ML20044D789

Amend 157 to License DPR-61,revising Footnote to TS 3.8.3.2.b to Identify Available Options for Providing Power to 480-volt Buses During Plant Shutdown & Changing Special Test Exception TS 3.10.3 Re Position Indication Sys

ML20044D789
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	05/17/1993
From:	Stolz J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20044D787	List: ML20044D789
References
NUDOCS 9305200295
Download: ML20044D789 (10)

Text

~

8(

[g UNITED STATES

+

NUCLEAR REGULATORY COMMISSION o

,j WASHINGTON, D. C. 20555

%...../

i CONNECTICUT YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-213 HADDAM NECK PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.157 License No. DPR-61 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Connecticut Yankee Atomic Power Company (the licensee), dated March 22, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission, C.

There is reasonable assurance (1) that.the activities authorized by i

this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

i 9305200295 930517 PDR ADOCK 05000213 P

PDR

, 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-61 is hereby amended to read as follows.

I (2) Technical Soecifications I

The Technical Specifications contained in Appendix A, as revised through Amendment No.157, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION J

F. Stolz, Direct roject Directorate I-4 t

Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation i

Attachment:

Changes to the Technical Specifications j

l i

Date of Issuance:

May 17, 1993 I

l i

i ATTACHMENT TO LICENSE AMENDMENT NO. 157 FACILITY OPERATING LICENSE NO. DPR-61 i

DOCKET NO. 50-213 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and i

contain vertical lines indicating the areas of change.

j 2

Remove Insert l

3/4 8-14 3/4 8-14 j

B3/4 4-3 B3/4 4-3 i

]

B3/4 7-2 B3/4 7-2

)

j B3/4 7-3 B3/4 7-3 1

B3/4 7-3a B3/4 7-3a 3/4 10-3 3/4 10-3 B3/4 10-1 B3/4 10-1 8

I l

i 1

ELECTRICAL POWER SYSTEMS I

SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner:

a.

4160-Volt Emergency Bus #8 or #9, b.

480-Volt Emergency Busses #4 and #5, if Emergency Bus #8 above is energized, or Emergency Busses #6, #7 and #11, if Emergency Bus #9 i

above is energized,*

120-Volt A.C. Vital Busses A and B or Vital Busses C, C1, D, and c.

D1 energized from their associated inverters connected to their respective D.C. busses, and d.

125-Volt D.C. Bus A energized from its associated battery bank if Emergency Bus #8 above is energized, or 125-Volt D.C. Bus B and BX energized from its associated battery bank if Emergency Bus #9 above is energized.

l APPLICABILITY: MODES 5 and 6.

ACTION:

With any of the above required electrical busses not energized in the a.

l required manner, immediately suspend all operations involving CORE i

ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, l

and initiate corrective action to energize the required electrical busses in the specified manner as soon as possible.

b.

Entry into Mode 5 pursuant to Specification 3.0.4 with any of the above required electrical busses not energized in the required manner is not permitted.

i Sj)RVEILLANCE REOUIREMENTS 4.8.3.2 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.

l Tie breaker 4T5 may be closed to power bus 4 or bus 5; 6T7 may be closed to power bus 6 or bus 7; 6T11 and 11T6 may be closed to power bus 11 from bus 6 only.

HADDAM NECK 3/4 8-14 Amendment No. UE,157 0099 1

~

REACTOR COOLANT SYSTEM BASES i

i 3/4.4.3 PRESSURIZER l

The limit on the water level in the pressurizer assures that the parameter is maintained within that assumed in the safety analyses. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored 2

to within its limit following expected transient operation. The requirement j

that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and l

establish natural circulation.

f 1

3/4.4.4 RELIEF VALVES I

Operation of the power-operated relief valves (PORVs) minimizes the undesir-able opening of the spring-loaded pressurizer Code-safety valves and provide an alternate means of core cooling.

Each PORV has a remotely operated block i

i valve to provide a positive shutoff capability should a PORV become inoper-abl e.

