ML20214F411

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Forwards Util 870416 Application for Amend to License NPF-30 & Amend 22 to License NPF-30,per Request.W/O Encl
ML20214F411
Person / Time
Site: Callaway Ameren icon.png
Issue date: 05/13/1987
From: Alexion T
Office of Nuclear Reactor Regulation
To: Kucera R
MISSOURI, STATE OF
References
NUDOCS 8705260069
Download: ML20214F411 (2)
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MAY 131987 - ,

' Docket No. 50-483:

l' .j Mr. Ronald A. Kucera, Deputy Director Department of Natural . Resources .

P. O. Box 176:

Jefferson City, Missouri 65102 '

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Dear Mr. Kucera:

I have enclosed the April 16, 1987 application for a license amendment from Union Electric Company that you requested. Also enclosed is License Amendment No. 22 in response to the111censee's application.

If I may assist you in any other way regarding licensing matters at Callaway Plant, please contact me.

Sincerely, hl Thomas W. Alexion, Project Manager Project Directorate III-3 Division of Reactor Projects

Enclosure:

As Stated Office: L PDhII-3 h

PM 3~ PD/PDIII-3 Surname: P eOffer TAlexich/tg DWigginton '

Date: 0 /j3/87 05/l3/87 05/j3/87 8705260069 870513 PDR P ADOCK 05000403 PDR

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(Docket files, .:i, Local PDR PDR PDIII-3 r/f PKreutzer TAlexion l

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1907 Grabot Stet. St. Love April 16, 1987 Donald F. Schnell Vce Prescent U.S. Nuclear Regulatory Commission ATTN Document Control Desk Washington, D.C. 20555 Gentlemen: ULNRC-1493 DOCKET NUMBER 50-483 CALLAWAY PLANT REVISION TO TECHNICAL SPBCIFICATION TABLE 3.3-5

Reference:

1. ULNRC-1207 dated 11/15/85 )

Union Electric Company (UE) herewith transmits an application for amendment to Facility Operating License No. NPF-30 for Callaway Plant. This license amendment request proposes to revise Technical Specification Table 3.3-5 to increase the ESF

- response times for Items 2.a. (Containment Pressure-High-1, Safety Injection) , 3.a. (Pressurizer Pressure-Low, Safety Injection) and 4.a. (Steam Line Pressure-Low, Safety Injection).

These changes are detailed in Attachment 1.

On April 13, UE received a letter from Westinghouse Electric Corporation (SCP 87-155 dated 4/8/87) khich informed us of a potential issue concerning the time required to change charging pump suction from the Volume control Tank (VCT) to the Refueling Water Storage Tank (RWST) following a postulated steam line break event. Westinghouse advised that the issue had been evaluated for Callaway and that there was negligible impact on the Main Steam Line Break analysis results. UE reviewed the information j and determined on April 15, 1987 that the issue was applicable to i Callaway Plant. It was determined that the situation which has I existed at Callaway is reportable within 30 days under  !

10CFR50.73 (a) (2) (B) as an operation or condition prohibited by the plant's Technical Specifications. The reason for non-compliance was a misinterpretation of the surveillance requirements in Technical Specification Table 3.3-5. Although all equipment had been determined to be operational as required, .

the timing of equipment operation to meet response time  !

requirements was misinterpreted. Notification of the situation was made to the NRC Resident Inspector and the NRC Licensing Project Manager on April 15, 1987.

Callaway has been in Mode 5 since April 3, 1987 for a maintenance outage. Callaway is scheduled to return to power in early May and will enter Mode 3 on May 6, 1987, at which time the requested change is required to be approved. Therefore, pursuant to 10CFR50. 91(a) (5) , UE hereby requests NRC emergency

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'B704200254-e70416 I PDR P

ADOCK 05000483 PDR 0 7[h g I L

Maihng Aodress: P.O. Box 149, SL Lous. MO 63166 C N $

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l authorization and approval of this proposed amendment to Facility Operating License No. NPF-30. -

Further, the requested emergency authorization is  !

appropriate because this amendment request involves In no addition, I significant hazards consideration (AttachmentAs3).indicated above,

the present situation could not be avoided. ,

} UE promptly notified the NRC of this situation and has pursued an The results of the Safety i

j expeditious resolution of this matter. ,

Evaluation (Attachment 2) show that there is sufficient margin ,

available in the analysis of record (Reference 1), such that {

conclusions reached therein remain valid.

. The Callaway Plant On-Site Review Committee and Nuclear i Safety Review Board have reviewed and approved this amendment 1 request.

In accordance with 10CFR50.91(b), UE will promptly provide the State of Missouri with a copy of this proposed amendment to 1 ensure their awareness of this request.

