 15).
'he only other change is renumbering of the section per the H STS format.
e <nificant Hazards Consideration - Sections 3/4.4.2.1. 3/4.4.2.2 and 3/4.4.3 i and 3/4.4.3 ! In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are , not compromised. The proposed changes do not involve a significant hazards ( consideration because the change would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The content of the RTS is the same as the '
previously approved version of the existing Technical Specifications. No i changes have occurred to this section other than the renumbering to achieve consistency with the H STS. Therefore, there is no increase in i probability or consequences of an accident previously analyzed. l
- 2. Create the possibility of a new or different kind of accident from any ;
previously evaluated. The proposed changes do not impact the operation ! of any component or system. The proposed changes do not introduce any . i
Attachment 2 Section3/4.4
- 813181/Page 9 new failures. Therefore, the proposed changes do not create the possi-bility of a new or different kind of accident from those previously analyzed.
- 3. Involve a significant reduction in a margin of safety. Since the pro-posed changes do not affect the consequences of an accident previously analyzed, there is no reduction in the margin or safety.
3/4.4.5. Steam Generators This proposed RTS section provides a new section for Steam Generator Opera-bility. There is no comparable section in the existing Technical Specifica-tions. However, surveillance requirements from the existing Technical .Speci-
.fications Section 4.10.1 have been incorporated into the RTS section. The new LC0 and Action Statement represent additional, new limitations on plant operation and are more restrictive than the existing Technical Specifications.
The Surveillance requirements are equivalent to those in the existing Tech-nical Specification Section 4.10.1 with the following exceptions:
- 1) The RTS (4.4.5.4.a.8) allows cold leg entry inspections as well as-hot leg entry inspections. The choice of entry side has no adverse effect on the inspection capability or results. The intent of Regulatory Guide 1.83 on tube inspection is still met using the alternate inspection method. Since the inspections performed from the cold leg side are equivalent to those performed from the hot leg side, this change does not impact the consequences of any design basis accident.
- 2) Section 4.4.5.3.c.2 specified unscheduled inspections after seismic occurrences that exceed 1/2 the Safe Shutdown Earthquake. The existing Technical Specifications specified the Operating Basis Earthquake. This is only a nomenclature difference, the two terms are equivalent.
- 3) The note in RTS section 4.4.5.2 on degraded tubes excludes imperfec-tions within the tube-to-tubesheet roll region whereas the existing Technical Specifications did not specifically exclude these. This change is provided to reflect a previously modified and approved definition of " degraded tube" per Amendment 96 to the existing Technical Specifications. The proposed RTS is equivalent to this and does not impact the consequences of any design basis accident.
- 4) A footnote has been provided for RTS table 4.4-2 that allows an exception to the requirements of Section 4.4.5.3.c.1 when it has been determined that the indicated inspections are not required.
This exception allows the plant to shut down prior to exceeding the 4.4.5.3.c.1 criteria to plug / repair tubes for operational conven-ience and avoid a full first sample inspection. This footnote represents further clarification of the evaluation provided for Section 4.4.5.3.c.1.
- 5) The RTS definition of " plugging limit" for imperfections in Section 4.4.5.4.a.6 has been revised to include repaired tubes as well as tubes removed from service, and deleted the reason for the plugging
~
- Attachment 2 Secticn 3/4.4.
L B13181/Page 10-L limit, i.e. "because it may become unserviceable prior to the next inspection."' This clarification was added to make the text of the definition in regards to repairing consistent with subsection 4.4.5.4.a.6.a which ' was previously revised. (Amendment 96 to the existing Technical Specifications.) 6). The RTS . sampling selection criteria in Section 4.4.5.2 provides clarification that 50% of the tubes inspected 'shall .come from s critical areas. The existing .-Technical Specifications did -not~ specify how much, it just specified that tubes shall be selected on a random basis unless prior experience indicates the need to-inspect critical areas. .The proposed change provides additional clarifica-tion on how to inspect these critical areas.
- 7) .The proposed RTS section also incorporates additional clarifications.
such as:
- Include reporting requirements of the existing Technical Specification sections.
- Deletion of information only applicable to previous cycles of operation.
- Format and layout of the specification.
The above items are judged to be inconsequential. The 'above changes represent clarifications to the existing surveillance requirements for the steam generator tube inspections. Significant Hazards Consideration - Section 3/4.4.5 In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration. The basis for.this conclusion is that the three criteria of 100FR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because the changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously analyzed. The proposed RTS (Section 4.4.5.4.a.2) will allow an alternate method for meeting the intent of Regulatory Guide 1.83 that a tube inspection include the hot leg and U-bend seg-ments. Specifically, the change will allow inspections from either the
- hot leg side or the cold leg side through to the tube end on the opposite side to be considered a valid tube inspection. As is currently the case, an inspection from the hot leg to the top support on the cold leg will still be considered a valid tube inspection. The proposed change will l provide the flexibility for steam generator tube inspection still meeting L the intent of Regulatory Guide 1.83. Therefore, this change does not impact the consequences of any design basis accident. The other change in the proposed RTS are strictly administrative in nature. Therefore, it is concluded that previously analyzed accidents are not affected.
Attachment 2 Section 3/4.4 B13181/Page 11
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. The proposed changes do not impact the operation of any component or system. The proposed changes do not introduce any i
new failures. Therefore, the proposed changes do not create the possi- , bility of a new or different kind of accident from those previously analyzed.
- 3. Involve a significant reduction in a margin of safety. Since the pro-posed changes do not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.
3/4.4.6.1. Leakaae Detection System This proposed RTS change provides requirements for the reactor coolant system (RCS) leakage detection system. The leakage detection system consists of the Volume Control Tank (VCT) level monitoring system, the containment main sump level monitoring system and the containment atmosphere gaseous radioactivity monitoring system. The proposed RTS section is equivalent to an amendment request which was submitted to the NRC on May 30, 1989. Significant Hazards Consideration - Section 3/4.4.6.1 In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because the changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously analyzed. The proposed changes will increase the surveillance and operability requirements on the Leakage Detection System (LDS). The LDS is not directly credited in any previously analyzed accidents. The changes will not impact any systems or equipment credited in previous analyses. Thus, the changes do not increase the probability or consequences of any accident previously analyzed.
- 2. Create the possibility of a new or different kind of accident from any previously analyzed. Since there are no changes in the way the plant is operated, the potential for an unanalyzed accident is not created. No new failure modes are introduced.
- 3. Involve a significant reduction in a margin of safety. The changes do not impact any previously analyzed consequences, safety analysis, or protective boundaries. Thus, the changes do not reduce the margin of safety.
3/4.4.6.2. Operational Leakaae This section will be submitted to the NRC by June 30, 1989.
Attachment 2 Sectien 3/4.4 B13181/Page 12 3/4.4.7. Chemistry The proposed RTS imposes steady state and transient limits on the concentra-tion of chloride, fluoride and dissolved oxygen for the different plant operating modes. Primary coolant is required to be sampled and analyzed every 72-hours to ensure that RCS chemistry is within the required limits. If steady-state limits are exceeded, the appropriate parameters must be corrected within 24 hours, or the plant must be brought to cold shutdown. If transient chemistry limits are exceeded, the plant must also be brought to cold shut-down. Exceeding chloride or fluoride limits during cold shutdown or refueling requires that RCS pressure be maintained below 500 psig. An engineering evaluation must then be conducted to determine whether RCS structural integ-rity has been compromised. The proposed RTS is also consistent with the M STS. Significant Hazards Consideration - Section 3/4.4.7 In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because the changes would not:
- 1. Involve a significant increased in the probability or consequences of an accident previously evaluated. The proposed RTS imposes limitations on RCS chemistry to ensure that corrosion of the RCS is minimized and potential for RCS leakage or failure due to stress corrosion is reduced.
Maintaining chemistry within the limits specified ensures structural integrity of the RCS over the life of the plant. Therefore, there is no increase in probability or consequences of an accident previously ana-lyzed.
- 2. Create the possibility of a new or different kind of accident from any plant hardware modifications. The proposed changes do not introduce new failures. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from those previously analyzed.
- 3. Involve a significant increase in a margin of safety. Since the proposed changes do not affect the consequences of an accident previously ana-lyzed, there is no reduction in the margin of safety.
3/4.4.8. Specific Activity This proposed RTS section provides requirements for reactor coolant specific activity and is equivalent to the existing Technical Specification Sections 3.2 and Table 4.2-2 item (1) with the following exceptions:
- 1) The limits provided in the proposed RTS section are on a per gram basis whereas the existing Technical Specification limits are on a per m1 basis. The H STS format utilizes the per gram basis units.
These are equivalent for all practical purposes.
- 2) The proposed RTS section provides additional requirements for I-131 activity whereas the existing Technical Specifications only 1
> Attachment 2 ! S:cticn 3/4.4 B13181/Page 13 addressed Gross Activity. This additional requirement is more restrictive than the existing Technical Specification.
- 3) The proposed RTS section also provides more restrictive surveillance requirements for Iodine Analysis during transient plant conditions.
- 4) The proposed RTS section provides a sampling frequency for Gross Radioactive Determination of once per 72 hours. The existing Technical Specification frequency per Table 4.2-2 item (1) is 5 days per week.
The first 3 items above are either equivalent or more restrictive. Item 4, however, provides a sampling frequency that is leli restrictive than the existing Technical Specifications. This has been evaluated by CYAPC0 and determined acceptable for the following reasons:
- 1) The LCO limit remains the same as the existing Technical Specifica-tion, only the surveillance frequency during steady state operation is slightly decreased.
- 2) The proposed surveillance is for steady state operation when RCS activities are not expected to change frequently.
- 3) Increases surveillance are required during power transients (start-up/ shutdown) when RCS activity is expected to change and when the fuel is more susceptible to failure.
- 4) Increases in coolant activity of any significance between sampling periods would be detected by the letdown radiation monitor.
- 5) The proposed surveillance frequency is consistent with the H STS.
Significant Hazards Consideration - Section 3/4.4.8 In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because the changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. As stated above, the proposed changes provide a sampling frequency for gross radioactivity determination of once per 72 hours. The existing Technical Specifications frequency per Table 4.2-2 item (1) is 5 days per week.
These analyses generally assume that the reactor coolant is at equilib-rium technical specification activity concentrations at the start of the transient. In many situations, such as the steam generator tube rupture, a coincident operating transient iodine spike is also assumed. The equilibrium RCS activity would have to increase above the Technical Specification value for the consequences of these design basis accidents to be impacted.
Attechment 2 Section 3/4.4 B13181/Page 14 The number of fuel failures required to achieve this level of activity in a short period of time (within the 3 day sampling per$od) could only be postulated as a result of an operational transient when increased sam-pling is required by the proposed change. This latter requirement actually makes the proposed change more conservative than the existing Technical Specification. Furthermore, it is inconceivable to postulate a manufacturing defect that would result in the essentially simultaneous failure of such a large number of rods during normal steady state plant operation. Therefore, H. is concluded that the consequences of the design basis accidents are not negatively impacted by the proposed change.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. The proposed changes do not involve any plant hardware modifications. The proposed changes do not introduce any new failures. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from those previously analyzed.
- 3. Involve a significant decrease in the margin of safety. Since the proposed changes do not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.
3/4.4.9.1. Pressure / Temperature Limits RCS This proposed RTS Section provides requirements for pressure and temperature limitations for the RCS and is equivalent to the existing Technical Specifica-tion Section 3.4.A with the following exceptions:
- 1) Figures 3.4-2, 3.4-3, 3.4-4, 3.4-5, 3.4-6, and 3.4-7 are deleted.
These figures are no longer applicable since their specified service period has been exceeded.
- 2) The RTS heatup rate limitation is 50*F in any one-hour period (1200*F) and 60*F in any one-hour period (1200*F) whereas the existing Technical Specification had specified a limit of 100*F/hr when averaged over a one-hour period and did not specify specific temperature ranges. The proposed RTS limitations are more restric-tive than the existing Technical Specification.
- 3) The RTS cooldown rate limitation is 100*F in any one-hour period (1200*F) and 30*F in any one-hour period (< 200*F) whereas the existing Technical Specification had specified a limit of 100*F/hr when averaged over a one-hour period and did not specify specific temperature ranges. The proposed RTS limitations are more restric-tive than the existing Technical Specifications.
- 4) The RTS includes an additional LCO that limits the maximum tempera-ture change to 110*F in any one-hour period during hydro and leak t e st i .~.g . This is an additional requirement and is more restrictive than the existing Technical Specification which didn't have any specific limitation invoked during hydro / leak testing. This will ensure that the vessel will not have thermal stresses across the wall in conjunction with pressure stresses induced during the
p
- Attachment 2 S*ctitn 3/4.4 B13181/Page 15 bydro/ leak test. This is also consistent with the proposed. Appen-dix XX of the ASME code for isothermal pressure transients and the M STS.
- 5) The RTS provides Action Statements and- Surveillance requirements-that were not specified in the existing' Technical Specification.
The',e new requirements are more restrictive that the existing Technical Specification.
- 6) The Reactor Vessel Surveillance Capsule surveillance requirements are included in this specification. However, the RTS Table 4.4-5 includes revised locations for capsules E and G and the surveillance schedule of the remaining capsules have been revised. The vessel locations have been changed as a result of damage to the capsule ~
chutes discovered during the thermal shield repair work. The surveillance schedule now has capsule B (type I) to be removed at 25 i EFPYs and capsule E being set aside in standby for future use. Significant Hazards' Consideration - Section 3/4.4.9 1 In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve a significant hazards consideration because they would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. Items (2), (3), (4), & (5) of the above represent more restrictive requirements. Item (6) is ' equivalent to the intent of the existing Technical Specification requirements and provides a more prudent use of the last remaining type II capsule -{ capsule E). It should be noted that ASTM E-185 standard dictates only the time schedule.
Therefore, the proposed change is still in compliance with the ASTM specification. The proposed RTS changes provide more limiting pressure / temperature heatup and cooldown requirements during plant operation and more accurate information regarding reactor vessel fracture toughness properties. A new requirement is also provided that will ensure hydro / leak testing in accordance with ASME isothermal guidelines that were not previously identified in the existing Technical Specifications. Therefore, this change does not impact the consequences of the design basis accident.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. Since the proposed changes do not involve any plant hardware modifications the performance of safety-related systems remains unaffected during operations. The proposed changes do not introduce any new failures. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from those previously analyzed.
Attachment 2 Section 3/4.4 B13181/Page 16
- 3. Involve a significant decrease in the margin of safety. Since the proposed changes do not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.
3/4.4.9.2. Pressure / Temperature limits Pressurizer This proposed RTS section provides heatup, cooldown, and operating temperature limitations for the pressurizer and is equivalent to the existing Technical Specification Section 3.4.B with the following exceptions:
- 1) The RTS provides Action Statements and Surveillance that were not included in the existing Technical Specification, therefore, the RTS is more restrictive than the associated existing Technical Specifi-cation Section. These requirements however, are consistent with the H STS.
- 2) The proposed RTS LC0 specifies heatup and cooldown limitations in terms of 'F in any one-hour period whereas the existing Technical Specification specifies these in terms of *F/hr. Step changes in X degrees per minute (if physically achievable by the plant) vs. X degrees in a one-hour period are still permissible as long as they are within the heatup/cooldown limit curves. The revised wording is also consistent with the H STS.
3/4.4.9.3. Overpressure Protection System This proposed RTS section provides requirements for the low temperature overpressure protection system and is equivalent to the existing Technical Specification Section 3.3.4.2 with the following exceptions:
- 1) The RTS Action Statement has an additional provision (item d) that specifies that the provisions of Specification 3.0.4 are not applic-able.
- 2) The Surveillance requirements in the RTS have been modified to include the requirements of the existing Technical Specification Table 4.2-1 item 24 and Table 4.2-2 items 12,13, and 14.
The only other change is renumbering of the section per the H STS. 3/4.4.10. Structural Inteority This proposed RTS section provides requirements for maintaining structural integrity of ASME components. This section is equivalent to the existing Technical Specification Section 4.10.A, B, & C with the following exception:
- 1) The RTS provides Action Statements that restrict plant operation when components do not meet their specified requirements. These limitations are not included within the existing Technical Specifi-cation, therefore, the RTS is more restrictive than the existing Technical Specification.
Attachment 2 Section 3/4.4 B13181/Page 17 3/4.4.11 - RCS Vents This section in the proposed RTS for the Haddam Neck Plant is the same as the existing Technical Specification Section 3.3. The existing Technical , Specification Section was approved by the NRC.g. No changes have occurred to this section other than the renumbering to achieve consistency with the M STS. The proposed Section is an enhancement to the existing Technical Speci-fications. It provides clear applicability, action and surveillance require-ments modeled after the M STS. Significant Hazards Consideration - Section 3/4.4.9.2. Section 3/4.4.9.3. 3/4.4.10. and 3/4.4.11 In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed RTS and has concluded that they do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed RTS do not involve a significant hazards consider-ation because the changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The determination of whether or not a proposed change is equivalent, more restrictive (or a new requirement),
or less restrictive is based on the Limiting Condition for Operation (LCO) and Applicability Requirements since it is these requirements which will impact the design basis accidents. In general, the conversion to M STS yields more extensive and/or restrictive Action and Surveillance Requirements. As described above, most of the changes are more restric-tive in that there are no comparable requirements in the existing Techni-cal Specifications, but the proposed changes are equivalent to the M STS. Based upon the above discussion, the proposed RTS will not increase the l probability or consequences of any accident previously analyzed.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. Because there are no hardware modifications associated with the proposed changes, the performance of safety-related systems remains unaffected during operations. The operability require-ments are in::reased over the current requirements thus enhancing the t
performance of safety systems. Therefore, the proposed RTS will not modify the plant response to the point where it can be considered a new l accident nor are any credible failure modes created. l
- 3. Involve a significant reduction in a margin of safety. Because the changes proposed herein provide acceptable results for the design basis l
accident, no additional burden will be placed on the protective bound-l aries for postulated accidents. In addition, there are no plant hardware I modifications associated with this change and hehce, there is no direct impact on the protective boundaries. The proposed RTS do not affect the l (16) A. B. Wang letter to E. J. Mroczka, Issuance of Amendment, dated April 24, 1989.
Attachment.2
. Section-3/4.4 B13181/Page :18 t,'
- safety limits of the protective boundaries and the bases of the proposed.
