ML20138H861

Proposed Tech Spec Changes,Consisting of Change 1 to Raise Spent Fuel Pool Radiation Monitor Setpoint & Change 2 to Modify Power Range Flux Rate Trip Setpoint

ML20138H861
Person / Time
Site:	Sequoyah
Issue date:	10/22/1985
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20138H841	List: ML20138H835 ML20138H861
References
TVA-SQN-TS-65, NUDOCS 8510290166
Download: ML20138H861 (18)

Text

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, ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGES SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 (TVA SQN TS 65)

CHANGE NO. 1 SPENT FUEL POOL RADIATION MONITOR SETPOINT CHANCES s

B510290166 851022 .

PDR ADOCK 05000327 P PDR l

TABLE 3.3-6 .

RADIATION MONITORING INSTRUMENTATION j

E MINIMUM CHANNELS APPLICABLE ALARM / TRIP MEASUREMENT OPERABLE MODES SETPOINT RANGE ACTION g INSTRUMENT

1. AREA MONITORS

~ 2% 4 Fuel Storage Pool Area

• 10"I - 10 mR/hr 26 l[

a. 1

% mR/hr

b. Containment Area 1 1, 2, 3 and 4 N/A 1 - 107 R/hr*** 30

2. PROCESS MONITORS ,

~3 28

a. Containment Purge Air 1 1, 2, 3, 4 & 6 5 8.5 x 10 pCi/cc 10 - 10 cpm

b. Containment y i. Gaseous Activity

-3 pCi/cc 10 - 10 7cpm 28 w a) Ventilation Isolation 1 ALL MODES 1 8.5 x 10 b)RCS Leakage Detection 1 1, 2, 3 & 4 N/A 10 - 10 cpm 27 g

ii. Particulate Activity -5 7 a) Ventilation Isolation 1 ALL MODES -< l.5 x 10 pCi/cc 10 - 10 cpm 28 ,16 a

b)RCS Leakage Detection 1 1, 2, 3 & 4 N/A 10 - 10 cpm 77 7

29 yf

s ;D

c. Control Room isolation 1 ALL MODES $ 400 cpm ** 10 - 10 cpm g d. Noble Gas Effluent Monitors s,

ot 2 I'[ "With fuel in the storage pool or building M 5 g ** Equivalent to 1.0 x 10 pCi/cc

• • • Measurement range by extrapolation 9

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_ _ _ . _ _ _ . . _ _ . - _ . _ . . - . - _ = _ . _ _ _ . _ . _ _ . _ . . _ _ . _ . _ _ . _ . . . . _ _ _ . . _ . . ___ . _ . . _ . _ . . __ _ ._ _ _ .

v. 1AutE 3.3-6 m ~

~

'@. RA01Afl0N MON!TORlflG INSTRUMENTATION

.S-

.E' MINIMUM ,

r

. , , , CilANNELS APPLICA8tE ALARM /lRIP MEASUREMENT INSTRUMENT -OPERABLE 'HODES SEIP0lNI RANGE ACIl0N g

'a

1. . AREA MON 110RS

• ~I 4

.. a . fuel Storage.1....I Area /c m:t/hr - 26 N 1 $ 10 - 10 mR/hr ll j~ b. Containment As.. I 1, 2, 3 & 4 N/A 1- 10 R/hr*** 30 i: 2. -PROCESS MONITOR 5 -

^

-3 pCi/cc 10 - 107 cpm 1 a. Containment I...ye Air 1 1, 2, 3, 4 & 6 5 8.5 x 10 28 ,

b. Containment 6

i. Gaseous i..tivity _3 7 i a) Ventilation isolation 1 ALL MODES i 8.5 x 10 pCi/cc 10 - 107 cpm 28

b)RCS Leakage D election "; I .1, 2, 3 & 4 N/A 10 - 10 cpm 27 i i

.M -ii. -Particulate Activity

-5 7 P .a) Vent.il.ition isolation 1 ALL MODES $ 1.5 x 10 pCi/cc 10,- 10 cm 28

.i' -b)RCS leakage Detection 1 1, 2, 3 & 4 N/A 10 - 10 cpm 27 I 7 4

( c. Control Room Isolation  :

1 ALL MODES -< 400 cpm ** 10 - 10 cpm 29

d. ' Noble Gas Erfluent Monitors i ,

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^ With fuel in the storagg pool or building

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• • Equivalent to 1.0:x 10 pCi/cc.

>^^^ Measurement rouge by extrapolation.

