Text

Proposed Tech Specs,Correcting Minor Discrepancies & Typos Identified During Various Reviews
	ML20216J679
Person / Time
Site:	Sequoyah
Issue date:	06/24/1987
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20216J671	List: ML20216J667; ML20216J679;
References
	TAC-R00187, TAC-R00188, TAC-R187, TAC-R188, NUDOCS 8707060256
	Download: ML20216J679 (56)
	v • d • e

{{#Wiki_filter:s i ^~ ENCLOSURE 1 PROPOSED technical specification CHANGE ~ SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-87-17) l LIST OF AFFECTED PAGES Unit 1 Unit 2 l I I j II II 2-7 2-1 2-9 2-9 B2-1 B2-1 3/4 1-14 3/4 1-14 3/4 1-21 3/4 1-21 3/4 3-13 3/4 3-4 3/4 3-56 3/4 3-7 3/4 3-73 3/4 3-13 3/4 7-1 3/4 3-57 3/4 7-6 3/4 6-4 3/4 7-10 3/4 6-18 t 3/4 7-37 3/4 7-1 3/4 8-4 3/4 7-6 3/4 8-5 3/4 7-10 3/4 11-12 3/4 7-25 3/4 12-1 3/4 7-50 3/4 12-2 3/4 9-1 3/4 12-10 3/4 11-9 B3/4 6-3 3/4 12-2 5-2 3/4 12-9 5-6 5-2 5-6 8707060256 870624 DR ADOCK 05000327 PDR o

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} 4 2.1 SAFETY LIMITS i BASES l 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and j possible cladding perforation which would result 19 the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented i by restricting fuel operation to within the nucleate boiling regime where the l heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform j and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, i defined as the ratio of the heat flux that would cause DNB at a particular i core location to the local heat flux, is indicative of the margin to DNB. l The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. The curves of Figure 2.1-1 end 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DN8R is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. The curves are based on an enthalpy hot channel factor, F gf1.55and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F at reduced power based on the expression: sH Fh=1.55[l+0.3(1-P)] M R23 ~ where P is the fraction of RATED THERMAL POWER v December 23, 1982 ^ SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19 (_ mm m_

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT-LIMITING CONDITION FOR OPERATION 3.1. 3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within + 12 steps (indicated position) of their group step counter demand position. APPLICABILITY: MODES 1* and 2* ACTION: a. With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUT 00WN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours. b. With more than one full length rod inoperable or misaligned from the group step counter demand position by more than + 12 steps (indicated position), be in HOT STANDBY within 6 hours. i c. With one full length rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group step counter demand height by more than + 12 steps (indicated position), POWER OPERATION may continue provide 3 that within one hour either: 1. The rod is restored to OPERABLE status within the above alignment requirements, 2. The remainder of the rods in the group with the g inoperable rod are aligned to within + 12 steps of the inoperable rod within one hour while maintaining the rod sequence and insertion limitt of Figuret 3.1-1 2M 2.'-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that: ^See Special Test Exceptions 3.10.2 and 3.10.3. r SEQUOYAH - UNIT 1 3/4 1-14 N SEP171980 $uu.qrw .d bh 1 *.fh! .h w4..a. u 2dRJ ~ d csZbd A -1

REACTIVITY CONTROL SYSTEMS m CONTROL R00 INSERTION LIMITS LIMITING CONDITION FOR OPERATION ~ 3.1. 3. 6 The control banks shall be limited in physical insertion as shown.in Figure 3.1-1. lR45 APPLICABILITY: MODES 1* and 2*#. Ev. ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either: Restore the control banks to within the limits within a. two hours, or b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position usingtheabovefigure/,or c. Be in HOT STANDBY within 6 hours. m SURVEILLANCE RE0UIREMENTS l i 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual i rod positions at least once per 4 hours. t ^5ee Special Test Exceptions 3.10.2 and 3.10.3. ~

With K,ff greater than or equal to 1.0.

l 9 'g. " /* .Sepcember 3, 1985 SEQUOYAH - UNIT 1 3/4 1-21 Amendment No. 43. i y .h d bdd$dn u.Mi OUSAGML i SN

TABLE 4.3-1 (Continued) m NOTATION ~I With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal. (1) If not performed in previous 7 days. (2) Heat balance only, above-15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent. AxzAt. FLUX DZFFEREnti (3) Compare incore to excere c: ':' .. di'f :r:: cbove 10% of RATED THERMAL POWER. Recalibrate if the absolute difference greater than or equal to 3 percent. (4) Manual ESF functional input check every 18 months. (S) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS. The test shall independently verify the RSS cpercbilit OPERADZLZh"yof the uncervoltage and automatic shunt trip circuits. (6) Neutron detectors may be excluded from CHANNEL CALIBRATION. (7) Below P-6 (Block of Source Range Reactor Trip) setpoint. (8) Logic only, each startup or when required with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal if not performed in previous 92 days. CQNNEL. fHNcr20NAL TEST (9) The thennel functicn test-shall independently verify the operability of the undervoltage and shunt trip circuits for the manual reactor trip function. (10) - Local manual shunt trip prior to placing breaker in service. Each R58 train shall be tested at least every 62 days on a -:tc;;; red tc:t b:.;i;.- STAGGER &D TEJT BMr1. (11) - Automatic and manual undervoltage trip. I ~ l t March 16, 1967 SEQUDYAH - UNIT 1 S/4 3-12 Amencment No. 54 a

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s { u; INSTRUMENTATION r TABLE 4-3-0 (Continued) TABLE NOTATION w During liquid additions to the tank. (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists: 1. Instrument indicates measured levels above the alarm / trip setpoint. 2. Circuit failure. 3. Downscale failure. R17 (2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists: 1. Instrument indicates measured levels above the alarm setpoint. 2. Circuit failure. 3. Downscale failure. R17 (3) The initial CHANNEL. CALIBRATION shall be performed using one or more of the ~ reference standards certified by the National Bureau of Standards or using ) standards that have been obtained from suppliers that participate in measure-ment c,ssurance activities with NBS. These standards shall permit calibrat-ing the system over its intended range of energy and measurement range. For j subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made, i &') ld $lblll/Vi/ r$//VC/iCNA?L i~~~$/~ .ib.9ll GkSG f/?fCn4b/I! s7 ~ O '1 rvo*J its/ sat:,>,seit i.i d e/;b-> c.: i h n p/itkiy/ /Ae $#csJx siis aish / in WW'- /> a/s/m sn,7unei.h N n seca/> s sny s 71-22-ot g,,y,,,,;g '"'"'Y /. l7?S/ramed is fic'an?.s />1FDSUNa/ /CN/J Q$ orc /dc W/WD. a/s/m,l/ rip serycid.

