ML20083C009

Proposed Tech Specs,Incorporating Proposed Revs Per GL 93-05 to Specs 2.3,3.1,3.2,3.3 & 3.6

ML20083C009
Person / Time
Site:	Fort Calhoun
Issue date:	05/08/1995
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20083C001	List: LIC-95-0100, Forwards Amend to License DPR-40,revising TSs 2.3,3.1,3.2, 3.3 & 3.6 of Plant TS ML20083C002 ML20083C009
References
GL-93-05, GL-93-5, NUDOCS 9505150066
Download: ML20083C009 (22)

Text

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-2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency core cooling system (continued)

.(2)

Modification of Minimum Requirements 4

.During power operation, the. Minimum Requirements may be modi-fied to allow one of the following conditions to be true at any one time.

If the system is not restored to meet the mini-mum requirements.vithin; the time period specified below, the

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reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the miMmum requirements are not met _vithin an addi -

tional h8 hours the reactor shall be placed in.a cold shutdown' condition within 24' hours.

a.

One low-pressure safety injection pump may be inoperable provided the pump is restored.to operable status within

'j-24 hours.

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b.

One high-pressure safety injection pump may be inoperable provided the. pump is restored to operable status within i

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l c.

One shutdown heat exchanger and two of four component -

cooling vater heat exchangers may be inoperabic for a period of no more than 2h hours.

. d.

Any valves, interlocks or piping directly ' associated with one of the above components and required to function -

during accident conditions shall be deemed to be part of that compongst and shall meet the same requirements as listed for that component.

e.

Any valve, interlock or piping associated with the safety injection and. shutdown cooling system which is not covered under d. above but which is required to function during accident conditions may be inoperable for a period of no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l f.

One safety injection tank may be inoperable for a period of n t

one hour.

AncUor g.

veland-pres e instrumentation on ne sa ety in tion 7

may b operable for a period f-eme. hour.

IA -

e

- Amendment No. k9' 2-21

.a TABLE 3-2 (Continued)

MINIMUM FREOUENCIES FOR CHECKS. CALIBRATIONS AND TESTING OF ENGINEERED SAFETY FEATURES. INSTRUMENTATION AND CONTROLS Channel Description Surveillance Function Freauency Surveillance Method Pump run to refill day tank.

12.

Diesel Fuel Trans-a.

Test M

a.

fer Pump Verify level indication between 13.

SIRW Tank F.ow a.

Check S

a.

Level Signal independent channels.

h.

Test Q

h.

A test pressure sirnulating the l

tank level is applied to each tank hubbler. one at a time.

c.

Calibrate R

c.

Known level signal applied to l

sensors and ST logic ve IEed

/ Verify that level and pressure are si&N limiIS.

14.

Safety injection a.

Check S

a.

Tank Level and Pres-

'ad;;;;;aa; ::: k wa.. la2+.-.

sure instruments in" h:;h ;ad lc,u a...

Tu.

t=:l :=d ; ::=::.

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3-10 Amendment No. 444A63-l l

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J TABLE 3 2 (Continued)

MINIMUM FPEOUENCIES FOR CHECKS. CALIBRATIONS AND TESTING OF ENGINEERED SAFETY FEATURES. INSTRUMENTATION AND COfRROLS Channel Description Eurveillance FunctN Freauency Surveillance Method 22.

Auxiliary Feedwater a.

Steam Generator a.

Check S

a.

Compare independent letti readmgs.

Water Level Low (Wide Range) h.

Calibrate R

h.

Known signal appikd to sensor.

l h.

Steam Generator a.

Check 5

m.

Compare independent pressure readmgs.

Pressure Low h.

Calibrate R

h.

Known signal applied to sensor.

l c.

Steam Generator a.

Calibrate R

a.

Known signal ar lied to sensor.

Differential Pressure liigh d.

Actuation Circuitry a.

Test Q

n.

Functional check of initiation circuits.

. l b.

Test R

h.

System functional test of AFW initiation circuits.

NOTES: (1)

Not required unless pressurizer pressure is above 1700 pria.

(2)

CRilS monitors are the containment atmosphere gaseous radiation monitor and the Auxiliary Building Exhaust Stack gaseous radiation monitor.

(3)

Not required unless steam generator pressure is above 600 psia.

(4)

QP - Quarterly during designated rmxles and prior to taking the reactor critical if not completed within the previous 92 days (not applicable _

l to a fast trip recovery).

Y 3 12a Amendment N(.4,55,!22443-G) Not re quire d 40 be don c on a S IT wi% inopec l,le.

level anha, y,,,,v,, 3,,3,,m 4,f7,,

a 1

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v.3-3 MINIMUM FREOUENCIES FOR CilECKS. CALIBRATIONS AND TESTING OF MISCELLANEOUS INSTRUMElWATION AND CONTROLS Surveillance Channel Description Function Freunency Surveillance Method I.

Primary CEA Position a.

Check S

a.

Comparison of output data with secondary CEAPIS.