One of two redundant PORV relief trains must be OPERABLE to adequate-ly 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 i

function to relieve RCS pressure during_ all design transients up to and 3

including the design step load decrease with steam dump. However, no credit is taken for the auto-open function of the PORVs in the design basis accident

analyses, i

(

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 The program for inservice inspection of steam generator tubes maintained.

is based on:

(a) a modification of Regulatory Guide 1.83, Revision 1 and l

(b) Previous Eddy Current Examination results. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of meenanical 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 j

cause of any tube 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 j

chemistry guidelines which minimize corrosion of the steam generator tubes.

j If the secondary coolant _ chemistry is not maintained within these limits, l

The extent of cracking during plant opera-l localized corrosion may occur.

l tion would be limited by the limitation of steam generator tube leakage l

between the primary coolant system and the. secondary system imposed in Operating plants have demon-Specification 3.4.6.2, Operational Leakage.

strated that primary to secondary leakage specified in Specification 3.4.6.2-can readily be detected by radiation monitoring of steam generator blowdown.

i Despite chemistry controls, several forms of steam generator corrosion have One of the most prevalent forms of been found throughout the, industry.

corrosion is wastage, which involves a general thinning of the tube wall HADDAM NECK B3/4 4-3 Amendment No. JU,157 4

0101

)

PLANT SYSTEMS BASES 4

AUXILIARY FEEDWATER SUPPLY (Continued) 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 J

transfer 200 gpm from the PWST to the DWST. An alternate supply can be provided from the 100,000 gallons Recycled Primary Water Storage Tank.

l 3/4.7.1.4 SPECIFIC ACTIVITY i

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 1

dose also includes the effects of a coincident 0.4 gpm reactor-to-secondary tube leak in the steam generator of the affected steam line.

These values are consistent with the assumptions used in the safety analyses.

i 3/4.7.1.5 MAIN STEAM LINE TRIP VALVES

~

i i

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 4

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 e

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

)

l The limitation on steam generator pressure and temperature ensures that the i

4 pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70*F and l

200 psig are based on a steam generator RTNDT of 10*F and are sufficient to l

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.

i 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.

For Cycle 16 Refueling Outage only, OPERABILITY of the Diesel Generator would include reliance on service water cooling via firehose for a maximum of 14 consecutive days, at a maximum inlet temperature of 80*F and a minimum service water header pressure of 23 psig.

The two service water pumps which are powered by the "A" EDG must be operable during the construction period. 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, the motor operated isolation valve that isolates nonessential turbine building service water loads, the air operated isolation valve that isolates nonessential primary i

HADDAM NECK B3/4 7-2 Amendment No. J2E,145,157 ozoo

=y

.-.g-

+

3 yy.+-

w

a l

4 l

PLANT SYSTEMS BASES 3/4.7.3 SERVICE WATER SYSTEM (Continued) auxiliary building service water loads, the motor operated isolation valve l

s l

that aligns service water flow to the RHR heat exchanger, and the

)

l safety-related piping and components.

Each service water header contains an in-line filter with a bypass line containing a motor operated isolation valve that supplies the containment

?

air recirculation unit coolers. The bypass line allows continued plant operation should the in-line filter become clogged.

Each bypass line and its isolation valve are sized such that only one bypass line is needed to To satisfy single failure pass the required service water flow.

J assumptions, an operable in-line filter bypass _ isolation valve is required in each service water header due to possible common mode failure of the in-line filters and at least one in-line filter must be in service whenever the service water system is required to be OPERABLE. The service water headers may be tied together by an open service water header cross-connect in the intake structure.

l 3/4.7.4 SNUBBERS All snubbers are required to be OPERABLE to ensure that the structural J

integrity of the reactor coolant system and all other safety-related systems l

is maintained during and following a seismic or other event initiating Snubbers excluded from this inspection program are those dynamic loads.

installed on nonsafety-related systems and then only if their failure, or i

l failure of the system on which they are installed, would have no adverse effect on any safety-related system.

i

.c 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 deter-i mined by the number of inoperable snubbers found during an inspection.

3 Inspections performed before that interval has elapsed may be used as a new However, the results of reference point to determine the next inspection.

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 inspec-tion interval will override the previous schedule.

}

i 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 Generically susceptible be exempted from being counted as inoperable.

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 environmental conditions, such as temperature, radiation, and vibration.