RTS have been modified to reflect the proposed changes'. f' [ '; . e. i 7 4 ..
-i -
e o t x' .s_ -_-_n'--_-..
t-Docket No. 50-213 191161 ( l Haddam Neck Plant Technical Specification Section 3/4.6, Containment Systems June 1989
' Attachment 2 Section 3/4.6 B13181/Page 1 Technical Specification Section 3/4.6 Containment Systems The proposed revised Technical Specification (RTS) Section 3/4.6 has been ,
prepared by converting the existing Technical Specification Sections 3.11 and 4.4 to a format consistent with the Westinghouse Standard Technical Specifica-tions (W STS). The proposed RTS section deals with the containment systems including containment integrity, containment leakage, containment air locks, internal pressure, air temperature, containment vessel structural integrity, containment ventilation, and containment isolation. A matrix summarizing this comparison is included in Attachment 3. 3/4.6.1 Primary Containment Definition 1.6 - Containment Intearity The definition of containment integrity has been modified to STS format. In addition, the two notes in the existing Technical Specification Section 1.8 have been incorporated into this revised definition. The two notes allow manipulation of certain containment isolation valves in order to perform diagnostic checks of safety systems. The ability to open certain vent and ! drain valves has been added to note 2. These additional changes make the l proposed RTS equivalent to the existing Technical Specifications. 3/4.6 2 1 - Containment Intearity The proposed action statement is more restrictive than the existing Technical Specifications in that it requires containment integrity to be restored within one hour if containment integrity is lost or be in at least hot standby within the next 6 hours and in cold shutdown within the following 30 hours. The existing Technical Specifications have the same time frame but they apply only to valves and their associated penetrations. l l The proposed RTS provides more detailed surveillance requirements to demon-strate containment integrity. The proposed RTS puts a surveillance frequency of at least once per 31 days on demonstrating containment integrity. An i exception is made for components in the primary auxiliary building pipe trench l and containment which are " locked, sealed or otherwise secured in the closed position." These components will be verified closed during each cold shutdown in excess of 48 hours. The existirg Technical Specifications do not place a surveillance frequency on any components. l An additional surveillance in the proposed RTS not required in the existing Technical Specifications requires a retest if a penetration has been opened. The test pressure designated in the proposed RTS is 39.6 psig not 40 psig. Although the proposed test pressure is lower, it meets the 10CFR50 Appendix J requirements of testing at the peak calculated pressure (39.6 psig), and is consistent with the M STS. 3/4.6.1.2 - Containment Leakaae The proposed applicability has been expanded from the existing Technical Specifications to include Modes 1-4. The existing Technical Specifications only include Modes 1 and 2. The overall integrated leak rate in the proposed RTS is 0.18 weight percent not 0.25 weight percent. This new leak rate is consistent with the current design basis analysis. Thus the proposed RTS is more restrictive.
Attachment 2 l Section 3/4.6 :
- .B13181/Page- 2 !
l The proposed action statement requires the RCS to be kept below 200*F if the j integrated leakage is above 0.75La or if the combined type B and C leakage is above 0.6 La. The existing Technical Specifications require shutdown and depressurization if the sum of the leakage from all penetrations is greater than 0.6La. Although not~ identical, these requirements are considered to be equivalent. The' existing Technical Specifications require type B and C tests be performed at 12-month intervals where practical. This is interpreted as each refueling. The proposed RTS requires type B and C tests to be conducted at intervals no greater than 24 months. Since these tests are performed during operation in I Mode 5 or 6, the proposed RTS is also interpreted as each refueling. Thus, even though the wording is different, the interpretation is considered the same and therefore, they are equivalent. The existing Technical Specification also requires that isolation valves . essential for plant operation shall be tested during shutdown when the reactor is depressurized if the valves have not been tested within the past 12 months. There is little benefit from this testing since it is dependent on unplanned shutdowr.s. The proposed RTS states that Technical Specification 4.0.2 does not apply. Thus, the stated testing frequencies (3 type A tests in 10 yrs., type B and C tests every 24 months) must be met' and the 25% grace period on frequency is not applicable. The existing specification does not mention anything about grace periods. 3/4.6.1.3 - Containment Air Locks The proposed LC0 3.6.1.3.a for an operable air lock is included in the exist-ing definition of containment integrity. The proposed Applicability is the same as the existing Technical Specification 3.11.B.(1). The proposed sur-veillance requirements are the same as the existing Technical Specification 4.4.II.D.3 except that the test pressure is reduced to 39.6 psig. Proposed surveillance requirement 4.6.1.3.a and b are the same as the existing Tech-nical Specification 4.4.II.D.3 except as follows. The existing Technical Specifications require the air lock to be tested within 72 hours of opening at a reduced pressure (P 10 psig). The results must be scaled to 40 psig and the sum of all leakagh =ust m continue to meet 0.6 La. Appendix J requires the scaler to be Pa/Pt (40/10 = 4). The proposed RTS has the same time and test pressure requirements, but the duration of the test and a specific limit for the airlock have been added. The leak rate limit is .01 La at 10 psig. This limit is less than the Appen-dix J limit of .05 La at 39.6 psig (.01
- 39.6/10 = .04), and is more restric-tive than proposed RTS 3.6.1.3.b. The proposed RTS still requires the air lock leak rate to be added to all other type B and C test results and the total to meet 0.6 La at 39.6 psig.
The remainder of the proposed specification is new. The proposed LC0 3.6.1.3.b places an overall leakage rate on the air lock of .05 La at 39.6 psig. If one containment air lack door is inoperable, the proposed Actions require:
o Attachment-2 Secticn 3/4.6 B13181/Page 3
- o. An inoperable door to be made operable within 24 hours or lock the
, operable door closed, o Verify the locked door is locked at least once per 31 days, o Otherwise be in Hot Standby in 6 hours and Cold Shutdown in the following 30 hours. If the air lock is inoperable for any reason other than an inoperable door, the proposed actions require one operable air lock door be closed and the air lock be made operable within 24 hours, or the plant must be brought to hot standby in 6 hours and cold shutdown within the following 30 hours. 3/4.6.1.4 - Internal Pressure The proposed LCO is the same as the existing Technical Specification 3.11.c. The proposed applicability (Modes 1, 2, 3, 4) is more restrictive than the existing Technical Specification. The proposed action and surveillance requirements are more restrictive in that there are no similar requirements in the existing Technical. Specifications. The action statement requires the containment pressure to be brought within the LC0 in one hour or be in hot standby within the next six hours and in cold shutdown within the following 30 hours. The proposed surveillance requires the containment pressure to be verified at least once per 12 hours. 3/4.6.1.5 - Air Temperature The proposed RTS is more restrictive in that there is no counterpart in the existing Technical Specifications. The LC0 requires average air temperature to be below 140*F. The proposed RTS is applicable in Modes 1, 2, 3, and 4. The Action requires the average air temperature to be brought within limits within 8 hours or be in Hot Standby in the next 6' hours and Cold Shutdown in the following 30 hours. The surveillance requirement states that the air temperature shall be taken every 24 hours and shall be a weighted average of ' all operable detectors. 3/4.6.1.6 - Containment Vessel Structural Intearity The proposed RTS is new except for surveillance requirement 4.6.1.6.1 which is equal to the existing Technical Specification 4.4.1.A (indirectly inferred by conformance to Appendix J), and requires a visual inspection of the contain-ment surfaces prior to any type A leakage test. The LC0 and Applicability statements require the structural integrity of the containment to be main-tained at a level consistent with surveillance requirement 4.6.1.6.1 while operating in Modes 1, 2, 3, and 4. The Action statement requires structural integrity to be maintained or restored within 24 hours or be in Hot Standby within the next 6 hours and Cold Shutdown in the following 30 hours. Proposed surveillance requirement 4.6.1.6.2 requires a report to be submitted to the NRC if any degradation in containment integrity is observed in the inspections per 4.6.1.6.1.
Attachment 2 Section 3/4.6 B13181/Page 4 3/4.6.1.7 - Containment Ventilation The footnote to the proposed LCO is the same as existing Technical Specifica-tion 3.11.F.(2) which allows the containment exhaust to be isolated by a blank flange inside containment for a period not to exceed 7 days. The remainder of , the proposed RTS is new. The LC0 and Applicability require the purge and I exhaust isolation valves to be operable and locked closed during operation in modes 1, 2, 3, and 4. The Action Statement requires that if the valves are not locked closed, they must be locked closed within 1 hour or be in Hot Standby in 6 hours and Cold Shutdown in the following 30 hours. The surveil-lance requirement states that the valves will be verified locked closed once per 31 days. Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed RTS and has concluded that the changes do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed RTS do not involve a significant hazards consideration because the changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The determination of whether or not a proposed change is equivalent, more restrictive (or a new requirement),
or less restrictive is based on the LC0 and Applicability requirements since it is these requirements which will impact the design basis acci-dent. In general, the conversion to W STS yields more extensive and/or restrictive action and surveillance requirements. As described above, the majority of the changes are more restrictive in that there are no comparable requirements in the existing Technical Specifications, and the proposed changes are at least equivalent to the H STS. Eauivalent Sections o 1.6 Definitions - Containment Integrity o 3/4.6.1.1 - Containment Integrity o 3/4.6.1.2 - Containment Leakage - LC0 0 3/4.6.1.3 - Containment Air Locks - LC0 o 3/4.6.1.4 - Internal Pressure - LC0 o 3/4.6.1.7 - Containment Ventilation More Restrictive Sections o 3/4.6.1.2 - Containment Leakage - Applicability o 3/4.6.1.3 - Containment Air Locks - Applicability o 3/4.6.1.4 - Internal Pressure - Applicability i o 3/4.6.1.5 - Air Temperature o 3/4.6.1.6 - Containment Vessel Structural Integrity Deviations Existing Technical Specification Section 3.11.B.(1) requires the shutdown margin to be 2 2600 pcm when containment integrity does not exist. It further states that containment integrity must be maintained when the RCS
l i ! Attachment 2 SGction 3/4.6 B13181/Page 5 is above 200*F and 300 psig. Thus, the existing Technical Specification i requires'a 2600 pcm shutdown margin in Modes 5 and 6 whenever containment integrity is lost. The proposed RTS section 3.1.1.3 requires at least a
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2600 pcm shutdown margin in Modes 4 and 5. Proposed RTS Section 3.9.1 requires Keff of 0.94 in Mode 6. A Keff of 0.94 is more restrictive than a 2600 pcm shutdown margin. Thus, the proposed RTS requires a shutdown ' margin'of 2600 pcm independent of containment integrity. Existing Technical Specification Section 3.ll.B(2) requires the reactor vessel to be borated to the refueling boron concentration when the vessel head is removed and containment integrity is not present. Proposed RTS Section 3.9.1 requires a boron concentration sufficient to ensure a Keff. of 0.94 during operation in Mode 6 (Refueling), independent of contain-ment integrity. Existing Technical Specification Section 3.II.B.(3) prohibits positive reactivity insertion via control rods or boron dilution whenever contain-ment integrity does not exist (Modes 5 and 6). The proposed RTS do not contain this requirement. The shutdown margin and Keff requirements are sufficient to prevent reactivity additions which could result in criti-cality and, thus, require containment integrity. Thus, deleting the prohibition of positive ' reactivity changes when containment integrity does not exist does not impact the consequences of any event. Existing Technical Specification Section 3.11.E requires the containment spray system to be operable whenever the reactor is critical. This requirement is not being included in the proposed RTS because credit is not taken for the containment spray in the design basis accident. The calculated peak containment pressure can be maintained at a safe level by using the containment air recirculation system. The specifications identified as more restrictive or new have no' negative impact on the' design basis accidents. Therefore, there is no increase in the probability or consequences of an accident previously analyzed.
- 2. Create the possibility of a new or different kind of- accident from any previously' evaluated. Because there are no hardware modifications associated with the proposed changes, the performance of safety-related systems remains unaffected during operations. The operability require-ments are increased over the current requirements thus enhancing the performance of safety systems. Therefore, the proposed RTS will not modify the plant response to the point where it can be considered a new accident nor are any credible failure modes created.
- 3. Involve a significant reduction in a margin of safety. Because the changes proposed herein provide acceptable results for the design basis accident, no additional burden will be placed on the protective bound-aries for postulated accidents. In addition, there are no plant hardware modifications associated with these changes and hence, there is no direct impact on the protective boundaries. The proposed RTS do not affect the safety limits of the protective boundaries and the bases of the proposed RTS have been modified to reflect the proposed changes.
Attachment 2 Section 3/4.6 B13181/Page 6 3/4.6.2 - Containment Air Recirculation This section will be submitted to the NRC by June 30, 1989. 3/4.6.3 - Containment Isolation The LCO, applicability and action statements in the proposed RTS are the same as those in the existing Technical Specifications. The one difference is the table of containment isolation valves. The valve list in the existing Tech-nical Specifications is not being included in the proposed RTS, but has been incorporated into a new table in the Final Safety Analysis Report (FSAR). The new containment isolation valve table has been expanded to include all containment system boundary valves (check valves, manual valves, drain / vent valves). The bases for the proposed RTS includes a reference to the FSAR table and further requires that addition or deletion of containment isolation valves from the table shall be made in accordance with 10CFR50.59 and requires approval by the Plant Operations Review Committee. This is considered accept-able since the FSAR is a controlled document which is readily accessible to all personnel. Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed RTS and has concluded that the changes do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed RTS do not involve a significant hazards consideration because the changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The determination of whether or not a proposed change is equivalent, more restrictive (or a new requirement),
or less restrictive is based on the LC0 and applicability requirements since it is these requirements which will impact the design basis acci-dent. The proposed RTS Section 3/4.6.3, Containment Isolation, is equivalent to the existing Technical Specifications in the areas of LCO, Applicability and Action statements. The major difference is the dele-tion of the table of containment isolation valves from the proposed RTS. Although this appears as a less restrictive requirement, the incorpo-ration of the revised and updated table in the FSAR continues to maintain it as a controlled document. Mc reover, the bases to the RTS further requires that any additions or deletions to the table be made in accor-dance with 10CFR50.59 and approved by the Plant Operations Review Commit-tee. Since there are no changes in the way the plant is operated, there is no increase in the probability or consequences of an accident pre-viously analyzed. 2 Create the probability of a new or different kind of accident from any previously evaluated. The proposed changes do not introduce any new failures. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from those previously analyzed.
- 3. Involve a significant reduction in a margin of safety. Since the pro-posed changes do not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.
[< }
- s. ,
i 1
-l Docket No. 50-213 I B13181 i 1
3/4.7 Plant Systems June 1989
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-l Attachment 2 1 Section 3/4.7 B13181/Page 1 Technical Specification Section 3/4.7 Plant Systems l The proposed revised Technical Specification (RTS) Section 3/4.7 has been prepared by converting the existing Technical Specification Sections 3.4.C, 3.8.A, 3.19, 4.8, 4.9, and 4.13 to a format consistent with the Westinghouse Standard Technical Specifications (W STS). The proposed RTS section deals with the ' Plant. Systems. A matrix summarizing this comparison is included in Attachment 3.
3/4.7.1 Turbine Cycle 3/4.7.1.1 - Safety Valves The proposed RTS provides more detailed information on actions to be taken in the event of main steam safety valve inoperability during Modes 1, 2, and 3 for both 3 and 4 loop operation. While in either 4 loop or 3 loop operation with one or more safety valves inoperable, continued operation in the applic-able mode is allowed for a period of 4 hours. If valve operability is not restored within this time period, the plant is required to be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. In addition to the more detailed action statement, each main steam line code safety valve associated with each steam generator is required to be demon-strated operable by checking its setpoint each refueling. Currently, there are no specific actions identified when a main steam line code safety valve is found to be inoperable. Also, while steam relieving capacity is specified in the existing Technical Specifications, the proposed RTS specifies valve designated number, lift settings and valve size. Basic-ally, the proposed RTS imposes more stringent restrictions on allowable power operation in the event of main steam line ' safety valve inoperability. The change does not render the valve incapable of performing its intended func-tion. There are no hardware modifications associated with the change. Implementation of the change will eliminate operator interpretation which could potentially result in safety valve actuation if the plant is tripped. No design basis accidents are adversely affected due to the change. 3/4.7.1.2 - Auxiliary Feedwater System The proposed requirement that two independent auxiliary feed pumps be " powered from an operable steam supply system and its associated flow path" is just an expansion of the existing requirement and better defines the system. For the pump surveillance requirements, a staggered test basis is proposed in order to avoid common mode failure due to testing. When there are two auxiliary feedwater pumps inoperable, the action has been reduced to immediately initiate corrective action to restore at least one auxiliary feedwater pump to operable status within 24 hours or be in Hot Standby within the next 6 hours and in Hot Shutdown within the following 6 hours. Both auxiliary feedwater pumps must be restored to operable status within 72 hours from the time of initial loss of the first auxiliary feedwater pump or be in at least Hot Standby within the next 6 hours and in Hot Shutdown within the next 6 hours. Also, each auxiliary feedwater pump shall now be ' demonstrated operable on a staggered test basis. This is based on the H STS to avoid common mode failures due to testing.