Enclosed is a check for the $150.00 application fee as required by 10CFR170.21

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' " Very truly yours, 8'

Do Schnell j DS/ mat ,

i j Attachments: 1 - Proposed Technical Specification Change 2 - Safety Evaluation

! 3 - Significant Basard Evaluation 4 - Application Fee 1

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STATE CF MISSOURI SS CITY OF ST. LOUIS .

Robert J. Schukai, of lawful age, being first duly swornfor upon oath says that he is General Manager-Engineering (Nuclear) I Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

By <

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Rob 4r chukai Gen tal M ager-Engineering Nuc SUBSCRIBEDandsworntobeforemethis/ day of , 198 7

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1m v._0' Wl BARGARA J. PFAFF IIOfARY PU4UC. STATE Of til1$0URI uv commissai Enrieu Arau. 22.1ses si. Louis COUNTY l

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v cc: Gerald Charnoff, Esq.

Shaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W.

  • Washington, D.C. 20037 Dr. J. O. Cermack CFA, Inc.

4 Professional Drive (Suite 110)

Gaithersburgh, MD 20879 W. L. Forney Division of Projects and Resident Programs, Chief, Section lA U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137

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Bruce Little Callaway Resident Office U.S. Nuclear Regulatory Commission RRil Steedman, Missouri 65077 Tom Alexion (2)

Office of Nuclear Reactor Regulation

- U.S. Nuclear Regulatory Ccamission Mail Stop 316 7920 Norfolk Avenue Bethesda, MD 20014 Ron Kucera, Deputy Director ,

Department of Natural Resources

  • P.O. Box 176 Jefferson City, MO 65102 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 l

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ULNRC-1493 Attcchment 1

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' PROPOSED TECHNICAL SPECIFICATION CHANGE Page 3/4 3-29 Page 3/4 3-30 Page 3/4 3-32 Page B3/4 3-2 4

Insert Page

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-8704ecogM 0416 ADOCK OQT lDR 500g3 b* w ass ha - e w -,hwe

TA8LE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TINES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECON05

1. Manual Initiation

' a. Safety Injection (ECCS) N.A.

b. Containment Spray N.A.
c. Phase "A" Isolation
  • N.A.
d. Phase "B" Isolation N.A. *
e. Containment Purge Isolation N/A.

Steam Line Isolation

f. N.A.

'. g. Feedwater Isolation M.A.

h. Auxiliary Feedwater N.A.
i. Essential Service Water N.A.

. j. Containeent Cooling N.A.

k. Control Room Isolation N.A.

. 1. Reactor Trip N.A.

, ) n. Emergency Diesel Generators N .' A .

n. ' Component Cooling Water N.A.
o. Turbine Trip N.A.
2. Containment Pressure-High-1 *
a. Safety Injection (ECCS) 1 29 N(7) /t g3d#)
1) Reactor Trip i2
2) Feedwater Isolation < 2(5)
3) Phase "A" Isolation i 1.5(5)
4) Auxiliary Feedwater '

i 60 i 5) Essential Service Water i 60(1) j 6) Containment Cooling 1 60 III

7) Component Cooling Water M.A. 1
8) Emergency Olesel Generators 50) 1 14
9) Turbine Trip N.A.

6 CALLAWAY - UNIT 1 3/4 3-29 Amenment No. '18

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. TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING $1GNAL AND FUNCTION RESPONSE TIME IN SECONDS

3. Pressurizer Pressure Low
a. Safety Injection (ECCS) 1 29 /2d4) g7
1) Reactor Trip <2
2) Feedwater Isolation l
3) Phase "A" Isolation I2((5)

~2 1 5)

4) Auxiliary Feedwater < 60 *
5) Essential Service Water 7 60(1) *
6) Containment Cooling

[60(1)

7) Component Cooling Water N.A.
8) Emergency Diesel Generators IO) 1 14
9) Turbine Trip N.A.
4. Steam Line Pressure-Low 31
a. SafetyInjection(ECCS), iK(3)j3g42
1) Reactor Trip <2
2) Feedwater Isolation 2(5)
3) Phase "A" Isolation h2(5)
4) Auxiliary Feedwater -

< 60

5) Essential Service Water h60(1)
6) Containment Cooling 1 60(1)
7) Component Cooling Water M.A.
8) Emergency Diesel Generators 1 14(6)
9) Turbine Trip N.A.
b. Steen Line Isolation 3 2(5)

CALLAWAY - UNIT 1 3/4 3-30 Amendment No. 18 1

REVislON I'*

  • TA8LE 3.3-5 (Continued)
  • )

ENGINEERED $AFETY FEATURES RESPONSE TIMES INITIATING $1GNAl, AND FUNCTION RESPONSE TIME IN $ ECON 05

12. Auxiliary Feedwater Puno Suction F're s s ure-1,ow ,

Transfer to Essential

.! Service Water N.A.

