ML20235M930
| ML20235M930 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 06/30/1987 |
| From: | Baxter D, Branson G EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20235M913 | List: |
| References | |
| CON-FIN-A-6824 EGG-NTA-7505, NUDOCS 8707170378 | |
| Download: ML20235M930 (23) | |
Text
_ _ _ _ _ _ _ - _ _ _ _ _ _
+
EGG-NTA-7505 June 1987
~
INFORMAL REPORT i
-Idaho;l L
^ Nationa/;
EVALUATION OF BRAIDWOOD STATION UNIT 1 Engineering TECHNICAL SPECIFICATIONS
. Laboratory-in
}
Managed-
. by the U.S:
D. E. Baxter
. Department G. L. Branson
. of Energy
~.
EGcG,..
p Prepared for the
- " '* "'ss'e"'C,,,""ll U.S. NUCLEAR REGULATORY COMMISSION No. DE-AC07-76tD0tS70 71 0379 9,9g9; p
UCCK 05000%$
l'DR
4 l
i 1
l 1
DISCLAIMER l
l This book was prepared as an account of work sponsored bv en agency of the United States Government. Neither the United States Government e/ar any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any l
information, apparatus, product or process disclosed, or represents that its use would
.i not infrnge pnvately owned nghts. References herein to an/ specific commercial product, process, or service by trade name, trademark, manuf acturer, or otherwise, does not necessanly constitute or imply its endorsement, recomtnendation, or favoring by the United States Government or any agency thereof. The siews and opinions of authors expressed herein do not necessanly state or reflect those of the United States Government or any agency thereof.
l l
l Y
l
- EGG-NTA-7505
[
EVALUATION OF.BRAIDWOOD STATION UNIT 1 1
- TECHNICAL SPECIFICATIONS l
l l
l D. E. BAXTER G. L. BRANSON
)
l Published June 1987 i
l Idaho National Engineering Laboratory-
.l EG&G Idaho, Inc.
l Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Under DOE-Contract No. DE-AC07-76ID01570 FIN A6824 O
ABSTRACT This document was prepared for the Nuclear Regulatory Commission (NRC)-
j to assist them in determining whether the Braidwood Station Unit 1 Technical Specifications (T/S), which govern plant systems configurations and' operations, are in conformance with the assumption of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative' audit of the FSAR l
as amended, and the SER as supplemented was performed with the Braidwood i
Unit 1 T/S.
Several discrepancies were identified and subsequently.
resolved by the NRC cognizant reviewer.
Resolutions ~to the discrepancies
]
noted in this report were achieved through conversations between the NRC l
l Reviewer and the Utility.
Resolutions were received for all discrepancies' j
that required resolution.
l i
i r
FOREWORD This report is supplied as part of the Power Reactor Technical Specifications Evaluations being conducted for the U.S. Nuclear Regulatory-Conunission, Office of Nuclear Reactor Regulation,-~ Division of Licensing by.
EG&G Idaho, Inc., NRR and I&E Support Unit.
The U.S. Nuclear Regulatory. Conunission funded work' under the.
authorization B&R 20 19 40 41 1, FIN No. A6824 - Power Reactor Technical Specification Evaluation.
9 0
a.
____________________---_-__a
L.. -
'l
'1 CONTENTS i
i ABSTRACT..............................................................
11
'l FOREWORD..............................................................
iii i
' 1.
. INTRODUCTION.....................................................
1 i
2.
REVIEW CRITERIA..................................................
.1 1
3.
SUMMARY
2 i
l 4.
BRAIDWOOD STATION UNIT ~1 TECHNICAL SPECIFICATIONS,
'FSAR,.SER CONSISTENCY COMPARISON.................................
3 Section.I.
Safety Limits......................................
3:
q Section II.
Reactor Protection System Setpoints................
3 Section III.
Engineered Safety Features Actuation System
'Setpoints~..........................................
'3 Section IV.
Pressure Boundary Isolation Valves.................
4 Section V.
Conta innent Isola t ion Va lves.......................
4 Section VI.
Containment Depressurization and Cooling System Limiting Conditions for Operation (LCO)...'.........
5-Section VII. Combustible Gas Control System Limiting Conditions for Operation...........................
