ML20246M344
| ML20246M344 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/04/1989 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20246M308 | List: |
| References | |
| NUDOCS 8905190092 | |
| Download: ML20246M344 (21) | |
Text
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TABLE 3.3-2 I-REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME 1.
Manual Rea'ctor Trip NOT APPLICABLE 2.
Power Range, Neutron Flux 1 0.5 seconds
- 3.
Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE 4.
Power Range, Neutron Flux, High Negative Rate 1 0.5 seconds
- I 5.
Intermediate Range, Neutron Flux NOT APPLICABLE 6.
Source Range, Neutron Flux NOT APPLICABLE (Below P-10) 05 7.
Overtemperature AT 1 Adf seconds
- 5'S
{<Aefsecondspl 8.
Overpower AT 9.
Pressurizer Pressure--Low
-< 2.0 seconds (Above P-7)
(
- 10. Pressurizer Pressure--High 1 2.0 seconds
- 11. Pressurizer Water Level--High NOT APPLICABLE (Above P-7)
- 12. Loss of Flow - Single Loop (Above P-8) 1 1.0 seconds
- 13. Loss of F1cw - Two Loop 1 1.0 seconds (Above P-7 and below P-8)
- 14. Steam Generator Water Level--Low-Low
< 2.0 seconds
~~
(Loop Stop Valves Open) t
- 15. Ste n/Feedwater Flow Mismatch and Low Steam Generator Water Level 110T APPLICABLE Imou s#*
- 16. Undervoltage-Reactor Coolant Pumps
< 1.5 ser.onds (Above P-7) y 'E E
- 17. Underfrequency-Reactor Coolant Pumps 1 0.9 seconds (Above P-7) i So gac gh
- Neutron detectors are exempt from response time testing.
Response
time shall be measured from detector output er input of first o<
2 electronic component in channel.
_ __ y '
m'a-BEAVER VALLEY - UNIT 2 3/4 3-8
?YoPoseci idord ing
ATTACHMENT D Safety Analysis Beaver Valley Power Station Unit No. 2 Proposed Technical Specification Change No. 25 Description of Amendment Request:
The proposed amendment would revise the reactor trip system overtemperature AT (OTAT) and overpower AT (OPAT) response times listed in Table 3.3-2 from 4.0 seconds to 5.5 seconds.
This change is based on a Westinghouse Electric Corporation evaluation of the Unit 2 FSAR accident analyses requirements for a
reactor trip initiated from the OTAT/OPAT protection systems.
This evaluation determined that sufficient margin exists to increase the reactor trip time delays as proposed in this change while continuing to meet the FSAR accident analyses requirements.
Discussion:
The current design basis requirements for the OTAT/OPAT reactor trip total response
- time, as provided in FSAR Table 15.0-4, is 6.0 seconds.
This total response time is defined as the delay from when the temperature in the reactor coolant loop exceeds the trip setpoint until the rods are free to fall into the core.
Included in this time response is the RTD bypass manifold fluid transport and heatup time delays along with RTD sensor time i
- delays, channel time delays and the reactor trip breaker and rod gripper release times.
By definition, the Technical Specification reactor trip response time includes the time from when the monitored parameter exceeds its setpoint at the sensor until loss of gripper coil voltage.
Therefore, for the Technical Specifications the RTD bypass manifold fluid transport and heatup times are not included in the response time requirement.
Two seconds is assigned for this portion of the total time delays which results in a 4.0 second time requirement for the Technical Specifications.
The OTAT/OPdT reactor trip time respense is initially verified during the startup test program and subsequently verified by the Technical Specifications surveillance program.
An evaluation of the results of the initial startup test indicated that the 6.0 second time response assumed in the FSAR was not met.
A Justification for Continued Operation (JCO) was developed by westinghouse to prcvide a technical basis for continued operation of Unit 2
pending determination of the cause and resolution of the design deviation.
This JCO concluded that sufficient margin was available in the FSAR accident analyses such that the analyses conclusions would remain valid for an OTAT/OPAT reactor trip time delay up to 7.5 seconds versus the 6.0 seconds (see Attachment B-1 for a detailed discussion of this JCO).
1
Attachment B j
Proposed T. S. Change No. 25 Page 2 The evaluation of the startup test program results was inconclusive as to the cause of the additional time delay and whether the Technical Specifications response time requirement for the OTAT/OPAT functions would be satisfied.