Indication System b.

Test M

b.

Test of power dependent insertion limits, deviation, and sequence monitoring systems.

c.

Calibrate R

c.

Physically measured CEDM position used to venfy system accuracy. Calibrate CEA position interlocks.

2.

Secondary CE4 Position a.

Check 5

a.

Comparison of output data with primary CEAPIS.

Indication Sy tem b.

Test M

h.

Test of power dependent insertion limit, deviation, out-of-sequence, and overlap monitoring systems.

c.

Calibrate R

c.

Calibrate secondary CEA position indication system a

.,a. lu,% m>.

1...A:CH NNEL CHECK 3.

Area and Post-Accident a.

Check D

a.

.c - -... i red.c.d...L.. -test....:.

Radiation Monitors'"

ad to n.;', :.....;.u..: ;ia; :;c.

Q C H ANNE L FUNCTIONAL TEST b.

Test

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- Oc:cc:cr cy J : sc;..u:c iis..;J c 1... dcd n..se cr :c;; :dg..1 c.

Calibrate R

c.

Secondary and Electronic calibration performed at refueling frequency.

Primary calibration with exposure to radioactive sources only when required by the secondary and electronic calibration.

RM-091 A/B - Calibration by electronic signal substitution is acceptable for all range decades abovs 10 R/hr. Calibration for at least one decade belcw I-R/hr. shall be by means of calibrated radia6on source.

(" Post Accident Radiation Monitors are: RM-063 RM-064, and RM-09: A/B. Area Radiation Monitors are: RM-070 thru RM-082, RM-084 thru l

RM-089, and RM-095 thru RM-098.

3-13 Amendment No

' M,93,!37,!52, M4- -

TABl.E 3-3 (Contmucil)

MINIMUhtMEQllENMS FOR CillI!iSJAL!llMTI!!NSAND TESTING OF MisrEl.I.ANEOllS INSTRUMENTATION ANILCONTROI.S Suncillame Chaa-a.!wri_a_tiaa Fumlion

Frruvenc, Suntillance Method DELETED W~

4.

-Emergency-PleRed-C-.h' : :c

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T >.

i.:

':.::eg h!:-

v C. H A NN E L CHECK 5.

Primary to secondary s.

chak D

t h a._: : d aip t.neJ :24 h:..T.::cs:dgni leak. Rete Detection

-wed 44 *erify E:.... ;.: iia. :. a Radiation Monitors Q

Cli A NNF. L FuMC TI ON AL T E ST (RM-054AIB, RM-057) b.

Test

-M-h.

--Deteetere;und :: res.c:: 44. : 2.--? S d.ed as:c;er:=:d;c".

c.

Calibrate R

c.

Secondary and Electmaic calibration performed at erfueling frequency. Prismary Calibration perfonned with esposure to radioactive sources only when required by the secondary sad electronic calibration.

6.

Pressurizer level s.

Check S

s.

Comparison of independent level readings.

Instruments b.

Cahbrate R

b.

Known differential pressure applied to sensor.

c.

Test M

c.

Signal # > eierm meter relay adjuted with test device u verify setting.

7.

CEA Drive System a.

Test R

a.

Verify proper operation of all CEDM system Interlocks interlocks, using simulesed signals where necessary.

b.

Test P

b.

If haven't been checked for three months and plant is shutdown.

3-14 Amendment No. -l ___

T/

!3-5 MINIMUM FREOUENCIES FOR EOUIPMENT TESTS

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FSAR Secties:

Tc;g Frecuency Reference 1.

Control Element Drop times of all full-length CEA's Each refueling operation 7.5.3 Assemblies 2.

Control Element Partial movement of all CEA's Oa.y :.m uh Q

7 Assemblies (Minimum of 6 in) 3.

Pressurizer Safety Set Point Once each refueling outage 7

Valves 4.

Main Steam Safety

.:t Point Each refueling outage 4

Valves 5.

Refueling System Functioning Prior to refueling outage 9.5.6 Interlocks 6.

DELETED g

7.

DELETED 8.

Reactor Coolant Evaluate Daily

• 4 System Leakage 9.

Diesel Fuel Supply Fuel Inventory Daily 8.4 los.

Charcoal and IIEPA

1. In-Place Testine**

9.10 Filters for Control Charcoal adsorbers and IIEPA Each refueling shutdown not to exceed 18 Room filter banks shall be leak months or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system tested and show 2.99.95%

operation or after each complete or Froon (R-1I or R-112) and partial replacement of the charcoal cold DOP particulates adsorber/IIEPA fiker banka, or after removal, respectively, any major structural nunatenance on the system housing and following significant painting, fire or chern-ical releases in a ventilation some commumcating with the system.

• Whenever the system is at or above operating temperature and pressure.

• • Tests shall be performed in accordance with applicable section(s) of ANSI N510-1980.