5 Section VIII. Technical Specifications Requirements Documented in the Safety Evaluation Report....................
6 i
1 l
1 i
i
)
iv
1 EVALUATION OF BRAIDWOOD STATION UNIT NO 1
]
TECHNICAL SPECIFICATIONS I
J
]
1.
INTRODUCTION The Braidwood Station Unit 1 is a Westinghouse Pressurized Water Reactor (PWR) plant.
It has been selected for an audit to determine if the j
Braidwood Technical Specifications (T/S) are consistent with the Braidwood Final Safety Analysis Report (FSAR) up to and including Amendment 47 and the Braidwood Safety Evaluation Report (SER).
The specific sections of the T/S which were audited are listed in Part 2.
Differences between these sections of the T/S and the FSAR and SER along with the resolutions are identified in Part 4 of this report.
2.
REVIEW CRITERIA The following T/S sections were reviewed for this evaluation.
l l
1.
Safety Limits 2.
Reactor Protection System (RPS) Setpoints 3.
Engineered Safety Features Actuation System (ESFAS) Setpoints 4.
Pressure Boundary Isolation Valves (PIVs) 5.
Containment Isolation Valves (CIVs) l 6.
Containment Depressurization and Cooling System Limiting Conditions for Operation (LCO) 7.
Combustible Gas Control System LCOs 8.
Technical Specification Requirements Contained in the Safety Evaluation Report (SER) l The sections of the T/S listed in Part 4 were compared to the FSAR and SER to determine if the T/S are CONSISTENT, CONSERVATIVE or DIFFERENT than the FSAR and SER.
Setpoints and lists of valves and instruments in the T/S-were checked against tables in the FSAR and SER.
1 i
The SER was reviewed to ensure that T/S requirements in the SER were addressed in the T/S.
A description of each difference between the T/S and the FSAR and SER is included in this report.
~
l 3.
SUMMARY
l 1
During the performance of this audit, several differences between the T/S, SER and FSAR were noted.
Subsequently, discussions were held with the cognizant NRC reviewer.
A resolution was provided for each item that required one.
The items are listed below and have been assigned a status code which indicates the status of the item.
These items are discussed in detail in Part 4 of this report. All other sections were evaluated and found to be consistent or conservative.
1 Item Title Page Status
- j Section V Containment Isolation Valves 4
2,4 Section VIII Containment Isolation Systems 6
5 l
Item 1 i
Section VIII Evaluation of Compliance to 10 CFR 50 9
5 l
Item 8 Appendix H l
Section VIII Water Level Measurement Errors 12 5
I Item 13 Section VIII safety System Trip setpoint 13 5
Item 16 Methodology Section VIII Auxiliary Feedwater System 15 6
Item 19 Section VIII undetectable Failure in Online 16 5
Item 20 Testing Circuitry for ESF Relays l
l Status Code 1
l.
Unresolved, awaiting NRC/ Utility action 2.
Resolved pending issuance of T/S revision 3.
Resolved pending issuance of SER Supplement 4.
Resolved pending issuance of FSAR Amendment 5.
Resolved, NRC accepts as-is 6.
Resolved, item clarified and accepted 2
l L________-_-___
4.
BRAIDWOOD STATION UNIT 1 TECHNICAL SPECIFICATION, FSAR, SER CONSISTENCY COMPARIS0N Section I.
Safety Limits This section covers the review of the safety limits as defined in Section 2.1 of the Technical Specifications.
It includes reactor core limits and RCS pressure.
FSAR SER Technical Specification Section Section Evaluation 2.1.1 Reactor Core 4.4.1.1 &
4.4.1.1-CONSISTENT Limits and 3/4 2.5 DNB 15.0 Parameters 2.1.2 Reactor Coolant 5.2.2 5.2.2 CONSISTENT System Pressure Section II.
Reactor Protection System Setpoints This section covers the review of the Reactor Protection System Setpoints to insure the T/S values agree with or are conservative to the values assumed in the safety analysis or defined in the SER.
FSAR SER Technical Specification Section Section Evaluation 2.2 Reactor Trip System 7.2.2 7.2 CONSISTENT
,1 Instrumentation 15.0 SetpointsSection III.
Engineered Safety Features Actuation System (ESFAS) Setpoints l
I This section covers the review of the ESFAS Setpoints to assure the T/S values agree with or are conservative to the values identified in the FSAR sections or as defined in the SER as required values.