Attachment 2-Section 3/4.7 B13181/Page 2 The proposed RTS also deletes the existing administrative requirement to test the auxiliary feedwater bypass and steam supply valve solenoid operators every 31 days. 1his was performed to verify a working system following Preventive Maintenance Procedure 9.2-45 which is intended to prevent the common valve ) . sticking problem. The six S0Vs associated with the auto initiation of l Auxiliary Feedwater were rebuilt and the modifications to the valves corrected the sticking problem, therefore, the existing administrative requirement to test the solenoid valve operators is no longer necessary. The valves have been satisfactorily tested under SUR 5.2-96 every 31 days for a year since the valve seats were replaced. The proposed RTS is consistent with the . existing Technical Specifications prior to implementation of the administrative requirement. In this manner, the Auxiliary Feedwater System's availability, integrity and proper operation is assured if needed during an accident. The functional testing of the steam generator load instrumentation channels and their logic was deleted from the auxiliary feedwater system section and moved to the Engineered Safety Features Actuation System Instrumentation Section 3/4.3.2. As such, the changes described above do not alter or modify the auxiliary feedwater system, but provide more stringent requirements relative to its surveillance and its operation and are more descriptive of the actual system. No design basis accidents are adversely impacted due to the changes. 3/4.7.1.3 - Auxiliary Feedwater Sunoly The new applicability requirements clarify the existing requirements by describing them in the standard " Mode" definition. A new surveillance requirement is added to verify an alternate source of water, i.e, recycle primary water storage tank (RPWST) to the primary water storage tank (PWST) if required. This surveillance does not increase the probability of a loss of auxiliary feedwater and it does not decrease the availability of the auxiliary feedwater system or the RPWST itself. The performance of the auxiliary feedwater system, as assumed in the design basis, is not compromised by the addition of this surveillance requirement. The remaining portions of the proposed RTS did not change and are equivalent to the existing Technical Specifications. 3/4.7.1.4 - Specific Activity The proposed RTS adds a new limit on secondary coolant activity and the limit is consistent with the assumptions made in the accident analysis. In addi-tion, the proposed surveillance requirements are being performed under present station procedures. The action statement requires a shutdown should the limit be exceeded. Based on normal blowdown rates, the primary to secondary leak rate limits are much more restrictive. Therefore, the new specification is not anticipated to result in additional shutdowns. The change is in addition to the existing Technical Specifications and is consistent with W STS. No design basis accident is adversely affected due to the change. -
Attachment 2 Section 3/4.7 B13181/Page 3 3/4.7.1.5 - Main Steam Line Trio Valves (MSLTVs) The proposed RTS provides detailed information on actions to be taken in the event of MSLTV inoperability during Modes 1, 2, and 3. The Mode 1 action allows the plant to continue in 3-loop operation by removing the respective reactor coolant loop from service if MSLTV operability can not be restored within 4 hours if the MSLTV is open and inoperable. The Mode 2 or 3 action allows the plant to continue in Mode 2 or 3 if either the inoperable MSLTV is maintained closed or the respective reactor coolant loop is removed from service (3-loop operation). Since 3-loop operation is an allowable operating condition and no required actions are identified for MSLTV inoperability in the current Technical Specification, the change is conservative with regards to existing plant operation. The proposed RTS surveillance requirements are consistent with the current , requirement and the proposed RTS are equivalent to the W STS. As such, the change will not adversely affect the design basis steam line break accident. Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed RTS and has concluded that the changes do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed RTS do not involve a significant hazards consideration because the changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The determination of whether or not a proposed change is equivalent, more restrictive (or a new requirement),
or less restrictive is based on the Limiting Condition for Operation (LCO) and Applicability Requirements since it is these requirements which will impact the design basis accident. In general, the conversion to W STS yields more extensive and/or restrictive Action and Surveillance Requirements. As described above, the majority of the changes are more restrictive in that there are no compartble requirements in the existing Technical Specifications, and the proposed changes are at least equiv-alent to the W STS. For those changes that deviate from the W STS, justification is provided for the change. Specifically, Technical Specification Section 3.8, Turbine Cycle, and Section 4.8, Auxiliary Steam Generator Feedwater System have been revised and combined into the proposed revised Technical Specification (RTS) Section 3.7.1.1, Auxiliary Feedwater System and 3.7.1.2, Auxiliary feedwater Supply. The proposed RTS Section 3.7.1.1 conforms to the recommendation of Generic Letter (GL) 83-37 and the Westinghouse Standard Technical Specifications (W STS) considering two independent steam driven auxiliary feedwater pumps. Differences between GL 87-37 and the proposed RTS are in the area of surveillance requirements and are described as follows: l
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Attachment 2 3 ..Sectisn 3/4.7 L B13181/Page 4 L L A. GL 83-37 specifies 'that when tested, the auxiliary 'feedwater pumps should be verified to deliver. a -particular flow at a particular , , discharge pressure. (See GL 83-37 Model . Technical Specification Section 4.7.1.2.a.2.) - In the proposed RTS, -the discharge pressure - is- verified on minimum recirculation rather than looking.at specific pressure and flow. B. . GL 83-37 Model Technical Specification Section 4.7.1.2.a.4 requires verification that each automatic valve in the flow path is in the fully open position whenever the auxiliary feedwater system is
. placed in automatic control or when above 10% RATED THERMAL POWER.
The two turbine driven pumps are provided with automatic initiation. These pumps will automatically start to supply water to the steam generators- via the main feedwater bypass valves. These- valves are normally closed - and open automatically to provide the - flow. path to the steam generators. Similarly, the steam admission valves.to the turbine are also normally closed and open automatically to provide the turbine driving force. Therefore, the GL 83-37 Model Technical
-Specification surveillance requirement is not applicable. for the Haddam Neck Plant and is not included in the proposed RTS.
C. GL 83-37 Model Technical Specification. Section 4.7.1.2.a.5 specifies
-that a dedicated individual be provided during surveillance testing who will be in communication with the control root. This individual shall be stationed near any manually realigned valves when only one-auxiliary feedwater train is available. During the surveillance testing of an auxiliary feedwater train, the same train is not taken out of _ service and, therefore, it is available and operable.
Therefore, this surveillance requirement is not applicable for the Haddam Neck Plant. D. GL 83-37 Model Technical Specification Section 4.7.1.3 specifies that an auxiliary feedwater flow path shall be demonstrated to be
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available prior to start-up after any refueling outage or other cold shutdown of lenger than 30 days. At the Haddam Neck Plant, the auxiliary feedwater system is used for start-up/ normal shutdown, therefore this surveillance requirement is not applicable.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. Because there are no hardware modifications associated with the proposed changes, the performance of safety related systems remains unaffected during operations. The operability require-ments are increased over the current requirements thus enhancing the performance of safety systems. Therefore, the proposed RTS will not modify the plant response to the point where it can be considered a new accident nor are any credible failure modes created.
- 3. Involve a significant reduction in a margin of safety. Because the changes proposed herein provide acceptable results for the design basis accident, no additional burden will be placed on the protective bound-aries for postulated accidents. In addition, there are no plant hardware modifications associated with these changes and, hence, there is no direct impact on the protective boundaries. The proposed RTS do not
Attachment 2 Section 3/4.7 B13181/Page 5-affect the safety limits of the protective boundaries and the bases of the proposed RTS have been enhanced to reflect the proposed changes. 3.4/7.2 - Steam Generator Pressure / Temperature limitation The current cooldown rate is given as 100*F/hr while the proposed RTS for cooldown rate is 1000F in any one hour period. These provide equivalent protection from thermal shock. Otherwise, the proposed RTS is essentially identical with the existing Technical Specification and is consistent with the W STS. The only differences between the proposed STS and the existing Tech-nical Specification are in format. However, the change adds a new surveil-lance requirements to the existing Technical Specification. No design basis accident is affected due to the change. Significant Hazards Consideration It accordance with 10CFR50.92, CYAPC0 has reviewed the proposed change and has concluded that the change does not involve a significant hazards consider-ation. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve a significant hazards consideration because the change would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The content of the RTS is the same as the existing Technical Specifications. The only differences between the proposed RTS and the existing Technical Specifications are in the fomat. The new surveillance requirement in the proposed RTS provides additional controls and thereby enhances plant operations. There is no increase in the probability or consequences of an accident previously analyzed.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. The proposed change does not introduce any new failures. Therefore, the proposed change does not create the possibility of a new or different kind of accident from those previously analyzed.
- 3. Involve a significant reduct.on in a margin of safaty. Since the pro-posed change does not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.
3/4.7.3 - Service Water System In the existing Technical Specification there is no specification covering the Service Water System. The proposed RTS requires a second service water header to be restored to operable status within 72 hours of entering single service water header operation. If this condition can not be met, the plant is required to be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. It also requires verification of correct valve position every 31 days and proper operation of automatic valves servicing safety-related equipment is required every 18 months by verifying that each safety-related valve that is required to actuate, goes to its correct position on a loss-of-normal power or safety injection test signal and each service water system pump starts automatically on a loss-of-normal power test signal. Presently, there are no requirements for valve surveillance. Restricting
- Attachment 2 Section 3/4.7 B13181/Page 6.
operation in this manner and adding surveillance requirements is consistent with the M STS. No design basis accidents are adversely affected due to the change. In addition, the proposed RTS t are considered an enhancement to the Technical Specifications. The new service water. system technical specification provides clear Applicability, Action, and Surveillance Requirements and are modeled after the W STS. Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed change and has concluded that ~ the change does not involve a significant hazards considera-tion. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve a significant hazards consideration because the change would not:
- 1. Involve a significant increase in the probability or consequences' of an accident previously evaluated. In general, the conversion to the W STS yields more extensive and/or restrictive Action and Surveillance Require-ments. As described above, the addition of a Technical Specification for the service water system provides a more restrictive requirement in that there are no comparable requirements in the existing Technical Specifica-tions. Moreover, the change proposed herein is equivalent to the W STS.
-2. Create the possibility of a new or different kind of accident from any previously evaluated. This specification provides restrictions for plant operations when in a single service water header condition. The existing Technical Specifications do not restrict plant operations when in this condition.' The proposed RTS is, therefore, more restrictive and will not create the possibility of any accident or malfunction of. a different type than any evaluated previously in the safety analysis report.
- 3. Involve a significant reduction in a margin of safety. Since the pro-posed changes do not affect the consequences of an accident previously analyzed, there is no reduction in the . margin of safety as defined in the basis of the existing Technical Specifications.
3/4.7.4 - Snubbers The existing Technical Specification applies to snubbers on the reactor coolant system, safety related systems, and on nonsafety-related systems if their failure or failure of the system on which they are installed would have an adverse affect on any safety-related system. All of these snubbers are included in the definition of safety-related snubbers, thus the applicability of the proposed RTS is equivalent to the existing Technical Specifications and to the W STS. The proposed RTS deviates from the existing Technical Specification in the area of functional tests. The current Technical Specification requires snubber functional testing well in excess of the standard ten percent sample size. Paragraph one of RTS Section 4.7.4.C requires the standard ten percent initial functional sample size, with additional five percent samples for each failure. This is equivalent to the existing Technical Specifications.
Section 3/4.7 813181/Page 7 The second paragraph of the existing Technical Specification Section 4.13.C deviates from the H STS resulting in an initial sample size greater than ten percent. The proposed RTS revises this section to more closely align it with the W STS functional test requirements. Although this change results in fewer snubbers being included in the initial functional test sample, the intent of the existing Technical Specifications and the H STS is met. The proposed RTS ensures that "the representative sample selected for the functional test sample plan shall be randomly selected from the snubbers of each type and reviewed before beginning the testing. The review shall ensure, as far as practicable, that they are representative of the various configurations, operating environments, range of size, and capacity of snubbers of each type." The proposed change requires a sample in excess of ten percent of the total snubber population to be included in the initial sample. This ten percent sample will be representative of the total snubber population being chosen in accordance with the change. Although this change differs from the existing Technical Specification, the intent of the specification is still met. As such, although the level of confidence in the overall snubber population will be slightly decreased, the reliability of the population will be maintained at or above existing Technical Specification levels. Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed RTS and has concluded that the change does not involve a significant hazards considera-tion. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed RTS do not involve a signifi-cant hazards consideration because the change would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed RTS initial snubber func-tional test sample size is consistent with the existing Technical Speci-fication. Therefore, the previously evaluated probability of occurrence or consequences of an accident or malfunction of equipment has not been increased. The proposed change does not impact any design basis acci-dents.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. Because there are no hardware modifications associated with the proposed change, the performance of safety-related systems remains unaffected. Thus, the proposed RTS will not modify the plant response to the point where it can be considered a new accident nor are any credible failure modes created.
- 3. Involve a significant reduction in a margin of safety. Since the pro-posed change does not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.
3/4.7.5 - Sealed Source Contamination The proposed RTS adds a new restriction on removable contamination for sealed sources. The new restrictions and surveillance requirements are in accordance with 10CFR and are presently being performed under station procedures. Since there is no change in the way the plant is operated, and the change provides
~ ' Attachment 2 SGctiCn 3/4.7 'B13181/Page 9
- 5. Verification of the operability of the fire pump diesel engine is required.
I i 6. Verification of -the operability of the fire pump diesel starting battery bank and charger is required. Relaxed / Deleted Requirements
- 1. The special reporting requirements have been - removed. Each potential reportable event will be reviewed in accordance with 10CFR50.73 as stated in proposed RTS Section 6.6.1.
- 2. An alternative source of water supply is now allowed if only one pump is inoperable.
- 3. The duration of testing for the electric pump has been reduced from 15 minutes to 5 minutes. A pump run time of 5 minutes is appro-priate for the existing pump configuration and is consistent with NFPA requirements.
- 4. Valve position verification now requires only those valves that are unsecured.
- 5. The specific reference to Chapter 5 of the 14th Edition of the Fire Protection Handbook has been replaced with the actual pump test requirements.
CYAPC0 has reviewed the proposed RTS and concluded that an equivalent level of fire protection will be maintained. The relaxed or deleted requirements do not significantly degrade the overall system reliability. 3/4.7.6.2 - Sorav and/or Sprinkler Systems This proposed RTS section provides the 1.00 requirements for spray and/or sprinkler systems operability whenever systems, structures, compenents, or equipment-protected by the stray and/or sprinkler systems are required to be OPERABLE. The requirements specified in this specification are, for the most part, equivalent to those required by the existing specifications. Some changes, both additional requirements and relaxed or deleted requirements, are proposed as follows: New/ Additional Requirements
- 1. A 31-day surveillance on the position of unsecured valves is now required.
- 2. There is now a requirement to cycle valves not testable during normal operation every 18 months.
- 3. There is now a requirement to verify each spray area is not obstructed.
Attachment 2 i S!cticn 3/4.7 B13181/Page 10 Relaxed / Deleted Requirement There is now the option of performing a water flow test, in lieu of just an air test, through each open head system. Allowing this option has been shown not to degrade the overall system reliability and has been concluded to provide an equivalent level of fire protection. In addition, the Special Reporting requirements have been removed. Each potential reportable event will be reviewed in accordance with 10CFR50.73 as stated in proposed RTS Sec-tion 6.6.1. 3/4.7.6.3 - C0 Systems 2 This proposed RTS section provides the LCO for the high pressure CO, systems operability whenever systems, structures, components, or equipment brotected by the CO, systems are required to be OPERABLE. The requirements specified in this propbsed RTS are equivalent to those required by the existing specifica-tion. In addition, the proposed RTS clarifies surveillance requirements associated with system interlocks, clarifies bottle weight requirements and deletes the special reporting requirements. Each potential reportable event will be reviewed in accordance with 10CFR50.73 as stated in proposed RTS Section 6.6.1. CYAPCO has reviewed this proposed RTS and has concluded that an equivalent level of fire protection will be maintained. 3/4.7.6.4 - Halon Systems This proposed RTS Section provides tne LCO for the Halon 1301 systems opera-bility whenever systems, structures, components or equipment protected by the Halon systems are required to be OPERABLE. The requirements specified in this proposed RTS are, at a minimum, equivalent to those required by the existing specification. In addition, the proposed specification has increased the required net storage container weight from 92% to 95% and now requires an air flow test through headers and nozzles rather than a visual inspection. The special reporting requirement of the existing specification have been removed and each potential reportable event will be reviewed in accordance with 10CFR50.73 as stated in RTS Section 6.6.1. CYAPC0 has reviewed this proposed RTS and has concluded that an equivalent level of fire protection will be maintained. 3/4.7.6.5 - Fire Hose Stations This proposed RTS section provides the LC0 for the fire water hose stations operability within the plant whenever systems, structures, components, or equipment protected by the fire water hose stations are required to be OPERABLE. The existing specification applies to both hose stations within the plant and outside hydrant hose houses. This proposed RTS is applicable only to fire hose stations inside the plant. Proposed RTS Section 3/4.7.6.6 addresses yard hydrants and hydrant hose houses. The requirements specified in this proposed RTS are, at a minimum, equivalent to those required in the existing specification. In addition, the proposed specification has increased the number of fire hose stations by five, and now requires a continuous fire watch in lieu of an hourly fire watch patrol if an inoperable hose station provides the primary fire suppression capability. Also, the proposed specification requires that an equivalent host *) provided
l
' Attachment 2 Secticn 3/4.7 B13181/Page 11 within 24 hours even if a fire watch is stationed, and a hose hydrostatic test is required every 12 months instead of every 3 years. These are enhancements of the protection provided by the existing specification.
CYAPC0 has reviewed this proposed RTS and has concluded that an equivalent level of fire protection will be maintained. 3/4.7.6.6 - Yard Fire Hydrants and Associated Hydrant Hose Houses This proposed RTS section provides the LCO for the yard fire hydrants and associated hydrant hose houses whenever systems, structures, components, or equipment in the areas protected by the yard fire hydrants and associated hydrant hose houses are required to be OPERABLE. The existing specification applies to both hose stations within the plant and outside yard hydrants. This proposed RTS is applicable only to yard hydrants and associated hydrant hose houses. Proposed RTS Section 3/4.7.6.5 addresses hose stations within the plant. The requirements specified in this proposed RTS are, at a minimum, equivalent to those required in the existing specification. Some changes, both addition-al requirements and relaxed or deleted requirements, are proposed as follows: New/ Additional Requirements
- 1. There are seven additional yard fire hydrants now included in this specification. In addition, fire hydrants FHH-5 and FHH-8 in the existing specification have been redesignated in this proposed specification.
- 2. Hydrants are now required to be accessible and unobstructed, not just available.
- 3. A requirement for a six-month inspection of each hydrant has been added.
- 4. Gasket inspections are required every 12 months now instead of every 18 months.
- 5. Under certain conditions, a continuous fire watch is required where only an hourly fire watch patrol was required previously.
- 6. When either a continuous fire watch or hourly fire watch patrol is stationed, a backup hose is now required to be established within 24 hours. The existing specification does not have this require-ment.