3-20 Annendment No. !5,21,!2",!T,4%-

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3.0 SURVEILLANCE REQUIREMENTS 3.3; Reactor Coolant System and Other Components Subject _ to ASME XI l

Boiler & Pressure Vessel Code Inspection and Testing Surveillance T

(Continued)-

}g condition for refueling, e h ' time ty plan is placed in a f

cold shutdown condition fo 72 har:; if ing has not been accomplished in the precedt s, and prior to return--

ing the valve to service after maintenance, repair or replace-ment work is perfomed.

+

b.

Whenever the integrity of a pressure isolation valve listed

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in Table 2-9 cannot be demonstrated the integrity of the r

remaining valve in each high pressure line having a leaking valve shall be determined and recorded daily.

In addition, the position of one other valve located in the high pressure line shall be recorded daily.

5 Basis Undetected prolonged leakage of borated reactor coolant onto carbon steel sets up an unusual corrosion mechanism. Detection of this r

leakage at an early stage can best be accommodated directly after an outage and before startup. The inspection program specified in 4

Specification 3.3(1) places major emphasis on the areas of highest l

stress concentration as determined by general design evaluation and l

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experience with similar systems. The inspections will rely on non-1 destructive analysis methods utilizing up-to-date analyzing equipment and trained personnel. Volumetric inspection of the reactor pressure.

vessel is to be perfanned completely from the outside diameter.

The testing techniques and acceptance criteria of Section XI of the ASME B&PV Code will be utilized, except where specific relief is granted l

by the' Comission.

Reference (1) USAR, Section 4.5.3

{

n!:=jsn Amendment No, f,p;';// ;/3/yW 3-22 9. -Wt i

(Next page is 3-27) l

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l 3.0 SURVEILLANCE REOUIREMENTS 3.6 Safety Inlection and containment Cooling Systems Tests j

Anplimbility Applies to the safety injection system, the containnymt spray system, the conmin===*

cooling system and air filtration system inside the conmin===*

l Obiective j

To verify that the subject systems will respond y.e..iy3y and perform their inW functions, if required.

specifications (1)

Safety Inlection System i

System tests shall be performed on a refueling frequency. A test safety featme l

actuation signal will be applied to initiate operation of the system. De safety l

injection and shutdown cooling system pump motors may be de-energized for this j

portion of the test.

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A second overlapping test will be considered - O'-- L-j if control board l

indication and visual observations indicate all components have reasived the safety feature actuation signal in the proper sequence and timing (i.e., the j

appropriste pump breakers shall have opened and closed, and all valves shall have i

completed their travel).

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(2)

Containment Sorav Svetan a.

System tests shall be performed on a refueling fi%w.cy. De test shall l

be performed with the isolation valves in the spray supply lines at the containment blocked closed. Operation of the system is initiated by

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tripping the norniid on instrumentation.

j

(+<n b.

At least the spray nozzles shall be verified to be open.

c.

De test will be considered satisfactory if:

j (i)

Visual observations indicate that at least 264 nozzles per spray header have operated satisfactorily.

(ii)

No more than one nozzle per spray header is missing.

d.

Undisturbed samples of Trisodium Phosphate Dodecahydrate (TSP) that have been exposed to the same environmental conditions as that in the mesh baskets shall be tested on a refueling frqssacy by:

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3-54 Amendment No.v.,,:21, 4 W l

3.0 SURYEILLANCE REOUIREMENTS 3.6 " Safety Injection and Containment Cooling Systems Tests (Continued) that the sprey piping and nozzles are open will be madejnitia-11y by' 3

A single cortairment spray hea erJ1o^d)

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Ysmokertes<t or other suitably sensithe method, and 4 east rear 4ffrofMe WO five-years-hereafter.

1551] pag atomized spray is redred to provide the containment specified in Section 2.4 of the Technical Specification:

response To achieve the 3155 gpm flow rate, no greater that ten (10) spray nozzles may be inoperable of which no more than one may be missing.

Since the material is all stainless stasimpnally in-a, dry-conc 4 tion,

-the-ratest-every4ive with no plugging mechanism availabl,Are4esHg.+ afprofgnYeears ds_.2-considered to be more than adequate Nev 5 Other systems that are also important to are the SI tanks, the component cooling system, the raw water system and the contairment air coolers. The SI tanks are a passive safeguard.

In accordance with the specifications, the water volume and pressure in the SI tanks are checked periodically. The other systems mentioned operate when the reactor is in operation and are continuously monitored for satisfactory performance.

The in-containment air treatment system is designed to filter the containment building atmosphere during accident conditions. Both in-containment air treatment systems are casigned to automatically start upon accident signals. Should one system fail to start, the recundant system is des 15nen to s g automatically. Each of the two systems has 100 percent capacity l

High efficiency particulate air (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the todine adsorbers..

The charcoal adscrbers are installed to reduce the potential release of radiciodine to the environment. The laboratory carbon sample test results should indicate a raoicactive methyl iodide removal efficiency of at least 85 percent. If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR Part 100 guidelines for the accidents analyzed.

Pressure drop across the ccmbined HEPA filters and charcoal adscrbers of less than 6 inches of water will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter.