As part of the continuing evaluation to determine the cause of the additional delay, the response time of the Reactor Coolant System (RCS) protection RTD's were recently measured using the loop current step response test method which was recommended by Westinghouse as the most appropriate test method for the Unit 2 RTDs.
The results of this testing indicated that the RTDs were responding slower than expected.
In
- addition, during the Westinghouse evaluation of our OTAT/OPAT response
- time, it was concluded that a more restrictive value than the 4.0 seconds provided in the Technical Specification is appropriate for the channel response time testing method used to ensure that the assumed time responses in our FSAR remain valid.
While the currently measured OTAT/OPAT channel response times meet the current 4.0 second Technical Specification time response requirements, they are not within the more restrictive limits provided by Westinghouse for our response time test method.
To account for the additional RTD time response delay and the reduced time response acceptance criteria provided by Westinghouse, it is necessary to include the additional time response margin provided by the 1.5 second increase in the assumed FSAR OTAT/OPAT reactor trip response time evaluated in the attached JCO.
The proposed Technical Specification change therefore increases the OTAT/OPAT response time from 4.0 to 5.5 seconds.
See Attachment B-2 for a further discussion on the basis for the proposed change.
With the additional time response margin provided by the revised l
FSAR assumed time response of 7.5 seconds, the currently measured channel time response will be within the Westinghouse acceptance criteria for demonstrating that the FSAR assumed time response is met.
Since the proposed revision to the OTAT/OPAT reactor trip l
time response will continue to ensure that the conclusions of the FSAR Safety Analyses remain valid, this change is considered safe.
It should be noted that this proposed change and the associated OTaT/OPAT reactor trip time response design basis change is required for the Unit 2 second fuel cycle only.
During the second refueling
Along with this design change, the design basis OTAT/OPAT reactor trip response time would be returned to 6.0 seconds.
Associated Technical Specification changes required for this design
- change, including revised OTAT/OPAT trip time
- response, will be submitted at a later date.
1
~Attcchment B-1
['
\\
,Proposad T.S. chengs No. 25 g
J NJ DLW-89-636 Westinghouse Energ Systems Nuclear and banced Electric Corporation Ton @p Dnnse Box 355 Pmsburp Pennsylvania 15230-0355 NS-0PLS-OPL-I-89-253 April 28, 1989 P.O.: D-071625 G.O.: PG 53165 Ref: DLW-89-537 Mr. N. R. Tonet Manager, Nuclear Engineering Duquesne Light Company Beaver Valley Power Stations P. O. Box 321 Shippingport, PA 15077 DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNIT N0. 2 JC0 FOR INCREASED RTD RESPONSE TIME
Dear Mr. Tonet:
e, Please find attached a revised justification for continued operation (JCO). This JC0 supersedes the previous JC0 provided via reference, and is to assess the impact of a measured RTD sensor time constant of 3.5 seconds for cycle 2 operation of the Beaver Valley Unit 2 plant. An l
increase in the RTD sensor time constant of 1.5 seconds could result in an increase of the total OTDT/0PDT reactor trip delays of 1.5 seconds. This i
JC0 also justifies that the OTDT/0PDT setpoints can be reset at the Tcchnical Specification value and no administrative penalty is added to the Cycle 2 operation in conjunction with a 1.5 second increase in RTD response time, and thus the conclusions of the FSAR analyses are still valid.
This JC0 concludes that the impact en the safety analysis design basis calculations of a 1.5 second increase in OTDT/0PDT reactor trip delay can be accommodated in safety analysis margins, and thus the FSAR thapter 15 and 6 conclusions still remain valid for cycle 2 operation.
Sections of the FSAR which do not reflect the increased response time are identified.
The effort required to provide safety analysis upgrades te reflect an increase in the OTDT/0PDT instrument channel response time from 4 seconds to 5.5 seconds has been provided in Reference 4.
It is notrf. that DLC0 intends to use this JC0 to support a relaxation of the Techr.ical Specification for OTDT/0PDT response time for Cycle 2 operation.
The RTD Bypass Elimination modification is scheduled for the Cycle 2/3 refueling outage.
In support of this plant modification, the Technical Specification OTDT/0PDT response time will be revised to reflect a total
DLW-89-636 NS-0PLS-0PL-I-89-253 Page 2 instrument channel response time of 6.0 seconds, which is directly comparable with the FSAR safety analysis assumptions. Therefore, this JC0 is intended to serve as an interim justification for continued operation until the RTD Bypass Elimination is in place.
If you have any questions on the information presented in this letter, please contact Aaron Jen at (412) 374-4357 or the undersigned.
Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION
/kjw
~
'J. N. Steinmetz, rojects Nanager i
Operating Plant Projects Y. A. Jen Attachments - JCO, SECL and Tech. Spec. mark-up for an increased RTD response time cc:
H. R. Tonet, IL, IA J. O. Crockett, IL J. D. Sieber, IL S. C. Ferr.er, IL W. S. Lacey, IL K. D. Grada, IL, IA G. A. Kammerdeiner, IL R. Zabowski, IL, IA T. P. Noonan, IL S. A. Nass, IL, IA C. D. Schmitt, IL P. W.
Dearborn,
IL NERU Records Section, 3L, 3A
!s
. ~., '
e l
ATTACHMENT TO DLW-89-636 SAFETY EVALUATION CHECKLIST 1.5 SECOND INCREASE IN RTD SENSOR TIME CONSTANT
SECL-89 700 Customer Reference No(s).
NDlMNE:4737 Westinghouse Reference No(s).
WAF NO.: B-03781 Rev. 4 WESTINGHOUSE NUCLEAR SAFETY EVALUATION CHECK LIST
- 1) NUCLEAR PLANT (S) Beaver Vallev Unit 2
- 2) CHECK LIST APPLICABLE TO:
1.5 second RTD Resoonse Time Increase For Cvele 2 Ooeration
- 3) The written safety evaluation of the revised procedure, design change or modification required by 10CFR50.59 has been prepared to the extent required and is attached.
If a safety evaluation is not required or is incomplete for any reason, explain on Page 2.
Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation performed.
CHECK LIST - PART A - 10CFR50.59 (a) (1)
(3.1) Yes.jL No A change to the plant as described in the FSAR?
(3.2) Yes No 1 A change to procedures as described in the FSAR?
(3.3)
Yes,__. No 1 A test or experiment not described in the FSAR?
(3.4) Yes l No _ A change to the plant technical specifications (See note on Page 2)
- 4) CHECK LIST - PART B - 10CFR50.59 (a) (2)
(Justification for Part B answers must be included on page 2.)
(4.1) Yes_ N01 Will the probability of en accident previously evaluated in the FSAR be increased?
(4.2) Yes No 1 Will the consequences of En accident previously evalcated in the FSAR be increased?
(4.3)
Yes___ Nc 1 May the possibility of an accident which is different than any already evaluated in the FSAR be created?
(4.4) Yes No 1 Will the probabili!.y of a en1 function of equipment important to safety previously evaluated in the FSAR be increased?
(4.5) Yes_ No.JL Will the consequences of a ma.1 function cf equipment important to safety previously evaluated in the FSAR be increased?
(4.6)
Yes___. No1 May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created?
(4.7) Yes No X Will the margin of safety as defined in the bases to any technical specification be reduced?
PAGE 1 0F 2
SECL NO.89-700 If the answers to any of the above questions are unknown, indicate under
- 5) REMARKS and explain below.
If the answer to question 3.4 of Part A or any of the above questions in i
Part B cannot be answered in the negative, based on written safety evaluation, the change review would require an application for license amendment as required by 10CFR50.59(c) and submitted to the NRC pt.rsuant to 10CFR50.90.
- 5) REMARKS:
The following sumarizes the justification upon the written safety evaluation (1) for answers given in ? 4) Part A and Part B.
See attached justification for egd nued coeration (JCO) for details (1) Reference to document (s) containing written safety evaluation:
FOR FSAR UPDATE Section:
Pages:
Tables:
Figures:
Reason for / Description of Change:
.See attachment - summary of FSAR imoacts See attached mark-uo of Tech. Soec. Table 3.3-2 SAFETY EVALUATION APPROVAL LADDER:
Prepared by (Nuclear Safety):
Y. A. Jan Date: W -Ff l
C/
l Coordinated with Engineer (s):
ON FILE
_._Date:.
Coordinated Group Manager (s):
ON FILE Date:
Nuclear Safety Group Manager:.
R. J. Sterdis Date: f-3-89
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n PAGE 2 0F 2
{
l l
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I ATTACHMENT TO DLW-89-636 BEAVER VALLEY UNIT 2 JUSTIFICATION FOR CONTINUED CYCLE 2 OPERATION 3.5 SECOND RTD SENSOR TIME CONSTANT i
This justification for continued operation (JCO) is written to provide technical justification for an increase in the RTD sensor time constant from 2 seconds to 3.5 seconds (e.g. 1.5 seconds increase) for Beaver Valley Unit 2.