13. RWST Level-Low-Low Coincident with *

, . Safety Injection Automatic Switchover to Containment -

1 40 .

Sump 4 . 14. Loss of Power .

a. 4 kV Sus Undervoltage- < 14

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Loss of Voltage

b. 4 kV Bus Undervoltage- < 144

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, i Grid Degraded Voltage s

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j 15. Phase "A" ! solation

a. Control Room Isolation M.A.

i b. Containment Purge Isolation i 2(5) ,

TA8LE NOTATIONS .

l (1) Olesel generator starting and sequence loading ddlays included.

l (2) Of esel generator starting delay agt, included. Offsite power available.

(3) Olesel generator starting and sequence leading dela RHR ,

pumps not included. S- ** 1- . af .a _._y included. w 4' vs.TM.8.in. i,2 A 2 1 , % g

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ve.TiiL Olesel generator MsT(twsv startingW.'M-pa and seque,nce loading delays no)t included.

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Offsite p e avatl kHR pumps n2.$ included. K----- 8 I

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  • (5) Does not include valve closure time. i.,- a- t 3 , ,

l (6) Includes time for diesel to reach full speed.

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g INSTRUMENTATION ,

BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continuto) rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for the actuation. R or Rack Error is the "as measurec" deviation, in percent span, for the affected channel from the specified Trip Setpoint. 5 or Sensor Error is either the "as measured" deviation of tne sensor from its calibration point or the value specified in Tac!e 3.3-4, in percent span, from the analysis assumptions. .

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of ,

operating within the allowances of these uncertainty magnitudes. Rack drift l

, in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statisitical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

3 The measurement of response time at the spectfied frequencies provides l

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/ assurance that the Reactor Trip and the Engineered Safety Features actuation 1 associated with each channel is completed within the time limit assumed in the ,

safety analyses. No credit was taken in the analyses for those channels with '

response times indicated as not applicable. Response time may be demonstrated i by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verificatioh may be demonstrated by either: (1) in place, onsite, or offsite test measurements or (2) utilizing replacement sensors with certified response times.

maA l The Engineered safety Features Actuation System senses selected plant parameters and determines whether or not predetermined ifmits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trips, (3) Feedwater System isolates, (4) the emergency diesel generators start, (5) containment spray pumps start and automatic valves position, (6) contain-ment isolates, (7) steam lines isolate (8) Turbine trips, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, (11) essential service water pumos start and automatic valves position, and (12) isolate normal control room ventilation and start Emergency Ventilation System.

CALLAWAY - UNIT 1 8 3/4 3-2 Amenoment No. 17

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4 INSERT A Engineered Safety Features response time specified in Table 3.3-5 which include sequential operation of the RWST and VCT valves (Notes 3 and 4) are based on values assumed in the non-LOCA safety analyses. These analyses take credit for injection of borated water from the RMST. Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening of the RWST charging pump suction valves. When the' sequential operation of the RWST and VCT valves is not included in the response times (Note 7), the values specified are based on the LOCA analyses. The LOCA analyses take credit for injection flow regardless of the source. Verification of the response time specified in Table 3.3-5 will assure that the assumptions used for the IOCA and non-IOCA analyses with respect to operation of the VCT and RWST valvert are valid.

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ULNRC-1493 Attachment 2 SAFETY EVALUATION

Reference:

1. U LNRC-l Z U 7 dated 11/15/85 ,

This safety evaluation is in support of a license amendment request to revise Technical Specification Table 3.3-5 to increase the ESF response times for Items: 2.a. (Containment Pressure-High-1, SI) ; 3.a. (Pressurizer Pressure-Low, SI) ; and 4.a. (Steam Line Pressure-Low, SI). These changes are contained in Attachment 1.

DACKGROUND In the normal configuration of the Chemical and Volume '

Control System (CVCS), the charging pumps take suction from the volume Control Tank (VCT). When a Safety Injection (SI) signal is generated from the protection logic, a signal is sent to start the high-head charging pumps and to begin opening the Refueling Water Storage Tank isolation valves, in order to align the borated water source for delivery to the RCS. Once the RWST isolation valves have repositioned and are indicated fully opened, the isolation valves on the VCT will begin to close.

. This sequential valve stroke time can be as long as 25 seconds.

Since the VCT is pressurized, it will be the source of the SI flow until the isolation valves are closed. This af fects the time assumed at which the 2000 ppm borated water in the RWST is available to the suction of the charging pumps.