4 i

l HADDAM NECK B3/4 7-3 Amendment No. J/E,157 0100 4

i l

i i

PLANT SYSTEMS 4

BASES i,

3/4.7.4 SNUBBERS (Continued)

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 I

determine if any safety-related component or system has been adversely affected by the inoperability of the snubber. The engineering evaluation j

shall determine whether or not the snubber mode of failure has imparted a I

significant effect or degradation on the supported component or system.

l i

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 i

snubbers to verify freedom of movement over the full stroke, restraining characteristics, and drag force (if applicable). Ten percent (107.) of the i

total of each type of snubber represents an adequate sampling for these Observed failures on these samples require testing of additional i

tests.

1 units.

i Hydraulic snubbers and mechanical snubbers may each be treated as a different entity for the above surveillance program.

1 i

i i

i 4

i.

1 4

l i

HADDAM NECK-B3/4 7-3a Amendment No. JEE.157 0100

I SPECIAL TEST EXCEPTIONS 3/4.10.3 POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual shutdown and control rod drop time measurements, calibration of the Analog Rod Position Indication System, or -

control rod drive slave cycler timing tests provided:

Only one shutdown or control bank is withdrawn from the fully a.

inserted position at a time, and b.

K is maintained 1 0.94, with the highest worth bank fully wi[kdrawn.

APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements, calibration of the Analog Rod Position Indication System, and control rod drive slave cycler timing tests.

ACTION:

0.94, or more than one bank of rods withdrawn, immediately open WithK((o>rtripbreakers.

there$

SURVEllLANCE RE0VIREMENTS 4.10.3 K

shall be determined to be 10.94 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the sIbt of and at least once per 24 i.ours thereafter during rod drop time measurements, Analog Rod Position Indication System calibration, and control rod drive slave cycler timing tests, by.

verification of adequate RCS baron concentration.

l i

l l

l l

HADDAM NECK 3/4 10-3 Amendment No. JEE, 157 0102 i

i i

d

.i 3/4.10 SPECTAL TEST EXCEPTIONS BASES 4

3/4.10.1 SHUTDOWN MARGIN This Special Test Exception provides that a minimum amount of control rod i

worth is immediately available for reactivity control when tests are per-formed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted q

i core reactivity condition occurring as a result of fuel burnup or fuel l

l cycling operations.

l 3/4.10.2 PHYSICS TESTS

)

This Special Test Exception permits PHYSICS TESTS to be performed at less

[

y than or equal to 5% of RATED THERMAL POWER with the RCS T slightly lower I

than normally allowed so that the fundamental nuclear chaNEteristics of the f

core and related instrumentation can be verified.

In order for various characteristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications.

j for instance, to measure the moderator temperature coefficient at BOL, it is s

necessary to position the various control rods at heights which may not normally be allowed by Specification 3.1.3.6 which in turn may cause the RCS T

to fall slightly below the minimum temperature of Specification i

i 3yg3,4, a

1 3/4.10.3 POSITION INDICATION SYSTEM-SHUTDOWN This Special Test Exception permits the Position Indication Systems to be inoperable during rod drop time measurements, position indication system calibration, and slave cycler testing. The exception is required during rod drop testing since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Position Indication Systems remain j

OPERABLE.

The exception is also necessary during position indication system

• calibration and slave cycler testing, since these activities require withdrawal of control rods in order to perform calibration and ensure proper operation of the control rod drive cyclers.

Since maintaining X 5 0.94 withthehighestworthbankfullywithdrawnensuresshutdownmarh requirements are met, position indication ic not required, and the reactivity requirements for rod drop testing are bounding.

3/4.10.4 INDIVIDUAL ROD POSITION INDICATION SYSTEM - OPERATING This Special Test Exception permits the IRPI system to be inoperable during the performance of data collection / verification / adjustment testing of the 3

IRPI. The testing is required to develop and implement correction factors for each individual rod position indicator. While the IRPI system is inoperable, the indicated individual rod position cannot be used to verify control rod alignment (Specification 3.1.3.1) or control rod insertion limits (Specifications 3.1.3.5, 3.1.3.6.1 and 3.1.3.6.2).

The actual rod position for banks C, D and A is, however, unaffected by this testing.

i HADDAM NECK B3/4 10-1 Amendment No. J2E, J27, 157 0103 w

.w-r-,.,m..

--