The scope of this JC0 is the FSAR Chapter 15 and 6 (Reference 1) transients within Westinghouse technical cognizance, the steamline break mass and energy releases calculated for equipment qualification outside containment (Reference 2) and subsequent safety evaluations performed with respect to these bases (Reference 5, 6, 7).
This JC0 is applicable for N-loop operation only.
~
For certain safety analyses, assumptions are made regarding the response times of the RTDs. As documented in Table 15.0-4 of Reference 1, transients which assume automatic protection from the overtemperature and overpower delta-T (OTDT/0PDT) functions account for a total trip time delay of 6 seconds.
This time delay includes the fluid transport time through the RTD bypass piping, the RTD time response and trip circuit channel electronics delay. The 6 seconds accounts for the time the temperature in the coolant. loops exceeds the trip setpoint to the time when the control rods are free to drop into the core. An increase in the measured RTD sensor time constant could result in an increase of the total RTD response time by 1.5 seconds. This JC0 assumes a worst case effect on the reactor trip delay for the OTDT/0PDT functions. However, no impact on the safety analyses would occur if the remaining portions of the trip channel (electronics delays and control rod release) could still substantiate the current Tech Spec limit for OTDT/0PDT instrument channel response time of 4 seconds..
l This JC0 addresses an increase in the safety analysis OTDT/0PDT trip delay', as indicated in Table 15.0-4, from 6 seconds to 7.5 seconds for cycle 2 operation (e.g. additional 1.5 seconds increase in RTD sensor time constant). Only those transients which assume OTDT/0PDT protection are potentially affected by an increase in the RTD sensor time constant.
It is assumed that the OTDT and OPDT trip setpoints are consistent with the current Technical Specification as documented in Table 2.2-1, Notes 1 through 6 (Reference 2).
FSAR Chapters 15 and 6 As documented in Reference 1 and 7, the following FSAR transients assume protection from either the overtemperature or overpower delta-T setpoints.
Assumed Protection FSAR Section Accident Function i
15.2.3 Turbine Trip OTDT with Pressurizer Control, Minimum Feedback 15.4.2 Uncontrol' led RCCA Bank OTDT Withrawal at Power 15.4.6 CVCS Malfunction that results OTDT in a decrease in the baron concentration in the RCS (Boron Dilution), Mode 1 - Manual Rod Control The effect of a 1.5 second increase in total RTD response time on these transients is a delay in the time of rod motion of 1.5 seconds from the transient times currently assumed for the above FSAR accidents. - _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _
The referenced Chapter 15.2.3 Turbine Trip analysis was examined to determine the rate of decrease in DNBR just prior to reactor trip.
Conservative extrapolation of the rate of decreasing DNBR was made to estimate the DNBR for an additional 1.5 second delay in rod motion.
It is concluded that the effect on transient minimum DNBR is minimal and the safety analysis DNBR acceptance criterion continues to be met. The transient peak pressurizer pressure is limited by the availabity of the pressurizer relief valves and occurs in the FSAR transient prior to rod motion. Therefore, peak pressure would not be affected by a 1.5 second increase in the RTD response time. Thus, the conclusions of FSAR Section 15.2.3 remain valid for a total trip delay of 7.5 seconds.
As documented in FSAR Figures 15.4-10, 11 and 12, some of the Section 15.4.2 RCCA Bank Withdrawal at Power cases have assumed OTDT protection.
l These cases of various power levels and reactivity insertion rates have I
been evaluated for an increase in the RTD response time and, therefore, delay of control rod motion, of 1.5 seconds. The rate of change in DNBR prior to time of rod motion has been used to conservatively estimate the impact on minimum DNBR of a 1.5 second delay in rod motion.
It is concluded that the effect is minimal and that the DNBR remains above the minimum safety analysis limit. Therefore, the conclusions of FSAR Section 15.4.2. remain valid for an increase in the trip delay of 1.5 seconds.
i The F5AR Chapter 15.4.6 8 von Dilution transient is analyzed to show that j
adequate time exists for operator action to terminate an inadvertent dilution prior to the ' loss of shutdown margin. The Mode 1 (power operation) case for manual rod control assumed reactor trip on OTDT. The j
analysis of record results (Reference 7) indicate that there are at least 16 minutes available for operator action following reactor trip prior to a return to criticality. The increased RTD response time would decrease the I
available operator action time by 1.5 seconds. The impact of this change J
is nearly imperceptible and does not affect the transient results and conclusions as presented for the analyris of record (Reference 7). _.