The FSAR Steam Line Break analysis (Reference 1) which supports the current Technical Specifications (Table 3.3-5) assumes the following delays for delivery of borated water to the RCS:

1. SI signal generation (2 seconds)
2. Diesel start-including time to como up to speed (12 reconds)
3. Valve stroke times and pumps to full speed (10, seconds)

This assumes, however, that the VCT and RWST isolation valves stroke simultaneously rather than sequentially. The valve interlock logic increases the delay time for the availability of borated water by 15 seconds (conservatively) to 27 seconds with of fsite power and 39 seconds without offsite power. The only non-LOCA transient impacted by the increased time delay is the steam line break event. No other Chapter 15 transient relies on short-term boration f rom the KWS'. to mitigate the event.

EVALUATION Based on the current steam line break analysis for the Callaway plant and sensitivities performed for other plants, the additional time delay is acceptable. Specifically:

.- ULNRC-1493 Attachment 2

1) The additional delay in the availability of borated water occurs early in the steam line break transient when RCS pressures are relatively high and SI flowrates are relatively small due to head vs. SI flow characteristics.
2) Previous sensitivities have shown that delays of this magnitude result in small changes in the analysis results. A comparison of cases with and without the additional SIS delay showed, over the limiting portien of the transient, maximum differences of 0.2% in power, 0.6 degrees F in temperature, and 10 psi in RCS pressure. A callaway specific review of the steam line break analysis demonstrated that there is sufficient

- margin available in the analysis such that the conclusions presented in Reference 1 remain valid.

3) The analysis assumes only one centrifugal charging pump is available. However, at the pressures characteristic of a steam line break, the centrifugal charging pump and safety injection pump of a given train would be available to deliver a significantly greater flowrate of borated water to the RCS.

From analyses performed for other Westinghouse plants, it has been shown that SI boron concentration reduction has little effect on the steam line break mass / energy release analysis inside containment. Since the additional time delay is a small i perturbation compared to a large change in the available boron concentration, there will be negligible impact on the steam line break mass / energy release inside containment analysis.

Sensitivities performed for the steam line break superheated mass / energy release outside containment analysis show that the results are not sensitive to large changes in SI flow (Reference WCAP-10961, Rev. 1). The additional time delay is a small perturbation compared to a large change in total SI flows therefore, it is concluded that the impact on the Callaway superheated-mass / energy releases outside containment is insignificant.

In the case of a Loss of Coolant Accident, the immediate safety function of SI is to supply water to the RCS, whether borated or not. The time at which water (from either the VCT or the RWST) is available to the suction of the high-head charging pumps is not affected. Thus, for those SI actuation signals that are only intended to provide protection against a LOCA, this additional delay is not required since boron is only required for maintaining suberiticality in the long term following a LOCA.

l

^[ I ULNRC-1493 Attachment 2 Ef fect on Design Basis Accident Analysis l

l A reference steam line break event for a four loop, 17 x 17

[ optimized fuel, PWR power plant was used to evaluate the

. sensitivity to SI flow. It was found that the difference in core l boron concentration, peak return to power, RCS temperature, RCS i pressure and DNBR were minimal with a 15 second delay in SI flow.

! The Callaway specific analysis was checked to ensure that sufficient margin existed.

Potential for Creation of an Unanalyzed Accident There are no new failure modes associated with this proposed change since no design changes have been made.

No new accident is created because the same equipment is assumed to perform in the same manner as before. Only the l* testing of the timing of the delivery of borated injection flow l is affec'ted. This can be adequately modeled in the current l

safety analysis.

Ef fect on the Marcin of Safety There is no impact on the consequences on protective boundaries. All acceptance criteria in Reference 1 are still met.

The proposed change is' intended to bring the Technical Specification surveillance in line with the basis. The basis is to mitigate a steam line break which requires injection of borated water into the RCS. The present Technical Specification surveillance ensures flow initiated to the core but did not test the time to provide borated water. The proposed change will i increase the time to initiate borated water flow to the core by 15 seconds. With the additional 15 seconds delay in supplying

borated water to the core, the DNB design basis is still met, and the conclusions in Reference 1 remain valid. There fore, the chango does not reduce the margin of safety as specified in the basis of any Technical Specification.

l Summary & Conc 1 unions l

The proposed change in the ESP response times for

! Containment Proosure-High-1, Low Pressurizer Pressure and Low Steam Line Pressure in Technical Specification Table 3.3-5, Items 2.a, 3.a and 4.a to incorporate an increase of 15 seconds is I

acceptable. Evaluation of the impact on the Callaway safety analysis licensing basis demonstrates that the conclusions in l Reference 1 remain valid.