FSAR SER Technical Specification Section Section Evaluation 3/4.3.2, Table 3.3-4 7.3, 15.1 7.3 CONSISTENT i
l 4
6 3
Section IV.
Pressure Boundary Isolation Valves (PIVs)
This review determines if all the PIVs identified in the FSAR and SER are included in the T/S.
FSAR SER' Technical Specification Section Section Evaluation 3/4.4.6, Table 3.4-1 5.2.2 5.2.5 CONSISTENT Q212-34 Q212-67 NOTE: There was no PIV lists in the SER Section V.
Containment Isolation Valves (CIVs)
This review determines if all the CIVs identified in the FSAR and SER are included-in the T/S.
FSAR SER Technical Specification Section Section Evaluation 3/4.6.3 Pg. 3/4 6-18 Table 6.2-58 6.2.4 DIFFERENT Table 3.6-1 3/4.3.7 Pg. 3/4 7-9 i
FSAR Table 6.2-58 identifies the following valve which does not appear l
in the T/S Table 3.6-1.
Penetration Valve Number 59 lSI8905 D*
1 It should be noted that T/S Table 3.6-1 lists valve 1SI88050 at Penetration 59.
Also, FSAR Table 6.2-58 identifies valve 1SI8968 as a CIV which receives type C leak testing and T/S Table 3.6-1 identifies it as one that l
doesn't receive type C leak testing.
RESOLUTION l
Per the Licensee, the T/S Table 3.6-1 is in error with valve ISI 8805D (it l
should be 1SI 89050) and will be corrected.
The FSAR will be amended to-remove the Type C leak testing requirement from Valve'1SI 8968.
This item is CONSISTENT 4
)
i
I l
l Section VI.
Containment Depressurization and Cooling System (CDCS)
L1mitina Conditions for Operation (LCO)
This section reviews the LCOs for the CDCS to insure they adequately l
l cover the operation of the CDCS during all required modes of plant I
operation.
i FSAR SER Technical Specification Section Section Evaluation 3/4.6.2 Pg 3/4 6-13 6.2.2 6.2.2 CONSISTENT LC0 3.6.2.1 l
S/R 4.6.2.1 l
S/R 4.6.2.2 LC0 3.6.2.3 S/R 4.6.2.3 The LCOs and S/R for these systems are applicable during Modes 1, 2, 3, and 4 and require all systems be operational.
Section VII.
Combustible Gas Control System-(CGCS)
Limitina Conditions for Operation (LCOs)
This section reviews the LCOs for the CGCS to insure they adequately cover the operation of the CGCS during all required modes of plant operation.
FSAR SER Technical Specification Section Section Evaluation 3/4.6.4 Pg 3/4 6-25 6.2.5 6.2.5 CONSISTENT LC0 3.6.4.1 S/R 4.6.4.1 LC0 3.6.4.2 S/R 4.6.4.2 The LCOs and S/Rs for these systems are applicable during Modes 1 and 2 and require all systems be operational.
l l
l l
5
)
i Section VIII.
Technical Specification Reautrements Documented' 1
in the Safety Evaluation Report This section covers the review ofEallithe items identified in the safety evaluation report (SER) and: supplements to the' safety evaluation.
j report (SSER)'as T/S required items and whether they have or have not been-4 adequately addressed in the T/S.
i 1.
SER Section:
6.2.4 Containment Isolation Systems (GDC 57)
Pg. 6-20 states:
The applicant has not,.however, adequately demonstrated;that.the containment setpoint pressure that initiates containment. isolation has
.been reduced to the minimum compatible with normal operating 3
conditions.
Therefore, the staff-~ concludes that.the applicant has complied with the provisions of NUREG-0737. Item II.E.4.2, with the:
exception of the. containment' isolation setpoint pressure which_will'be covered in the Technical: Specifications ~ review for (Braidwood).
l T/S Section 3/4.3.2,
'Pg. 3/4 3-23' 1
T/S Table 3.3-4 surveillance lists the necessary'testir.g requirements and frequencies and the setpoints.
The high pressure setpoint was reviewed and found acceptable for Byron by the NRC.
However, the Braidwood setpoint is higher and may not be-acceptable.
This item is DIFFERENT 1
RESOLUTION I
The Staff has reviewed the setpoint and accepts it as-is.
This item is CONSISTENT
]
2.