SGTR Evaluation OTDT/0PDT protection is assumed to be available for the Steam Generator Tube Rupture (SGTR) analysis; however, a review of the FSAR Steam Generator Tube Rupture analysis reveals that reactor trip occurs due to the low pressurizer pressure trip signal and thus the 1.5 second increase in the RTD response time will not change the reactor trip signal nor reactor trip time. Since the reactor trip signal and reactor trip time assumed in the FSAR do not change, the primary to secondary break flow and atmospheric steam release results are unaffected. Therefore, the reported offsite doses for the Steam Generator Tube Rupture (SGTR) event do not change. Additionally, the conclusion that the DNB limits are met for the SGTR event remains valid. Therefore, it is concluded that the results reported for the SGTR event in FSAR Section 15.6.3 remain valid.
LOCA Evaluation None of the FSAR LOCA analyses or LOCA related design calculations are affected by an increase in the OTDT/0PDT instrument channel response time of 1.5 seconds.
In conclusion, only those FSAR transients identified above are affected by the RTD response time increase. The balance of the FSAR Chapter 15 and 6 transients do not assume OTDT/0PDT protection and are, therefore, unaffected. Additionally, response time does not affect the FSAR Figure 15.0-1 illustration of OTDT/0PDT protection of the core thermal limits.
WCAP 10961. Rev. 1 Reference 2 documents safety analysis calculations done in addition to the FSAR to support equipment environmental qualification outside containment. The steamline break mass and energy releases applicable to Beaver Valley Unit 2 are reported in Reference 2 under the Category 4 results. As indicated in Tables III.E-1, 3 and 5, some of the cases of various break sizes and power levels have assumed OPDT protection. _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ -
Westinghouse sensitivity studies of the impact of reactor trip delays on the Reference 2 results indicate that a delay in rod motion of 1.5 seconds has no significant effect on the reported peak enthalpies or mass releases. Therefore, the Reference 2 data presented for Category 4 plants continues to be applicable for Beaver Valley Unit 2 assuming an increase in reactor trip delay of 1.5 seconds.
WCAP-9226. Rev. 1 Generic 3 loop plant studies were performed in Reference 3 to arrive at conclusions concerning safety analysis methodology for reactor core response to excessive secondary steam releases. A specific steamline break size was examined for hot full power initial conditions to determine core response assuming OPDT protection. The WCAP concludes that this case is not more limiting than the case examined for Section 15.1.5 of the FSAR. This conclusion can be expected to remain valid for increases in RTD response time within the limit that the FSAR transients which assume OTDT/0PDT protection are not significantly impacted. On this basis, the conclusions of Reference 3, as they apply to Beaver Valley Unit 2, are considered to remain valid for an increase trip delay of 1.5 seconds.
Precautions. Limitations and Setooints Document Consistent with the Precautions, Limitations and Setpoints document for Beaver Valley Unit 2 (Reference 2), the OTDT K1 and OPDT K4 coefficients should be administrative 1y reduced by 3% until the time that plant trip test results are evaluated to meet all applicable acceptance criteria. A Westinghouse safety evaluation was performed to justify the relaxation of the E setpoint penalty to 2.5% in conjunction with a 1.5 second increase in RTD response time during Cycle 1 operation (Reference 9). The Safety evaluation provided herein has been performed assuming no administrative penalty to the Tech Spec OTDT/0PDT setpoints (Reference 4). The conclusions presented in this safety evaluation are not affected if the 1.5% OTDT/0PDT administrative penalty is removed and the RTD response time is less than or equal to 3.5 seconds. ___________________ _______-______- - _
Conclusion In conclusion, the Beaver Valley Unit 2 safety analysis design bases as defined in References 1, 2, 3, 5, 6 and 7 have been evaluated for a measured RTD sensor time constant of 3.5 seconds and the existing OTDT/0PDT protection system setpoints as defined in the current Technical Specifications (Reference 4). As a worst case, an assumed corresponding increase in the safety analysis assumption for total reactor trip delay
. for the OPDT/0 TDT protection functions of 1.5 seconds has been evaluated.
Only those transients which assume protection from the OTDT/0PDT functions are affected. This evaluation has identified these transients and has determined that the effect is minimal and can be accommodated within the margins of the safety analyses. The conclusions of References 1, 2, 3, 5, 6 and 7 remain valid for Cycle 2 operation and no reanalyses are required.