If significant painting, fire or chemical release occurs in a ventilation zone communicating with the system that could lead to the degradation of charcoal adsorbers or HEPA filters, testing will be performed to assure system integrity and performance.

3-57 Amendment No. E_121-

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U.S. Nuclear Regulatory Comission LIC-95-0100 ATTACHMENT B i

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N DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT HAZARDS CONSIDERATIONS DISCUSSION AND JUSTIFICATION:

In accordance with the guidance of Generic Letter (GL) 93-05 "Line Item Technical Specifications Improvcaients to Reduce Surveillance Requirements for Testing During Power Operation," dated September 27, 1993, Omaha Public Power District (OPPD) proposes to revise the Fort Calhoun Station (FCS) Unit No. 1 Technical Specifications (TS).

The proposed TS changes listed below are consistent with Station operating experience and NUREG-1366, " Improvements to Technical Specifications Surveillance Requirements," dated December 1992.

)

t GL 93-05. Item 4.2. Control Rod Novement Test: Specification 3.2. Table 3-5.

I Item 2 On March 1, 1995, OPPD submitted an application for amendment that proposed l

numerous revisions to the FCS Unit No. 1 TS.

In order to introduce consistent use of the notation defined in Specification 3.0.2, OPPD proposed to change i

the surveillance interval notation for Specification 3.2, Table 3-5, Item 2 i

(partial movement of all control element assemblies) from "every two weeks,"

)

to "BW" (biweekly). A review of operating experience was recently completed following the guidance of GL 93-05 and OPPD now proposes to extend this j

surveillance to a quarterly frequency.

Previous surveillance tests were reviewed and personnel familiar with the test were interviewed; no surveillance test failures were identified.

This test has resulted in reactor i

trips, dropped rods and unnecessary challenges to safety systems at other plants.

GL 93-05. Item 5.14. Radiation Monitors: Specification 3.1. Table 3-3. Items

3. 4 and 5 OPPD proposes to replace descriptive wording in Specification 3.1, Table 3-3, Items 3a/b (area and post-acci'.it radiation monitors) and Sa/b (primary to secondary leak-rate detection endiation monitors) with defined terms. OPPD also proposes to extend the su.veillance frequency of Specification 3.1, Table 3-3, Item 3b and Item 5b from monthly to qwrterly in accordance with the recommendation of GL 93-05, Item 5.14.

These monitors are new, having been installed within the last two operating cycles (except for RM-091 A/B). The i

value of monthly testing is greatly reduced as the new radiation monitors include self checking circuitry that will provide notification if a failure occurs. Although post accident radiation monitors RM-091 A/B are not new monitors, Station operating experience has shown that they are reliable.

In cases where new components interface with older components, the older components have a history of reliable operation.

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1 DISCUSSION AND JUSTIFICATION (Continued):

Replacing descriptive words with defined terms ensures consistency and that the surveillance accomplishes its purpose.

The proposed frequency (quarterly) is the same frequency currently specified for the containment radiation high signal (CRHS) monitors (Specification 3.1, Table 3-2, Item 6b), which generate an engineered safeguards signal. A quarterly surveillance frequency conserves resources and increases the availability of the area, post-accident and primary to secondary leak-rate detection radiation monitors.

OPPD also proposes to delete Table 3-3, Item 4 on surveillance testing of the emergency plan radiation instruments.

These are portable instruments stored in specified locations for use by emergency response personnel in the event of an accident.

These instruments may be used to survey onsite/offsite areas for i

radioactivity or to facilitate the decontamination of personnel following an accident. Although not specifically mentioned in GL 93-05, these instruments do not have a limiting condition for operation (LCO) action statement associated with them. As a result, there is no basis for the TS to contain a surveillance requirement on these instruments.

In addition, retaining this surveillance in the TS is unnecessary since it does not meet criteria 1 through 4 of the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993.

GL 93-05. Item 6.1. Reactor Coolant System Isolation Valves: Specification 3.3(21a Following the recommendation of GL 93-05, Item 6.1, OPPD proposes to revise Specification 3.3(2)a (surveillance of reactor coolant system (RCS) pressure isolation valves) based upon Station operating experience.

Currently, this surveillance test requires that the high pressure safety injection and low pressure safety injection loop isolation valves be leak tested each time the Station is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, if testing has not been accomplished in the preceding nine months. A review of previous surveillance tests and interviews with personnel familiar with the test did not identify any prior surveillance test failures. Therefore, OPPD proposes to extend the time that the plant can be in cold shutdown before this test is required from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days.

GL 93-05. Item 7.4. Accumulator Water Level and Pressure Channel Surveillance Reauirements: Specification 2.3(210. Specification 3.1. Table 3-2. Item 14a OPPD proposes two revisions to Specification 2.3(2)g.

OPPD proposes to clarify that safety injection tank (SIT) level a_nd/or pressure instrumentation may be inoperable; and OPPD proposes to extend the LCO action time from I hour to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Clarifying that SIT level and/or pressure instrumentation may be inoperable does not alter the intent of the Specification and it more accurately defines when the Specification applies.