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DESIGN FEATURES

5. 6 FUEL STORAGE CRITICALITY - SPENT FUEL

5. 6.1.1 The spent fuel storage racksbr Sel enru'chetl fo . v.o wej$rfettf47 O*23S with: are designed and shall be maintained ll

a. Ak g7 equivalent to less than 0.95 when flooded with unborated.

water, which includes a conservative allowance of 1.42% delta k/k for uncertainties.* -

b.

A nominal 10.375 assemblies placedinch center-to-center in the storage ranks.distance between fuel CRITICALITY - NEW FUEL 5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with that ka nominal 21.0 center-to-center distance between new fuel assemblies suc eff will n t exceed 0.98 when fuel having an- enrichment of 4.5 11 weight percent U-235,is in place ,and optimum achievpble moderation is assumed New feel entochorer\t n breoled k V.0 wegfrt percent ces ncHd on S.2,/ and SG,h /, .

DRAINAGE I 5.6.2 The spent fuel pit is designed and shall be maintained to prevent inaovertent draining of the pool below elevation 722 f t.

, CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies.

5. 7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

"For some accident conditions, the presence of dissolved boron in the pool water may be taken into account by applying the double contingency principle which R17 requires two unlikely, independent, concurrent events to produce a criticality accident.

)

SEQUOYAH - UNIT 1 un 4 s982 5-5 Amendment No. 13

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9 DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY - SPENT FUEL],, [ el er,r.;chg) o ,o urg lst percent U-235 5.6.1.1 Thespentfuelstorageracksaredesigneda)ndshallbemaintained with: 4 ll

a. Ak gf equivalent to less than 0.95 when flooded with unborated p.r.

water, which includes a conservative allowance of 1.42% delta k/k for uncertainties.* "

b.

A nominal 10.375 assemblies inch placed in thecenter-to-center storage racks. distance between fuel CRITICALITY - NEW FUEL 5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a nominal such that k 21.0 inch center-to-center distance between new fuel df will n t exceed 0.98 when fuel having anmanames enrichment of 9 4.5 weight percent U-235 is in place and optimum achievable moderation is assu NeW lutl S'dmtaf ts I'h*'kl Yo '/.0 Wrohli DRAINAGE t

ff'ct'rf, as / tens h 5:1 {GdYS*

r C.f h

/.

ll R4 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 722 ft.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies.

5. 7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

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} l *For some accident conditions, the presence of dissolved boron in the pool wa L may be taken into account by applying the double contingency principle which requires accident. two unlikely, independent, concurrent events to produce a criticalit R4 SEQUOYAH - UNIT 2 g ' 4 QQ 5-5 l

• Amendment No. 4 i .

1 PROPOSED TECHNICAL SPECIFICATION CHANGES SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 (TVA SQN TS 65)

CHANGE NO. 2 POWER RANCE FLUX RATE TRIP SETPOINT CHANCES

-w.m.-m

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,' TABLE 2.2-1 u,

E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

~

8 s3: FUNCTIONAL UNIT TRIP SETPOINT , ALLOWABLE VALUES Not Appilcable

~

1. Manual Reactor Trip Not Appitcable E

Low Setpoint - $ 26% of RATED

2. Power Range, Neutron Flux Low Setpoint - 5 25% of RATED THERMAL POWER THERNAL POWER High Setpoint - $ 110% of RATED High Setpoint - $ 109% of RATED THERMAL POWER 1

THERMAL POWER

< of RATED THE R

< 5% of RATED THE L POWER with onds

3. Power Range, Neutron Flux, aconds Gith a time constan High Positive Rate a time constant 3 a i

n. < of RATED THERMA R

4. Power Range, Neutron Flux, < % f RATED THE M L POWER with econd, Gith a time constant 1 High Negative Rate itimeconstantlysacondr

5. Intermediate Range, Neutron $ 30% of RATED THERMAL POWER

5 25% of RATED THERMAL POWER Flux 5 5 i 1.3 x 10 counts per second

6. Source Range, Neutron Flux 5 10 counts per second See Note 3

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 f 1 1970 psig 1 1960 psig

9. Pressurizer Pressure--Low

^

i $ 2395 psig

10. Pressurizer Pressure--High $ 2385 psig i

f 93.0% of instrument span

11. Pressurizer Water Level--High i 92% of instrument span

) > 89% of design flow

12. Less of Flow 2 90% of design flow jier loop

• per loop" l

• Design flow is 91,400 gpm per loop.