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775c dlG)/V/ VEL RWAN./44 i~fW skwl/ a/Je ormeru /r.de Stof c:n/>r/ /?:m uniuarish:a c c o u r.s,; f f/,, fi/ca,,,,,r &rMheh occa,,.. l* 0040!!.SdQ/c ?'ON2//c, gu. MM f. s. SEQUOYAH - UNIT 1 3/4 3-73 Amendment No. 13 ~ ~ " ~~

4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE l SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-4. APPLICABILITY: MODES 1, 2 and 3. ACTION: a. With 4 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Setpoint trip is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. 'n b. With 3 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves associated with an operating loop inoperable, operation in unamme M00E -2;;33W 3 may proceed provided, that within 4 hours, either the inoperable j valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint trip is reduced per Table 3.7-2; otherwise, be in at least 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 REOUIREMENTS ~ i 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5. b SEQUOYAH - UNIT 1 3/4 7-1 gg;py; A -o lf;. }Nf '. -.QY jf fj.:il' 5..:.. k),Nhh[Mkhdh b M $ b l d} M NA 21 4 i, 2;ddi$ ht.y . w,u n i butsdQL

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.~ PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 31 days by is<Wisy 4d Al lead pec once

3. 2 each automatic control valve in the flow path is OPERABLE R16 whenever the auxiliary feedwater system is placed in automatic control or when above 10% of RATED THERMAL POWER.

l b. At least once per 18 months during shutdown by: j l 1. Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an auxiliary feedwater actuation test signal and a low auxiliary feedwater pump suction ) pressure test signal. 2. Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an auxiliary feedwater actuation test signal. At least once per 7 days by verifying that each non-automatic valve c. in the auxiliary feedwater system flowpath is in its correct position. e h A MAR 251982 SEQUOYAH - UNIT 1 3/4 7-6 Amendment No. I2

1 6 - PLANT SYSTEMS m [MAINSTEAMLINEISOLATIONVALVES _ p. y LIMITING CONDITION FOR OPERATION s 3.7.1.5 Each said steam line' isolation valve shall be OPERABLE. ~ APPLICABILITY:, MODES 1, 2 and 3. .4 % -ACTION: MODE 1 - Withonemainsteamlineisolatihnvalveinoperable,POWEROPERATION. may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours; Otherwise,'be in at least HOT STAN0BY'within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. MODES 2 - With one main steam line isolation valve inoperable, subsequent and 3 operation.in MODES A 2 or 3 may proceed provided: I a. The isolation valve is maintained closed; b. The provisions of-Specification 3.0.4 are not applicable. A Otherwise, be in at least HOT STAND 8Y within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS f4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5. l ) i 1 ej.l.. f l.1 : > ' - dd(MC E,s SEQUOYAH - UNIT l' 3/4 7-10 J-. SM < i+ $E! :. MM j

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~ ~ FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION 3.7.11.4 The fire hose stations shown in Table 3.7-)5d shall be OPE ( APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. ACTION: With one or more of "the fire hose stations shown in Table 3.7-10 a. inoperable, route an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within 1 hour if the inoperable fire hose is the primary means of fire suppression; otherwise route th.e additional hose within 24 hours. Restore the fire hose station to OPERABLE status.within 14 days or, in lieu of any other report required by Specification 6.6.1, prepare and submit R40 a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the station to OPERABLE status. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. u.,/ SURVEILLANCE REOUIREMENTS l 4.7.11.4 Each of the fire hose stations shown in Table 3.7-10 shall be demonstrated OPERABLE: At least once per 31 days by visual inspection of the stations a, accessible during plant operations to assure all required equipment is at the station. b. At least once per 18 months by: ~ l. Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station, 2. Removing the hose for inspection and re-racking, and 3. Inspecting all gaskets and replacing any degraded gaskets in the couplings. ~ At least once per 3 years by: c. 1. Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage. 2. Conducting a hose hydrostatic test at a pressure of 150 psig or ,-g' at least 50 psig above maximum fire main operating pressure, whichever is greater. j November 23, 1984 SEQUOYAH - UNIT 1 3/4 7-37 Amendment No. 36

[ -{ ELECTRICAL POWER SYSTEMS { SURVEILLANCE REQUIREMENTS (Continued) W _. ) 4. Simulating a loss of offsite power by itself, and- \\ a) Verifying de energization of the shutdown boards and load shedding from the shutdown boards. b) Verifying the diesel starts on the auto-start signal, energizes the shutdown boards with permanently connected loads within 10 seconds, energizes the auto-connected shutdown loads through the load sequencers and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady state voltage and frequency of the shutdown boards shall'be maintained at 6900 i 690 volts and 60 x 1.2 Hz during this test. 5. Verifying that on a ESF actuation test signal (without~ loss of offsite power) the diesel generator start's on the auto-start signal and operates on standby _ for greater than or equal to 5 minutes. The generator voltage and frequency shall be 6900 1 690 volts and 60 1 1.2 Hz within 10 seconds after the auto-start signal; the steady state generator voltage and fre-quency shall be maintained within these limits during this test. 6. Simulating a loss of offsite power in conjunction with an ESF RS3 actuation test signal, and , ~.. a) Verifying de-energization of the shutdown boards and load 'j, shedding from the shutdown boards. b) Verifying the diesel' starts +rer -Mert condition-on the auto start signal, energizes the shutdown boards with permanently connected loads within 10 seconds, energizes the auto connected emergency (accident) loads through the load sequencers and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady state voltage and frequency of the emergency busses shall be maintained at 6900 1 690 volts and 60 1 1.2 Hz during this test. l c) Verifying that all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the shutdown board and/or safety injection actuation signal. 7. Verifying the diesel generator operates for at least 24 hours. lR53 _ During the first 2 hours of this test, the diesel generator shall be loaded to greater than or equal to 4400 kw and during the remaining 22 hours of this test, the diesel generator shall be loaded to greater than or equal to 4000 kw. October 28, 1986 SEQUOYAH - UNIT 1 3/4 8-4 Amendment No. 49 ) L

1 ELECTRICAL POWER SYSTEMS

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SURVEILLANCE REQUIREMENTS (Continued) [ Within 5 minutes aftn ting this 24 hour test, perform nmn i SpecificationQ.8.1.1.2.d. The generator voltage and fre-lR56 quency shall be byuu 2 690 volts and 60 1.2 Hz within 10 seconds after the start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test. 8. Verifying that the auto-connected loads to each o;eir' cenerator do not exceed the 2000 hour rating of 4000 kW. 9. Verifying the diesel generator's capability to: a) Synchronize with the offsite power source while the genera-tor is loaded with its emergency loads upon a simulated restoration of offsite power. b) Transfer its loads _to the offsite power source, and c) Be restored to its shutdown status. 10. Verifying that the automatic load sequence timers are OPERABLE with the setpoint for each sequence timer within f, 5 percent of its design setpoint. 11. Verifying that the following diesel generator lockout features prevent diesel generator starting only when required: a) Engine overspeed b) 86 GA lockout relay At least once per 10 years or af ter any modifications which could e. affect diesel generator interdependence by starting the diesel generators simultaneously, during shutdown, and verifying that the diesel generators accelerate to at least 900 rpm in less than or equal to 10 seconds. f. At least once per 10 years

by:

1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypoclorite solution, and 2. Performing a pressure test of those portions of the diesel fuel oil system design to Section III, subsection ND of the ASME Code at a test pressure equal to 110 percent of the system design pressure.