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b 4-ult!RC-1493 Attachment 2 Based on the foregoing assessment, the change proposed herein is considered safe and does not represent an unreviewed safety question as defined in 10CFR50.59 since is does not:

1. Increase the frequency of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report;
2. Create the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis r9 ports

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3. Reduce the margin of safety as defined in the basis for 4 any technical specification.

This amendment request would not adversely affect or endanger the health and safety of the general public and does not involve an unreviewed safety question.

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ULNRC-1493 Attcchment 3 SIGNIFICANT HAZARD EVALUATION .

This significant hazard evaluation is in support of a license amendment request to revise Technical Specification Table 3.3-5 to increase the Engineered Safety Features (ESF) responso times for Items: 2.a. (Containment Pressure-!!igh-1, SI) ; 3.a.

(Pressurizer Pressure-Low, SI); and 4.a. (Steam Line Pressure-Low, SI).

In accordance with 10CFR50.92, Union Electric Company has I reviewed the proposed changes and has concluded they do not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised, a conclusion which is supported by our determinations mado pursuant to 10CFR50.59. The proposed chango

- does not involve a significant hazards consideration because the .

chango would not

1) Involvo a significant increase in the probability or consequences of accident previously evaluated. An increase in the acceptance criterion for the ESP j

response time is acceptable since the evaluation of the impact of the increased delay on the steam line break event demonstrated that the DNB design basis la still met. The conclusions presented in the ULNRC-1207 dated November 19, 1985 remain valid.

2) , Create the possibility of a new or dif ferent kind of accident from any previously evaluated. There are no new failure modes associated with this proposed change, an no design changes have been made. No new accident is created because the same equipment is assumed to perform in the same manner as before. Therefore, an increane in the ESr response times for high containment preneure, low steam line pressure, and low steam line pressure does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report.

I

3) Involvo a significant reduction in a margin of safety.

The proposed change is intended to bring the Technical Specification surveillance in line with the basis. An

( otated beforo, there is no impact on the consequences on protective boundaries, and all acceptance criteria in the analysis of record, submitted by ULNRC-1207 dated November 15, 1985, are still met. Therefore, the safety limits will still be met.

Moreover, the Commission han provided guidance concerning the application of standards in 10CFR50.92 by providing certain exampion (March 6, 1986, IB7751) of amendments that are considered not likely to involve significant hazards l

'T ULNRC-1493 Attachment 3 consideration. Although the proposed change herein is not enveloped by a specific example, the proposed change would not involve a signiflcant increase in the probability or consequences of an accident previously analyzed. The results of the safety evaluation show that there is suf ficient margin available in the I analysis such that the conclusions presented in ULNRC-1207 dated November 15, 1985 remain valid.

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ULNRC-1493 Attachment 4 I

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! Applicatton Fee UMON ELECTIUC COMPANY sr,w um.ousu "af 315355 10.o41 .. . orrRoy,,,ssouni 3211 .

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US NUCLEAR'RCGULATORY COMMISSION '04/16/87 $******150 00 l' WASHINGTON'DC 20555

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UNITED STATES g g NUCLEAR REGULATORY COMMISSION t ' -l wasHiwovoN. o. c. rosss k.'..../ May 4, 1987 Docket No. STN 50-483 Mr. Donald F. Schnell Vice President - Nuclear Union Electric Company '

Post Office Box 149 l St. Louis, Missouri 63166

Dear Mr. Schnell:

)

SUBJECT:

CALLAWAY PLANT, UNIT 1 - AMENDMENT NO.22 TO FACILITY OPERATING LICENSE NPF-30 The Commission has issued the enclosed Amendmenj No. 22 to Facility Operating

, License NPF-30 for the Callaway Plant, Unit 1. The amendment consists of a change to the Technical Specifications (TS) in response to your application dated April'16,1987.

The amendment revises Table 3.3-5 of the TS to increase the Engineered Safety Features (ESF) response times by fifteen seconds for Items: 2.a. (Containment Pressure-High-1, Safety Injection); 3.a. (Pressurizer Pressure-Low, Safety injection);and4.a.(SteamLinePressure-Low,SafetyInjection). These TS revisions are being issued before the expiration of the notice period to preclude any unnecessary delay in plant startup from the current outage. The amendment is effective as of its date of issuance.

A copy of the related Safety Evaluation is enclosed. Notice of issuance will be included in the Comission's next regular bi-weekly Federal Register notice.

Sincerely, M 0l %l)s,/,.*n s wMnmw bi.

Thomas W. Alexion, Project Manager Project Directorate !!!-3 DivisionofReactorProjects

Enclosures:

1. Amendment No. 22 to License No. NPT-30
2. Safety Evaluation ,

cc w/ enclosures:

See next page l

4 7 4 G B g'g b F .2 ,

Mr. D. F. Schnell Callaway Flant .

