ML20027A863
| ML20027A863 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/13/1982 |
| From: | Mucha E Franklin Research Ctr, Franklin Institute |
| To: | Eccleston K Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20027A864 | List: |
| References | |
| TAC 42224, TAC 42225 TER-C5506-63-65, NUDOCS 8208040257 | |
| Download: ML20027A863 (67) | |
Text
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TECHNICAL EVALUATION REPORT BWR SCRAM DISCHARGE VOLUME LONG-TERM MODIFICATIONS COMMONWEALTH EDIS0N COMPANY QUAD CITIES STATION UNITS 1 AND 2 4 NRCDOCKETNO. 50-254. 50-265 F#C PROJECTN 5
NRC TAC NO. 42224, 42225 FRC ASSIGNMENT 2 NRC CONTRACT NO. NRC-03 81 130 FRCTASKS 63, 65
. Preparedby Author: E. Mucha Franklin Research Center i
20th and Race Street FRC Group Leader: E. Mucha Philadelphia, PA 19103 Pronar=dfor i
Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: K. Eccleston l
July 13, 1982 t
This report was prepared as an account of work sponsored by an agency of the United States Govemment. Neither the United States Govemment nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or
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responsibility for any third party's use, or the results of such use, of any information, appa-ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.
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TECHNICAL EVALUATION REPORT BWR SCRAM DISCHARGE VOLUME LONG-TERM MODIFICATIONS COMMONWEALTH EDISON COMPANY p
l QUAD CITIES STATION UNITS 1 AND 2
- NRC OOCKET NO. 50-254, 50-265 FRC PROJECT C3506 l NRC TAC NO. 42224, 42225 FRC ASSIGNMENT 2 NRC CONTRACT NO. NRC-03-81 130 FRCTASKS 63, 65 g
Prepared by Author: E. Mucha Franklin Research Center 20th and Race Street FRC Group Leader: E. Mucha Philadelphia, PA 19103 Prepared for l
Nuclear Regulatory Commission l
Washington, D.C. 20555 Lead NRC Engineer: K. Eccleston I
July 13, 1982 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, appa-e l
ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.
Reviewed by:
Approved by:
'kM Ne d
Group Leader Department Director i
d Franklin Research Center A Division of The Franklin Institute The Bengtrun Franklin Penn ey, PNia Pa. 19103(215) 448-1000 t
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TER-C5506-63/65 CONTENTS Section Title Page
SUMMARY
1 l
1 INTRODUCTION 3
1.1 Purpose of Review.
3 1.2 Generic Issue Background 3
l.3 Plant-Specific Background.
5 2
REVIEN CRITERIA.
7 2.1 Surveillance Requirements for SDV Drain and Vent valves 7
2.2 LCO/ Surveillance Requirements for Reactor Protection System SDV Limit Switches 8
2.3 LCO/ Surveillance Requirements for Control Rod Withdrawal Block SDV Limit Switches 10 3
METHOD OF EVALUATION 13 4
TECHNICAL EVALUATION 14 4.1 Surveillance Requirements for SDV Drain and Vent Valves 14 4.2 LCO/ Surveillance Requirements for Reactor Protection System SDV Limit Switches 16 4.3 LCO/ Surveillance Requirements for Control Rod Withdrawal Block SDV Limit Switches 21 5
CONCLUSIONS.
26 6
REFERENCES.
30 APPENDIX A - NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS APPENDIX B - COMMONWEALTH EDISON LETTER OF OCTOBER 14, 1980 AND SUBMITTAL WITH PROPOSED TECHNICAL SPECIFICATIONS CHANGES FOR QUAD CITIES STATION UNITS 1 AND 2 APPENDIX C - COMMONWEALTH EDISON LETTER OF OCTOBER 22, 1981 WITH ANSWER TO RFI FOR QUAD CITIES STATION UNITS 1 AND 2 iii A
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s TER-C5506-63/65 FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Commission (Office of
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Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.
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SUMMARY
This technical evaluation report reviews and evaluates proposed Phase 1 changes in the Quad Cities Station Units 1 and 2 Technical Specifications for scram discharge volume (SDV) long-term modifications regarding surveillance requirements for SDV vent and drain valves and the limiting condition for operation (LCO)/ surveillance requirements for reactor protecti'on system and control rod withdrawal block SDV limit switches. Conclusions were based on the degree of compliance of the Licensee's submittal with criteria from the Nuclear Regulatory Commission (NRC) staff's Model Technical Specifications.
The revised page 3.3/4.3-3, with the Licensee's agreement to incorporate a revision into the proposed specifications changes that requires cycling each valve at least one complete cycle of full travel at least quarterly, complies with the NRC staff's Model Technical Specifications requirements of paragraphs 4.1.3.1.la and 4.1.3.1.lb.
The original pages 3.1/4.1-12 and 3.1/4.1-13, Table 4.1.1, of the Quad Cities Station Units 1 and 2 Technical Specifications, which provide for the reactor protection system SUV limit switches water level-high Channel Functional Test to be performed once per 3 months, do not meet the surveillance requirement (paragraph 4.3.1.1, Table 4.3.1.1-1, of the NRC Staff's Model Technical Specifications) for the test to be performed monthly. However, the Licensee is installing a second instrument volume containing four additional i
limit switches, for a total of eight limit switches, for the reactor
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protection system.
This increases significantly the reliability of the system i
and provides technical bases for acceptance of the proposed surveillance
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requirements to perform the Channel Functional Test quarterly.