SER Section:
4.4.6 Loose Parts Monitoring System Pg. 4-11 states:
The staff has reviewed the (Braidwood] LPMS by comparing _it with the j
equipment and procedures used on other comparable plants, taking into 4
account pertinent differences and the requirements of Regulatory.
Guide 1.133. The staff will require that the limiting conditions for operation and surveillance requirements be included in the Technical Specifications in accordance with Regulatory Guide 1.133.
T/S Section:
3/4.3.3 Pg. 3/4 3-64 T/S 3.3.3.8 lists the LC0 and surveillance requirements for the Loose Parts Monitoring System.
This item is CONSISTENT 6-4
_-------_A
3.
SER Section:
3.9.6 Periodic Leak Testing Requirements Pg. 3-44 states:
The staff will require the [Braidwood] pressure isolation valves to be categorized as A or AC according to IWB-1400 of Section XI of the ASME Code for safety injection and shutdown cooling systems.
The staff also requires that the [Braidwood] Technical Specifications contain limiting conditions for operation that will require plant shutdown or system isolation when the leakage limits are not met.
The Technical Specifications will include surveillance requirements which state the acceptable frequency of leak rate testing and the acceptable values for leakage rates. The above Technical Specifications will be based on the latest revision of NUREG-0452, " Standard Technical Specifications for Westinghouse Pressurized Water Reactors." Based on these Technical Specifications and the applicant's commitment to perform periodic leak rate testing of pressure isolation valves
'between the reactor coolant system and low pressure systems, the staff
- oncludes that there is reasonable assurance that the design pressure of the low-pressure systems will not be exceeded, and therefore, an intersystem LOCA will not occur.
This meets in part, the requirements of GDC 55 of Appendix A to 10 CFR 50.
T/S Section:
3/4.4.6 Pg. 3/4 4-21 T/S 3.4.6.2 lists the allowable leakage limits and the monitoring frequencies and provides limiting conditions for operation as required.
This item is CONSISTENT 4.
SER Section:
4.3.1 Nuclear Design Discussion Pg.4-4 states:
[Braidwood] will use the improved load-follow package.
The constant axial offset control (CAOC) band will be +3 to -12 A flux difference for this control mode.
The analysis performed by Westinghouse has indicated that the peaking factor limit cannot be met at the beginning of life of cycle 1 due to the wide AI band.
This has resulted in limiting the width of the band to the value of iS% AI until 3000 mwd /MTU burnup for [Braidwood) Units 1 and 2.
The 15% AI is the value previously justified by the CAOC analysis.
These features will be incorporated in the (Braidwood] Technical Specifications.
T/S Section:
3/4.2.1 Pg. 3/4 2-1 T/S 3.2.1 requires the maximum AI for cycle 1 to be 5% AI until 5000 mwd /MTU which is conservative with respect to the SER requirements.
This item is CONSERVATIVE 7
5.
SER Section:
4.4.2 Fuel Rod Bowing, SER Supplement No. 1, Section 4.4 Pg. 4-9 states:
TABLE 4.1.
R00 BOW PENALTIES Burnup DNBR Penalty (mwd /MTU)
(%)
0 0
3500 0
5000 0
10000 2.15 15000 4.64 20000 6.74 25000 8.59 30000 10.27 35000 13.07 40000 19.09 l
Before the Technical Specifications are issued, the staff will ensure that the thermal margin reductions given above have been accommodated l
using an acceptable method.
For those plants using the Westinghouse standard 17 x 17 fuel design (RESAR-3S), the staff has previously approved plant-specific and generic margins which could be used to compensate the DNBR reductions given above.
For the (Braidwood] units, the applicant has thermal i
margin available which could be used to offset DNBR reductions.
If the applicant intends to use this available thermal margin, a description of the margin and the amount of reduction in the rod bow penalties would be included in the basis of the Technical Specifications.
T/S Section:
B 3/4.2.2 and B 3/4.2.3 Pg. B 3/4 2-4 These T/S bases discuss rod bowing and include a discussion on the thermal margin limits.
This item is CONSISTENT 6.
SER Section:
4.4.3 N-1 Loop Operation Pg. 4-10 states:
N-1 loop operation is when one reactor coolant loop is out of service, leaving three coolant loops available to supply coolant to the reactor Core.
8
In response to a staff question, the applicant stated that he did not wish to exercise the option to operate in the N-1 mode.
The staff will require that the Technical Specifications include appropriate provisions to ensure that this type of operation is prohibited.