These requirements are waived for the initial surveillance.

I February 3, 1987 SEQUOYAH - UNIT 1 3/4 8-5 Amendment No. 52 Ua, , as

- E b

b S I TABLE 4.11-2 (Continued) O %,.: i . TABLE NOTATION 4 The LLD is defined, for the purposes of these specifications, as the a. smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95% proba-R46 bility with only a 5% probability of falsely concluding that a blank observation represents a "real" signal. 1 For a' particular measurem' ent system (which may include radiochemical separation): 1 4.66 sb E*V 2.22x10* Y exp (. ) $#00 ~ Where: - A Ad 1 LLD is the "a priori" lower limit of detection as defined above in microcurie per unit mass or volume, lF46 is the standard deviation of the background counting rate or of s h tne counting rate of a blank sample as appropriate (as counts per minute), E is the counting efficiency as counts per disintegration, m 'lRF) V is the sample size in units of mass or volume, 2.22 x 108 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable), A is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between midpoint of sample collection and ^ time of counting (midpoint). R46 It should be noted that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. January 14, 1986 SEQUOYAH - UNIT 1 3/4 11-12 Amendment No. 42 R41

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION ' ~ 3.12.1 The radiological environmental monitoring program shall be conducted ~ as specified in Table 3.12-1. APPLICABILITY: At all times. ACTION: With the radiological environmental monitoring program not being con-a. ducted as specified in Table 3.12-1, in lieu of a LER, prepare and submit to the Commission, in the Annual Radiological Environmental R46 l Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. b. With the level of radioactivity in an environmental sampling medium exceeding'the reporting levels of Table 3.12-2 when averaged over any calendar quarter, in lieu of any other report required by Specifica-tion 6.9.1, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Report pursuant to Speci-fication 6.9.2, a special report that identifies the cause(s) for ex-ceeding the limit (s) and defines the correctivo actions to be taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of specifications of 3.11.1.2, 3.11.2.2, and 3.11.2.3. When more than R46 one of the radionuclides in Table 3.12-2 is detected in the sampling medium, this report shall be submitted if: concentration (1), concentration (2) + ***> 1*0 limit level (1) limit level (2) When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a member of the public is equal to or lR46 greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents;- however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report. c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12-1,. identify loca-tions for obtaining replacement samples and add them to the radio-R46 logical environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. In lieu of a licensee event report (LER) and pursuant to Specification C.^.2, identify the cause(s) of the 1 unavailablity of samples and identify the new locations for obtaining 'b.9.l.7. January 14, 1986 SEQUOYAH - UNIT 1 3/4 12-1 Amendment No. 42

RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION (Continued) 4 ~ \\ replacement samples in the Annual Radiological Environmental Operat-ing Report. A revised figure (s) and table (s) for the ODCM reflecting the new location (s) shall be included in the next anaus! radie!egica! q u' effluent release report pursuant to Specification 6.9.1.9. T d.- The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. C SURVEILLANCEREQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the table and figure in the ODCM and shall be analyzed pursuant to the requirements of Table 3.12-1 dnd the detection capabilities required by Table 4.12-1. l n/a ^ m January 14,1986 SEQUOYAH - UNIT 1 3/4 12-2 Amendment No. 42 ~

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A Land Use Census shall be conducted and shall identify within a dis-i tance of 8 km (5 miles) the location in each of the 16 meteorological sectors of'the nearest milk animal d the nearest residence and the nearest garden

of 2 (500 f t ) producing fresh leafy vegetables.

greater than 50 m APPLICABILITY: At all times. -{ ACTION: With a Land Use Census identifying a location (s) that yields a calcu-a. lated dose or dose commitment 20% greater than the values currently being calculated in Specification 4.11.2.3, identify the new loca-tion (s) in the next Semian'ual Radioactive Effluent Release Report n pursuant to Specification 5.0.1.7.

6. 9. /. 9, b.

With a Land Use Census identifying a location (s) that yields a I calculated dose or dose commitment (via the same exposure p&thway) 4 20% greater than at a location from which sample: are currently being obtained in accordance with Specification 3.12.1, add the new location (s) within 30 days to the Radiological Environmenta.1 Mont-R46 toring Program given in the ODCM, if samples are available. The sampling location (s), excluding the control station location, having the lowest calculated dose or dose commitment (s), via the same expo-sure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted. Pursuant to Specification C.14, submit in the next Semiannual Radio-active Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflect-ing the new location (s) ylth information supporting the change in sampling locations. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. c. SURVEILLANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during.the growing season at-least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, mail survey, telephone survey, aerial survey, or by consulting local agriculture authorities. The results of R46 the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 0.0.1.0. (o. 9. ' 7. /

Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census.

Speci-fications for broad leaf vegetatioi. sampling in Table 3.12-1.4c. shall be followed, including analysis of control samples. I SEQUOYAH - UNIT 1 3/4 12-10 Amendment No.42 V January 14, 1986

CONTAINMENT SYSTEMS BASES 3 /4. 6.1. 8 EMERGENCY GAS TREATMENT SYSTEM (EGTS) The OPE 3 ABILITY of the EGTS cleanup subsystem ensures that during LOCA conditions, containment vessel leakage into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is necessary to meet tha assumptions used in the accident analyses and limit the site boundary radiation deses to within FP the limits of 10 CFR 100 during LOCA conditions. Cumulative operation of the system with the heaters on for 10 hours over a 31 day period is sufficient to reduce the buildup of moisture on the absorbers and HEPA filters. *M30-N510-1975 will be used as a procedural guide for surveillance testing. A y,gy 3/4.6.1.9 CONTAINMENT VENTILATION SYSTEM Use of the containment purge lines is restricted to only one pair (one supply line and one exhaust line) of purge system lines at a time to ensure that the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss of coolant accident during purging operations. The analysis of this accident assumed purging through the largest pair of lines (a 24 inch inlet line and a 24 inch outlet line), a pre-existing iodine spike in the reactor coolant and four second valve closure times. 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that containment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses. 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 j pressurization of the containment. Containment isolation within the time i limits specified ensures that the release of radioactive material to the j environment will be consistent with the assumptions used in the analyses for a 1 LOCA. By letters dated March 3, 1981, and April 2, 1981, TVA will submit a l report on the operating experience of the plant no later than startup after R8 the first refueling. This information will be used to provide a basis to re-evaluate the adequacy of the purge and vent time limits. I I SEQUOYAH - UNIT 1 D 3/4 6-3 Amendment 5 4/15/81 . $ wn~f, bkh M)dsh$hNM$Ed m.,;,,, %i

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i e I INDEX 4a.: C. : OEFINITIONS .a... SECTION PAGE /? comber. A. n 3 M" "J8 '.':/ d# 1.0 DEFINITIONS 1.1 ACTI0N............................................................ 1-1< 1.2 AXIAL FLUX DIFFERENCE..............*............................... 1-1 I.3 B ypMs 1.sm & B PA TH................., .......s..,.....s.,. /W 4ra CH ANN E L C A LI B RAT I O N......i........................................ 1-1 I /,r.b4 CHANNEL CHECK..................................................... 1-1 i 1 l 1,6 5 CH ANNE L FUNCT I ONAL TE ST........................................... 1-1 4 /.7.h & CONTAINMENT INTEGRITY............................................. 1-2 /,84r7' CONTROLLED LEAKAGE................................................ 1-2 ) /,9.be-C O R E A LTE R AT I O N................................................... 1-2

l. Io -1d DO S E EQUI VA L ENT I-131.............................................