Affected Documentation A change in the total OTDT/0PDT reactor trip delay from 6 to 7.5 seconds is a change to the plant as described in the FSAR. Within FSAR Chapters 15 and 6 Table 15.0-4, Table 15.2-1 (Item 1), Figure 15.2-1, Figure 15.2-2, and Figures 15.4.10-12 do not reflect a 7.5 second delay in the OTDT/0PDT reactor trips.
The RTD sensor time constant is one of the components of the Beaver Valley Unit 2 Techncial Specification (Reference 4) total time response surveillance requirement for the OTDT/0PDT protection functions. The requirement documented in Table 3.3-2 of page 3/4 3-8 is "I 4 seconds".
An increase in the ATD sensor time constant of 1.5 reconds suggests the need for a relaxation in this specification to "I 5.5 seconds". A mark-up of Reference 4 Table 3.3-2 is provided.
The associated Technic.a1 Specification bases specify a relationship betw en the safety analycis assumptions and the response time requirement. A broad interpretation of the bases text has been assumed such that evaluations performed subsequent to the FSAR safety evaluations are encompassed in the terms " safety analyses" and " safety analysis assumptions"..
References:
1.
Updated Final Safety Analysis Report, Beaver Valley Power Station.
Unit 2, Revision 0, April, 1988.
2.
WCAP-10961, Rev.1, Steamline Break Mass / Energy Releases for Equipment Qualification Outside Containment, Report to the Westinghouse Owners Group High Energy Line Break /Superheated Blowdowns Outside Containment Subgroup, October 1985.
3.
WCAP-9226, Rev. 1, Reactor Core Response to Excessive Secondary Steam 1
Releases, January 1978.
4.
Beaver Valley Power Station, Unit 2, Technical Specifications, Amendment 12, January 1989.
5.
890L*-G-0016, Reload Safety Evaluation, Beaver Valley Power Station, Unit 2 Cycle 2, March 1989.
g 6.
89DL*-G-0034, Evaluation of Beaver Valley Power Station Unit 2, Cycle 2 Extended Operation, April 1989.
7.
DLW-89-601, Beaver Valley Power Station Unit 2, Cycle 2 RSE FSAR Updates, March 1989.
8.
Beaver Valley Power Station Unit I, Precautions, Limitations and Satpoints for Nuclear Steen Supply Systems, Revision 1, June 1983, l
and subsequent amendments through January 1985, DMW-D-4715.
9.
DLW-68-724, Duquesae Light Co., Beaver Valley Power Station Unit No.
2, JC0 for Increased RTD Response Time and Reduced OTDT/0PDT Reactor Trip Setpoints, September 1988.
- 10. DLW-89-537, Duquesne Light Co., Beaver Valley Power Station Unit No.
2, JC0 for increased RTD Response Time and Reduced OTDT/0PDT Reactor Trip Setpoints, January 1989. __-_ ___-__.
?1 L.
9
' ATTACHMENT TO DLW-89-636 TECHNICAL SPECIFICATION MARKUP i
1.5 SECOND INCREASE IN RTD SENSOR TIME CONSTANT f..
q TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME 1.
Nanual Reactor Trip NOT APPLICABLE
^
2.
Power Range, Neutron Flux
< 0.5 seconds
- 3.
Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE 4
Power Range, Neutron Flux, High Negative Rate 1 0.5 seconds" 5.
Intermediate Range, Neutron Flux NOT APPLICABLE 6.
Source Range, Neutron Flux NOT APPLICABLE (Below P-10)
!L5 1
7.
Overtemperature AT SfetF seconds *
$Nseconds*
8.
Overpower AT stGV 9.
Pressurizer Pressure--Low
< 2.0 seconds
~
(Above P-7)
- 10. Pressurizer Pressure--High
$ 2.0 seconds
- 11. Pressurizer Water Level--High NOT APPLICABLE (Above P-7)
- 12. Loss of Flow - Single Loop (Above P-8)
$ 1.0 seconds
- 13. Loss of Flow - Two Loop
< 1.0 seconds (Above P-7 and below P-8)
- 14. Steam Generator Water Level--Low-Low
< 2.0 seconds
~
(Loop Stop Valves Open)
- 15. Steam /Feedwater Flow Mismatch and Low Steam Generator Watee level NOT APPLitA8LE
- 16. Undervoltage Reactor Ccolant Pumps
< 1.5 seconds
~
(Above P-7)
- 17. Underfrequency-Reactor Coolant Pumps 5 0.9 seconds (Above P-7)
" Neutron detectors are exempt from response time testing. Response time shall be measured from detector output er input of first electronic component in channel.