T/S Section:
3/4.4.1 Pg. 3/4 4-1 T/S 3.4.1.1 states that all reactor coolant loops must be in operation during Modes 1 or 2.
This item is CONSISTENT 7.
SER Section:
4.4.4 Crud Deposition Pg. 4-10 states:
Based on the information given above, the staff concludes that the applicant has adequately addressed the staff's concerns relative to uniform or preferential crud depositions in the core.
The staff will ensure that appropriate surveillance requirements to recognize rapid crud buildup are included in the Technical Specifications.
T/S Section:
3/4.2.3 Pg. 3/4 2-9 S/R 4.2.3.3 requires flowrate verification at least one per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
This item is CONSISTENT 8.
SER Section:
5.3.1.2 Evaluation of Compliance to 10 CFR 50, Appendix H Pg. 5-1 states:
On the basis of its review of the applicant's submittal that described the extent of compliance of Braidwood with Appendix H, 10 CFR 50, the staff has determined that (1) Braidwood 1 has met all the requirements of Appendix H, but has not indicated the schedule for removal of reactor vessel beltline surveillance capsules.
(2) Braidwood 2 has met all the requirements of Appendix H, but has not indicated the schedule and the materials in the the reactor vessel beltline surveillance program.
The applicant indicated that the reactor vessel beltline surveillance program will comply with ASTM E-185 and Appendix H, 10 CFR 50.
The applicant also indicated that he will provide the information identified above at a later date.
The staff will review this report during its review of the applicant's Technical Specifications to confirm that the surveillance program complies with ASTM E-185 and Appendix H, 10 CFR 50.
T/S Section:
3/4.4.9 Pg. 3/4 4-32 and 4-37 9
e n
1
-i S/R 4.'4.9.1'.2. specifies' surveillance' specimen removal.and examination.
in accordance with the. schedule in Table 4.4-5 on page 3/4 4-37.
The NRC staff needs to insure this is in compliance with ASTM E-185 and:
1 Appendix H, 10 CFR 50.
This item is NOT EVALUATE 0 RESOLUTION The Staff has. reviewed this schedule and accepts it'as-is.
This-item is CONSISTENT 9.
-SER Section:
6.2.6-Containment Leakage Testing Pg. 6-25' states:
If the periodic 6-month test of paragraph III.D.2(b)(i) and the. test required by paragraph-III.D.2.(b)(iii) are current, no sr.aintenance has been performed on the air lock, and.the' air lock is properly sealed, there should be no reason to expect the air lock toLleak excessively l
l just because it has been. opened in Mode 5 or Mode 6.
Ll l
Accordingly, the. staff concludes that the applicant's proposed approach of substituting the seal leakage test of paragraph III.D.2(b)(iii) is acceptable when no maintenance has been performed on an air lock.
Whenever maintenance has been performed on.
an air lock, the requirements of paragraph II.D.2(b)(ii) of Appendix J must still be met by the applicant.
I Therefore, an exemption from this requirement [10 CFR 50, Appendix J,.
l Paragraph III.0.2(b)(ii)) is justified and acceptable for [Braidwood)
Units 1 and 2, and appropriate requirements will be added.to the plant Technical Specificatioini,.
i T/S Section:
3/4.6.1 Pg. 3/4 6-5.
S/R 4.6.1.3.b.1 and 2 state that the airlock need only be t'ested if maintenance has been performed or six months has elapsed since its last test.
This item is CONSISTENT
- 10. SER Section:
6.3.1 Heat Tracing of RWST Vent Lines Pg.'6-28 states:
The RWST minimum inventory is 350,000 gal of 2000-ppm borated water.
To maintain the RWST water:above the temperature of boron precipitation and freezing, the' applicant has provided the RWST with a heating system.
In addition, heat tracing will be'added to the RWST vent line to minimize the potential for ice or snow blockage. The 1
staff requires that this heat tracing be included with the cold I
weather surveillance items in the. Technical Specifications.
Confirmation of this specification will be made during the review of i
Technical Specifications prior to power operation.
j
)
l l
l 10 l
l 2
. 1 T/S Section:
3/4.1.2 Pg. 3/4 1-13 S/R 4.1.2.6.b specifies that the vent line temperature must be1 l
verified to be greater than 35'F-whenever outside air temperature.is-i 35'F or less.