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1-3 \\ I./.14rli-ENGINEERED SAFETY FEATURE RESPONSE TIME........................... 1-3 i /. /3 -Irle F R EQU E N CY N0TATI O N................................................ 1-3 J /./Y4-It GASE005 RADWASTE TREATMENT SYSTEM................................. 1-3 /,/.r d-it' I D ENT I F I ED LE A KAG E................................................ 1-3

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/./7 -h 5-OFFSITE DOSE CALCULATION MANUAL................................... 1-4 ' li // -b46 O P E RAB LE - O P E R AB I LITY............................................1-4 //9 1-19 OPERATIONAL MODE - M0DE......'..................................... ~ l-4 l /, 2 0 h P HY S I C S T E ST S.................................................... 1-4 i 4 /. 2 /.L-a PRE S 5 URE SOUND ARY L EAKAG E....................................... 1-4 ^ I /.12 _ bee-P RO C E S S CO N T RO L P R0G R AM...................................... 1-4 1.13 -hel P U R G E - P U RG I NG..................................................... 1-5 /, 2 V 20- QU AD RANT POWER T I LT RATI 0........................................ 1-5 V d.7.e) u SEQUOYAH - UNIT 2 .I Y e

t { INDEX l DEFINITIONS '" l SECTION PAGE l f'ene.< bee /L <> J 1.0 DEFINITIONS (Continued) jpr,tran ey as 4251.2C RATED THERMAL POWER. 1-5 { 1 l 424 4-44 REACTOR TRIP SYSTEM RESPONSE TIME...... 1-5 l l EVCNT~ i l 627 AreE R E PO R TA B L E -0 C C 'J R R E N C E............................... ~............. 1-5 { ^2B -kn8E SHIELD BUILDING INTEGRITY... 1-5 49 1-af-SHUTDOWN MARGIN.. 1-6 t.30.sz 7g BodNMRS... \\ l /. 3 / 1rES SOLIDIFICATION.. 1-6 ^32-kr&& SOURCE CHECK... 1-6 1 433 J J4b STAGGERED TEST SASIS.. 1-6 /,3Y J-St THERMAL POWER. 1-6 I.Jf.p,gg. UNIDENTIFIED LEAKAGE.,. 1-6 428 yyREs7RZCTED ARSA 437 1-41 VENTILATION EXHAUST TREATMENT SYSTEM........ 1-7 438 4-3+ VENTING., 1-7 l OPERATIONAL M00E5 (TABLE 1.1). 1-8 il l FREQUENCY NOTATION (TABLE 1.2). 1-9 1 l l 1 i l (.. SEQUOYAH - UNIT 2 JI l

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BASES ~ 1 2.1.1 REACTOR CORE The restrictions of thi.s Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding.is prevented by restricting fuel operation-to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the co_olant' saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction ii. heat transfer coefficient. ONB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation. The W-3 ONB correlation has been developed to predict the ONB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, i defined as the ratio of the heat flux that would cause DNB at a particular l core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the ONBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30. This /^ value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin'to DNS for all operating conditions. The curves of Figure / 2.1-1 M 2.'-2 show the. loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. These curves are based on an enthalpy hot channel factor, f H' 0 1 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F at reduced power basea on the_expressioni H FSH=1.55[1+0.3(1-P)] where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the fi (delta I) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits. SEQUOYAH - UNIT 2 B 2-1 Amendment No. 21 'SEP 2 91983 M

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES '~- GROUP HEIGHT ' 'Il LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods shall be OPERABLE and positioned within : 12 steps (indicated position) of their group step counter demand position. APPLICABILITY: M0. DES 1* and 2*. ACTION: a. With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to De untriopable, determine that the SHUTDOWN MARGIN require-ment of Scecification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours. b. With more than one full length rod inoperable or. misaligned from tne j group step counter demand position by more than t 12 steps (indicated ] position), be in HOT STANDBY within 6 hours. c. With one full length rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its grouD step counter demand height by more tnan t 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either: 1. The rod is restored to OPERABLE status within the above alignment requirements, or 2. The remainder of the rods in the group with the inoperable rod are aligned to within + 12 steps of the inoperable rod while g-maintaining the red sequence and insertion limitt of Figuret 3.1-12-d 2. '_-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3. The rod is declared inoperable and the SHUTDOWN MARGIN reouirement of Specification 3.1.1.1 is satisfied. POWER. OPERATION may then continue provided that: ~ a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions. "See Special Test Exceptions 3.10.2 and 3.10.3. i SEQUOYAH - UNIT 2 3/4 1-14 I i t '[ , d' k 56pA g gghfhg%gg[g l

REACTIVITY CONTROL SYSTEMS ~ CONTROL ROD INSERTION LIMITS ~

~~

LIMITING CONDITION FOR OPERATION f 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3.1-1. R33 APPLICABILITY: Modes 1* and 2*#. + ACTION: With the control banks inserted beyond the above insertion limits, except for { surveillance testing pursuant to Specification 4.1.3.1.2, either: Restore the control banks to within the limits within two hours, or a. b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group positionusingtheabovefigure/,or c. Be in at least HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS / 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours. "See Special Test Exceptions 3.10.2 and 3.10.3.

With K,ff greater than or equal to 1.0.

f %, september 3, 1985 SEQUOYAH - UNIT 2 3/4 1-21 Amendment No., ;33 up; { .b;bh -.a ; ir m a l M ab bisdC.I a

Am 6 4 .R_ 5 N 1, O e I 2 2 6 2 6 a b c d b 4 s T 1 1 1 1 1 B 8 8 8 8 1 CA 5 5 d d d d d n n n n n E a a a a a L d BS n AE 2 2 4 2 4 a 2 2 CD IO LM 1 1 3 1 3 2 1 1 1 1 1 i P P A SE MLL UEB MNA INR NAE 2 2 1 2 1 2 3 3 3 2 2 3 N IHP O MCO I TA T ) N d E S e M LP u U EI n .R NR i T NT 1 1 0 1 0 1 2 2 2 1 1 a t S A n N HO o I CT C ( M E 1 T S 3 YS S 3 .L P OE E I NN i L R N B T LA 2 2 2 2 2 2 4 4 4 2 2 4 A AH T R TC O O T TF C O AE R n n o o i i s t t k a a c r r r o e e e l b p p r m O O e n n n a t e o o o h t r r n g6 r r r C 4 o" u se e I n-t t t p rw cw aP u u u e P rd n eo io mR e e e s3 I kP gP e N N N l1 Ve a o tex u-p / n ed Ld st u e e e pP i o rn n yal g g g0 m r e y i Ba pa SiF n7 n8 n1 I T yA t n i n d a-a- a-e o c ppw rpw pen RP RP RP er r A e iuo Tuo imo nu o T j rtd td rrr r, r, r, is t I n Trt crt Tet ex ex ex bs c N IF au iau t u wu wu wu re a ?o% w U S rth tth rne ol ol ol ur e yE oSS ASS oIN PF PF PF TP R L t t m t A em c o c N f o a t a O ar e.. u.. e. [ f. I Sf RAB AAB RA B C D E TC N U 9 0 1 2 F 1 2 2 2 ,. [ - G5 m@Sg, c5' N ,s +g y'