Union Electric Company Unit No. I cc:

J. O. Cemak, CFA Inc. Lewis C. Green. Esq.

4 Professional Dr., Suite 110 Green, Hennings & Henry Gaithersburg, MD 20879 Attorney for Joint Intervenors St. Louis, Missouri 65251 Cerald Charnoff, Esq.

Thomas A. Baxter, Esq. Ms. Marjorie Reilly Shaw, Pittman, Potts & Trowbridge Energy Chaiman of the League of 1800 M Street, N. W. .

Women Voters of Univ. City, M0 Washington, D. C. 20037 7065 Pershing Avenue

, University City, Missouri 63130 Mr. G. L. Randolf ,

General Manager, Operations Mr. Donald Bollinger, Member Union Electric Company Missourians for Safe Energy  ;

Post Office Box 620 6267 Delmar Boulevard .

Fulton, Missouri 65251 University City, Missouri 03130  !

  • ' l U. S. Nuclear Regulatory Comission' Mr. Dan I. Bolef, President l Resident Inspectors Office Kay Drey, Representative '

RRf1 Board of Directors Coalition 1 Steedman, Missouri 65077 for the Environment i J

St. Louis Region Mr. Donald W. Capone, Manager 6267 Delmar Boulevard i Nuclear Engineering University City, Missouri 63130 Union Electric Company

, Post Office Box 149 ,

St. Louis, Missourt 63166 Chris R. Rogers, P.E.

Manager - Electric Department 301 W. High Post Office Box 360 Jefferson City, Missouri 65102

, Regional Administrator U. S. NRC, Region !!!

799 Roosevelt Road l Glen Ellyn, Illinois 60137 '

Mr. Ronald A. Kucera, Deputy Director Department of Natural Resources P. O. Box 176 Jefferson City, Missouri 65102 Mr. Bart D. Withers Pesident and Chief Executive Officer Wolf Creek Nuclear Operating Corporation P. 0. Box 411 -

Burlington, Kansas 66839 e e .

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  1. 'o UNITED STATES E h NUCLEAR REGULATORY COMMISSION 3  : j WASHINGTON, D. C. 20655

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UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 1

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DOCKET NO. STN 50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 22 License No. NPF-30

1. The Nuclear Regulatory Connission (the Commission) has found that:

A. The application for amendment filed by Union Electric Company (the licensee)datedApril 16, 1987, complies with the standards and requirementsoftheAtomicEnergyAct.of1954,asamended(theAct).

and the Connission's rules and regulations set forth in 10 CFR

. Chapter I;

8. The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Connission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Connission's regulations; D. The issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Connission's regulations and all applicable requirements have been satisfied.
2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-30 is hereby amended to read as follows:

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T (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained ir. Appendix A, as revised through Amendment No.22 , and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license. UE shall operate the facility in accordance with the Technical Specifications and tLe Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance. l FOR THE NUCLEAR REGULATORY COMMISSION

'i David L. Wi nton.. Acting Project Director Project Directorate III-3 Division of Reactor Projects Attachrent:

Changes to the Technical

. Specifications Date of Issuance: hay 4, 1987 l

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ATTACHMENT TO LICENSE AMEf DFENT NO.22 OPERATIhG LICENSE NO. NPF-30 DOCKET NO. 50-483 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. 'The revised pages are identified t3 the captiened amendment number and contain marginal lines indicating the area of change. Corresponding overleaf pages are provided to maintain docur.ent completeness.

REMOVE PAGES INSERT PAGES 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-32 -

3/4 3-32 3/4 3-32a B 3/4 3-2 B 3/4 3-2 B 3/4 3-2a (repositioned) 1 i

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TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES

!NITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONCS

1. Manual Initiation
a. Safety Injection (ECCS) N.A.
b. Containment Spray N.A.
c. Phase "A" Isolation N.A.
d. Phase "B" Isolation N.A.

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e. Containment Purge Isolation N.A.

. f. Steam Line Isolation N.A.

g. Feedwater Isolation N.A.
h. Auxiliary Feedwater .

N.A. i

i. Essential Service Water N.A.

,- j. Containment Cooling N.A.

k. Control Room Isolation N.A.
1. Reactor Trip N.A.