j

TABit 2.2-1 g RIAC10R IRIP SYSlEM INSIRUMENTA110N IRIP SETP0lNTS o .

g FUNCTIONAL UNIT TRIP SETPOINT , All0WABLE VALUES

1. Manual Reactor Trip Not Applicable Not Applicable N 2. Power Range, Neutron flux Low Setpoint 5 257,of RATED Low 5etpoint 1 26% of RATED N TilERMAL POWER THERMAL POWER High Setpoint 1 109% of RATED High Setpoint 1 110% of RATED TifERilAL POWER THERMAL POWER 6.3%

3. Power Range, Neutron Flux, 1 5% of RATED THE L PUWER with 1 +-W of RATED IllERMA 30VER l seconds with a time constant 3 seconds J High Positive Rate a time constant 1

4. Power Range, Neutro.1 Flux, < of RATED THE MAL POWER with < of RATED TilERMAL WER I High Negative Rate a time constant 1 beconds with a time constant 1 % econds j m 5. Intermediate Range, Neutron i 25% of RATED THERMAL POWER 1 30% of RATED

• THERMAL POWER m Flux 5 5

6. Source Range, Neutron Flux 1 10 counts per second i 1.3 x 10 counts per second

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 3 l

9. Pressurizei- Pressure--Low 3 1970 psig 1 1960 psig l 10. Pressurizer Pressure--High 1 2385 psig i 2395 psig

! 11. Pressurizer Water Level--liigh 1 92% of instrument span i 93% of instrument span l 12. Loss of Flow 3 90% of design flow per loop

• 1 89% of design flow per loop

• l

" Design tiow is 91,400 gpm per loop.

1 ENCLOSURE 2 JUSTIFICATION FOR PROPOSED CHANGE IN TECHNICAL SPECIFICATIONS FOR UNITS 1 AND 2 SEQUOYAH NUCLEAR PLANT (TVA SQN TS 65)

CHANGE NO. 1 SPENT FUEL POOL RADIATION MONITOR SETPOINT CHANCES 6

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1 JUSTIFICATION FOR INCREASING THE SPENT FUEL AREA RADIATION MONITOR SETPOINTS There are two safety concerns involved in increasing the setpoints of RM-90-102 and -103; inadvertent criticality in the fuel pools and a fuel handling accident. The existing plant and site features are sufficient to preclude an inadvertent criticality and to adequately mitigate the consequences of a fuel handling accident with a 200 mr/hr setpoint on RM-90-102 and -103.

INADVERTENT CRITICALITY CONSIDERATIONS TVA has designed and built the new fuel pool and the spent fuel pool to preclude an inadvertent criticality event as required by 10 CFR 50 Appendix A Criterion 62. These features are described in the design features section of the Sequoyah Nuclear Plant (SQN) technical specifi-cations. Specifically, sections 5.3.1, 5.6.1.1, and 5.6.1.2 ensure that no criticality will occur even under the worst-case assumptions of unborated, cold water flooding to the new and spent fuel pools. This technical specification change adds additional wording to section 5.6.1.1 and 2 to further clarify the assumptions used in the fuel rack design analysis.

Part 10 of the Code of Federal Regulations Section 70.24 requires each licensee with 1500 grams U235 at less than 4 percent enrichment to maintain a monitoring system. The system shall be capable of detecting a criticality that produces an absorbed dose in sof t tissue of 20 rads of combined neutron and gamma radiation at an unshielded distance of two meters from the reacting material, within one minute. Coverage is to be provided by two detectors. 10 CFR 70.24 subsection (d) allows the licensee to apply for exemption to the requirements of 10 CFR 70.24.

TVA cited the design features used to preclude the possibility of an inadvertent criticality, and NRC granted relief from this requirement.

This relief was specifically stated in the SQN Special Nuclear Materials (SNM) license. The SNM license was subsequently incorporated into the Operating License (0/L) in 1980.

FUEL HANDLING ACCIDENT CONSIDERATIONS SQN has two radiation monitors in the auxiliary building spent fuel pool area (RM-90-102 and -103) . The physical locations are shown on the attached drawing. The monitors are positioned to detect gap activity which may be released by a fuel handling accident, as required. This configuration satisfies 10 CFR 50 Appendix A Criterion 63 for spent fuel storage monitoring.