TABLE 3.3-1(Continued), ACTION 8 - With less than the Minimum Number of Channels OPERABLE,. declare ~ the interlock inoperable and verify that all affected channels. of the functions listed below are OPERABLE or apply the appro-q priate ACTION statement (s) for those functions. Functions to-t be evaluated are: Source Range Reactor Trip. a. b. Reactor Trip Low Reactor Coolant Loop Flow (2 loops) Undervoltage Underfrequency Turbine Trip Pressurizer Low Pressure Pressurizer High Level c. Reactor Trip Low Reactor Coolant Loop Flow (1 loop) d. Reactor Trip Intermediate Range Low Power Range Source Range ACTION 9 - Dele g ACTION 10 - Deleted L//a Dr ACTION 11 - Deleted ACTION 12 - With the number of OPERABLE channels one less than required the Minimum Channels OPERABLE re STANDBY'within 6 hours; however,quirement, be in at least HOT one channel may be bypassed for I up to 2 hours for surveillance testing per Specification 4.3.1.1.1 'provided the other channel is OPERABLE. i V March 16, 1987 SEQUOYAH - UNIT 2 3/4 3-7 Amendment No. 46

TABLE 4.3-1 (Continued) NOTATION'

w:

With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal. (1) If not performed in previous 7 days. (2) Heat balance only, above 15% of RATED THERMAL POWER. if absolute difference greater than 2 percent. . Adjust channel l A)(ZAL /2tAK OWEAfhts (3}' - f \\ .Comoare incore to excore ex.i:1 ' lux differ:nce above 15% of RATE THERMAL POWER. Recalibrate if the absolute difference greater than I or equal to 3 percent. (4) Manual ESF functional input check every 18 months. Each train or logic channel shall be tested at least every 62 days (5) on a STAGGERED TEST BASIS. The test shall independently verify the cperchil"ky of the undervoltage and automatic shunt-trip R46I g circuits. OPEAADILZTy I (6) Neutron detectors may be excluded from CHANNEL CALIBRATION. (7) Below P-6 (Block of Source Range Reactor Trip) setpoint. (8) Logic only, each startup or when required with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal if not performed in previous 92 days. O M W EL /~pAMTZO4'AL TE.fr (9) - The chann:1 functi:n t :t shall independently verify the f operability of the undervoltage and shunt trip circuits for the manual reactor trip function. (10) - Local manual shunt trip prior to placing breaker in service. Each train shall be tested at least every 62 days on a R46 -t:ggered test b fi:r JTAGSKgGD 7*gfr DAfts (11) - Automatic and manual undervoltage trip. i SEQUDYAH - UNIT 2 March 16, 1967 3/4 3. 3' Amencment. No. 46 1 a

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e CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) 3. Valves pressurized with fluid from a seal system. e. The combined bypass leakage rate shall be determined to be less than or g equal to 0.25 L by applicable Type B and C tests at least once a per 24 months except for pene,trations which are not individually testable; penetrations not individually testable'shall be determined: to have no detectable leakage when tested with soap bubbles while ~ the containment is pressurized to P,,12 psig, during each Type A test. f. Air locks shall be tested and demonstrated CPERABLE per Surveillance Requirement 4.6.1.3. g. Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,. Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P,, 13.2 f psig, and the seal system capacity is adequate to maintain system pressure (or fluid head for the containment spray system valves at penetrations 48A and 48B) for at least 30-days. h. Type B tests for penetrations employing a continuous leakage monitoring system shall be conducted at P,,12 psig, at-intervals no greater than once per 3 years. i. All test leakage rates shall be calculated using observed data converted to absolute values. Error analyses shall be performed to select a balanced integrated leakage measurement system. j. The provisions of Specification 4.0.2 are not' applicable. q m> SEQUOYAH - UNIT 2 3/4 6-4 0 ' ti e ' q ?,,s : ' 9,hi el Y$1dNP d3En$ 'p, aufGgg. a, mad 24.idMMg ' ];1g u 4 G

a ~ 4 4 ~' CONTAINMENT SYSTEMS .-,o wg - ??%' SURVEILLANCE RE0VIREMENTS (Continued) " ' ~ ' 4.6.3.2 Each isolation valve specified in Table 3.6-2 shall be demonstrated OPERABLE during the COLO SHUT 00WN or REFUELING MODE at least once per 18 months by: a. Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuat,es to its isolation position, b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position. J c. Verifying that on a Containment Ventilation isolation test signal, each Containment Ventilation valve actuates to its' isolation position. 4 4.6.3.3 The isolation time of each power operated or automatic valve of Table 3.6-2 shall be determined to be within its limit when tested pursuant to Specification 4.0.5. 4.6.3.4 Each containment purge isolation valve shall be demonstrated OPERABLEE ) within 24 hours after each closing of the valve, except when the valve is being used for multiple cyclings, then at least once per 72 hours, by verifying i that when the measured leakage rate is added to the leakage rates determined i pursuant to Specification ' S.!.2d. for all other Type 8 and C penetrations, the J comoined leakage rate is less than or equal to 0.60 L 'a u.s. o. a. s = 9* 1 SEQUOYAH - UNIT 2 3/4 6-18

1 4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES l LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-3. APPLICABILITY: MODES 1, 2 and 3. ACTION: ~ With 4 reactor coolant loops and associated steam generators in a. operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. e n b. With 3 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves t associated with an operating loop inoperable, operation in MCDEE -t, MODE j fksped 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-2; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. The provisions of Specification 3.0.4 are not applicable. c. O SURVEILLANCE REOUIREMENTS 4.7.1.1 No additional Surveillance Requiremen,ts other than those required by Specification 4.0.5. J SEQUOYAH - UNIT 2 3/4 7-1 .__._m_

i PLANT SYSTEMS %;W -- SURVEILLANCE REQUIREMENTS (Continued) ~ 3/ A i 3 wl$5 $+ /]} /ead. ome pe< b f

3. A each automatic control valve in the flow path is OPERABLE whenever the auxiliary feedwater system is placed in automatic control or when above 10% of RATED THERMAL POWER.