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m. Emergency Diesel Generators N.A.
n. Component Cooling Water 'i!. A .
o. Turbine Trip N.A.
2. Containment Pressure-High-1
a. Safety Injection (ECCS) 29(7)/27I4)
1) Reactor Trip

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2) Feedwater Isolation 2(5)
3) Phase "A" Isolation 1.5(5)'
4) Auxiliary Feedwater < 60
5) Essential Service Water < 60(I)
6) Containment Cooling 60(I)
7) Component Cooling Water N.A.
8) Emergency Diesel Generators < 14(6)
9) Turbine Trip N.A.

s CALLAWAY - UNIT 1 3/4 3-29 Amendnent No. 78. 22

l TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES i

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

5. Coatainment Pressure-High-3
a. Containment Spray -1 32(1)/20(2)
b. Phase "B" Isolation 1 31.5
6. ~ Containment Pressure-High-2 Steam Line Isolation 1 2(5)
7. Steam Line Pressure-Negative .

Rate-High Steam Line Isolation 5 2(5)

8. Steam Generator Water Level-Hiah-Hich
a. Feedwater Isolation i 2(5)
b. Turbine Trip 1 2.5
9. Statam Generator Water Level-Low-Low
a. Start Motor-Driven Auxiliary Feedwater Pumps 1 60
b. Start Turbine-Driven Auxiliary Feedwater Pump

$ 60

10. Loss-of-Offsite Power _

Start Turbine-Driven Auxiliary Feedwater Pump N.A.

11. Trip of All Main Feedwater Pumps _ ..

Start Motor-Driven Auxiliary Feedwater Pumps N.A.  !

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CALLAWAY - UNIT 1 3/4 3-31 Amendment No. 18

l TABLE NOTATIONS (Continued)

(7) Diesel generator starting and sequence loading delays included. Sequen .

tial transfer of charging pump suction from the VCT to the RWST (RWST valves open, then VCT valves close) is not included. Response time assures only opening of RWST valves. .

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CALLAWAY - UNIT 1 3/4 3-32a Amendment No.22

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f 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION i The OPERABILITY of the Reactor Trip System and the Engineered Safety l Features Actuation System instrumentation and interlocks ensures that: (1) the associated action and/or Reactor trip will be initiated when the parameter {

i monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is main-  ;

tained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

The OPERASILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility '

design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems.is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system fonctional capability is main-tained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. Specified ' surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service times' for the Reactor Protection Instrumentation System," supplements to that report, and the NRC's Safety Evaluation dated February 21, 1985. Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

I To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated,  !

Allowable Values for the Setpoints have been specified in Table 3.3-4.

Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERA 8ILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 3.3-1, Z + R + S < TA, the interactive effects of the errors in the rack and the sensor, an'd the "as measured" values of the errors are considered. Z, as specified in Table 3.3-4, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with thessensor and rack drift and the accuracy of their measurement.. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the7alue used in the analysis for the actuation. R or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified Trip CALLAWAY - UNIT 1 8 3/4 3-1 Amendment No. 17

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l INSTRUMENTATION l

BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continuec) those Engineered Safety Features components whose aggregate function best l serves the requirements of the condition. As an example, the following acticns  ;

may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trips, (3) Feedwater System isolates, (4) the emergency diesel generators start, (5) containment spray pumps start and automatic valves position. (6) contain-ment isolates, (7) steam lines isolate, (8) Turbine trips, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, (11) essential service water pumps start and automatic valves position, and (12) isolate normal control room ventilation and start Emergency Ventilation System.

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CALLAWAY - UNIT 1 B 3/4 3-2a - Amendment No. 4W/,22

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  1. o, UNITED STATES 8' 3 , i NUCLEAR REGULATORY COMMISSION j g .p WASHINGTON, D. C. 20555 1

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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l RELATED TO AMENDMENT NO. 22 TO FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. STN 50-483 1 .

INTRODUCTION By letter dated April 16, 1987, Union Electric Co'apany (the licensee) informed l

, the staff that Westinghouse Electric Corporation recently discovered that it '

had assumed simultaneous, rather than sequential, operation of valves in the calculation of time it takes to get a safety injection (SI) of 2000 ppm j boratedwaterintotheReactorCoolantSystem(RCS). Since the valves that  !

- i transfer the charging pump suction from the Volume Control Tank (VCT) to the j Refueling Water Storage Tank (RWST), which contains 2000 ppm boron, are i operated sequentially, it was found that safety injection (ECCS) response l times listed in Table 3.3-5 of the Technical Specifications (TS) were not achievable. There were too short by the 15-second delay encountered by the sequential operation of the two valves.

! EVALUATION I

The primary function of the ECCS is to supply water to the RCS in the event of i

i a loss of coolant accident (LOCA). Since a LOCA is not a reactivity induced 1 accident, the 2000 ppm boron is not immediately needed. It is only needed to i maintain subscriticality in the long term. Therefore, for those SI actuation d

signals that are only intended to provide protection against a LOCA, this I 15-second delay in the delivery of 2000 ppe borated water has no effect on the safety analysis.