This item is CONSISTENT
- 11. SER Section:
6.5.1 ESF Atmospheric Cleanup System Pg. 6-37-states:
The applicant has not connitted. to record flow rate through the system as noted under acceptance criteria II.2.e of SRP Section 6.5.1.
This j
will be acceptable provided the fans are fixed-speed fans and the applicant has a curve of flow rate versus pressure drop, with pressure drop verified during plant operation on a routine basis.
Such a
{
requirement
,n be made a part of the (Braidwood) Technical j
Specifications 3/4.7.6, 3/4.9.12 and'3/4.7.7.
)
l l
T/S Section:
T/S 3/4.7.6 Pg. 3/4 7-14, T/S 3/4.7.7 Pg. 3/4 7-17 and T/S 3/4.9.12 Pg. 3/4 9-14 Each of these technical specifications surveillance requirements states that flow rates and pressure drop shall be verified once per 18 month interval.
This item is CONSISTENT j
I
- 12. SER Section.
7.2.2.1 Testing the Reactor. Trip Breakers and Manual Trip Switches Pg. 7-5 states:
l Testing of the undervoltage coil operation is carried out with a trip l
signal from the solid-state logic protection system.
Testing of the manual reactor trip channel does not allow independent verification of l
the operability of the shunt coil and the undervoltage coil because l
the operation of the manual trip switch results in a simultaneous trip action by both coils.
The staff will include.in the station's Technical Specifications a requirement to periodically and independently verify the operability'of the undervoltage and shunt trip functions at least once each refueling outage.
T/S Section:
3/4.3.1 Pg. 3/4 3-12 Table 4.3-1 Table Notations (Note 11) states that once every 18 months the reactor trip breakers undervoltage'and shunt trips shall be i
verified operable.
This item is CONSISTENT O
t 11
13.
SER Section:
7.2.2.3 Water Level Measurement Errors Pg. 7-6 states:
The steam generator and pressurizer water level measurement channels utilize differential pressure transmitters.
The measurement accuracy of such a system is affected by several. factors. Of primary importance is the increase in the indicated water leve) caused by a decrease of the water density in the reference 1eg resulting from an increase in the ambient temperature due to a high-energy-line break.
For such an accident, the steam generator water level provides the primary trip function and the trip setpoints need to'lue selected to ensure that the action required by the safety analyses will be initiated throughout the range of temperatures that can be expected.
This issue was addressed for operating reactors in IE Bulletin 79-21.
In response to this concern, the applicant has committed in a letter dated January 26, 1982, to evaluate the effect of high temperature on the reference legs of water level measurement systems following a high-energy-line break and to factor the measurement errors in the trip setpoints.
The effects of environmental errors in level measurement will be reviewed as part of the staff review of setpoint methodology (see Section 7.3.2.4).
T/S Section:
Pg.
This item requires staff attention as we do not have access to the staff's review of the applicants setpoint methodology.
This item is NOT EVALUATED RESOLUTION The Staff has reviewed the methodology and accepts it as-is, l
This item is CONSISTENT 1
14.
SER Section:
7.2.2.4 Lead, Lag, and Rate Time Constant Setpoints Used in Safety System Channels Pg. 7-6 states:
Several safety system channels make use of lead, lag, or rate signal compensation to provide signal time responses consistent with assumptions in the Chapter 15 analyses.
The time constants for these signal compensations are adjustable within the analog portion of the safety system.
The applicant has committed to incorporate testing of the time constants in the station Technical Specifications.
T/S Section:
3/4.3.2 Pg. 3/4 3-29 i
Table 3.3-4 Table Notations
- and ** state that the time constants utilized shall be channel calibrated to the specified values.
This item is CONSISTENT l
12 1
1
- 15. SER Section:
7.2.2.7 Verification of RTD Loop Bypass flow Pg. 7-7 states:
i The reactor coolant system hot-and cold-leg resistance temperature
.)
detectors (RTDs) used for reactor protection are located in reactor l
coolant bypass loops. A bypass loop from upstream of the steam generator to downstream of the steam generator is used for the hot-leg RTD and a bypass loop from downstream of the reactor coolant pump to upstream of the pump is used for the cold-leg RTD. The flow rate affects the overall time response of the temperature signals provided J
for reactor protection and, thus, should be monitored at appropriate intervals. The staff requires that the magnitude of the RTD bypass
' loop flow rate be verified to be within required limits at each H
refueling period. This requirement will be incorporated in the plant-Technical Specifications.