b. At least once per 18 months during shutdown by: 1. Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an auxiliary feedwater actuation test signal and a low auxiliary feedwater pump suction. pressure test signal. q 2. Verifying that each auxiliary feedwater pump starts as designed 1 automatically upon receipt of each auxiliary feedwater actuation test signal. At least once per 7 days by verifying that each non-automatic valve c. in the auxiliary feedwater system flowpath is in its correct position. c j l 1 ~a i SEQUOYAH - UNIT ?. 3/4 7-6

l i PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES A O_@ n, LIMITING CONDITION FOR OPERATION _2

3. 7.1. 5 Each main steam line isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3. ACTION: MODES 1 - With one main steam line isolation valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours; Otherwise, be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. MODES 2 - With one main steam line isolation valve inoperable, subseq'uent I and 3 operationinMODESjgg 2 or 3 may proceed provided: The isolation valve is maintained closed. a. b. The provisions of Specification 3.0.4 are not applicable. Otherwise, be in HOT STANDBY within the next 6 hours and in HOT '.4 SHUTDOWN within the following 6 hours. 'syr SURVEILLANCE REQUIREMENTS } 4.7.1.5 Each main steam line isolation valve shall'be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5. I ~ 1 se . s,>- SEQUOYAH - UNIT 2 3/4 7-10 N .i s i

1 0 PLANT SYSTEMS b SURVEILLANCE RE0VIREMENTS (Continued) u... g. Functional Test Failure - Attached Component _ Analysis For snubbers (s) found inoperable, an engineering evaluation shall be performed on the components which are restrained by the snut'ber(s). The purpose of this engineering evaluation shall be to determine if the components restrained by the snubber (s) were adversely affected by { the inoperability of the snubbers (s), and in order to ensure that the restrained component remains capable of meeting the designed service. h. Functional Testing of Repaired and Spare Snubbers Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the funtional test criteria before installation in the unit. These snubbers shall have met the acceptance criteria subsequent to their most recent service, and the functional test must have been performed within 12 months before being installed in the unit. i. Snubber Service Life Program The seal service life of hydraulic snubbers shall be monitored to ensure that the seals do not fail between surveillance inspections. The maximum expected service life for the various seals, seal materials, and applications shall be estimated based on engineering information, and the seals shall be replaced so that the maximum expected service life does not expire during a period when the r snubber is required to be(.pgerabvTD The seal replacements shall be documented and the documentatioli shall'be retained in accordance wth 6.10.2.n. gg Mechanical snubber drag force increases greater than 50 percent of previously measured values shall be evaluated as an indication of - impending failure of the snubber..These evaluations and any associated corrective action, shall be documented, and the documentation shall be retained in accordance with 6.10.2.n. j. _Exemotion From Visual Inspection or Functional Tests Permanent or other exemptions form the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and if applicable snubber life destructive testing was performed to qualify snubber operability for the applicable design conditions at either the completion of their fabrication or at a subsequent date. R3 June 20, 1985 SEQUOYAH - UNIT 2 3/4 7-25 Amendment No. 31 .+ g Mh e' (1 v ,) 4 4

4 Table 3.7-5 (Continued) FIRE HOSE STATIONS LOCATION ELEVATION HOSE RACK # Control Building 685 0-26-1188 Control Building .685 0-26-1193 Control Building 669 - 0-26-1189 Control Building 669 0-26-1194 d. . Diesel Generator Building Corridor 722 0-26-1077 Corridor 740.5 0-26-1080 Air Exhaust Rm. 740.5 0-26-1082 e. Additional Equipment Building - Unit 2 North Wall 740.5 2-26-687 l North Wall 706 2-26-686 f. Auxiliary Building 759 2-26-669 i s 749 2-26-644'664 cP' 749 1-26-664 734 2-26-670 l 734 0-26-684 734 1-26-670 734 0-26-682 734 Siamese Outlet 2-26-671 734 2-26-672 1 734 2-26-665 714 0-26-660 714 2-26-666 ~ 714 0-26-677 ~ 706 0-26-658 690 0-26-690 690 0-26-661-690 Siamese Outlet 2-26-674 690 2-26-675 669 2-26-667 669 2-26-668 669 0-26-662 669 0-26-680 653 0-26-663 653 0-26-691 l SEQUOYAH - UNIT 2 ' 3/4 7-50 ~ $ Ti/

1 s ,8 "32;g) 3/4.9 REFUELING OPERATIONS ,l 3/4.9.1 BORON CONCENTRATION j LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that th,e more restrictive of the following reactivity conditions is met: Either a K,ff of 0.95 or less,.which includes a 1% delta k/k a. q conservative allowance for uncertainties, or b. A boron concentration of greater than or equal to 2000 ppm, which' includes a 50 ppm conservative allowance for uncertainties. APPLICABILITY: MODE 6* ACTION: y With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until K is reduced to less than or equal to 0.95 or the baron eff concentration is restored to greater than or equal to 2000 ppm, whichever is the more restrictive. The provisions of Specification 3.0.3 are not applicable. ' SURVEILLANCE RE0VIREMENTS

4. 9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

Removing or unbolting the reactor vessel head, and a. b. Withdrawal of any full length control rod in excess of'3 feet from its fully inserted position within the reactor pressure vessel.

The reactor shall be maintained in MODE 6 whenever fuel is in the reactor 73

.y ;) vessel-heed with the vessel head closure bolts less than fully tensioned or with the head removed. SEQUOYAH - UNIT 2 3/4 9-1 l

i{i t 4 i lI 4 3 4 4 3 R 3 3 a g g ) fD oL( >y t() i~ l mn m 4 8 i o / 4 8 00 4 8 r Li i 8 11 t C 0 0 0 3 t. 8 rc 1 1 ee (p 0 0 1 1 xx 1 1 0 0 0 x x x wt 1 1 1 11 1 1 x x x x x 1 1 1 oe LD 1 1 1 M ARG O R P 9 9 3 9 s S s r s s 9 I r r r s Vt e e S e e r t Y t t t e t L t t t i A s m m m m i i i i t N i E E E E m A s y a E D a a a m m m m l N a a) A f n a mr a s oA G G a a G G ae N ey l Gh a I l pt a a t h l l L yi p a a lO p M 'T v p p p a P i i i l 2A c i i p c c c i1 A t n n n n 1 S c i c3 s 1 i 3 i3 i A r r r-r 1 3 3 n1 s 1 G N P P H PH P H rI r i - o 4 I 1 R P( G E O L T c TO B I A N M y k eg e e sc n r E mi n a u t y t T use PT 1P D d a d u iu a a ea l S myu D l tl A il q h h h cl Woe Wce Msce W nae c c c rp tp pt p il oil inr a a Ja am rm mrm S MAF E E DE M ha aa oaa UO E CS PS CPS SA e G y k ge l I I gc n r p s s s EV nn a u ym u u u i e T e P e DS e ur ur u r aa o o o I l u P l i l T pq hbp Phbp hb bp il il i l l ne-ne n e C me cam cam ca A ar ara 0 SF EGS arajar am tp tp t p ra nm nm n m EGSDEG MGS oa oa o a I DA CS~ CS' C S R_ gn g e s i g in e g u d n g d g p a u l i n i l n y r d c i d i t.i i T o n a u l 1J t a V B i r u d e S t a B l p Bu { P s ne s r i' a s u y u e eg t e a mr n g d r B s s tatet t n at t l G nu e a' n s1 scsds a' is ds e iP V Gme u1 uiul u sl u l u R e a ud a1 avaea eeia ea tk t ei n h x h r hih nt xh ih . s sn n l t o xuxexh x iaux hx u aa o. biC. E A.E S.E S.E dlA.E S. E o WT C1 2 N T1 2 3 4 Ic1 2 or e ou sa G A B g C D MSS $'EUN E w} h* k=kg 2P y aB2Q ? GR

l RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION (Continued) ~ms ~ r 6 4 /. 7 deleted from the monitoring program. Inlieuofdalicensee~ event of the unavailability of samples and identify the new lo obtaining replacement samples in the Annual Radiological Environmen-tal Operating Report. A revised figure (s) and table (s) for the ODCM reflecting the new location (s) shall be included in the next maaue! radic!c-ical eff"!cn+t release report pursuant to Specification 6.9.1.9. r'a2icod?ve a fffuen d. The provisions of Specifications 3.0.3 and 3.0.4'are not applicable. Jem o'-a.,na / SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the locations given in the table and figure in the ODCH and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1. e ~. i i January 14, 1986 SEQUOYAH - UNIT 2 3/4 12-2 Amendment No. 34'

i RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION ~ - i 3.12.2 A Land Use Census shall be conducted and shall identify within a dis-tance of 8 km (5 miles) the location in each of the 16 meteorological sectors the nearest milk animal, the nearest residence and the nearest garden

of greater than 50 m2 (500 ftz) producing fresh leafy vegetation.

i APPLICABILITY: At all times. ACTION: With a Land Use Census identifying a location (s) which yields a calcu-a. lated dose or dose commitment 20% greater than the values currently being calculated in Specification 4.11.2.3, identify the new locations (s) in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification E 9.1.7. 6.9 b. With a Land Use Cens. /.9 us identifying a location (s) which yields a calcu-13l lated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, add the new loca-tion (s) within 30 days to f.he Radiological Environmental Monitoring Program given in the ODCM, if samples are available. The sampling i location, excluding the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. Pursuant to Speci-fication 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflecting the new loca-tion (s) with information supporting the change in sampling locations. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. c. SURVEILLANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during the growing season ~ at-least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, mail survey, telephone survey, aerial. g; survey, or by consulting local agriculture authorities. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating-Report pursuant to Specification -E.0.1.0

4. 9, /. 7

Broao leaf vegetation sampling of at least three different kinds of vegeta-tion may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census.

fications for broad leaf vegetation sampling in Table 3.12-1.4c shall beSpeci-followed, including analysis of control samples. Januaryi 14, 1986 SEQUOYAH - UNIT 2 3/4 12-9 Amendment No. 34 336 t

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i l ENCLOSURE 2 ~ PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-87-17) DESCRIPTION AND JUSTIFICATION FOR CORRECTION OF MINOR DISCREPANCIES AND TYPOGRAPHICAL ERRORS I i

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\\ ENCLOSURE 2 Description of change Changes That Affect Unit I and Unit 2 Item page Description 1. I correct alphabetical listing of definition II section index. 2. 2-7 Add Laplace variable to the las compensation and correct typographical error in definition of ti. 3. 2-9 Correct typographical errors in definition of K. 4 4. B2-1 Delete references to nonexistent figure. 5. 3/4 1-14 Delete references to nonexistent figure. 6. 3/4 1-21 Delete references to nonexistent figure. 7. 3/4 3-13 Correct typographical errors in table 4.3-1 notation. 8. 3/4 3-56 (U1) Correctly identify the required number of 3/4 3-57 (U2) channels and minimum number of channels operable for auxiliary feedwater flow in the list of accident monitoring instrumentation to be consistent with plant design. 9. 3/4 7-1 Correct an inconsistency between action b. and technical specification 3.4.1.1 for operation with less than four reactor coolant pumpc running. 10. 3/4 7-6 Correct an error from a previous amendment that i inadvertently omitted the surveillance frequency for verification of control valve operability. 11, 3/4 7-10 Correct an inconsistency within the action statement for modes 1 and 3. 12. 3/4 12-1 (U1 only) Correctly identify radiological reporting 3/4 12-2 requirements of specification 3.12.1.c. 13. 3/4 12-10 (U1) Correctly identify radiological reporting 3/4 12-9 (U2) requirements of specifications 3.12.2 action a. and 4.12.2.

. Changes That Affect Unit 1 and Unit 2 (continued) Item Page Description 14. 5-2 Add the location of the meteorological t'ower to figure 5.1-1. 15. 5-6 Correct the hydrostatic test pressures for the reactor coolant system and secondary side in table 5.7-1. Changes That Affect Unit 1 Only 16. B2-1 Correc't typographical errors in FN terms. AH 17. 3/4 3-73 Correct typographical error in page heading, j i 18. 3/4 7-37 Correct an incorrect reference in the limiting condition for operation. 19. 3/4 8-4 Correct an inconsistency in surveillance requirement 4.8.1.1.2.d.6.b and the corresponding requirement in the unit 2 technical specifications and NRC Standard Technical Specifications (STS). 20. 3/4 8-5 Correct an inconsistency in surveillance requirement 4.8.1.1.2.d.7 and the corresponding requirement in the unit 2 technical specifications and NRC STS. 21. 3/4 11-12 Correct an inconsistency between table 4.11-2, note s., and the corresponding requirement in the unit 2 technical specifications and NRC STS. 22. B3/4 6-3 Correct a typographical error in item 3/4.6.1.8. Changes That Affect Unit 2 Only 23. 3/4 3-4 Correct an inconsistency between table 3.3-1,_ item 22, and the corresponding item in the unit I technical specifications and NRC STS. 24, 3/4 3-7 Correct an inconsistency between table 3.3-1, action 8, and the corresponding item in the unit 1 technical specifications as a result of change number 18. 25. 3/4 6-4 Correct a typographical error in item e. 26. 3/4 6-18 Correct a typographical error in specification 4.6.3.4

4

Chanres That Affect Unit 2 Only (continued)

Itea Page Description 27, 3/4 7-25 Correct a typograr.hical error in item i. 28. 3/4 7.50 Correct a typographical error in one valve number. i 1 29. 3/4 9-1 Correct a typographical error in the footnote. I 30. 3/4 11-9 Correct inconsistencies in table 4.11-2 and the i corresponding requirement in the unit 1 technical specifications and NRC STS. i 31. 3/4 12-2 Correct typographical error in specification j 3.12.1.c. I