Extending the time limit for inoperability of SIT instrumentation to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, following the recommendation of GL 93-05, Item 7.4 is justified based on a review of historical data.

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a DISCUSSION AND JUSTIFICATION (Continued):

As stated in NUREG-1366: "While technically inoperable, the accumulator [ SIT]

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would be available to fulfill its safety function during this time, and thus, this change would have a negligible increase on risk." Currently, Specification 2.3(2)g allows only one hour for SIT level and pressure instrumentation to be inoperable, which is insufficient time to initiate

)

repairs. A review of historical data found that SIT water level stays l

relatively constant while pressure decreases slightly over time.

It is unlikely that SIT pressure would decrease below the limit (240 psig) of Specification 2.3(1)c during the proposed 72-hour LCO.

SIT pressure is i

normally maintained around 255 psig (Updated Safety Analysis Report (USAR),

Section 6.2.3.5).

]

1 OPPD also proposes to revise Specification 3.1, Table 3-2, Item 14a to require shiftly verification that SIT level and pressure are within limits and remove reference to verifying " indications are between independent high and low alarms for level and pressure." This revision to Item 14a is consistent with the guidance of GL 93-05, Item 7.4, which recognizes that SIT instrumentation operability is not directly related to the capability of a SIT to perform its safety function. OPPD proposes to suspend this surveillance on the affected SIT while the instrumentation is being repaired. As stated above, SIT level and pressure are expected to stay within the limits of Specification 2.3(1)c during the proposed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO.

GL 93-05. Item 8.1. Containment Spray System: Specification 3.6(2)b This Specification currently requires the containment spray nozzles to be verified open at five year intervals. OPPD proposes to revise Specification 3.6(2)b and associated Specification 3.6 basis statements to extend this surveillance to a ten year frequency following the recommendation of GL 93-05, Item 8.1.

Personnel familiar with this surveillance test were interviewed and previous surveillance tests were reviewed; no prior nozzle clogging has been identified. The problem that caused several containment spray nozzles to become clogged at San Onofre Unit 1 is not a concern at FCS, since the FCS containment spray system piping and valves are constructed of stainless steel (USAR Table 6.3-2).

ADMINISTRATIVE CHANGES It is proposed to add Amendment 41 to the list of amendments revising Page 3-12a.

It is proposed to remove Amendment 14 and add Amendment 8 to the list of

)

amendments revising Page 3-13.

TS Change 14 revised Page 3-13 but this change was issued with Amendment 8.

It is proposed to remove Amendment 99 and revise " Order 4/20/87" to " Order 4/20/81," as items revising Page 3-22.

It is proposed to add " Change 7" to the list of items revising Page 3-54.

Prior to amendments the TS were revised by TS Changes.

Change 7 was issued February 23, 1974.

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I BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION:

i The proposed Technical Specification (TS) changes do not involve significant l

hazards considerations because operation of Fort Calhoun Station (FCS) Unit No. 1 in accordance with these changes would not:

(1)

Involve a significant increase in the probability or consequences of an accident previously evaluated.

l GL 93-05. Item 4.2. Control Rod Novement Test: SDecification 3.2. Table 3-5. Item 2 Omaha Public Power District (0 PPD) proposes to extend the control element assembly (CEA) partial movement surveillance test of l

Specification 3.2, Table 3-5, Item 2 from a biweekly to a quarterly l

frequency.

This change is based on operating experience and the recommendation of Generic letter (GL) 93-05, Item 4.2.1.

A review of previous surveillance tests and interviews with personnel familiar with the test did not identify any prior surveillance test failures.

Industry experience has shown that this test can cause reactor trips, dropped rods and unnecessary challenges to safety systems at stated in NUREG-1366, " Improvements to Technical Specification Requirements,"

i dated December 1992. Therefore, extending the frequency of conducting this surveillance test may be beneficial to plant operations and does not involve a significant increase in the probability or consequences of an accident previously evaluated.

GL 93-05. Item 5.14. Radiation Monitors: Specification 3.1. Table 3-3.

Items 3b. 4 and 5b OPPD proposes to replace descriptive wording in Specification 3.1, Table 3-3, Items 3a/b and Sa/b with defined terms. OPPD also proposes to extend surveillance of the area, post-accident and primary to secondary leak-rate radiation monitors (Specification 3.1, Table 3-3, Items 3b and 5b) from a monthly to a quarterly frequency as recounended by GL 93-05, Item 5.14.

Most of these monitors are new (i.e., installed within the last two cycles) or contain many new components.

The value of monthly testing is greatly reduced as the new monitors include self checking circuitry that will indicate monitor failure, loss of power, or loss of background. Although post accident radiation monitors RM-091 A/B are not new, Station operating experience has shown that they are reliable.

In cases where new components interface with older components, the older components have a history of reliable operation.