1 T/S Section:
3/4.3.1-Pg. 3/4 3-12 3
Table 4.3-1 Table Notation 13 requires the bypass flow be verified at each channel calibration which is done at each refueling outage.
This item is CONSISTENT 16.
SER Section:
7.3.2.4 Safety System Trip Setpoint Methodology Pg. 7-14 states:
The methodology followed in setting the safety system trip points has not been described in the fSAR.
In response to a request for information concerning this item, the applicant stated that the setpoint study has not been completed for [Braidwood).
Because the j
primary function of this information is to confirm the adequacy of setpoints specified in the plant Technical Specifications, the staff I
will audit this information on setpoint methodology, includino the l
effect of water level measurement errors (Section 7.2.2.3), at the j
time the Technical Specifications are available for review.
T/S Section:
Pg.
This item is the same as item 13 of this section and requires Staff Review.
This item is NOT EVALUATED RESOLUTION The Staff has reviewed this methodology and accepts it as-is.
This item is CONSISTENT
?
13
i 17.
SER Section:
8.4.3 Power Lockout to Motor-0perated (MO) Valves Pg. 8-14 states:
The. applicant has provided a list of valves that require power lockout to meet the single failure criterion in the fluid systems.
BTP ICSB 18 (PSB) requires that all such valves and their required position be
.c listed in the Technical Specifications and that the position indications for these valves meet the single-failure criterion.
To meet the staff requirements on the power lockout to the motor-operated 3
valves, the applicant has provided two breakers in series for these i
valves.
To meet the staff requirement that redundant valve status indication be provided to the control room operator, the applicant has provided redundant and separate valve position switches.
One position switch is mounted on the valve stem.
If one position switch is inoperable, the second will be available to provide position indication. The.
j position switch on the valve operator actuates an indication light in the main control room, and the position switch on the stem gives annunciation in the main control room.
These circuits are powered from separate power supplies.
As required, the applicant has agreed to list all such valves and their required position in the Technical Specifications. 'The staff finds that the design of power lockout to motor-operated valves conforms to BTP ICSB 18 and is, therefore, acceptable.
I T/S Section:
3/4.5.1 and 3/4.5.2 Pg. 3/4 5-1 and 3/4 5-5 i
S/R 4.5.1.1 and 4.5.2 list the valves that must have power removed to meet single failure criterion.
This item is CONSISTENT 18.
SER Section:
10.4.9 Auxiliary Feedwater System Pg. 10-16 states:
1 The staff has reviewed the auxiliary feedwater system (AFWS) against the specific acceptance criteria of SRP Section 10.4.9 as follows:
GDC 46 as related to design provisions made to permit appropriate functional testing of the system and components to ensure structural integrity and leak tightness, operability and performance of active components, and capability of the integrated system to function as intended during normal, shutdown, and accident conditions.
In meeting this criterion, the Technical Specifications should specify that the i
monthly AFWS pump test shall be performed on a staggered test' basis to reduce the likelihood of leaving more than one pump in a test mode following the tests.
?
T/S Section:
3/4.7.1 Pg. 3/4 7-4 j
14 a
i S/R 4.7.1.2.1 requires that the pumps be tested once per 31 days on a staggered basis.
l This item is CONSISTENT 19.
SER Section:
10.4.9 Auxiliary feedwater System Pg. 10-18 states:
The applicant has committed (by letter dated January 14, 1982) to revise the proposed plant Technical Specifications to-state that one essential AFWS pump train may be inoperable for no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
If this time is exceeded, the. unit affected must be placed in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and hot shutdown within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The Technical Specifications will.further state that cold shutdown will begin immediately if both essential AFWS pump trains are
{
determined inoperable. Further, the applicant has incorporated the capability to manually transfer power to the essential motor-driven AFWS pump from the corresponding emergency (Class IE) diesel generator power supply in the opposite unit in order to improve AFWS availability as discussed further in this SER section.