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SEAVER VALLEY - UNIT 2 3/4 3-8
1. D.
1 Attachment'B-2
(~
Propossd T.S. Changa No. 25 BVSET-89-013 FROM:
Westinghouse Support Engineering Team 393-5611 DATE:
March 17, 1989
REFERENCES:
1 ) DLW-89 -537 January 31, 1989
- 2) DLW-88-724 September 2, 1988
- 3) DLW-88-758. October 20, 1988
SUBJECT:
Description of Tech Spec Change for OTDT/OPDT Instrumentation Channel Response Time To:
R. Zabowski, Admin. Bldg.
cc:
N. R. Tonet. ERF K. D. Grada, Admin. Bldg.
R. E. Martin, SEB-3 J.V. Vassello, Admin. Bldg.
S. A. Nass ERF H.J. Kahl, SEB-3 G. L. Shildt, SOSB-6 Via Reference 1. Westinghouse provided a revised Justification for Continued Operation (JCO) of BVPS-2 with an OTDT/OPDT reactor trip.
time delay of 7.5 seconds.
Reference 1 included a Technical Specification markup to reflect the associated increase in OTDT/OPDT instrumentation channel response time to 5.5 seconds.
Subsequent to receipt of Reference 1, you requested an expanded description / basis for the Technical Specification change that could' be used in the letter to be issued to the NRC.
In response to your request, I have developed the attached description based on information provided by Westinghouse via References 1, 2 and 3.
In this description. I have focused on the Technical Specification aspect of this overall topic and have provided a basis for the Technical Specification change while acknowledging that questions regarding RTD sensor response time testing methods (i.e., process noise vs. loop current step response) and OTDT/OPDT instrumentation channel response time acceptance criteria methodology (i.e.,
startup program vs. surveillance program) remain to be resolved.
All information in the Attachment is documented in References 1, 2 and 3 with the exception of the statement in the fourth paragraph that compliance with Technical Specificat. ions was verified in the Spring of 1988 after sensor response time testing.
This statement is based on information provided by DLC Operations / Testing personnel that verification was performed after sensor response time testing using accepted practice, i.e..
use of process noise test results for sensor response time and use of 4 second acceptance criteria from the Technical Specifications.
Please advise if there are any questions on the attached information or if you desire additional details.
Ralph Surman Beaver Valley SET
4 ATTACHMENT TO BVSET-89-013 Page 1 of 2 i
Description / Technical Basis for Revision to Technical Specification for OTDT/OPDT Instrumentation Channel Resconse Times Beaver Valley Unit 2 Technical Specification 3.3.1.1 defines the channel, interlock and response time requirements for the reactor trip system instrumentation.
Instrumentation channel response time requirements are defined in Technical Specification Table 3.3-2.
Compliance with these response time requirements is necessary to ensure that Beaver Valley Unit 2 reactor trip time delays are in compliance with FSAR accident analyses requirements.
The FSAR accident analyses requirements for a reactor trip initiated from the Overtemperature/ Overpower Delta Temperature (OTDT/OPDT) protection system require that the reactor trip time delay be less than or equal to 6 seconds.
This response time requirement applies to the protection system as modeled in the FSAR. including delays due to fluid transport in the Reactor Coolant Loop (RCS) bypass manifold, heatup of the bypass manifold and response time of the OTDT/OPDT instrumentation channel.
Technical Specification Table 3.3-2 identifies that portion of the overall OTDT/OPDT reactor trip time delay assigned to the OTDT/OPDT instrumentation channel.
The instrumentation chann.el response time requirement is less than or equal to 4 seconds.
The Beaver Valley Unit 2 Startup Program included several tests which together were intended to verify that the OTDT/OPDT reactor trip time delay satisfies the design basis requirement of less than or equal to 6,.0 seconds.
These tests yielded acceptable results with the exception of the Plant Trip From 1004 Power test which was postponed to be performed prior to the first refueling.
Pending receipt of results from the plant trip test and confirmation that the OTDT/OPDT reactor trip time delay satisfied design basis requirements. Duquesne Light Company continued to operate Beaver Valley Unit 2 with conservative OTDT/OPDT turbine runback and reactor trip setpoints consistent with Startup Program requirements.
2 Sensor response time testing was performed at Beaver Valley Unit in the Spring of 1988 to establish baseline data for the Surveillance Program.