Readings and internal test signals are used to verify instrument operation on a daily basis.

In addition, the proposed frequency (quarterly) is Ge same frequency currently specified for the I

containment radiation high signal (CRHS) monitors (Specification 3.1, Table 3-2, Item 6b), which generate an engineered safeguards signal.

Replacing descriptive words with defined terms ensures consistency and that the surveillance test accomplishes its purpose.

4

. r BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

i i

A quarterly surveillance conserves resources, increases the availability of the area, post-accident and primary to secondary leak-rate detection

)

radiation monitors and is consistent with CRHS monitor testing.

These proposed changes do not involve a significant increase in the i

probability or consequences of an accident previously evaluated.

]

OPPD proposes to delete Specification 3.1, Table 3-3, Item 4 on l

surveillance testing of the emergency plan radiation instruments. These are portable instruments stored in specified locations for use by emergency response personnel in the event of an accident.

The instruments may be used to survey onsite/offsite areas for radioactivity or to facilitate the decontamination of personnel following an accident.

No limiting condition for operation (LCO) action statement is associated i

l with these instruments. As a result, there is no basis for the TS to i

contain a surveillance requirement for them.

In addition, re'aining this surveillance in the TS is unnecessary since it does not meet criteria 1 through 4 of the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, i

1993. Therefore, since these instruments are not utilized until after an accident has occurred, and do not assist in accident mitigation, 1

deleting this surveillance requirement does not involve a significant 1

increase in the probability or consequences of an accident previously i

evaluated.

GL 93-05. Item 6.1. Reactor Coolant System Isolation Valves:

Specification 3.3(2)a The reactor coolant system (RCS) pressure isolation valves have proven to be very reliable. Therefore, OPPD proposes to extend the time that the plant can be in cold shutdown before the test is required (Specification 3.3(2)a) from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days, following the recommendation of GL 93-05, Item 6.1.

A review of previous surveillance tests and interviews with personnel familiar with the test did not identify any prior surveillance test failures. This proposed change will reduce radiation exposure and does not involve a significant increase in the probability or consequences of an accident previously evaluated.

GL 93-05. Item 7.4. Accumulator Water Level and Pressure Channel Surveillance Reouirements: Specification 2.3(2)a. SDecification 3.1.

Table 3-2. Item 14a OPPD proposes to revise Specification 2.3(2)g following the recommendation of GL 93-05, Item 7.4.

This revision will clarify that the safety injection tank (SIT) level and/or pressure instrumentation may be inoperable, which does not alter the intent of the Specification, but is more accurate in defining when the Specification applies. This revision also extends the time limit for inoperability of SIT instrumentation from I hour to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, which is justified based upon a review of historical data.

As stated in NUREG-1366: "While technically 5

s>

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

inoperable, the accumulator (SIT] would be available to fulfill its safety function during this time, and thus, this change would have a negligible increase on risk."

Currently, Specification 2.3(2)g allows only one hour for SIT level and pressure instrumentation to be inoperable, which is insufficient time to initiate repairs. A review of historical data determined that SIT water level stays relatively constant while pressure decreases slightly over time.

It is unlikely that SIT pressure would decrease below the Specification 2.3(1)c limit of 240 psig during the proposed 72-hour LCO, since SIT pressure is normally maintained around 255 psig (Updated Safety Analysis Report (USAR), Section 6.2.3.5).

OPPD's proposal to revise Specification 3.1, Table 3-2, Item 14a to require shiftly verification that SIT level and pressure are within limits and remove reference to verifying " indications are between independent high and low alarms for level and pressure," is consistent with the gu @ nce of GL 93-05, Item 7.4.

As stated in GL 93-05, Item 7.4, the operability of SIT instrumentation is not directly related to the capability of a SIT to perform its safety function. OPPD proposes to suspend this surveillance on the affected SIT while the instrumentation is being repaired, since as stated above, SIT level and pressure are expected to stay within the limits of Specification 2.3(1)c during the proposed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO.

Therefore, these proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

GL 93-05. Item 8.1. Containment Soray System: Specification 3.6(2)b OPPD proposes to extend the surveillance frequency for verifying that the containment spray nozzles are open (Specification 3.6(2)b) from five to ten years following the recommendation of GL 93-05, Item 8.I.

Minor revisions to statements in the basis of Specification 3.6 that refer to conducting this test at five year intervals are proposed also. OPPD has not experienced problems with obstructions in the containment spray nozzles as determined by a review of previous surveillance tests and personnel interviews.

Of the three instances reported in NUREG-1366 concerning obstructions of containment spray nozzles, all were problems related to construction errors. Any construction errors in the FCS i

containment spray system would have been found by previous surveillance tests.

The problem that occurred at San Onofre Unit 1 (clogging of several containment spray nozzles following the application of a coating material to the carbon steel piping) is not a concern at FCS since the FCS containment spray system piping and valves are constructed of stainless steel (USAR Table 6.3-2).