At the request of the staff, the applicant has committed (Tramm, January 14, 1982)'to revise the plant Technical Specifications to include a limiting condition for operation and action statement to ensure maximum AfWS reliability based on the'above power supply transfer feature as follows:
With the "A" diesel generator in either unit inoperable for 7 days, j
immediately restore it to an operable condition or place the opposite' j
unit in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
T/S Section: 3/4.7.1 Pg. 3/4 7-4 T/S 3.7.1.2 allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to get to hot shutdown if both AFWS pumps are inoperable with no action statement concerning diesel-generator status. The SER requires that an LCO and action statement be included concerning operability of the diesel-generators and immediately going to cold shutdown if both AFWS pump trains are inoperable.
This item is DIFFERENT RESOLUTION The Staff has r.eviewed the T/S as written and feels it adequately covers the SER connitment.
It is accepted as-is.
This item is CONSISTENT l
e 15
_ __ __a
I i
- 20. SSER No. 3 Section:
7.3.2.12 Undetectable failure in Online Testing l
Circuitry for Engineered Safeguards Relays Pg. 7-1. states:
]
On August 16, 1982 Westinghouse notified the staff of a potential undetectable failure on online test circuity'for the master relays in the engineered safeguards systems.
The undetectable failure involved the output (slave) relay continuity proving lamps and their associated shunts provided by test pushbuttons.
If after testing, a shunt is not provided for any proving lamp because of a switch contact failure, any j
subsequent safeguards actuation could cause the lamp to burn open before its associated slave relay is energized.
This would then' prevent actuation of any associated safeguards devices on that slave relay. Westinghouse has provided test procedures that ensure that the l
slave relay circuits operate normally when testing of the master relays is completed.
In a March 28, 1983 letter to James G. Keppler of Region III, the applicant provided details of a permanent circuit change and the staff has not completed its review of this information.
Until an acceptable circuit modification is installed, the staff will require Technical Specifications to include monthly tests (instead of quarterly) of any slave relay that has a proving lamp.
These tests should be performed immediately following the monthly test of an associated master relay.
T/S Section:
3/4.3.2 Pg. 3/4 3-34 T/S Table 4.3-2 does not indicate any slave relays with' proving lamps that require monthly testing. This may be because there has been an acceptable modification installed.
The licensee has demonstrated that the modifications have been made and accepted by the staff in Supplement 3 of the SER. This item requires staff review to ensure that an acceptable modification has been completed.
This item is NOT EVALUATED RESOLUTION The Staff has reviewed this item and has accepted it as-is.
This item is CONSISTENT O
16
asmC pomm am u t. NUCLSA A #80utaf oA V Comenstessoas e es pont Numeta vam. pass a, fl0C. ear we me.,# ea,#
(2 646
- 'Z"3;;$'-
BIBLIOGRAPHIC DATA SHEET EGG-NTA-7505
$85 th57 AUCTIONS ON f M4 86VER14 2 fif Lt ANo SuefiTLt J L8AvtSLANE EVALUATION OF BRAIDWOOD STATION UNIT 1 TECHNICAL SPECIFICATIONS 4 oaTE meroRY CoMPLtYto es0 set M vlAA June 1987-
..o,.o.,,,
D. E. Baxter e oari aeroar eauso G. L. Branson "o"'"
l
" ^a June 1987 7 Pimpommino omGamstaTioM maut Amo maskamG acomiss parevede Ceau 8 PROJ8CTITA8ssfoAE uNat NUedGGR NRR and I&E Support EG&G Idaho, Inc.
' ' ' " "o"*"'"*"""
P. 0. Box 1625 Idaho Falls, ID 83415 A6824
. 3,o o.,~
o.o..a.r o.
..t.~o...L..o.oo... u
- i. Co.,,
ii. rv. o, as, oat Division of Licensing Final Technical Office of Nuclear Reactor Regulation Evaluation Report U.S. Nuclear' Regulatory Commission
' "'a co""'o a-~ ~~
Washsington, DC 20555 12 SUPPLt. TNT ART NOf f 5 IJ at$TR ACT 1200 wores e
- eses Final technical evaluation report on the audit of the Braidwood Station Unit 1 j
Technical Specifications performed for the NRC in connection with the issuance of Low Power and Full Power License for the applicant. All identified discrepancies have been resolved.
l
.. ooc w...,.~. 6....
..t
.o os,ouc...To..
STAftMENT Unlimited 18 8tCyniTV CLAS8ielCAflOCl trane espw
,,etNr,,,t.s,o t=e~otorgaus Unclassified
< r.
n Unclassified I i 7 w.et a v,..c.n l
j is en,Cs
_________.