This data in combination with the OTDT/OPDT electronics response time data from the startup tests was subsequently evaluated to determine if OTDT/OPDT channel response times were in compliance with Technical Specifications.
This evaluation confirmed that Technical Specification requirements were satisfied.
Plant data needed to evaluate the plant trip startup test acceptance criteri,a was. obtained from inadvertent reactor trips of
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ATTACHMENT TO BVSET-89 -013 Page 2 of 2 Beaver Valley Unit 2 that occurred on July 27 and September 20, 1989.
Evaluation of the plant trip data by Westinghouse indicated that the OTDT/OPDT reactor trip time delay on Beaver Valley Unit 2 was less than or equal to 7.5 seconds in lieu of 6.0 seconds.
Upon arriving at this conclusion, the FSAR accident analyses were reviewed to determine the impact of this design deviation.
It was concluded that margin available in the accident analyses was sufficient that the FSAR accident analyses conclusions remained valid for an OTDT/OPDT reactor trip time delay of 7.5 seconds.
A Justification for Continued Operation (JCO) was developed by Westinghouse to provide a technical basis for continued operation of Beaver Valley Unit 2 pending determination of the cause and resolution of the design deviation.
The Beaver Valley Unit 2 Startup Program was inconclusive relative to the cause of the additional time delay and whether the additional time delay was due to the OTDT/OPDT instrumentation channel.
Startup tests confirmed that the delay was not due to bypass manifold fluid transport time or the electronics portion of the OTDT/OPDT instrumentation channel.
However, the startup tests,
inconclusive with respect to whether the additional time delay were was due to bypass manifold heatup or RTD sensor response.
The separate sensor response time testing performed on Beaver Valley Unit 2 as part of the Surveillance Program indicated that the OTDT/OPDT instrumentation channel response times were in compliance with Technical Specifications.
A comparison of the Startup Program and Surveillance Program was then performed to identify any differences in test methodology or acceptance criteria.
Evaluation is continuing to resolve several apparent differences prior to performing surveillance tests of the OTDT/OPDT instrumentation channels at the first refueling outage.
Independent of the results of this ongoing evaluation, it is appropriate at this time to revise Technical Specification 3.3.1.1 for OTDT/OPDT instrumentation channel response times to be consistent with the analytical and technical basis of the JCO under which Beaver Valley Unit 2 is presently operating.
The JCO was developed by Westinghouse consistent with accepted practice on Westinghouse designed plants which attributes the additional time delay to the RTD sensor element of the OTDT/OPDT instrumentation channel.
Since the JCO attributed the additional 1.5 second delay to the OTDT/OPDT instrumentation channel, the associated Technical Specification needs to be revised from 4.0 seconds to 5.5 seconds.
Surveillance testing to be performed at the first refueling outage will verify that the OTDT/OPDT instrumentation channel response times meet this design basis requirement consistent with the JCO that defines the technical basis for Beaver Valley Unit 2 operation.
The portion of the OTDT/OPDT reactor trip delay time assigned to bypass manifold fluid transport and heatup remains at 2 l
seconds.
Delay times due to these elements of the OTDT/OPDT reactor trip time delay are not included in the instrumentation channel surveillance tests.
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r ATTACHMENT C
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.o No Significant Hazard EvaluOtion Proposed Technical Specification Change No. 25 Basis for Proposed No Significant Hazards Consideration Determination:
The Commission has provided standards for determining whether a
significant hazards consideration exists (10 CFR 50.92(c)).
A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (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 proposed change does not involve a_ significant hazards consideration because:
1.
The effect of increasing the OTAT/OPAT trip time response has been evaluated for all FSAR accident analyses which assumes protection from OTAT/OPaT trip functions.
This evaluation has determined that the proposed increased time response can be accommodated within the margins of the safety analyses.
Since the conclusions of the safety analyses will remain valid this change would not involve a
significant increase in the probability or consequences of an accident previously evaluated.
2.
The proposed change does not involve any plant equipment or operating configuration changes within the plant.
Therefore the probability of an accident or a malfunction of a different type than previously evaluated would not be created.
3.
The proposed change will not involve a significant reduction in a margin of safety.
The effect of the proposed OTAT/OPAT trip time response on the Turbine Trip and Uncontrolled Rod Withdrawal at Power FSAR accident analyses have been evaluated.
This evaluation determined that the effect on transient DNBR is minimal and the safety analysis DNBR acceptance criteria will continue to be met.
Based on the above considerations, it is proposed to characterize the change as involving no significant hazards consideration.
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