If either detector exceeds the trip setpoint, an auxiliary building isolation (ABI) signal is generated, and the auxiliary building secondary containment enclosure (ABSCE) is isolated. The normal ventilation system is stopped, and the auxiliary building gas treatment system is initiated. There were 13 spurious ABI signals generated in 1984 due to the present conservatively-low setpoints on the spent fuel radiation monitors.

l

_2 FUEL HANDLING ACCIDENT CONSIDERATIONS (Continued)

In each case the operator must follow Abnormal Operating Instruction (AOI) 29. This procedure is used to verify the shutdown of normal ventilation fans and startup of the ABGTS. The damper configuration is then verified. Af ter the signal is determined to be spurious, the operator must stop personnel traffic while System Operating Instruction (SOI) 30 is followed to reinstate normal fan and damper alignments.

This is an unnecessary burden on the operators, and distracts from more important duties.

Due to the new licensco event report (LER) rule 10 CFR 50.72 and 73. the plant staff is then required to file an LER on the spurious ABI with NRC. NRC in turn ' puts the LER into its data bank on plant events.

These reports have little useful information, and processing them diverts TVA and NRC resources from more important tasks.

An alternative to increasing the setpoint to 200 mrem was to apply an electronic time delay to RM-90-102 and -103 in order to filter spurious signals and prevent inadvertent auxiliary building isolations. The length of this time delay is governed by the response time of the system (from signal receipt until damper isolation is complete). An evaluation of the allowable time delay that could be added to the monitors determined that this was not a feasible alternative since damper isolation may not occur before a release to the atmosphere.

The plant design does not lend itself to performing large refueling outage maintenance tasks and associated decontamination without af fecting the spent fuel pool radiation monitors.

The offsite dose conseq'uences of a design basis fuel handling accident were reanalyzed in an ef fort to justify a higher setpoint. The analysis demonstrated that a change in the monitor setpoint of 10 mr/hr corresponds to a change in the unfiltered offsite thyroid dose of .5 rem. This leads to a possible setpoint of 6000 mr/hr without violating the criteria of 10 CFR 100. The ANS 51.1 guidelines reduce the 10 CFR 100 values by a factor of 10 which would still allow a setpoint of 600 mr/hr. A setpoint of 200 mr/hr on RM-90-102 and -103 will provide the required fuel handling accident mitigation features actuation while keeping a 5800 mr/hr margin for error.

This setpoint change is not expected to change the accident dose as presented in FSAR table 15.5.6-2.

This dose will remain well below the Regulatory Guide 1.25 acceptance criteria of 75 rem thyroid. This is due to the relatively instantaneous rise of radiation readings expected at the detectors to a reading well above the new setpoints.

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. SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION 10 CFR 50.91 AND 92 (SPENT FUEL POOL RADIATION MONITOR SETPOINT CHANGES)

1. The inadvertent criticality consideration has been addressed by passive design features which preclude any approach to criticality in either the new or spent fuel pool. NRC has previously reviewed this item as part of the SNM license review and has granted relief from the requirements of 10 CFR 70.24 " Criticality Monitors." The fuel nandling accident consideration has been reanalyzed. The results show that the monitors will perform their safety function with a setpoint of up to 6000 mr/hr. The proposed 200 mr/hr setpoint will initiate mitigation with a 5800 mr/hr margin for error. There is no significant increase in the probability of occurrence or the consequences of an accident previously evaluated in the safety analysis report.

2. Spent fuel pool area radiation monitors RM-90-102 and -103 will remain operable. The new setpoint will not degrade the capability of the plant to suf ficiently mitigate the ef fects of a design basis fuel handling accident. No new accident will be created by operation with a higher setpoint.

3. With the new setpoints, RM-90-102 and -103 will continue to recognize and initiate mitigation of a design basis fuel handling accident. The setpoint will be sufficiently low to ensure the timely operation of the secondary containment isolation features and the initiation of the auxiliary building gas treatment system.

These features will continue to provide their design safety function, and they will continue limit offsite effects to below the guidelines of 10 CFR 100.

CONCLUSION The spent fuci pool area radiation monitors were installed to address two safety concerns. The first concern was an inadvertent criticality in the spent fuel pool. This concern was previously reviewed by NRC and an exemption to 10 CFR 70.24 " Criticality Monitoring" was granted. The second safety concern involves the release of the fue' gap activity following a postulated fuel handling accident. Analysis has shown that a spent fuel pool radiation monitor setpoint below 6000 mr/hr will initiate an auxiliary building isolation which will limit the of fsite dose consequence below the limits detailed in 10 CFR 100. TVA fec1s that a setpoint of 200 mr/hr is more than reasonably conservative and that no risk to the public health and safety is created by raising the setpoint from 15 to 200 mr/hr. This change should eliminate spurious actuations of the auxiliary building gas treatment system; reduce the burden on the unit operators; reduce the burden of needed licensee event reports on the plant staff; and eliminate the input of valueless data into the NRC data bank. This change will prevent unnecessary challenges to this engineered safety feature and reduce the probability of system failure.