Relaxed / Deleted Requirements This proposed RTS allows 24 hours to place a backup system in-service if the hose / hydrant provides a back-up capability. This requirement is less restric-tive than what was allowed in the existing specification. However, the proposed specification gives credit for operating automatic systems and lessens the importance of back-up hose houses / hydrants. CYAPC0 has reviewed this proposed RTS and concluded that an equivalent level of fire protection
1. l Attachment 2 Section 3/4.7 B13181/Page 12 will be maintained and that the overall system reliability will not be degrad-ed. Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed RTS sections and has concluded. that they do not involve a significant hazards considera-tion. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed RTS do not involve a signifi-cant hazards consideration-because the changes would not: l 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The determination of whether or not a proposed change is equivalent, more restrictive (or a new requirement), or less restrictive is based on the Limiting Condition for Operation and Applicability Requirements since it is these requirements which will impact the design basis accidents. In general, the conversion to the M STS yields more extensive and/or restrictive Action and Surveillance Requirements. As described above, most of the changes are more restric- , tive in that they are a conservative change and there are no comparable requirements in the existing Technical Specifications. This will help ensure the operability and reliability of the systems covered under the proposed RTS. For few changes that are less restrictive, justification is provided for the changes. Based upon the above discussion, the l proposed RTS will not increase the probability or consequences of any accident previously analyzed. i
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. Since there are no hardware modifications asso-ciated with the proposed changes, the performance of safety-related systems remains unaffected during operations. The operability require-i- ments are increased over the current requirements thus enhancing the performance of safety systems. Therefore, the proposed RTS will not modify the plant response to the point where it can be considered a new accident nor are any credible failure modes created.
- 3. Involve a significant reduction in a margin of safety. Because the i changes proposed herein provide acceptable results for the design basis accideat, no additional burden will be placed on the protective bound-aries for postulated accidents. In addition, there are no plant modifications associated with these changes and hence, there is no direct impact on the protective boundaries. The proposed RTS do not affect the safety limits of the protective boundaries and the bases of the proposed RTS have been modified to reflect the proposed changes.
3/4.7.7 - Fire Rated Assemblies
'This proposed RTS section provides the LC0 requirements for operability of all l fire rated assemblies at all times unless otherwise determined that the separation of safety-related fire areas or separating portions of redundant systems important to safe shutdown with a fire area is not required based on the MODE of operation. The requirements specified in this specification are, for the most part, equivalent to those required by the existing specification.
E 1 h j L Attachment 2- l L S:cticn 3/4.7 j B13181/Page13 Some changes, both L additional requirements and relaxed or deleted require-ments, are proposed as follows: New/ Additional Requirements
, 1. An 18-month fire damper test has been added.
- 2. A 6-month test of fire doors has been added.
Relaxed / Deleted Requirements
- 1. The requirement for inspection following repair is deleted. Inspec-tion is considered to be an integral part of the installation / repair process, and specifying it in technical specifications is not considered necessary.
l
- 2. Credit is given for automatic fire suppression located on both sides of a degraded fire barrier when a fire patrol is initiated.
- 3. The 'use of a qualified temporary seal is now allowed for up to 30 days. This allows use of qualified seals that are installed on a
-temporary basis while work to the penetration is still on-going.
Once work is complete, a permanent seal will be installed. The overall rating of the fire barrier will not be reduced as these temporary seals will be fully qualified.
- 4. A rotating 10% inspection is allowed every 18 months instead of the 100% inspection now required. Based on operating experiences, it' has been concluded that reducing fire barrier penetration seal inspections from 100% to a 10% sampling program is acceptable. This conclusion is based on the following:
o A review of manufacturers' technical data pertaining to the fire barrier penetration sealing material indicates that there is no degradation with age and there are virtually no mainte-nance requirements once properly installed. o A review of previous Significant Operating Experience Reports (S0ERs) indicates that degradation of the barriers is attribut-ed mostly to construction and maintenance personnel who often fail to re-establish barrier integrity after work completion. o In order to prevent the above referenced condition (failure to restore penetration seals) from occurring, Plant Design Change Records (PDCRs) are specifically reviewed for penetration concerns. In addition, Administrative Control Procedures specifically address the need to reinstate any new or existing fire barrier penetration (s) as part of the work close-out requirements, o Station Surveillance Procedures provide for periodic inspec-tions of all safety and nonsafety-related penetration fire barriers. The Administrative Control Procedures also ensure that any work performed on barriers necessitating removal or
L l Att'achment 2 S:cti:n3/4.7? [. B13181/Page 14-degradation of any barrier material, requires restoration' of-affected barriers and subsequent QA inspection.. o Ten percent (10%) surveillance requirements are consistent with
-the W STS.
In . summary, with the infrequent' occurrence of degraded barrier events and existing technical specification requirements and admin-istrative controls, it is CYAPCO's conclusion that the reduction from 100% to 10% inspection will not affect the fire safety concerns within the plant. CYAPC0 has reviewed the proposed RTS and concluded that an equivalent level of fire protection will be maintained and the relaxed or deleted requirements do not significantly degrade the overall system reliability. 3/4.7.8 - Flammable liauids Control This proposed RTS section provides the LC0 requirements for introduction of flammable liquids in volumes greater than one pint into the control room at any time. The requirements of this specification are equivalent to the requirements of the existing specification. In addition, an action statement has been added to ensure that flammable liquids- are immediately removed from the control-room when-the conditions of the proposed RTS are not met. CYAPC0 has reviewed the proposed RTS and concluded that an equivalent level of fire I protection will be maintained. Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed RTS sections and has concluded that they do not involve a significant hazards considera-tion. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed RTS do not involve a signifi-cant hazards consideration because the changes would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The determinatico of whether or not a proposed change is equivalent, more restrictive (or a new requirement),
or less restrictive is based on the Limiting Condition for operation and Applicability Requirements since it is these requirements which will impact-the' design basis accidents. In general, the conversion to the H STS yield more extensive and/or restrictive Action and Surveillance Requirements. As described above, most of the changes are more restric-l tive in that they are a conservative change and there are no comparable requirements in the existing Technical Specifications. This will help ensure the operability and reliability of the systems covered under the proposed RTS. For the few changes that are less restrictive, justifica-tion is provided for the changes. Based upon the above discussion, the proposed RTS will not increase the probability or consequences of any accident previously analyzed.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. Since there are no hardware modifications
[
Attachment 2' Section3/4.7 B13181/Page 15 t l associated with the proposed changes, the performance of safety-related f' systems remains unaffected during operations. The operability require-l ments are increased over the current requirements thus enhancing the performance of safety systems. Therefore, the proposed RTS will not modify the plant response to the point where it can be considered a new accident nor are any credible failure modes created.
- 3. Involve a significant reduction in a margin of safety. Because the changes proposed herein provide acceptable results for the design basis accident, no additional burden will be placed on the protective bound--
aries for postulated accidents. In addition, there are no plant modifi-cations associated with these changes and hence, there is no direct impact on the protective boundaries. The proposed RTS do not affect the safety limits of the protective boundaries and the bases of the proposed RTS have been modified to reflect the proposed changes. 3/4.7.9 - Feedwater Isolation Valves Currently, there are no specific required actions identified in the event that a feedwater isolation valve is found to be inoperable. The proposed RTS provides detailed information on action to be taken when the feedwater isola-tion valve is inoperable during modes 1, 2, 3, and 4. Continued operation in the applicable mode is allowed for a period of 72 hours with one or more feedwater isolation valves inoperable. If after this period, each valve is not restored to operable status, the plant is required to be in at least HOT STANDBY within the next 6 hours, and in COLD SHUTDOWN within the following 30 hours. Since no required actions are identified for feedwater isolation valve inoperability in the existing Technical Specification, the change is conserva-tive with regards to existing plant operation. In addition, the surveillance requirements of the proposed RTS provide a testing frequency 9f the valves to demonstrate operability and valve closure time. The maximum of 70 seconds specified for the valve closure time is consistent with the assumption made for the design basis steam line break analysis. These requirements increase the level of confidence that the valves will perform their ini. ended function in the event of an accident. There are no hardware modifications associated with the change. As such, the consequences of the design basis accident will not be changed by implementation of the proposed RTS. Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed change and has concluded that the change does not involve a significant hazards consider-ation. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve a significant hazards consideration because the change would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously analyzed. The new surveillance requirements increase the level of confidence that the valves will perform their intended function in the event of an accident. The consequences of an accident previously analyzed have not been increased. The implementation of this
Attachment 2 Section 3/4.7 B13181/Page 16 technical specification will increase the reliability of the feedwater isolation valves.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. There are no failure modes that will create a new unanalyzed accident. The main steam line break and steam generator tube rupture have been previously anal,vzed and addressed in the updated Final Safety Analysis Report. The change is more conservative with respect to the increased probability that the nonsafety-related portion of the main feedwater system will be isolated in the event of either design basis accident identified above through implementation of the technical speci-fication requirements.
- 3. Involve a significant reduction in a margin of safety. There is no impact on the protective boundaries. Since the proposed change does not affect the consequences of an accident previously analyzed, there is no reduction in a margin of safety as defined in the basis for any technical specification.
3/4.7.10 External Flood Protection The proposed change imposes external flood protection restrictions on the pl ant. No current external flooding technical specification exists. The restrictions require the initiation and completion within 8 hours of the flood protection measures outlined in the plant procedures when the water level of the Connecticut River is at 16 feet above mean sea level and forecasts indi-cate a level of 19 feet or more. When the water level reaches 19 feet above mean sea level, a shutdown to Mode 3 is required, if applicable, considering plant operating status. Also, the change adds the surveillance requirement which assures the water level at the Haddam Neck Plant does not exceed the limits. The proposed restrictions are presently incorporated in the station procedures and are equivalent to the !! STS. Significant Hazards Consideration In accordance with 10CFR50.-92, CYAPC0 has reviewed the proposed change and has concluded that the change does not involve a significant hazards considera-tion. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve a significant hazards consideration because the change would not:
- 1. Involve a significant increase in the probability or consequences of an I accident previously evaluated. The proposed RTS will not affect the l intended functions of the existing flood control systems, other plant l systems, or their associated supports. The proposed change imposes additional requirements not in the existing Technical Specifications.
There is no impact on the design basis accidents. Therefore, there is no increase in the probability or consequences of an accident previously analyzed.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. The proposed change does not impact the operation
.______________~
S::cticn 3/4.7 B131SI/Page 17 of any component or system. No new failure modes are introduced. Therefore, the proposed change does not create the possibility of a new or different kind of accident from those previously analyzed.
- 3. Involve a significant reduction in a margin of safety. Since the pro-posed change does not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.
3/4.7.11 Primary Auxiliary Buildina (PAB) Air Cleanuo System The proposed change adds an LC0 on the PAB Air. Cleanup System and associated surveillance requirements. The LC0 is needed to ensure operability of the PAB filtration system during operations involving movement of fuel assemblies or control rods within containment. This function is required to limit release of radioactivity to the environment during a postulated fuel handling accident in the containment. The LC0 during Modes I through 4 is desirable to limit radioactivity release to the environment through containment penetrations and PAB equipment operating during postulated design basis events. The surveillance requirements are also based .on the H STS. The in-place penetration and bypass leakage testing acceptance criteria of "less than 1%" is specified consistent with the guidance provided in GL 83-13. criteria is based on a 90% filter efficiency assumed by the Staff.gs leakage The proposed Technical Specification flowrate of " greater than 20,000 cfm" is based on system testing. This flowrate is adequate to maintain containment at a negative- pressure during a postulated fuel handling accident with the containment hatch closed or with the containment hatch open and purge supply line closed. The flow rate is also sufficient to keep the PAB at a slightly negative pressure, and is well below the 52,000 cfm design capacity of the filtration units. The maximum allowable methyl . iodide penetration acceptance criteria of "less than 10%" is. consistent with GL 83-13 leakage criteria. As described above, this is based on the 90% filter efficiency assumed by the NRC. Surveillance requirement 4.7.11.a(2) was modified to reflect standard environmental condi-tions specified by Table 5-1 of ANSI N509-1980. A revised test condition of 40 feet / minute will be specified to agree with ASTM D3803 test procedure and the original design specification. The maximum allowable pressure drop for the combined HEPA filters and charcoal adsorber banks is revised to 1.6 inches Water Gauge at 20,000 cfm 10%. This pressure drop is based on a 1.25 inch w.g. maximum pressure drop for the charcoal filter plus a 3 inch w.g. maximum pressure drop recommended by the vendor for HEPA filter replacement. at rated flow (52,000 cfm). Since HEPA filter pressure drop is directly proportional to flow rate at low (1) U.S. NRC letter to W. G. Counsil, "SEP Topic XV Radiological Consequences of Fuel Damaging Accidents," dated June 9, 1981.
Attachment 2 Section 3/4.7 B13181/Page 18 velocities,(2) the 1.6 inch w.g. maximum pressure drop is conservatively estimated for the 20,000 cfm condition (i.e., total dp - (20,000/52,000) x 4.25 = 1.63). Specification 3/4.7.11.d is added to the meet the NRC requirement to implement M STS No. 2 per GL 83-13. This specification will provide assurance of proper function of the charcoal adsorbers which are operated frequently at the Haddam Neck Plant. The proposed RTS as described above is acceptable and meets the requirements of ANSI N510-1980 to the extent that the existing equipment at the Haddam Neck Plant allows. These proposed changes do not modify the equipment or its operation, but institute surveillance requirements which assure that the system is operating properly. Implementation of the proposed RTS Section 3/4.7.11 will reduce the potential for a significant dose to the public. The proposed change will have no negative impact on the design basis doses to the public, and will not create the possibility of a new release path. Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed change and has concluded that the change does not involve a significant hazards considera-tion. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve a significant hazards consideration because the change would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The determination of whether or not a proposed change is equivalent, more restrictive (or a new requirement),
or less restrictive is based on the LC0 and Applicability Requirements since it is these requirements which will impact the design basis acci-dents. In general, the conversion to the H STS yields more extensive and/or restrictive Action and Surveillance Requirements. As described above, this change is more restrictive in that there are no comparable requirements in the existing Technical Specifications.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. Because there are no hardware modifications associated with the proposed change, the performance of safety related systems remains unaffected during operations. The operability require-ments are increased over the current requirements thus enhancing the performance of safety systems. Therefore, the proposed RTS will not modify the plant response to the point where it can be considered a new accident nor are any credible failure modes created.
- 3. Involve a significant reduction in a margin of safety. Because the change proposed herein provides acceptable results for the design basis (2) " Survey of Loading Performance of Currently Available Types of HEPA -
Filters Under In-Service Conditions," 16th DOE Nuclear Air Cleaning Conference, Gunn & McDonough, Mine Safety Appliances Company, at page 16.
p b : Attachment' 2 Sectitn 3/4.7-
.B13181/Page:19 accident, no additional burden will be placed. on the protective bound-artes for postulated accidents. In addition, there are no plant hardware
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modifications associated'with this change and, hence, there is no direct [ ._ impact.on'the' protective boundaries. The' proposed RTS do not affect the .L -safety limits of the protective boundaries. L. f
Attachment 2 Section3/4.7 B13181/Page 8 L additional -technical specification requirements, the change will not influence sources with any possible safety significance. Therefore, this change does not involve an unreviewed safety question in that there is no safety signifi-cance. Significant Hazards Consideration In accordance with 10CFR50.92, CYAPC0 has reviewed the proposed change and has concluded that the change does not involve a significant hazards considera-tion. .The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve a significant hazards consideration because the change would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The surveillance requirements of the RTS are in accordance with 10CFR and are currently being performed under station procedures. Since there are no changes in the way the plant is operated, and the surveillance requirements are currently being performed by procedures, there is no increase in probability or consequences of an accident previously analyzed.
- 2. Create the possibility of a new or different kind of accident from any previously evaluated. The proposed change does not introduce - any new failures. Therefore, the proposed change does not create the possibility of a new or different kind of accident from those previously analyzed.
- 3. Involve a significant reduction in a margin of safety. Since the pro-posed change does not affect the consequences of an accident previously analyzed, there is no reduction in the margin of safety.
3/4.7.6 FIRE SUPPRESSION SYSTEMS 3/4.7.6.1 Fire Water Supolv/ Distribution System This proposed RTS section provides the LC0 requirements for the fire water supply and distribution system operability status at all times. The require-ments specified in this specification are, for the most part, equivalent to those required by the existing specification. Some changesi both additional requirements and relaxed or deleted re;direments, are proposed as follows: New/ Additional Requirements
- 1. The required capacity of the fire suppression pumps, with respect to pump pressure and speed, is now specified.
- 2. The flow path requirements are now specified.
- 3. The duration for testing the diesel pump has been increased from 15 minutes to 30 minutes.
- 4. Verification of automatic and sequential start and discharge pres-sure is now required for each pump.
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.I k-Docket No. 50-213 B13181 r Attachment 3 Technical Specification Comparison Matrix June 1989
Attachment 3 Page 1 TECHNICAL SPECIFICATION COMPARISON MATRIX Introduction The Technical Specification Comparison Matrix (TSCM) was prepared to facilitate the revision of the existing Haddam Neck Technical Specifications (T.S.). The TSCM is set up denoting the proposed Technical Specification section numbers in the left hand column followed by a short description. The next column lists the corresponding existing T.S. section number. The final two columns compare the requirements contained in the proposed section with the existing T.S. and the Westinghouse STS, respectively. The key at the bottom of each page provides an explanation for the symbols located in the two comparison columns. The equivalent notation "E" may either denote that exact wording has been transposed from the existing T.S. or different verbage conveying equivalent requirements has been used. In many cases, there was not a one-for-one relationship, but rather multi-section relationships, whereas the requirements in a given T.S. section may be divided between several different sections in the proposed Technical Specification. The additional requirement notation "++" denotes that the proposed Technical Specification is more restrictive because it is an entirely new requirement as compared to the existing T.S. or it is more restrictive in . the sense that the existing T.S. requirements have been changed such that they are more restrictive.
Attachment i 3 Matrix 3/4.4
.B13181/Page 1 i-3/4.4 REACTOR COOLANT SYSTEM TECHNICAL SPECIFICATION COMPARIS0N MATRIX Comparison Comparison h Existing With Existing- With M T.S.# Description T.S. # T.S. STS 3.4.1.1- S/U & Power Ops
- a. 4 loops 3. 3.1.1 - E ~E
- b. 3 loops- 3.3.1.1 E *(1)
Applicability 3.3.1.1 'E E Action 3.3.1.1 E E P 4.4.1.1.1 Verified Operable 3.3.1.1 E E 4.4.1.1.2 Cycle Stop Valves 3.3.1.1 .E 3.4.1.2 Hot Standby-
- a. Reactor Trip- . 3.3.1.2 E E Breakers closed p b. Reactor Trip 3.3.1.2 E E.