The only non-LOCA transient impacted by this increased response time is the

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steam line break event. No other FSAR Chapter 15 transient takes credit for short-term boration from the RWST.

The licensee compared ediculations of the steam line break accident with and I without the additional !ti delay. The calculations showed no significant change in the consequences. Ode of the reasons for this is that the additional delay occurs early in the steSm line break event when the RCS pressure is high and the SI' flow rate is relatively small. In addition, the licensee stated that studiejofthesteamlinebreakaccidenthavegenerallyshownthatthe consectences are not sensitive to large changes in SI flow or borpn concentration.

The licensee, therefore, concluded that the departure from nucleate boiling design basis for the steam line break analysis is still met and that the conclusions presented in the FSAR remain valid.

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The staff has reviewed the licensee's analysis and finds the licensee's cerclusions acceptable. Thus, the staff concludes that the requested technical specification changes satisfy the applicable regulatory requirements anc are acceptable.

EMERGENCY CIRCUMSTANCES These TS changes are being issued before the expiration of the notice period to prelcude an unnecessary delay in plant startup from the current outage. On April 13, 1987, the licensee received a letter from Westinghouse informing them of a poter.tial issue concerning the time required to change charging pump suction from the VCT to the RWST following a postulated steam line break event (see introduc. tion section). The licensee completed their own evaluation and detennined on April 15, 1987 that the issue was indeed applicable to Callaway Plant and that changes to the TS were needed for plant startup from the current outage. The licensee then promptly notified the staff (on April 15, 1987) of the situation at Callaway and followed-up with a license amendment application dated April 16, 1987. .

The Commission has determined that emergency. circumstances exist in that swift action is necessary to avoid a delay in startup not related to safety and finds that for the reason stated above, and an accelerated outage schedule, emergency circumstances exist.

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In connection with a request indicating an amergency, the Commission expects its licensees to apply for license amendments in a timely fashion. However,  !

with this consideration in mind,.it has been determined that a circumstance has arisen where the licensee and the Commission must act quickly, and~the licensee has made a good effort to make a timely application.

FINAL NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION In accordance with 10 CFR 50.92, the Commission may make a final determination I that a license amendment involves no significant hazards consideration if operation of the facility in accordance with the amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or. (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. ' The infonnation in this section provides the staff's evaluation of this license amendment against the three criteria.

The staff has confirmed the basis of the no significant hazards findings described in the notice published in the Federal Register on April 22, 1987 (52FR13367). The amendment does not involve a significant increase in the probability or consequences of accidents previously evaluated. An increase in the acceptance criterion for the ESF response time is acceptable since the

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evaluation of the in: pact of the increasec delay on the steam line break event demonstrated that the departure from nucleate boiling design basis is stili met. The conclusions in the FSAR remain valid.

The amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. There are no new failure modes associated with this proposed change, as no design changes have been made. No new accident is created because the same equipment is assumed to perform in the same manner as before. Therefore, an increase in the ESF response times for high containment pressure, low pressurizer pressure, and low steam line pressure does not create the possibility of an accident or malfunction of a

, 'different type than any evaluated previously in the safety analysis report.

The amendment does not involve a significant reduction in a margin of safety.

There is no impact on the consequences on protective boundaries, and all j acceptance criteria in the analysis of record are still met. Therefore, the safety limits will still be met.

Therefore, the staff concluded that:

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(1) Operation of the facility in accordance with the amendment would not significantly increase the probability or consequences of an accident previously evaluated.

(2) Operation of the facility in accordance with the amendment would not create the possibility of a new or different kind of accident from L

any accident previously evaluated.

(3) Operation of the facility in accordance with the amendment would not involve a significant reduction in a margin of safety.

Therefore, we conclude that the amendment to Facility Operating License No.

NPF-30 to support operation of the Callaway Plant, Unit 1, which revises Table  ;

3.3-5 of the TS to increase the response time by 15 seconds for certain SI  !

functions involves no significant hazards considerations, j STATE CONSULTATION In accordance with the Commission's regulations, consultation was held with the State of Missouri by telephone. The State expressed no concern, either from the standpoint of safety or of our no significant hazards consideration determination.

ENVIRONMENTAL CONSIDERATION

' This amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part.20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual g

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e or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR -

51.22(c)(9). Pursuant to 10 CFP. 51.22(b) no environmental impact statement er environmental assessment need be prepared in connection with the issuance of this amendment. ,

C0hCLUSION We have concluded, based on the considerations discussed above, that: (1)

there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: T. Alexion, PD33, DRSP, E. Lantz, SRXB, DEST Dated: May 4, 1987 l

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