Thus, extending the surveillance frequency of Specification 3.6(2)b from five to ten years does not involve a significant increase in the probability or consequences of an accident previously evaluated.

6

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BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

(2)

Create the possibility of a new or different kind of accident from any accident previously evaluated.

E 93-05. Item 4.2. Control Rod Movement Test: Specification 3.2. Table 3-5. Item 2 OPPD's proposal to extend the CEA partial movement surveillance test (Specification 3.2, Table 3-5, Item 2) to a quarterly frequency is based on operating experience and the recommendation of GL 93-05, Item 4.2.1.

The proposed change only lengthens the time between surveillance tests and will not result in any physical alterations to the plant configuration, changes to setpoint values, or changes to the application of setpoints or limits.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

GL 93-05. Item 5.14. Radiation Monitors: Specification 3.1. Table 3-3.

Items 3b. 4 and 5b OPPD proposes to replace unnecessary wording in Specification 3.1, Table 3-3, Items 3a/b and Sa/b with defined terms and to extend the surveillance frequency of Items 3b and 5b from monthly to quarterly j

based on the recommendation of GL 93-05, Item 5.14.

Most of the area, post accident and primary to secondary leak-rate detection radiation monitors are new or contain new components. The new monitors include self checking circuitry that provides failure notification. Although post accident radiation monitors RM-091 A/B are not new, they have an

)

excellent operating history. The proposed changes introduce consistent use of terminology and lengthen the time between surveillance tests and will not result in any physical alterations to the plant configuration, changes to setpoint values, or changes to the application of setpoints or limits.

Therefore, these proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

OPPD's proposal to delete Specification 3.1, Table 3-3, Item 4 on surveillance testing of the emergency plan radiation instruments will not result in any physical alterations to the plant configuration, changes to setpoint values, or changes to the application of setpoints or limits. Since these instruments are not utilized until after an accident has occurred, and do not assist in accident mitigation, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

GL 93-05. Item 6.1. Reactor Coolant System Isolation Valves:

Specification 3.3(2)a The RCS pressure isolation valves have proven to be very reliable.

As a result, OPPD proposes to extend the time that the plant can be in cold shutdown before the test is required (Specification 3.3(2)a) from 72 7

s,

a o a e r

. BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

hours to 7 days following the recommendation of GL 93-05, Item 6.1.

The proposed change will reduce radiation exposure and does not result in any physical alterations to the plant configuration, changes to setpoint values, or changes to the application of setpoints or limits.

Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

GL 93-05. Item 7.4. Accumulator Water Level and Pressure Channel Surveillance Recuirements: Specification 2.3f 2)c. SDecification 3.1.

Table 3-2. Item 14a OPPD's proposal to revise Specification 2.3(2)g following the guidance of GL 93-05, Item 7.4 more accurately states when the specification should apply and extends the time limit for inoperability of SIT i

instrumentation from I hour to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> based upon a review of l

historical data. The proposed change will not result in any physical alterations to the plant configuration, changes to setpoint values, or changes to the application of setpoints or limits. As stated in NUREG-1366: "While technically inoperable, the accumulator [ SIT] would be available to fulfill its safety function during this time, and thus, this change would have a negligible increase on risk."

0 PPD's proposal to revise Specification 3.1, Table 3-2, Item 14a to require shiftly verification that SIT level and pressure are within limits and remove reference to verifying " indications are between independent high and low alarms for level and pressure," is consistent with the guidance of GL 93-05, Item 7.4.

As stated in GL 93-05, Item 7.4, the operability of SIT instrumentation is not directly related to the capability of a SIT to perform its safety function.

OPPD proposes to suspend this surveillance on the affected SIT while the instrumentation is being repaired, since SIT level and pressure are expected to stay within the limits of Specification 2.3(1)c during the proposed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO. Therefore, since these proposed changes do not result in any physical alterations to the plant configuration, changes to setpoint values, or changes to the application of setpoints or limits, they do not create the possibility of.a new or different kind of accident from any accident previously evaluated.

GL 93-05. Item 8.1. Containment Sorav System: Specification 3.6(2)b OPPD's proposal to extend the surveillance frequency for verifying that the containment spray nozzles are open (Specification 3.6(2)b) from five to ten years as recommended by GL 93-05, Item 8.1 is justified by operating experience. OPPD has not experienced problems with obstructions in the containment spray nozzles as determined by a review of previous surveillance tests and personnel interviews. The problem that occurred at San Onofre Unit 1 (clogging of several containment spray nozzles following the application of a coating material to the carbon steel piping) is not a concern at FCS since the FCS containment spray system piping and valves are constructed of stainless steel (USAR Table 6.3-2).

8

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-*)

$ g BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

The proposed change only extends the time between surveillance tests and revises associated basis statements to support the extension. The proposed change will not result in any physical alterations to the plant configuration, changes to setpoint values, or changes to the application of setpoints or limits. Therefore, OPPD's proposal to extend the surveillance frequency of Specification 3.6(2)b from five to ten years does not create the possibility of a new or different kind of accident from any accident previously evaluated.