Breakers open Applicability 3.3.1.2 E E-Action 3.3.1.2 E ++(2) 4.4.1.2.1 Reactor Coolant Pumps 3.3.1.2 E E 4.4.1.2.2 SGs 3.3.1.2 E E 4.4.1.2.3 ' Reactor Coolant Loops 3.3.1.2 E E 4.4.1.2.4 Reactor Trip Breakers 3.3.1.2 E ++ 3.4.1.3 Hot Shutdown 3.3.1.3 & 1.27 E ++(3) Applicability 3.3.1.3 E Action 3.3.1.3 E ++(2)
- 4.4.1.3.1 Reactor Coolant Pumps 3.3.1.3 E E 4.4.1.3.2 SGs 3.3.1.3 E E 4.4.l~.3.3. Reactor Coolant Loop 3.3.1.3 E E 4.4.1.3.4 Reactor Trip Breakers 3.3.1.3 E E 3.4.1.4.1 Cold Shutdown - Loops Filled 3.3.1.4.1 E (4)
, a. RHR Loop 3.3.1.4.1 E E
- b. SG water levels 3.3.1.4.1 E E Applicability 3.3.1.4.1 E E Action 3.3.1.4.1 E (4) 4.4.1.4.1.1 SG Water levels 3.3.1.4.1 E E 4.4.1.4.1.2 RHR Loop' 3.3.1.4.1 E E 4.4.1.4.1.3 RHR Loop not in operation 3.3.1.4.1 E ++
4.4.1.4.1.4 Reactor Trip Breakers 3.3.1.4.1 E ++ 3.4.1.4.2 Cold Shutdown - Loops Not 3.3.1.4.2 E ++(4) Filled Applicability 3.3.1.4.2 E E Action 3.3.1.4.2 E ++(4) 4.4.1.4.2.1 RHR Loop 3.3.1.4.2 E E
.4.4.1.4.2.2 RHR Loop not in operation 3.3.1.4.2 E ++
4.4.1.2.2.3 Reactor Trip Breakers 3.3.1.4.2 E ++
c . . -_ __ ___ __ _ _ __ _ __ . _ _ _ _ _ - l Attachment l3
-Matrix 3/4.4' B13181/Page 2
. Comparison Comparison
. Existing With Existing With H
.T.S.# Description T.S. # T.S. STS 3.4.1.5 Isolated Loop 3.3.1.5 E '++(5)
Modes 1, 2 Applicability- 3.3.1.5 E E Action . 3.3.1.5 E ++(5) 4.4.1.5.1 Removal of Power 3.3.1.5 E ++ 4.4.1.5.2 Boron Concentration 3.3.1.5 E *(6) 3.4.1.6 Isolated Loop 3.3.1.5 E ++(5) Modes 3, 4, 5 & 6 Applicability 3.3.1.5 E E Action 3.3.1.5 E ++(5) 4.4.1.6.1 Removal of Power 3.3.1.5 E ++ 4.4.1.6.2 Boron Concentration 3.3.1.5 E *(6) 3 . 4 .1. 7 -- Isolated Loop - S/U 3.3.1.6 E .E Applicability 3.3.1.6 E *(7). Action 3.3.1.6 E 'E 4.4.1.7.1 Isolated Loop Cold Leg 3.3.1.6 E E Temperature 4.4.1.7.2 Operability Table 4.2-1 E ++ Item 22 4.4.1.7.3 Boron Concentration 3.3.1.6 E ++ 4.4.1.7.4 Loop Stop and Bypass Valve Table 4.2-1 E 4+ Interlock Item 23 3.4.1.8 Idle Loop -3.3.1.7 E ++ Modes 1, 2 Applicability 3.3.1.7 E ++ Action 3.3.1.7 E ++- 4.4.1.8.1 Removal of Power 3.3.1.7 E ++ 4.4.. 8.2 Boron Concentration 3.3.1.7 E ++ 3.4.1.9 Idled Loop 3.3.1.7 E ++ Modes 3, 4, 5 & 6 Applicability 3.3.1.7 E ++ Action 3.3.1.7 E ++- 4.4.1.9.1 Removal of Power 3.3.1.7 E ++
- 4.4.1.9.2 Boron Concentration 3.3.1.7 E +F 3.4.1.10 Idled Loop - S/U. 3.3.1.8 E ++
Modes 1, 2 Applicability 3.3.1.8 E ++ Action 3.3.1.8 E ++ 4.4.1.10.1 Idled Loop Cold Temperature 3.3.1.8 E ++ 4.4.1.10.2 Reactor Thermal Power 3.3.1.8 E ++ 4.4.1:.10.3 Boron Concentration 3.3.1.8 E ++
4 D ' Attachment 3 l' Matrix 3/4.4 B13181/Page 3. Comparison Comparison L Existing With Existing With W T.S.# Description T.S. # T.S. STS 3.4.1.11 Idled Loop - S/U 3.3.1.8 E . ++- Modes 3, 4, 5 & 6 Applicability 3.3.1.8 E ++ ! ' Action 3.3.1.8 E ++ 4.4.1.11.1 Idled Loop Cold Leg 3.3.1.8 E ++ Temperature 4.4.1.11.2 Boron Concentration 3.3.1.8 E ++ 4.4.1.11.3 Operability Table 4.2-1 E ++ Item 22 4.4.1.11.4 Loop Stop and Bypass Valve Table 4.2-1 E ++ Interlock Item 23 E ++ 3.4.2.1 Safety Valves - S/D 3.3.2.1 E E Applicability 3.3.2.1 E *(8) Action 3.3.2.1 E E 4.4.2.1 IST Surveillance 3.3.2.1 E E 3.4.2.2 Safety Valves - ops 3.3.2.2 E E Applicability 3.3.2.2 E E Action 3.3.2.2 E E 4.4.2.2 IST Surveillance 3.3.2.2 and E E Table 4.2-2 Item 6 3.4.3 Pressurizer 3.3.3 E E Applicability 3.3.3 E E Action 3.3.3 E E 4.4.3.1 Pressurizer Water Level 3.3.3 E E 4.4.3.2 Pressurizer Heaters 3.3.3 E E 4.4.3.3 Emergency Power Supply -
++ E 3.4.4 Relief Valves 3.3.4.1 E ++(9)
Applicability 3.3.4.1 E E Action 3.3.4.1 E (10)&(11) 4.4.4.1 PORV 3.3.4.1 E ++(12) 4.4.4.2 PORV Block Valve 3.3.4.1 E ++ 4.4.4.3- PORV Block Valve 3.3.4.1 E E 4.4.4.4 Control Air Supply 3.3.4.1 E ++ 4.4.4.5 Air Supply 3.3.4.1 E ++ 4.4.4.6 Emergency Air and Power Table 4.2-2 E E Supply Item 15 3.4.5 Steam Generators -
++ E Applicability - ++ E Action - ++ E 4.4.5.1 Determined Operable 4.10.1A E E 4.4.5.2 Sample Size & Action 4.10.1B ++(13) E 4.4.5.3 Inspection Frequencies 4.10.1C ++(15) E
Attachment 3 Matrix 3/4.4 B13181/Page 4 I Comparison Comparison Existing With: Existing With W T.S.# ' Description T.S. # -T.S. STS 4.4.5.4 Acceptance Criteria 4.10.1D ++(14) E 4.4.5.5 Reports 4.10.1E,6.9.1.b E ++(16) 3.4.6.1 ' Leakage Detection -
++ E Applicability -
++ E Action -
++ E 4.4.6.1 Demonstrated Operable
- a. Containment Atmosphere -
++ E
- b. Sump Level Table 4.2-1, ++ E Item 21 ++(17) E
- c. Volume Control Tank Table 4.2-1,
++(17) E Item 17 ++(17) E 3.4.7 Chemistry -
++(18) E Applicability -
++ E Action- --
++ .E 4.4.7 Sample Frequency -
++- E 3.4.8 Specific Activity 3.2 ++(19) ++(21).
Applicability 3.2 ++ E l Action 3.2 ++ E 4.4.8 . Sample Frequency & Modes Table 4.2-2, *(20) E Item 1 3.4.9.1 Press / Temp - RCS
- a. max. heatup 3.4.A ++(22) ++(22)
- b. max. cooldown 3.4.A E E
- c. max. temp. change during -
++ E testing Applicability Action 4.4.9.1.1 Heatup, Cooldown & Testing -
++ E 4.4.9.1.2 Irradiation Surveillance 4.10D (23) E Specimens 3.4.9.2 Press / Temp - Prz
- a. max. heatup 3.4.B E E
- b. max. cooldown 3.4.B E E
- c. max. RCS temp. 3.4.B E ++
difference
- d. pressurizer pressure 3.4.B E ++
greater than 500 psig Applicability -
++ E Action -
++ E l
E j Attachment 3 ] Matrix 3/4.4' l B13181/Page 5 j Comparison- Comparison Existing With Existing With W T.S.# Descrio.tiRD T.S. # T.S. STS 1 L 4.4.9.2.1 Heatup & Cooldown -
++ E
'4.4.9.2.2 Spray Water Temperature -
++ E
.4.4.9.2.3 RCS Temperature -- ++ ++
4.4.9.2.4 ' Pressurizer Pressure -
++ ++
3.4.9.3 Overpress Protect
- a. SLRV 3.3.4.2 E (24)
- b. RCS_ Vent 3.3.4.2 E E
, Applicability 3'3.4.2
. E ++(25) .
Action 3.3.4.2 E E l 4.4.9.3.1 LTOP SLVR 3.3.4.2 E (24) j Table 4.2-2 L Item 13 4.4.9.3.2- RCS Vent 3.3.4.2 E E . Table 4.2-2 l Item 14 I 4.4.9.3.3 LTOP SLVR Tabl e. 4.2-2 E (24) 1 Item 12 4.4.9.3.4 LTOP Operability Tabl e ' 4.2-1 E (24) Item 24 1 Structural Integ. 4 10.A, B, C
~
- 3.4.10 E E Applicability -
++ E Action -
++ E 4.4.10 RCP flywheel 4.10.C E E I 3.4.11 RCS Vents 3.3.5.1 E *(26)
Applicability 3.3.5.1 E E Action 3.3.5.1 E *(27) 4.4.11 Demonstrated Operable 3.3.5.1 E E Notes:
-E = Equivalent requirements
~
* = Less restrictive requirements
++ = Additional requirements I L
q
' Attachment-3 Matrix 3/4.4 1 B13181/Page 6 Section 3/4.4 Notes 1
(1) The M STS requires that all loops be in operation while the proposed revised Technical Specification (RTS) allows for three loop operation.
-(2) Reactor Trip Breakers are required to be opened in the proposed RTS for the case with no reactor coolant loop in operation. H STS does not state l: this requirement.
(3) The proposed RTS requires 3 loops to be operable and 2 loops in operation l under given conditions while the M STS does not require this many loops. The more stringent requirement is imposed because of the potential power , excursion due to an inadvertent rod withdrawal. ' (4) The proposed . RTS has an additional requirement of the reactor trip , breakers being open or the lift coils being deenergized. 1 (5) The proposed RTS requires power to be removed from the valve operators of the RCS stop valves. The W STS does not allow this. (6) There is an additional surveillance requirement in the M STS requiring
' determination of baron concentration within 30 minutes prior to opening an isolated loop.
(7) The W STS is applicable for all modes, whereas the proposed RTS is . applicable for Modes 3, 4, 5, and 6. i l (8) The W STS is applicable to Modes 4 & 5 while the proposed RTS is applicable only to Mode 4. (9) The proposed RTS provides restrictions on PORV setpoints while the M STS i does not. !
)
(10) An additional Action Statement is included in the proposed RTS addressing emergency control air supply pressure. (11) The W STS includes an Action Statement on inoperable block valves. (12) The proposed RTS requires an analog channel operational test while the W ! STS requires only channel calibration. (13) The proposed RTS has a requirement that at least 50% of the tubes inspected be from critical areas while the existing T.S. does not give a specific number. Al so, action is prescribed for the case where the selected tube does not permit passage of the eddy current probe. (14) The proposed RTS provides a long-term acceptance criteria for tubes with defects in the rolled region (bottom 4 inches of the tube) and updates the bases for this criteria. These proposed changes are consistent and meet the intent of Regulatory Guide 1.121.
l..
- Attachment 3' Matrix 3/4.4.
B13181/Page 7 (15) The proposed' RTS includes an additional condition (i.e., primary-to-secondary tube leaks in excess of specified limits) under which further unscheduled inservice inspections may be required. (16) The proposed RTS requires a Special Report within 90 days of inspection completion while the M STS requires the Report within 12 months.
-(17).The proposed RTS requires a channel check.
(18) These additions' reduce the possibility of stress corrosion cracking in the reactor coolant system. The current sampling system is designed for such purposes and is described in the safety analysis report. (19) The proposed RTS has an additional limitation on the dose equivalent I-131 which the current T.S. does.not have. (20) The existing T.S. requires radio-chemical analysis 5 days a week vs. once per 72 hours under the proposed RTS. However, additional measurements (i.e., isotopic analysis) are included in the proposed RTS and require increased sampling during transient conditions. Thus, reduced sampling under steady state conditions where conditions are not expected to change. is offset by increased sampling under transient condition. (21) The specific activity requirement on gross radioactivity is less in the proposed RTS than that shown in the W STS. (22) The proposed RTS uses a maximum heatup rate of 60*F/hr. vs. 100*F/hr. (23) The existing T.S. requires capsule "E" removal time at 25 years. with capsule "B" on standby while the proposed RTS has the reverse. (24) This requirement is not comparable to the W STS since the W system design uses the PORVs. (25) The proposed RTS has an additional requirement for placing the LTOP system in service prior to placing the RHR system in service. (26) The proposed RTS does not address block valves while the W STS does. (27) The W STS requires that the reactor be brought to cold shutdown under given situations while the proposed RTS does not have this requirement.
b a : Attachment 3 Matrix 3/4'.6' B13181/Page'l 3/4.6 CONTAINMENT SYSTEMS TECHNICAL SPECIFICATION COMPARISON MATRIX Comparison Comparison Existing With Existing With M
-T.S.# Description T.S. # T.S. STS 3.6.1.3 Cont. Air Locks L a. Both doors 1.8.4 E E closed
- b. Leakage rate -
++ E Applicability 3.11.B.1 ++ E.
Action -
++(1) E-4.6.1.3 Demonstrated 4.4.II.D.3 E E Operable 3.6.1.4- Internal Pressure 3.11.C Adm E E Applicability 3.11.C Adm ++(2)' E Action -
++ E 4.6.I'.4 Determined within -
++ E Limits 3.6.1.5 Air Temperature -
++ E Applicability -
++ E' Action- -
++ E 4.6.1.5 Determination -
++ E 3.6.1.6 Structural Integrity - ++ E Applicability -
++ E
' Action - ++ E 4.6.1.6.1 Exposed Surfaces 4.4.1.A E E 4.6.1.6.2 Reports -
++ E 3.6.1.7 Cont. Ventilation 3.11.F.2 E E(3)
Applicability 3.ll.F.1&2 E E Action -
++ E(3)
'4.6.1.7.1 Locked Closed -
++ E(3) 4.6.1.7.2 Demonstrated Operable -
++ E(3) 3.6.3 Cont. Isolation 3.11.G *(4)+(5) E App'licability 3.ll.G E E Action 3.11.G E E 4.6.3.1 Return to Service 4.10(B),4.3.F *(6)++(5) E 4.6.3.2 Every 18 Months 4.10(B) E E Table 4.2-2, Item 9 4.6.3.3 Isolation Time 4.10(B) E E Notes E - Equivalent Requirements
++ = Additional Requirements
* - Less Restrictive Requirements Section 3.6.2, Containment air recirculation to be submitted at a later date. !
l Attachment 3 Matrix 3/4.6 B13181/Page 2 SECTION 3/4.6 NOTES (1) An air lock leakage rate of 0.01 La is specified in the proposed RTS at a test pressure of 10 .psig. The existing T.S. does not specify acceptable ' leakage . for just the airlock, but rather provides total acceptable leakage for all containment penetrations. (2) The proposed RTS restrictions are applicable to both critical and non-critical modes while the existing Technical Specification applies to critical modes only. (3) The proposed Revised Technical Specification is based on . the plant specific design and meets the intent of M Standard Technical Specifications. (4) The following valv' es are contained'in existing Technical Specifications, but not included in the proposed RTS, but have been relocated in the FSAR Table 7.3-1. (SG-TV-1312-1, 2, 3 and 4, MS-TV-1212 and 13, CH-MOV-331). (5) The proposed Revised Technical Specifications do not include the CIV list presently included in the existing Technical Specifications. This list of valves is included in the FSAR Table 7.3-1. (6) Leak testing of ECCS check valves CD-CV-872A and CD-CV-872B are required
.in existing Technical Specifications, whereas the proposed Revised Technical Specification does not list these two valves. Also valves.
MS-TV-1212, 1213 and CM-M0V-331 contained in current T.S. are deleted in the proposed RTS. The list of CIVs is included in FSAR Table 7.3-1.
L ' Attachment 3 Matrix 3/4.7 B13181/Page 1-3/4.7 PLANT SYSTEMS TECHNICAL SPECIFICAI.10H COMPARISON MATRIX Comparison Comparison Existing With Existing With H
.T.S.# Description T.S. # T.S. STS 3.7.1.1 Safety Valves 3.8.A.1, 3.8.A.4 (1) E Applicability 3.8.A.1 +(2) E
. Action - ++ E 4.7.1.1 Spec. 4.0.5 Table 4.2-2, Item 7 3.7.1.2 ' Aux. Feedwater System 3.8.A.2.a, 3.8.A.4 E (5)
Applicability 3.8.A +(2) E Action 3.8.A.2 b & c (3) E 4.7.1.2.1 Staggered Test Basis 4.8.1 a & c E(4) E
-4.7.1.2.2 Demonstrated Operable 4.8.3 E (5) 3.7.1.3 Aux..Feedwater Supply 3.8.A.3.a. 3.8.A.4 E (5)
Applicability 3.8.A +(2) E Action 3.8.A.3.b & c E (5) 4.'7.1.3.1 DWST & PWST 4.8.2 E E 4.7.1.3.2- RPWST - ++ E 3.7.1.4 Specific Activity - ++ E Applicability -
++ E Action -
++ E 4.7.1.4 Sampling and Analysis -
++(6) E 3.7.1.5 Main Steam Line Trip Valve 3.8.A.4 E E Applicability 3.8.A +(2) E Action - ++ E 4.7.1.5 Demonstrated Operab'e 4.9 E (5)
Table 4.2-2, item 8 3.7.2 SG Press / Temp Limitation
- a. RC temperature 3.4.C.1 E E
- b. SG vessel temperature 3.4.C.4 E ++
- c. max. heatup or cooldown 3.4.C.2 E ++
- d. tube sheet temperature 3.4.C.3 E ++
Applicability 3.4 E E Action -
++ E 4.7.2.1 Pressure - ++ E 4.7.2.2 Heatup & Cooldown Rates - ++ ++
3.7.3 Service Water System - ++ E Applicability - ++ E Action - ++ E 4.7.3 Two Service Water Headers - ++ E
Attachment 3 Matrix 3/4.7 B13181/Page 2 L Comparison . Comparison Existing With Existing With H J_4J. Descriolign T.S. # T.S. STS
'3.7.4' Snubbers 3.19 E E Applicability 3.19A E E Action 3.19B E E 4.7.4 Augmented ISI Program 4.13- E' (5) l 3.7.5 ' Sealed Source Contamination -
++(6) E Applicability -
++(6) E Action -
++(6). E 4.7.5.1 Test Requirements -
++(6) E 4.7.5.2 Test Frequencies -
++(6) E
'4.7.5.3 Reports -
++(6) E 3.7.6.1 Fire Water Supply 3.22.A.1 ++(7) ++(8)
Applicability 3.22.A.1 E E Action 3.22.A 2 & 3 *(9)++(10) ++(10) 4.7.6.1.1 Supply / Distribution 4.15.A- (11)++(12)(32) (11)(13) Sys.