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( l JUSTIFICATION FOR PROPOSED CHANGE IN TECHNICAL SPECIFICATIONS FOR UNITS 1 AND 2 SEQUOYAH NUCLEAR PLANT l

(TVA SQN TS 65)

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CHANGE NO. 2 l

POWER RANGE FLUX RATE TRIP SETPOINT CHANGES i

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TECHNICAL SPECIFICATION CHANCE No. 65 REACTOR TRIP SYSTEM INSTRUMENTATION FLUX RATE TRIP SETPOINTS REASON FOR CHANCE Westinghouse Technical Bulletin NSID-85-13. " Flux Rate Trip Setpoint,"

was issued to address the alignment of the nuclear instrumentation system power range positive and nega tive rate trip bistables and to recommend a method of achieving proper calibra tion. Given the current technical specification setpoint, implementing this calibration method for the negative rate trip function would significantly increase the chances of inadvertent trip actuations caused by process noise. The proposed change for the negative rate trip setpoint will provide the margin required to allow the Westinghouse-recommended procedure to be implemented without causing inadvertent actuations and unnecessa ry challenges to the associated safety systems.

For the positive rate function, the time constant portion of the trip setpoint must be changed for consistency with the proposed negative rate value (2 seconds versus I second) because the associated instrumentation rate circuitry is common to both functions.

The revised " allowable values" for both functions will provide more flexibility in the calibration of the instrumentation. These values are consistent with the la test setpoint methodology calcula tions performed by Westinghouse.

1 JUSTIFICATION FOR CHANGE The proposed trip setpoint changes are consistent with the latest Westinghouse

" Precautions, Limitations, and Setpoints Nuclear Steam Supply System" document. Westinghouse WCAP-10297-P-A, " Dropped Rod Methodology for Negative Flux Rate Trip Plants," provides the technical justification'for this change.

The WCAP has been reviewed and approved by NRC (March 31, 1983).

The proposed changes to the " allowable values" are consistent with the latest setpoint methodology calculations performed by Westinghouse (see attached Westinghouse letter No. TVA-85-158 dated August 30, 1985).

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- ' .;u; 30,85 13:59 WESTIllGHOJSE $51D F ROJECTS SOi:THERn y;t,.g,;.: p , ,y,,

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TVA 85-paa J. P. Darling August 30, 1985 If you have any questions concerning the above please feel free in contacting me.

Very truly yours. .

WE5 NGHOUSE ELECTRIC CORPORATION fI/htU gf b U

. L. Williams, Manager NSID Projects Mid South Area cc: H. L. Abercromble R. U. Mathleson I. R. Williamson F. Mashburn J. H. Sullivan e a e. .

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l SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION 10 CFR 50.91 AND 92 (FLUX RATE TRIP SETPOINT CHANGES)

1. The purpose of the negative flux rate trip function is to pro-ide protection against control rod drop accidents by ensuring the miniraum DNBR is maintained above 1.3. WCAP-10297-P-A provides a ecmprehensive methodology to support use of the revised setpoint. This WCAP demonstrates that all dropped rod events will result in either a reactor trip or will not exceed DNB limits. The DNB design basis is satisfied.

Both the safety' analysis and technical specification bases remain satisfied, and there are no adverse effects on the probability of occurrence or the consequences of an accident previously analyzed.

The purpose of the positive flux rate trip is to provide protection against rapid flux increases which are characteristic of rod ejection events. The proposed positive rate trip setpoint change affects the time constant portion only. Current technical specifications require the time constant to be 1 second. The proposed change to 2 seconds is more conservative, and therefore satisfies the criteria for finding that no significant hazards considerations are created by this change.

The proposed changes to the allowable values for both the positive and negative rate trip functions are consistent with the latest setpoint methodology calculations performed by Westinghouse.

2. The positive and negative rate trip functions provide protection against rod ejection events and dropped rod accidents, respectively. Both of these events are considered in the safety analysis report. No new or different type accidents are created.

3. With the revised setpoints, the instrumentation will continue to function as assumed in the safety analysis and technical specifications. For the reasons given in item 1 above, the reactor trip system remains operable and no safety margins are compromised.

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