~

(3)

Involve a significant reduction in a margin of safety.

GL 93-05. Item 4.2. Control Rod Movement Test: SDecification 3.2. Table 3-5. Item 2 OPPD's proposal to extend the CEA partial movement surveillance test of Specification 3.2, Table 3-5, Item 2 to a quarterly frequency is based on operating experience and the recommendation of GL 93-05, Item 4.2.1.

A review of previous surveillance tests and interviews with personnel familiar with the test did not identify any prior surveillance test failures.

Industry experience has shown that this test can occasionally cause reactor trips, dropped rods and unnecessary challenges to safety i

systems as stated in NUREG-1366.

Therefore, extending the frequency of conducting this surveillance test may be beneficial to plant operations and does not involve a significant reduction in a margin of safety.

GL 93-05. Item 5.14. Radiation Monitors: SDecification 3.1. Table 3-3.

Items 3b. 4 and Sb OPPD proposes to replace descriptive wording in Specification 3.1, Table 3-3, Items 3a/b and Sa/b with defined terms and to extend the surveillance frequency of Items 3b and 5b from monthly to quarterly i

based on the recommendation of GL 93-05, Item 5.14.

Most of the area, post accident and primary to secondary leak-rate detection radiation monitors are new or contain new components.

Post accident radiation i

monitors RM-091 A/B are not new but have a history of reliable operation.

The value of monthly testing is greatly reduced since the

)

new monitors include self checking circuitry that provides failure notification.

The proposed changes introduce consistent use of terminology and lengthen the time between surveillance tests and therefore do not involve a significant reduction in a margin of safety.

OPPD's proposal to delete Specification 3.1, Table 3-3, Item 4 is justified because the emergency plan radiation instruments are portable instruments that are not utilized until after an accident has occurred.

The instruments are checked for proper operation before use and since these instruments do not assist in accident mitigation, the deletion of this surveillance requirement does not involve a significant reduction in a margin of safety.

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3t, 8 ASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

i GL 93-05. Item 6.1. Reactor Coolant System Isolation Valves:

Specification 3.3(2)a j

The RCS pressure isolation valves have proven to be very reliable.

Therefore, consistent with the guidance of GL 93-05, Item 6.1, OPPD proposes to revise Specification 3.3(2)a and extend the time that the j

plant is allowed to be in cold shutdown before this surveillance test is required from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. This change will reduce radiation-exposure and does not involve a significant reduction in a margin of i

safety.

GL 93-05. Item 7.4. Accumulator Water Level and Pressure Channel i

Surveillance Reauirements: Specification 2.3(2)a. SDecification 3.1.

Table 3-2. Item 14a j

OPPD's proposal to revise Specification 2.3(2)g following the guidance I

of GL 93-05, Item 7.4 more accurately states when the specification applies and extends the time limit for inoperability of SIT I

instrumentation from 1 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> based upon historical data. As l

stated in NUREG-1366: "While technically inoperable, the accumulator

[ SIT] would be available to fulfill its safety function during this time, and thus, this change would have a negligible increase on risk."

OPPD's proposal to revise Specification 3.1, Table 3-2, Item 14a to require shiftly verification that SIT level and pressure are within limits and remove reference to verifying " indications are between i

independent high and low alarms for level and pressure," is consistent with the guidance of GL 93-05, Item 7.4.

As stated in GL 93-05, Item i

7.4, the operability of SIT' instrumentation is not directly related to i

the capability of a SIT to perform its safety function.

OPPD proposes i

to suspend this surveillance on the affected SIT while the instrumentation is being repaired, since SIT level and pressure are expected to stay within the limits of Specification 2.3(1)c during the proposed 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO. Therefore, these proposed changes do not involve a significant reduction in a margin of safety.

GL 93-05. Item 8.1. Containment Spray System: Specification 3.6(21b OPPD's proposal to extend the surveillance frequency for verifying that the containment spray nozzles are open (Specification 3.6(2)b) from five to ten years as recommended by GL 93-05, Item 8.1 is justified by operating experience.

OPPD has not experienced problems with obstructions in the containment spray nozzles as determined by a review of previous surveillance tests and personnel interviews.

The problem that occurred at San Onofre Unit 1 is not a concern at FCS since the FCS containment spray system piping and valves are constructed of stainless steel (USAR Table 6.3-2).

Therefore, OPPD's proposal to extend the surveillance frequency of Specification 3.6(2)b from five to ten yoars and revise associated basis statements does not involve a significant reduction in a margin of safety.

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. Based on the above considerations, it is OPPD's position that this proposed i

t amendment does not involve significant hazards considerations as defined by 10

'CFR 50.92..The proposed changes will not result in a condition that j

significantly alters the impact of the Station on the environment.

Thus, the i

proposed changes meet the eligibility criteria for ' categorical exclusion set forth in 10 CFR 51.22(c)(9) and pursuant to 10 CFR 51.22(b) no environmental j

assessment need be prepared.

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