-4.7.6.1.2 Fire Pump Diesel -
++ -++(14) 4.7.6.1.3 Battery Bank & Charger -
++ E 3.7.6.2 Spray / Sprinkler Sys
- a. Sprinkler Sys 3.22.G.1 E E
- b. Deluge Spray Sys 3.22.G.1 E E
- c. Preaction System 3.22.G.1 E E Applicability 3.22.G.1 E E Action 3.22.G.2 & 3 *(15) E 4.7.6.2- Demonstrated Operable 4.15.G ++(16) E 3.7.6.3 Co, SYS 3.22.B.1 E E Applicability 3.22.B.1 E E Action- 3.22.B.2 & 3 *(15) E 4.7.6.3 Demonstrated Operable 4.15.B E *(30)-
.3.7.6.4 Halon SYS 3.22.C.1 E E Applicability 3.22.C.1 E E Action 3.22.C.2 & 3 (17) E 4.7.6.4 Demonstrated Operable 4.15.C ++(18) E 3.7.6.5 Fire Hose Stations 3.22.D.1 ++(19) E Applicability 3.22.D.1 E E Action 3.22.D.2 E E 4.7.6.5 - Demonstrated Operable 4.15.D ++(21) ++(22) 3.7.6 6 Fire Hydrants 3.22.D.1 ++(23) E Applicability 3.22.D.1 E E Action 3.22.D.1 ++(20) (20) 4.7.6.6 Demonstrated Operable 4.15.D ++(24) E 3.7.7 Fire Rated Assemblies 3.22.F.1 ++(25) E Applicability 3.22.F.1 E E
Attachment 3 Matrix-3/4.7. t 'B13181/Page 3 Comparison Comparison Existing With Existing With.)( T.S.# Description T.s. # T.s. sTs Action- 3.22.F.2 *(26)(27) E(26)++(27)- 4.7.7.1 Visual Inspection 4.15.F.1 *(28)++(29) E
, .4.7.7.2 Operational Test. -
++ ++
4.7.7.3- Inspecting . -
++ *(31) ,
3.7.8 Flammable Liquid Control 3.22.H E ++ Applicability 3.22.H E- ++ Action 3.22.H E ++ ' 4.7.8 - 3.22.H E ++ 3.7.9 Feedwater Isolation Valves - ++ ++ '! l_ Applicability -
++ ++-
't Action -- ++ ++ 4.7.9.1 Repair and Maintenance -
++ ++
4.7.9.2 Refueling -
++ ++ !
4.7.9.3 Isolation Time _
++ ++ !
3.7.10 Ext. Flood Protection -
++(6) E Applicability -
++(6) E- l
-Action -
++(6) E 4.7.10 : Measurements -
++(6) E :
l
-3.7.11 PAB Air. Cleanup System -
++(6) (5) l Applicability ~ -
++(6) (5) !
Action -
+1 (5) "
j 4.7.11 Demonstrated Operable -
++(6) (5)
Notes Eo Equivalent Requirements
++ = Additional Requirements i
l 1
Att'achment 3 Matrix 3/4.7 B13181/Page 4 (1) The steam relieving capability is specified' in the current Technical Specifications while lift set points of code safety valves are speci-fied in proposed revised Technical Specifications. (2)' The ' Applicability is expanded from current Technical Specifications to include a non-critical mode in proposed revised Technical Specifica-- .J tions.
-(3) The proposed Revised Technical Specification deletes the portion of the action statement requiring a= reduction of power to lowest stable power level by virtue of a M recommendation that' a power reduction is to be avoided since such reductions would be more likely to cause a trip.
(4) A staggered test basis is proposed in order to avoid common mode failure due.to testing. (5) ~ The proposed revised- Technical Specifications are different from the W Standard Technical Specifications, but are consistent with the current Technical Specifications. (6) These requirements are being performed under present station proce-
.dures.
(7) . The proposed Revised Technical Specification (RTS)' includes a require-ment for operable flow paths which are not included in' current T.S. (8) Automatic start. capability of the high pressure fire suppression pumps are included in the LC0 of the proposed RTS while the-M STS does not include this requirement.
'(9) The proposed RTS will delete the reporting requirements to NRC. Each reportable event will be reviewed in accordance with 10CFR 50.73 as stated in proposed RTS Section 6.6.1.
(10) The proposed RTS requires an hourly fire w:tch patrol for the affected areas while the H STS and current T.S. do not. (11) The proposed RTS does not contain a reference to Chapter 5 of the Fire Protection Handbook for performing a flow test _ of the - Distribution System. Although the existing T.S. and M STS include this reference, the RTS testing requirements are equivalent. (12) The requirement for a system functional test with verification that the high pressure pump starts automatically is not contained in the existing T.S. (13) The proposed RTS does not require verification of water supply volume and performance of a system functional test with verification that each automatic valve actuates to its correct position while the M STS does. (14) The proposed RTS requires verifying that the diesel starts from ambient conditions on an auto-start signal and operates while loaded with the fire pump. The M STS does not contain this requirement.
o At.tachment '3-Matrix 3/4.7 B13181/Page 5 (15) . The proposed RTS ' deletes the requirement ' for submitting a Special Report' to NRC. Each reportable event-will be reviewed in accordance with the requirement of 10CFR50.73 as stated in the proposed RTS. Section 6.6.1.
'(16) Verifying that each valve is in its correct position and cycling each valve that is not testable during plant operation are two requirements-included in the proposed RTS which are not included in the existing T.S.
g (17) The existing T.S. requires submitting a report to the NRC if the system is not restored to operable status within 14 days while the proposed RTS requires that a fire watch be established within one hour if one or more of the systems is determined to be inoperable. Each event will be reviewed in accordance with the requirements of.10CFR50.73 as stated in the proposed RTS Section 6.6.1. (18) A flow' test is required by the proposed RTS - to assure no blockage, whereas visual inspection is all that is required in the existing T.S. (19) The proposed RTS lists fire hoses for the Control Room and Switchgear Room not contained in the existing T.S.
-(20) The proposed RTS contains a more detailed action statement in the event one or more yard fire' hydrants (and-the associated hydrant hose houses) become inoperable.
(21) The proposed RTS uses a 12-month hydrostatic hose test vs. the existing T.S. requirement of once per 3 years. (22) The proposed RTS requires hydrostatic testing on 12-month intervals vs. the 3-year intervals in the W STS. (23) Additional yard fire hydrants are included in the proposed RTS. (24) The _ proposed RTS -includes a visual inspection to verify that the hydrant barrel is dry and requires more frequent surveillance than the existing T.S. requirements for the hydrostatic. test and the gasket inspection. (25) The proposed RTS applies to all fire-rated assemblies including pene-trations while the existing T.S. requirements apply to penetrations only. (26) The proposed RTS allows for temporary repair as an acceptable remedy for up to 30 days. The existing T.S. and W STS do not contain this provision. (27) Monitoring and establishing a fire watch on both sides of the affected fire-rated assembly are required in the proposed RTS while the W STS i and existing T.S. require monitoring and establishing a fire watch on just one side of the assembly. (28) The existing T.S. requirement for verifying that the barrier is func-tional following repairs has been deleted from the proposed RTS.
i
Attachment:
3 Matrix 3/4.7 B13181/Page 6 .i (29) More ' description is provided in the proposed RTS- for~ what is to be visually inspected than is contained in the existing T.S. i-p .(30) .M STS~ requires to verify that; once per 31 days, each valve in the flow ', path is in correct position while the RTS does not. l-' (31) ' While the M STS requires that .the operability of the fire door super-vision system for each electrically supervised: fire door be -verified, the RTS does not because there are no electrically supervised fire doors at the Haddam Neck Plant. (32)' . The proposed RTS requires operation of the electric ' pump.for 5 minutes-and the diesel pump for 30 minutes in accordance with recommendations in NFPA. l 1
Dpcket No. 50-213 B13181 Attachment 4 Haddam Neck Plant Existing Technical Specifications l Cross Reference June 1989
b Attachment 4 B13181/Page 1 Existina T.S. # Description- Proposed RTS 1.0 Definition 1.1 Defined Terms 1.0 1.2 Thermal Power 1.32 1.3 Rated Thermal Power 1.25 1.4 Operation Mode 1.18 1.5 - - 1.6 Operability 1.17 1.7 Reportable Event 1.26 1.8 Containment Integrity 1.6 1.9 Channel Calibration 1.4 1.10 Channel Check 1.5 1.11 Channel Functional Test 1.2 1.12 Core Alteration 1.8 1.13 Shutdown Margin 1.28 1.14 Identified Leakage 1.13 1.15 Unidentified Leakage 1.34 1.16 Pressure Boundary Leakage 1.20 1.17 Controlled Leakage 1.7 1.18 Quadrant Power Tilt Ratio 1.22 1.19 - - 1.20 - - 1.21 Frequency Notation 1.12 Table 1.1 1.22 - - 1.23 - - 1.24 Axial Offset 1.3 1.25 Low Power Physics Test 1.19 1.26 Action 1.1 ' 1.27 Channel Calibration 1.4 1.28 Channel Check 1.5 1.29 Channel Functional Test 1.2 1.30 Dose Equivalent I-131 1.9 1.31 Member (s) of the Public 1.15 1.32 Operable 1.17 1.33 Purge - Purging 1.21 1.34 Radioactive Waste Treatment Systems 1.23 1.35 Radiological Effluent Monitoring 1.24 and Off-site Dose Calculation Manual 1.36 Site Boundary 1.29 1.37 Source Check 1.30 1.38 Unrestricted Area Note (17) 1.39 Venting 1.35 Table 1.1 Operational Modes Table 1.2 Table 1.2 Frequency Notation Table 1.1
t ~ Attachment 4' K ;B13181/Page 2 Existino T.S. # Description Proposef RTS 2.0 Safety Limits and Maximum Safety
- Settings 2.1 Introduction 2.1 Bases Section 2.2- Safety Limits 2.2.1 Reactor Core. 2.1.1 e 2.3. Reactor Coolant System Pressure 2.1.2 L 2.4 Maximum Safety Settings Protective 2.2.1 E
Instrumentation
' Specifications Trip Setpoints l: Item 1 Pressurizer Pressure Table 2.2-1 Item 5 Item 2 Pressurizer Level Table 2.2-1 Item 6 Item 3 Variable Low Pressure Table 2.2-1 Item 4 Item 4 Nuclear Overpower Table 2.2-1 Item 2 Item 5 Low Coolant Flow Table 2.2-1 Item 7 Item 6 Reactor Coolant Loop Valve Section 4.4.1.7.2
- Temperature Interlock Item 7 High Steam Flow Table 2.2-1 Item 8 Item 8 High Start-up Rate Table 2.2-1 Item 3 3.0 Limiting Conditions for Operation 3.01
-3.1 Introduction 3.01 3.2 Reactor Coolant System Activity 3/4.4.8 3.3.1.1 Start-up & Power Opr. ration 3.4.1.1
- a. 4 Loops 3.4.1.1.a
- b. 3 loop 3.4.1.1.b Applicability 3.4.1.1 Action 3.4.1.1 Surveillance (1) 4.4.1.1.1 Surveillance (2) 4.4.1.1.2
g 3 ' Attachment 4 4~" . B13181/Page 3 Existina T.S. # Description Proposed RTS L3.3.1.2- ' Hot Standby . 3.4.1.2 a .- Reactor Trip Breakers '3.4.1.2.a Closed , , b. Reactor Trip Breakers 3.4.1.2.b Open-Applicability 3.4.1.2 Action . 3.4.1.2 Surveillance (a)- 4.4.1.2.1 Surveillance (b) 4.4.1.2.2-Surveillance (c) 4.4.1.2.3-Surveillance (d) 4.4.1.2.4 3.3.1.3 Hot Shutdown 3.4.1.3 and l'. 27 . Applicability 3.4.1.3-Action 3.4.1.3-Surveillance-(a) 4.4.1.3.1 F Surveillance (b); 4.4.1.3.2 Surveillance (c) 4.4.1.3.3 Surveillance (d)- 4.4.1.3.4 , 3.3.1.4.1 ' Cold Shutdown ~- Loops Filled 3.4.1.4.1
- a. RHR Loop 3.4.1.4.la
- b. SG Water Levels 3.4.1.4.lb Applicability 3.4.1.4.1 Action 3.4.1.4.1 Surveillance (a) 4.4.1.4.1.1 Surveillance (b) 4.4.1.4.1.2
. Surveillance (c) 4.4.1.4.1.3 Surveillance (d) 4.4.1.4.1.4 3.3.1.4.2 Cold Shutdown - Loops Not Filled 3.4.1.4.2 Applicability 3.4.1.4.2 Action 3.4.1.4.2 Surveillance (a) 4.4.1.4.2.1 Surveillance (b) 4.4.1.4.2.2 Surveillance (c) 4.4.1.4.2.3 3.3.1.5 Isolated Loops 3.4.1.5 & 3.4.1.6 Applicability 3.4.1.5 & 3.4.1.6 Action 3.4.1.5 & 3.4.1.6 Surveillance 3.4.1.4 & 3.4.1.6
-3.3.1.6 Isolation Loop Start-up 3.4.1.7 Applicability 3.4.1.7 Action 3.4.1.7 Surveillance (a) 4.4.1.7.1 Surveillance (b) 4.4.1.7.3 Surveillance (c) 4.4.1.7.3
_____-______________-___-_=___-
- Attachment;4 B13181/Page 4
.Existina T.S. # Description Proposed RTS
=3.3.1.7 Idled Loop 3.4.1.8 & 3.4.1.9 Applicability 3.4.1.8 & 3.4.1.9 Action 3.4.1.8 & 3.4.1.9-Surveillance (a) 4.4.1.8.1 & 4.4.1.9.1 Surveillance.(b) 4.4.1.8.2 & 4.4.1.9.2 3.3.I'.8 Idled Loop Start-up 3.4;1.10 & 3.4.1.11 Applicability 3.4.1.10 & 3.4.1.11 Action . 3.4.1.10 & 3.4.1.11
. Surveillance (a) 4.4.1.10.1 & 4.4.10.11.1 Surveillance (b) -4.4.1.10.2 Surveillance-(c) 4.4.1.10.3 & 4.4.1.11.2 3.3.2.I' Safety Valves-Shutdown 3. 4. 2.1 -
Applicability 3.4.2.1 Action 3.4.2.1 Surveillance 4.4.2.1 3.3.2.2 Safety Valves - Operation 3.4.2.2 Applicability 3.4.2.2 Action 3.4.2.2-Surveillance 4.4.2.2 3.3.3 Pressurizer 3.4.3 Applicability. 3.4.3 Action
- 3. 4.3 .
Surveillance (a) 4.4.3.1 Surveillance (b) 4.4.3.2 3.3.4.1. Relief Valves 3.4.4 Applicability- 3.4.4 Action . 3.4.4 Surveillance (1) 4.4.4.1 Surveillance (2) 4.4.4.2 Surveillance (3) 4.4.4.3 Surveillance (4) 4.4.4.4 Surveillance (5) 4.4.4.5 3.3.4.2 Low Temperature Overpressure 3.4.9.3 Protection System
- a. SLRV 3.4.9.3a
- b. RCS Vent. 3.4.9.3b Applicability 3.4.9.3 Action 3.4.9.3 Surveillance (a) 4.4.9.3.1 Surveillance (b) 4.4.9.3.2
Aitachment4.
'B13181/Page'5' Existina T.S.'# Description' , Proposed RTS ]!
'3.3.5 . .
'f 3.3.5.1 Reactor Coolant System Vents 314.11
.. Applicability '3.4.11
- Action 3.4.11 Surveillance (a). 4.4.11.a
. Surveillance (b).~
4.4.ll.b Surveillance.(c) 4.4.11.c
'3.4 Combined Heatup, Cooldown'and Pressure Limitations 3.4.A Reactor Vessel
'3.4.A.1 RCS pressure and' temperature 3.4.9.lc l
During hydrostatic and leak testing.
'3.4.A.2 RCS pressure and temperature 3.4.9.1 heatup and cooldown 3.4.A.3 . Average rate of RCS temp Change 3.4.9.1.a and b of.RCS Temp. Change 3.4.A.4 Allowable Pressure - Temp -3.4.9.I' Combinations 3.4.B Pressurizer 3.4.B.I' 500 psig.. Limit 3.4.9.2.d.