1 C Reason for Change _m 1 Twelve changes correct typographical errors that were found during a review of the technical specifications. One change corrects the alphabetical order of an index section. Five changes correct references to figures or the figure itself. The reasons for the changes are simply to correct those errors. The remaining changes are explained separately below. Change No. 8 is required to correctly identify the location and number of auxiliary feedwater flow rate instrument channels. The instrument loops are shown on TVA diagram 47W610-3-3 (Final Safety Analysis Report [FSAR) figure 10.4.7-7). The flow elements are located in the common portion of the auxiliary feedwater piping such that they measure the injection flow from both the motor-driven and turbine-driven pumps. The flow rate indicators are designated FI-3-163A, FI-3-155A, FI-3-147A, and FI-3-170A for steam generators 1 through 4, respectively. Only the turbine-driven pump has its own flow indicator, FI-3-142A. The technical specifications are being revised to conform to plant design and NRC STS. Change No. 9 corrects an inconsistency between action b for specification 3.7.1.1 and specification 3.4.1.1. Specification 3.4.1.1 requires four reactor coolant pumps (RCPs) to be in operation in modes 1 and 2. Action b currently allows operation in modes 1 and 2 with three RCPs. This change to action b would only permit operation with less than four RCPs in mode 3. In this manner, specifications 3.4.1.1, 3.4.1.2, and 3.7.1.1 will be consistent. Change No. 10 adds the surveillance frequency to the verification of operability for the auxiliary feedwater control valves. The frequency statement was inadvertently omitted by a previous amendment. The valves have always been verified operable on this 31-day interval. Change No. 11 corrects an inconsistency in the mode 2 and 3 action for. specification 3.7.1.5. Operation in modes 2 and 3 is permitted with one main steam isolation valve (MSIV) inoperable. However, extended operation in mode 1 with an inoperable MSIV is not permitted by the action for mode 1. Change No. 12 corrects the administrative section radiological reporting requirements referenced in specification 3.12.1.c. The change corrects the references to annual and semiannual reports as described in section 6 of the specifications. The proper reporting requirements have been followed to date. i Change No. 13 in the same manner corrects the radiological reporting requirements referenced in specifications 3.12.2, action a., and 4.12.2. Change No. 15 corrects the hydrostatic test pressures for the reactor coolant ~ system and secondary side. The reactor coolant system design pressure is 2,485 psig. The steam generator design pressure is 1,085 psig. The I hydrostatic test pressures of 125 percent of design pressure are 3,107 psig and 1,356 psig, respectively. The reactor coolant system design parameters are listed in FSAR table 5.4.2-1. The steam generator design data is listed in FSAR table 5.5.2-1. Related stress analysis data is listed in FSAR table 5.2.1-20. I l l

4 e Change No. 19 corrects an inconsistency between the unit 1 and unit 2 surveillance requirements for diesel generator testing. It also corrects an j inconsistency with the same requirement in the NRC STS. The NRC STS and the ~ unit 2 specification require verification of a diesel start on the auto-start signal. The unit i requirements incorrectly require the start to be from ambient conditions. This change will make the unit I requirement consistent j with both unit 2 and the NRC STS. The test has been performed under ambient conditions as currently required by the specification. Change No. 20 corrects an inconsistency between the unit 1 and unit 2 i surveillance requirements for diesel generator testing. It also corrects an inconsistency with the same requirement in the NRC STS. The NRC STS and the unit 2 specifications require that a hot start be performed within five j minutes of completing the 24-hour test. The unit I requirements incorrectly require another verification of the load shedding. This change will make the unit I requirement consistent with both unit 2 and the NRC STS. The additional load shedding test has been performed as required under the current specification. Change No. 21 corrects an inconsistency between unit 1 and both unit 2 and the NRC STS. The minus sign was omitted from the exponential function in the equation for the lower limit of detection. This change will make the unit 1 equation consistent with both unit 2 and the NRC STS. Change No. 23 corrects an inconsistency between the unit 2 requirements and both unit I and NRC STS requirements for reactor trip system interlocks. The Power Range Neutron Flux P-9 interlock was omitted from table 3.3-1. This change will make the unit 2 table consistent with both unit 1 and the NRC STS. Change No. 24 corrects an inconsistency between the unit 2 action statements for table 3.3-1 and those for unit 1. The appropriate action is added for the addition of the P-9 interlock (change number 18). This change will make the unit 2 action statements consistent with those of unit 1. Change No. 30 corrects inconsistencies between the unit 2 requirements and both unit 1 and the NRC STS for the activity analysis requirements for contaittment vent. The wording is in the singular tense, and the footnota defining principal gamma emitters was omitted from the unit 2 specifications. This change will make unit 2 consistent with both unit 1 and the NRC STS. Justification for Channe Twelve changes correct typographical errors. One change corrects the alphabetical listing of an index section. Five changes correct references to figures or the figure itself. Six changes correct inconsistencies between the technical specifications for one unit and the other. In each case, the proposed changes conform with the NRC STS. Correction of these inconsistencies will eliminato confusion over applicable requirements and eliminate the potential for error.

, Four changes correct inconsistencies between action statements and other requiremento in the technical specifications. Correction of these ~. . ~ inconsistencies will eliminate confusion over applicable requirements and eliminate the potential for error. Two changes correct discrepancies between the technical specifications and the design of the plant. Correction of these discrepancies will eliminate confusion over applicable requirements and eliminate the potential for error. One change corrects an error from a previous amendment. Correction of the error will properly identify the correct surveillance intervals for a surveillance requirement. e i 1 1

6 I e i 1 ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGES SEQUOYAH' NUCLEAR PLANT UNITS 1 AND 2 ^ DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-87-17) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS J FOR CORRECTION OF MINOR DISCREPANCIES AND TYFCGRAPHICAL ERRORS i l i i 1 1 + e' I e . j

t* o ENCLOSURE 3 . ;=, ) SIGNIFICANT HAZARDS CONSIDERATIONS c. 1. Is the probability of an occurrence or the consequences of an accident previously evaluated in the safety analysis report significantly 1 increased? No. These changes are to correct typographical errors, inconsistencies between requirements, discrepancies between plant design and requirements, and to remedy an error from a previous amendment. Correcting these problems will eliminate confusion over applicable requirements and eliminate the potential for error. Eliminating the potential for error will reduce the probability of an occurrence. These l changes have no effect on the consequences of an accident previously evaluated. 2. Is the possibility for an accident of a new or different type than evaluated previously in the safety analysis report created? No. These changes are to correct typographical errors, inconsistencies between requirements, discrepancies between plant design and requirements, and to remedy an error from a previous amendment. No hardware changes were made to the plant. Correcting the inconsistencies between certain action statements and other requirements in the technical specifications will eliminate confusion over applicable requirements and eliminate the potential for error. In this case, the error would be to operate in a plant configuration not previously analyzed. Correcting these inconsistencies should eliminate the potential for this type of error. 3. Is the margin of safety significantly reduced? No. These changes are to correct typographical errors, inconsistencies between requirements, discrepancies between plant design and requirements, and to remedy an error from a previous amendment. Correcting these problems will eliminate confusion over applicable. ~ requirements and eliminate the potential for error. The nargin of safety will be increased with the elimination of the potential for error. _}}

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