3.4.B.2 Heatup Rate 3.4.9.2.a
'3.4.B.3 Cooldown Rate 3.4.9.2.b 3.4.B.4: Temperature Difference 3.4.9.2.c l 3. 4 . C .1 ' . Steam Generator Pr/ Temp 3.7.2.a
-3.4.C.2' Max heat up/cooldown 3.7.2.c 3.4.C.3 Tube sheet temp 3.7.2.d
-3.4.C.4' SG vessel temp 3.7.2.b
'3.4 Applicability 3.7.2 3.5 Chemical and Volume Control System 3.5.A.1 Charging Pumps 3.1.2.2.a &
3.1.2.4 l 3.5.A.2 Boric Acid Pumps 3.1.2.2.b h 3.5.A.3' Boric Acid Tank 3.1.2.6 3.5.A.4 Maintenance 3.1.2.6
- 3.5.A.5~ Flow Paths 3.1.2.2.a 3.5.A.6- Valve BA-V-399 3.1.2.1 &'
3.1.2.2 3.5.8 RCS Cold Legs Less than 315'F 4.1.2.3.3 and 3.1.2.4 3.8 ' Turbine Cycle 3.7.1 3.8.A.~1 Safety Valves-Steam 3.7.1.1/ relieving capability Table 3.7-1 3.8.A.2.a Steam driven AFW pumps- 3.7.1.2 3.8.A.2.b One AFW pump inoperable 3.7.1.2.a 3.8.A.2.c Two AFW pumps inoperable 3.7.1.2.b 3.8.A.3.a DWST/PWST min. vol. 3.7.1.3 3.8.A.3.b DWST inoperable 3.7.1.3.a \,
- u. __1_1______________ _ . _ _ _ _ _ _ _ _ _ _
p~ .. i 1 Attachment'4 -j B13181/Page _. 6 ' i I! ! Existino T.S. #' Description- Proposed'RTS ( L 3,8.A.3.c PWST inoperable 3.7.1.3.b 3.8.A.4 System piping 3.7.1.1 and 3.7.1.2 and 3.7.1.3 and 3.7.1.5-3.8.B.1- AFW actuation system instrumentation Table 3.3-2 Item 3 l, 3.8.B.2 AFW actuation contacts and relays Table 3.3-2 Item 3-Table 3.8-1 AFW actuation system instrumentation Table 3.3-2 Item 3 3.9 Operational Safety Instrumentation and Control Systems A Logic Required for Full Power Table 3.3-1 Operations Table 3.3-2 B Required Action if Logic Falls Below Table 3.3-1 Limit ' Table 3.3 - C Neutron Monitoring Note (3) D Accident Monitoring Inst. Channel Table 3.3-7 E Required Action Table 3.3-7 Table 3.9-1 Minimum Instrumentation Operating Conditions Item 1 Nuclear Overpower Reactor Trip Table 3.3-1 Item 2 Item 2 Pressurizer Variable Low Pressure Table 3.3-1 Reactor Trip Item 4 Item 3 Pressurizer Fixed High Pressure Trip Table 3.3-1 Item 5 Item 4 Pressurizer High Water Level Table 3.3-1 Reactor Trip Item 6 Item 5 Reactor Coolant Flow Table 3.3-1 Item 7 Item 6 Pressurizer Pressure Low Table 3.3-2 Item 1 Item .7 -- Deleted Item 8- Manual Trip Table 3.3-1 Item 1 Item 9 Steam - Feedwater Flow riismatch Table 3.3-1 Item 9
-Item 10 High Steam Flow Table 3.3-1 Item 8 Item 11 Containment High Pressure Table 3.3-2 Item 5 Start-up Eouipment Intermediate Range SUR Reactor Trip Table 3.3-1 Item 3 Source Range SUR Rod Stop Note (4)
Attachment 4 B13181/Page 7 j j Existino T.S. # Description- Proposed RTS Refuelino Requirement p 3.9.2
~
l Shutdown High Neutron Level Alarm Table 3.9-2 Accident Monitoring Instrumentation- , Item 1 Pressurizer Level Table 3.3-6 : L Item 5 i Item 2 Aux. Feedwater Flow Rate Table 3.3-6 1 Item 11 < Item 3 Delete I Item 4 PORY Position Indicator Table 3.3-6 ) l Acoustic Flow Monitor Item 14 Item 5 PORV Block Valve Position Table 3.3-6 l , Indicator Item 13 , Item 6 Safety Valve Position Indicator, . Table 3.3-6 ! Acoustic Flow Monitor. Item 14 i 3.10 Reactivity Control System 4 1 3.10.1.1 Shutdown Margin - Modes 1, 2 3.1.1.1 Applicability 3.1.1.1 ; Action- 3.1.1.1 l Surveillance la 3.1.1.1 Surveillance Ib' 4.1.1.1.1.a 1 Surveillance Ic 4.1.1.1.1.b j Surveillance ld 4.1.1.1.1.c I Surveillance 2 4.1.1.1.2 j 3.10.1.2 Shutdown Margin - Mode 3 3.1.1.2 Applicability 3.1.1.2 l Action 3.1.1.2 ; Surveillance (a) 4.1.1.2a { Surveillance (b) 4.1.1.2b' ; 3.10.1.3 Shutdown Margin - Modes 4, 5 3.1.1.3 ; Applicability 3.1.1.3 l Action 3.1.1.3 ' i Surveillance (a) 4.1.1.3.a Surveillance (b) 4.1.1.3.b . 3.10.1.4 Shutdown Margin - three loop 3.1.1.4 1 Applicability 3.1.1.4 1^ Action 3.1.1.4 Surveillance (1) 4.1.1.4.1 . Surveillance (2) 4.1.1.4.2 l 3.10.1.5 Moderator Temperature Coefficient 3.1.1.5 j Applicability 3.1.1.5 ' Action 3.1.1.5 Surveillance (a), (b), (c) 4.1.1.5.a,b,c 3.10.1.6 Minimum Temp. for Criticality 3.1.1.6 Applicability 3.1.1.6 Action 3.1.1.6 a Surveillance (a)(b) 4.1.1.6.a, b l;
# : Attachment.4 B13181/Page 8 p ,
Existina T.S. #' Description Proposed RTS 3.10.2~ Movable Control Assemblies , .3.10.2.1 Bank Height 3 .1. 3 .1 - Applicability 3.1.3.1-Action 3.1.3.I' Surveillance (a)(b) 4.1.3.1.1 and: 4.1.3.1.2 3.10.2.2 Positive Indication System'- Operating 3.1.3.2 Applicability 3.1.3.2 o Action 3.1.3.2 Surveillance Requirement. 4.1.3.2 3.10.2.3 Positive Indication Systems - Shutdown 3.1.3.3 Applicability 3.1.3.3
-Action' 3.1.3.3 Surveillance 4.1.3.3 3.10.2.4 Rod Drop Time 3.1.3.4 Applicability 3.1.3.4 Action 3.1.3.4 Surveillance 4.1.3.4 3.10.2.5 Shutdown Insertion Limits 3.1.3.5 Applicability 3.1.3.5 Action 3.1.3.5
- i. = Surveillance Requirement .
4.1.3.5 3.10.2.6 Control Group Insertion Limits - 3.1.3.6.1 Four Loops i Applicability .3.1.3.6.1 Action 3.1.3.6.1 Surveillance 4.1.3.6.1 3.10.2.7 Control Group Insertion Limits - 3.1.3.6.2 Three Loops Applicability 3.1.3.6.2 Action 3.1.3.6.2 Surveillance 4.1.3.6.2 3.11 Containment Administrative Tech. Spec. 3.11.A. Leakage Limit 3.6.1.2.a 3.11.B.2 Containment Integrity with reactor 3.9.1 vessel head removed 3.11.C Internal Pressure 3.6.1.4 3.ll.D.1 Air Recirculation System Performance 4.6.2.c Requirement 3.ll.D.2 Air Recirculation System Cold Shutdown 3.6.2 Requirement 3.11 Containment
-3.11A. Leakage Limit (see 3.11A Admin.) 3.6.1.2.a 3.11B Containment Integrity 3.11 B.1 RCS above 300 psig. and 200*F 3.6.1.1 3.ll.B.2 See Admin. 3.11.B.2 3.9.1 3.11.B.3 Positive Reactivity Changes 3.6.1.1 - 3.9.4 3.11.C Internal Pressure (See Admin. 3.11.c) 3.6.1.4
Attachment 4 813181/Page 9 i Existina T.S. # Description Proposed RTS 3.11.0 See Admin. 3.11.D.1 and 3.11.D.2 4.6.2.c and 3.6.2 3.ll.E Containment Spray System Note (5) 3.11 F Containment Venting 3.11.F.1- Post-Accident Hydrogen Venting 3.6.1.7, Table 3.3-10 item Ic 3.11.F.2 Purge Capability 3.6.1.7 3.11.G. Containment Isolation Valve 3.11.G.1 Restore Inoperable Valve 3.6.3.a 3.11.G.2 Isolate by use of automatic valve 3.6.3.b 3.11.G.3 Isolate by use of manual valve 3.6.3.c 3.11.G.4 Hot Standby 3.6.3.d 3.11.H Trip Setpoint 3.3.2
- r. 3.13 Refueling 3.13A Monitoring Radiation Levels Note (2) 3.13B Monitoring Neutron Flux 3.9.2 3.13.C.1 Water Level in the Refueling Cavity 3.9.8.2, 3.9.10 3.13.C.2 RHR Pump & Heat Exchanger in Operation 3.9.8.1 3.13D Baron Concentration 3.9.1 3.13E Charging Pump 3.1.2.3 3.13F Verification of Subcriticality Note (1) 3.13G Direct Communication 3.9.5 3.13H Handling of Spent Fuel Cask 3.9.7 3.13I Loading of Fuel for Offsite Lab Study -
3.15 Intentionally left Blank 3.16 Intentionally Left Blank 3.17 Power Distribution Limits 3.17.1 Axial Offset - 3.17.1.1 Axial Offset - four Loops 3.2.1.1 Applicability 3.2.1.1
' Action 3.2.1.1 Surveillance (a) 4.2.1.1.1 Surveillance (b) 4.2.1.1.2 Surveillance (c) 4.2.1.1.3 Surveillance (d) 4.2.1.1.4 3.17.1.2 Axial Offset - three loops 3.2.1.2 Applicability 3.2.1.2 Action 3.2.1.2 Surveillance (a) 4.2.1.2.1 Surveillance (b) 4.2.1.2.2 Surveillance (c) 4.2.1.2.3 Surveillance (d) 4.2.1.2.4 3.17.2 Linear Heat Generator Rate 3.2.2.1 L 3.17.2.1 Four Loops Operating 3.2.2.1 Applicability 3.2.2.1 Action 3.2.2.1 i Surveillance (1) 4.2.2.1.1 Surveillance (2) 4.2.2.1.2
t, h m Attachment 4 LB13181/Page.10 Existino T.S. # ~ Description Proposed RTS 3.17.2.2. Three Loops Operating 3.2.2.2 L Applicability ~ 3.2.2.2 l ' Action 3.2.2.2 L Surveillance (1) 4.2.2.2.1 Surveillance (2) 4.2.2.2.2 3.17.3 Nuclear Enthalpy Rise Hot Channel Factor 3.17.3.1 Four Loops Operating 3.2.3.1 Applicability 3.2.3.1 Action 3.2.3.1 Surveillance (a) 4.2.3.1.1 l Surveillance (b) 4.2.3.1.2-3.2.3.2 13.17.3.2 Three Loops Operating Applicability. 3.2.3.2 Action 3.2.3.2. Surveillance (a) 4.2.3.2.1 Surveillance (b) 4.2.3.2.2 3.17.4 Quadrant Power Tilt Ratio 3.2.4 Applicability 3.2.4 Action 3.2.4 Surveillance (a) 4.2.4.1
-Surveillance (b) 4.2.4.1 Surveillance (c). 4.2.4.1
-3.17.5 DNB Parameters 3.2.5-Applicability 3.2.5 Action 3.2.5 Surveillance.(a) 4.2.5.1 Surveillance (b) 4.2.5.2 Surveillance (c) 4.2.5.3 3.18 Intentionally Left Blank 3.19 Snubbers 3.7.4 3.19.A- Applicability 3.7.4 3.19.B One inoperable 3.7.4 3.20 Intentionally Left Blank 3.21 Safety-Related Equipment Flood Protection 3.21.1 Operability Requirement 3.3.4 3.21.2 Condensate Return Pump Operability 3.3.4 3.21.3 Screenwell House & D.G. Room Operability 3.21.4 Actions for 3.21.1 and 3.21.2 3.3.4 Can't Be Met 3.22.A.1 Fire Water System / Operability 3.7.6.1 3.22.A.2 One Pump Inoperable 3.7.6.1.a (Action) Note (9)_
3.7.6.1.b 3.22.A.3 Two Pumps Inoperable (Action) Note (9) ; 3.22.B.1 C0 System /0perability 3.7.6.3 1 3.22.B.2 Ackion 3.7.6.3.a (Action) l 3.22.B.3 Action - Deportability Note (9) 3.22.C.1 Halon System /0perability 3.7.6.4
Attachment 4
.B13181/Page 11 Existina T.S. # Description Proposed RTS 3.22.C.2 Action 3.7.6.4.a (Action) 3.22.C.3 Action - Deportability Note (9) 3.22.D.1 Fire Water Stations / Operability 3.7.6.5/3.7.6.6 3.22.D.2 Action 3.7.6.5.a/3.7.6.6.a Note (13) 3.22.E.1 Fire Detection System / Operability 3.3.3.6 3.22.E.2.a Action 3.3.3.6.b 3.22.E.2.b Action - Deportability Note (9) 3.22.F.1 Penetration Fire Barriers / Operability 3.7.7 3.22.F.2 Action 3.7.7.a Note (15) 3.22.G.1 Spray and/or Sprinkler Systems 3.7.6.2 3.22.G.2 Action 3.7.6.2.a (Action) 3.22.G.3 Action - Deportability Note (9) -
3.22.H Flammable Liquids Controls 3.7.8 3.22.H.1 Action - Written Permission 3.7.8.a 3.22.H.2 Action - Container 3.7.8.b 3.22.H.3 Action - Fire Watch 3.7.8.c Table 3.22-1 Fire Water Stations Table 3.7-4/3.7-5 Table 3.22-2 Fire D'tection Instruments Table 3.3-8 3.23 Post-accident Monitoring Instrumentation 3.3.3.5 Table 3.23-1 and Accident Monitoring Instrumentation Table 3.3-7 and Table 3.23-2 Table 4.3-6 Item 1 Containment Pressure 1 Item 2 RCS - Cold Leg Temp. 2 Item 3 RCS - Hot Leg Temp. 3 Item 4 RCS Pressure 4 Item 5 Containment Water Level 15 Item 6 CET 17 Item / Main Stack Wide Range Noble Gas Monitor 18 Item 8 Containment Atmosphere High Range Radiation 19 Monitor Item 9 Reactor Vessel Water Level 20 Itea 10 RCS Subcooling Maring Monitor 12 3.24 Special Test Exceptions 3.10.1 3.24.1 Shutdown Margin 3.10.1 Applicability 3.10.1 Action 3.10.1 Surveillance (a) 4.10.1.1 Surveillance (b) 4.10.1.2 3.24.2 Physics Test 3.10.2 Applicability 3.10.2 Action 3.10.2 Surveillance (a) 4.10.2.1 Surveillance (b) 4.10.2.1 Surveillance (c) 4.10.2.3 3.24.3 Position Indication System - Shutdown 3.10.3 Applicability 3.10.3 , Action 3.10.3 i Surveillance 4.10.3 i j~ )
' Attachment 4 1 '
- B13181/Page 12' u '
Existina T.S. #- Description Proposed-RTS
~ . 4 .1'. Introduction to Surveillance Requirements 4.01, 4.02 4.2 Administrate'ive Operational Safety Items l . Table 4.2-2 PORV's and Block Valves Demonstrated Operable ' Item 15 A Demonstrated Operable . 4.4.4.1 B Block Valve Demonstrated . 4.4.4.3 C The Emergency Air and Power Supply Demonstrated Operable L
C.1 Transfer From Normal to Emergency Power 4.4.4.6 C.2 Operate through Complete Cycle 4.4.4.6 D- Demonstration of Minimum Pressure on 4.4.4.5 Emergency Air Supply. 4.2 Operational Safety Items
' Table 4.2-1 . Minimum Frequencies for Testing, Calibrating and/or Checking Instrument Channels 1 Nuclear Power . Table 4.3-1, Item 2 2 ' Intermediate Range Table 4.3-1, Item 3 3- Source Range -
4 . Reactor Coolant Temperature Table 4.3-6 Note (8) 5 Reactor Coolant Flow -Table 4.3-1, Item 7, 3.2.5 6 Pressurizer Level Table 4.3-1, Item 6 7 Pressurizer Pressure Table 4.3-1, Item 5 8 Variable Low Pressure Trip Setpoint Table 4.3-1, Item 4 Calculator . 9 Rod Position Digital Voltmeter 3.1.3.2 10 Rod Position Counters 3.1.3.2 11 Steam Generator Level Table 4.3-2 12 Steam Generator Flow Mismatch Table 4.3-1 13 Charging Flow _ Note (6) 14 Residual Heat Pump Flow 4.5.1.g(7) 15 Boric Acid Tank Level 4.1.2.5a, 4.1.2.6.1.b, Table 4.3-6 16 Refueling Water Storage Tank Level Table 4.3-6, Item 11 17 Volume Control Tank Level 4.4.6.1.c 18 Blank i Radiation Monitoring System Table 4.3-3 20 Boric Acid Control Note (6) 21 Blank 22 Valve Temperature Interlocks 4.4.1.7.2 and 4.4.1.1:.3 , 23 Pump-Valve Interlock 4.4.1.7.4 and 4.4.1.11.4 1 _ - _ _ - - - _ - . _ _ - _ =
Attachment 4. f( ;B13181/Page 13-
;Existina T.S. #'
Description Proposed RTS-. 4 24- Reactor Coolant System OPS 4.4.9.3.4 125 Auxiliary Feedwater Flow Rate Table 4.3-6, Item 11 26 Blank 27 .PORV Position Indication Table 4.3-6, (Acoustic Monitor) Item 14 28- PORV Block Valve Indication . Table 4.3-6, Item 13 29 Safety Valve Position Indication Table 4.3-6, Item 14 2-(Acoustic Monitor). 4.2 Operational Safety Items Table 4.2-2. Minimum Equipment Check and Sampling Frequency 1 Reactor Coolant Sample Table 4.4-4, Item 1 2 Reactor Coolant Boron 3.1.1.2,.3.1.1.3 3 Refueling Water Storage Tank Water Sample 4.1.2.5a 4 Control Rods 3.1 3.4 5 Control Rods 3.1.3.1 6 Pressurizer Safety Valves 4.4.2.1 and 4.4.2.2 Main Safety 4.7.1.1 7 8 Main Steam Isolation Valves 4.7.1.5 : 4.6.3 9 Reactor Containment. Trip Valves 10 Refueling System Interlocks - 3 11 Boric Acid Pumps 4.1.2.1.b, 4.1.2.2.b ! 12 RCS Overpressure Protection System 4.4.9.3.3 !
-Isolation Valve Interlocks and Alarms l 13 RCS Overpressure Protection Isolation 4.4.9.3.1 l Valves 1 14 RCS Vent (s) 4.4.9.3.2 ;
4;4- Containment Testing Administrative Tech. Specs. i I.B.1 Acceptance Criteria 3.6.1.2a IV.A.4- Demonstrated condition for filteration 4.6.2.e unit-IV.B.1 Acceptable filter efficiencies - IV.C.3 Corrective Actions for Unusual Conditions - IV.D Test Frequency IV.D.1 18-month test frequency 4.6.2.c IV.D.2 Visual Inspectiu - 1 IV.D.3 Damper test - ; IV.D.4 Charcoal Spray Valve - IV.D.5 Halogenated Hydrocarbon Testing 4.6.2 9 IV.D.6 Cold D0P Test 4.6.2.f IV.D.7 15-Minutes Operational Requirement 4.6.2.a.1 ] 4.4 Containment Testing 4.4.I.A Integrated Leakage Test 4.6.1.2, 4.6.1.6.1 4.4.I.B.1 See Admin. Spec. 3.6.1.2.a I t _ .___-__-___-_a
p Attachment 4-E 'B13181/Page 14 h Existina'T.S. # Description Proposed RTS 4.4.I.B.2
- Max.- Allowable Reduced Pressure Test -
(P Leakage Rate 4.4.II.A Inb)ividual Leak Detection Test 4.6.1.2, 4.6.1.3, 4.6.1.7.2-4.4.'II.B Acceptance Criterion 3. 6.1.'2. b ; 4.4.II.C Corrective Action 3.6.1.2
.4.4.II.D.1- Equipment hatch and fuel transfer- 4.6.1.1.c, Tube 4.6.1.2 4.4.II.D.2 Isolation Valves 4.6.1.2.d
'4.4.II.D.3 - Personnel Air-lock Assembly 4.6.1.3,-
4.6.1.1.b, 4.6.1.2.d 4.8 AFW system 3/4.7.1.2
.. 4.8.1 AFW operability every 31 days 4.7.1.2.1 4.8.1.a Discharge pressure 4.7.1.2.1.a 4.8.1.b S/G level instrumentation Tabl es . 4.3-2/
4.2-1 4.8.1.c Verify correct valve position 4.7.1.2.1.b 4.8.2- DWST/PWST operability every 12 hours. 4.7.1.3.1 4.8.3 AFW operability every refueling 4.7.1.2.2 ( 4.8.3.a Pump capability 4.7.1.2.2.a 4.8.3.b Verify correct valve position upon 4.7.1.2.2.b AFW actuation test signal 4.8.3.c Verify AFW pump starts upon AFW 4.7.1.2.2.c actuation test signal 4.9 MSIVs 3.7.1.5 4.10 Inservice Inspection and Reactor Vessel Surveillance 4.10A ISI of Class 1, 2, 3 Component . 4.0.5a 4.4.10 4.10B ISI of Class 1, 2, 3 Pumps and Valves 4.0.5a. 4.4.10 4.10C RCP Flywheel 4.4.10. 4.10D Reactor Vessel Surveillance Capsule Table 4.4-5 4.10.1 In-service Inspection of Steam Generator lobes 4.10.1A SG Sample Selection and Inspection 4.4.5.1 4.10.1B SG Tube Sample Selection and Inspection 4.4.5.2 4.10.1.B.1 Areas to Be Inspected 4.4.5.2.a 4.10.1B.2 First Sample 4.4.5.2.b 4.10.1.B.3 Second and Third Sample 4.4.5.2.c 4.10.1.C Inspection Frequencies 4.4.5.3 t 4.10.1.D Acceptance Criteria 4.4.5.4 f 4.10.1.E Reports 4.4.5.5 Table 4.10.1-1 Minimum Number of SG Tube Inspected Table 4.4-1 Table 4.10.1-2 SG Tube Inspection Table 4.4-2 L lz N - - - _ - _ _ _ _ _ _
b JAttachment 4
.B13181/Page 15 L '
Existino T.S.'#- Description Proposed RTS-4.11 Deleted 4.12 High Energy Piping System 4.0.6 4.12A -Augmented Inservice Inspection Program .4.0.6 4.12.A.1- FirSt Ten-Year Inspection Program 4.0.6 4.12.A.2 Successive Inservice Inspection Program 4.0.6 4.12.A.3 Repairs, Reexamination and Test 4.0.6 4.13 Snubbers 4.7.4 4.13.A Visual inspection schedule 4.7.4.a 4.13.B- Visual inspection criteria 4.7.4.b 4.13.C Functional tests 4.7.4.c 4.13.D Hydraulic snubbers test criteria 4.7.4.d 4.13.E Mechanical snubbers test criteria 4.7.4.e 4.13.F Snubber service life monitoring 4.7.4.f 4.14 Flood Protection Annunciators 4.14A Test 4.3.4 4.148 - Acceptance Criteria 4.3.4 4.14C Corrective Action 3.3.4 4.14D- Test Frequency 4.3.4
'4.15.A.1 Fire Water System Operability 4.7.6.1.1 4.15.A.I.a Pump Operability 4.7.6.1.1.a Note (10) 4.15.A.I.b Valve Operability 4.7.6.1.1.b Note (11)
, 4.15.A.I.c Valve Operability 4.7.6.1.1.c 4.15 A.I.d.1 Auto Actuation 4.7.6.1.1.d 4.15.A.I.d.2 Pump Flow / Pressure 4.7.6.1.1.d 4.15.A.I.d.3 Valve Operability 4.7.6.1.1.d 4.15.A.I.e Flow Test 4.7.6.1.1.e Note (12) 4.15.B.1 System Operability 4.7.6.3 4.15.B.I.a CO,inder CyT Weight 4.7.6.3.a 4.15.B.I.b.1 Component Operability 4.7.6.3.b.1 4.15.B.I.b.2 Flow Test 4.7.6.3.b.2 4.15.C.1 Halon System Operability 4.7.6.4 4.15.C.I.a Cylinder Weight / Pressure 4.7.6.4.a 4.15.C.I.b.1 Component Operability 4.7.6.4.b.1 4.15.C.I.b.2 Visual Inspection 4.7.6.4.b.2 4.15.D.1 Fire Hose Station Operability 4.7.6.5/4.7.6.6 4.15.D.I.a Visual Inspection 4.7.6.5.a/ 4.7.6.6 a 4.15.D.I.b Revmoval/ Inspection 4.7.6.5.b/ 4.7.6.6.c 4.15.D.I.c Flushing 4.7.6.5.c/ 4.7.6.6.c 4.15.D.I.d Valve Operability 4.7.6.5.c/ I.- 4.7.6.6.c 4.15.D.I.e Hose Hydrostatic Test 4.7.6.5.b.1/ 4.7.6.6.c.1 4.15.E.1 Channel Functional Test 4.3.3.6.1 Note (14) l
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l Attachment 4* B13181/Page16 Existino T.S.'# i . Description Proposed RTS 4.15.E.2' Circuit Supervision 4.3.3.6.2
.4.15.F.1 Penetration Fire Barrier Operability 4.7.7.1 4.15.F.1.a Visual Inspection- 4.7.7.1.a 4.15.F.1.b. Post-Repair Inspection .
Note (16) 4.15.G.I. Spray and/or Sprinkler Operability 4.7.6.2 . 4.15.G.I.a Valve Operability 4.7.6.2.b'
.4.15.G.I.b.1 Functional Test 4.7.6.2.'c.1 4.15.G.I.b.2 Visual Inspection - Headers. 4.7.6.2.c.2 , 4.15.G.I.b.3 Visual . Inspection - Nozzles 4.7.6.2.c.3 4.15.G.I.c Flow Test 4.7.6.2.d Note (18) 5.0 Design Features 15.1 Introduction -
5.2 Site Description 5.1.1 5.3.A- Reactor Core 5.3.I' 5.3.8 Reactor Coolant System 5.4.1 5.4 Containment 5.2.1 6.0 Administrative Controls 6.1- Responsibility 6.1
6.2 Organization
6.2
~6.2.1 Offsite Organization 6. 2.1 -
6.2.2 Facility Staff 6.2.2-6.3 Facility Staff Qualification 6.3 6.3.I' Facility Staff Qualification 6.3.1 6.3.1.1 Health Physics Supervisor 6.3.1.1 6.3.1.2 STA 6.3.1.2 6.4 Training 6.4 6.4.1 Petraining and Replacement 6.4.1 Training Program 6.4.2 Fire Brigade Training Program 6.4.2 6.5 Review and Audit 6.5
- 6. 5.' PORC 6.5.1 6.5.1.1 PORC Function 6.5.1.1 6.5.1.2 Composition 6.5.1.2 6.5.1.3 Alternate 6.5.1.3 6.5.1.4 Meeting Frequency 6.5.1.4 6.5.1.5 Quorum 6.5.1.5 6.5.1.6 Responsibilities 6.5.1.6 6.5.1.6a Responsibilities 6.5.1.6a 6.5.1.6b Responsibilities 6.5.1.6b 6.5.1.6c Responsibilities 6.5.1.6c 6.5.1.6d Responsibilities 6.5.1.6d 6.5.1.6e Responsibilities 6.5.1.6e 6.5.1.6f Responsibilities 6.5.1.6g 6.5.1.6g Responsibilities 6.5.1.6h
,.- 6.5.1.6h Responsibilities 6.5.1.6i 6.5.1.61 Responsibilities 6.5.1.6j 6.5.1.6j Responsibilities 6.5.1.7b
, y L h ~ Attachment 41 813181/Page 17 Existino T.S. # Description- Proposed RTS 6.5.1.7a. Authority- 6.5 1.7a 6.5.1.7b Authority 6.5.1.7b L 6.5.1.8 Records 6.5.1.8 6.5.2- .NRB- 6.5.2 6.5.2.1 Qualification 6.5.2.1 6.5.2.2 Composition 6.5.2.2 6.5.2.3' ~ Consultants 6.5.2.3 6.5.2.4- Meeting Frequency 6.5.2.4 6.5.2.5 Quorum 6.5.2.5 6.5.2.6 . Review 6.5.2.6 6.5.2.7 Audits- 6.5.2.7 6.5.2.8 Authority 6.5.2.8~ 6.5.2.9 Records 6.5.2.9
- 6.6 Reportable Event l Action 6.6 6.7 Safety Limit Violation 6.7 6.8.1 Written Procedures 6.8.1 6.8.2 Approval of Procedures 6.8.2 6.8.3 .
Temp. Changes to Procedures 6.8.3 6.8.4 Admin. Written Procedures 6.8.1.d, e,'f
~6.8.5 Admin. Written Procedurec 6.8.4 6.8.6 Admin. Written Requirements 6.8.5 6.9 Reporting Requirements 6.9.1 6.9.la Start-up Reports 6.9.1.1, 6.9.1.2, and 6.9.1.3-L 6.9.1.b SG Tube Inspection 4.4.5.5.b 6.9.1.c Occupational Exposure Report 6.9.1.5.a 6.9.1.d Monthly Operating Report 6.9.1.8 6.9.1.e 10CFR50.59b~
6.9.1.f Admin. Annual Radiological . 6.9.1.6 Environmental Operating Report 6.9.1.g Admin. Semiannual Radioactive Effluent 6.9.1.7 Release Report 6.9.2 Special Reports 6.9.2 6.1.2a ISI Results 4.0.5 6.9.2c Reactor Vessel Material Surveillance 4.4.9.1.2 Specification Examination 6.9.2.d SG Tube Report 4.4.5.5.a 6.9.2.e Post-Accident Operability Table 3.3-7 6.9.2.f Fire Protection System Operability - 6.9.2.g RCS Vent 3.4.11 6.9.2.h Radiological Effluent Reports 3.11.2.2 3.11.1.2 3.11.2.3 3.11.3 Y _ . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ -
b " ~ Attachment 14~ j ' (,. . B13181/Page~ .18 ' l
'Existina T.S.' #
' Description '
Proposed RTS 6.10.1 Record Retention '6.10.2
- 6.10.2: Record Retention 6.10.3.
,. 6.11. .
Radiation ~ Protection Program: 6.11.1
, 6.12- : Deleted -
6.13 High Radiation Area 6.1.2
~6.14' Deleted -
6.15 , System Integrity 6.15-
.6.16 lodine Monitoring ,
.6.16 x
6.17 -. REM 0DCM _ .. 6.13 6.18 Radioactive Waste Treatment Systems 6.14
- 6.19 - PASS / Sampling and Analysis Plant 6.16
_ . Effluents 7/8' .. Radioactive Effluents 3.11 7/8.1.1-- Liquid Effluents 3.11.1.1 4 7.1.1.1: Concentration : 3.11.1.1'
' Applicability 3.11.1.1
. Action: . 3.11.1.1 8.1.'1.1.1 Sampled and Analyzed 4.11.1.1.1 8.1.1'.1.2 Assure Limits 4.11.1.1.2 7.1.1.2 Dose-Liquid 3.11.1.2 Applicability. 3.11.1.2:
Action ;3.11.1.2
' 8.1.1. 2.1 ' Determination - 4.11'. l . 2.1 8.1.1.2.2 Confirmation- 4.11.1.2.2-
'7.1.2.1- Dose Rate - Gas 3.11.2.1
, Applicability 3.11.2.1
. Action '3.11.2.1' 8.1.2.1.1 Determination 4.11.2.1.'1 8.1.2.1.21 Control of Release Rates 4.11.2.1.2 8.1.2.1.3 Release Rate of I-131, etc. 4.11.2.1.3
,7.1.2 ~ 2 . Dose-Noble Gas 3.11.2.2 Applicability 3.11.2.2-Action 3.11.2.2
'8.1.2.2.1 Cum.. Dose 4.11.2.2.1 8.1.2.2.1 . Confirmation 4.11.2.2.2 7.1.2.3 : Dose-Iodine 3.11.2.3 Applicability 3.11.2.3 Action 3.11.2.3 8.1.2.3.1 Cum. Dose Contributions 4.11.2.3.1
'8.1.2.3.2 Confirmation' 4.11.2.3.2 7.1.3 Total Dose 3.11.3 Applicability 3.11.3 Action 3.11.3 8.1.3 Determination 4.11.3
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ii : Attachment 4 B13181/Page-19 J LExistinaT.SI~# Description ~- ~ Proposed RTS' 7/8.2; Instrumentation he 7.2.1.1- Radioactive Liquid Effluent Instrumentation 3.3.3.7 Applicability- 3.3.3.7. Action .3;3.3.7. t 8.2.1.1 Demonstrate Operable -4.3.3.7.1 [ 7.2.2.1 - Radioactive Gaseous Effluent Monitoring 3.3.3.8< Instrumentation
' Applicability 3.3.3.8 Action. 3.3.3.8:
D 8.2.2.1 - Demonstrated Operable '4~.3.3.8.1
, , t s
n-I l'
- - - . . - - - - - - . _ - - - - - - - - - - - - - u --_ -,-_.--_ _ . - - -_- _. - - - - - - -- - _ _ - ------.u-_ - - _ . . - _ - - - - - - - - - - - - - - _ - - _ - - _ _
Attachment 4' B13181/Page 20 Notes: (1) The proposed deletion of existing Technical Specification Section 3.13(F) could potentially affect the detection of an inadvertent criticality. However, adequate maintenance and monitoring of the core reactivity . condition is assured by the. refueling boron concentration requirements (the proposed RTS 3.9.1) and via the audible indication of neutron flux (the proposed RTS 3.9.2). (2) The deletion of Section 3.13(A), which requires continuous radiation monitoring during fuel movement, does not mitigate the consequences of any accident and is more appropriately located in plant operating procedures rather than Technical Specifications. (3) The operability requirements of the existing Technical Specification Section 3.9.c are related to plant reliability and not reactor safety. All reactor safety requirements are covered in the proposed _ Table 2.2-1 and 3.3-1. (4) The source range rod stop is - a contro1 grade circuit which is not credited in any safety analysis. (5) The containment spray is not credited in any accident. The calculated peak containment pressure can be maintained at a safe level by using just the air recirculation system alone. (6) These items are not credited in the design basis analyses. In addition, the W STS format does not require these parameters in the Technical Specifications. (7) The RHR pump flow test is performed once per 60 months instead of once per refueling. (8) The RCS hot leg and cold leg temperature channel surveillance is performed ~once per month instead of once per shift. (9) The proposed RTS deletes the requirement for submitting a Special Report to NRC. Each reportable event will be reviewed in accordance l with the requirements of 10CFR50.73 as stated in the proposed RTS I Section 6.6.1. (10) The proposed RTS requires operation of the electric pump for 5 minutes and the diesl pump for 30 minutes in accordance with recommendations in NFPA. (11) The proposed RTS requires verification of position only for valves which are not locked, sealed, or otherwise secured. (12) The proposed RTS deletes the reference to Chapter 5 of the Fire Protection Handbook for performing flow testing of the Distribution System. The RTS testing requirements are equivalent. (13) The proposed RTS allows several different options if a fire water hose station is out of service. k- --
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[ L ' Attachment 4 . t B13181/Page 21- _ (14)' The proposed RTS requires'a Trip Actuating Device Operational Test in
- lieu of the Channel Functional Test. -This is an enhancement because .;
the trip device will now be tested along with the channel circuit. 1 Detectors which cannot- be- reset are exempted from this . testing,. however the associated channel circuit:will have- been tested and is
- i. equivalent to the current specification requirements.
(15) The proposed RTS allows the use of qualified temporary seals for up to i 30 days. ; t-(16) The requirement for verifying that the barrier is functional following ;
- repairs ' have been deleted in the proposed RTS. An inspection- after i repairs or replacement is an integral part of' the repair. and replacement procedures and is not necessary.
(17) For the' Haddam Neck Plant, the UNRESTRICTED AREA boundary coincides , with the EXCLUSION AREA boundary as defined in 10CFR100.3(a), but the H UNRESTRICTED AREA does not include areas over water bodies. (18) An option for performing the water flow test in lieu of just an air - test, is now included. Allowing this option- has been shown not to degrade the overall system reliability and has been concluded to provide an equivalent level of fire protection. I 1 s 1 i 1 a l
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