To meet the NRC staff's Model Technical Specifications requirements of paragraph 4.3.1.1 and Table 4.3.1.1-1, the Licensee agreed to revise page 3.1/4.1-14 of the present Quad Cities Station Units 1 and 2 Technical Specifications to incorporate in Table 4.1-2 the Calibration Test "each refueling" for " Instrument Channel - SDV Water Level High."
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TER-C5506-63/65 f the level switches Test will consist of physical inspection and actuation o using water columns.
i ments The NRC staff's Model Technical Specifications surveillance requ re f
6 1 for in paragraph 3.3.6, Table 3.3.6-1, paragraph 4.3.6, and Table 4.3. -
l the operational control rod block SDV scram trip bypassed are not applicable to l rod conditions of startup, run, and refuel with more than one contro
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SDV Therefore, the Licensee agreed to delete " Instrument Channel -
withdrawn.
ii of page High Water I4 vel Scram Trip Bypassed" from the proposed rev s on 3.2/4.2-16, Table 4.2-1.
3.2/4.2-14, Table 3.2-3, and page t
The existing SDV system has only one trip system with one instrumen i h and is channel containing one control rod withdrawal block SDV limit sw tc To reflect this, the Licensee agreed to revise the first sentence acceptacle.
3.2/4.2-14.
of Note 1 in Table 3.2.3 on original page f
To meet the NRC staff's Model Technical Specifications requirements o te into paragraph 4.3.6 and Table 4.3.6-1, the Licensee agreed to incorpora 3.2/4.2-15 the Calibration Test "each refueling" Table 4.2-1 on revised page instead of "Not applicable" for " Instrument Channel-Rod Blocks, High Water Channel Calibration with the Magnetrol Level in Scram Discharge Volume."
itch level switch will consist of physical inspection and actuation of the sw using a water column.
3.2/4.2-14, The remaining surveillance requirements are met by revised pages
/ 1 12, and original, unrevised pages 3.1 4. -
3.2/4.2-16, 3.3/4.3-3, 3.3/4.3-9, of the Quad Cities Station Un ts i
1 and 2 3.1/4.1-13, 3.1/4.1-14, and 3.3/4.3-10 Table 5-1 on pages 28 and 29 of this report Technical Specifications.
summarizes the evaluation results.
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TER-C5506-63/65
- 1. INTRODUCTION l.1 PURPOSE OF THE TECHNICAL EVALUATION The purpose of this technical evaluation report (TER) is to review and evaluate the proposed changes in the Technical Specifications of the Quad Cities Station Units 1 and 2 boiling water reactor (BWR) in regard to "BWR Scram Discharge Volume Long Term Modification," specifically:
o surveillance requirements for scram discharge volume (SDV) vent and drain valves limiting condition for operation (LCO)/ surveillance requirements o
for the reactor protection system o LCO/ surveillance requirements for the control rod withdrawal block SDV limit switches.
The evaluation used criteria proposed by the NRC staff in Model Technical Specifications (see Appendia A of this report). This effort is directed toward the NRC objective of increasing the reliability of installed BWR scram discharge volume systems, the need for which was made apparent by events described below.
1.2 GENERIC ISSUE BACKGROUND 1
On June 13, 1979, while the reactor at Hatch Unit 1 was in the refuel mode, two SDV high level switches had been modified, tested, and found inoperable.
The remaining switches were operable.
Inspection of each inoperable level switch revealed a bent float rod binding against the side of the float chamber.
On October 19, 1979, Brunswick Unit I reported that water hammer due to' slow closure of the SDV drain valve during a reactor scram damaged several pipe supports on the SDV drain line.
Drain valve closure time was approximately 5 minutes because of a faulty solenoid controlling the air supply to the valve.
After repair, to avoid probable damage from a scram, the unit was started with the SDV vent and drain valves closed except for periodic draining.
During this mode of operation, the reactor scrammed due to a high water level in the l
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TER-C5506-63/65 I
SDV system without prior actuation of either the high level alarm or rod block switch.
Inspection revealed that the float ball on the rod block switch was bent, making the switches inoperable. The water hammer was reported to be the cause of these level switch failures.
As a result of these events involving common-cause failures of SDV limit switches and SDV drain valve operacility, the NRC issued IE Bulletin 80-14,
" Degradation of BWR Scram Discharge Volume Capability," on June 12, 1980 [1].
In addition, to strengthen the provisions of this bulletin and to ensure that the scram system would continue to work during reactor operation, the NRC sent a letter dated July 7, 1980 [2] to all operating BWR licensees requesting that they propose Technical Specifications changes to provide surveillance require-ments for reactor protection system and control rod block SDV limit switches.
The letter also contained the NRC staff's Model Technical Specifications to be used as a guide by licensees in preparing their submittals.
Meanwhile, during a routine shutdown of the Browns Ferry Unit 3 reactor on June 28, 1980, 76 of 105 control rods failed to insert fully. Full inser-tion required two additional manual sces:is and an automatic scram for a total elapsed time of approximately 15 minutes between the first scram initiation and the complete insertion of all the rods. On July 3, 1980, in response to both this event and the previous events at Hatch Unit 1 and Brunswick Unit 1, the NRC issued (in addition to the earlier IE Bulletin 80-14) IE Bulletin 80-17 followed by five supplements. These initiated short-term and long-term programs described in " Generic Safety Evaluation Report BWR Scram Discharge System," NRC Staff, December 1, 1980 [9] and " Staff Report and Evaluation of Supplement 4 to IE Bulletin 80-17 (Continuous Monitoring Systems)" (10].
Analysis and evaluation of the Browns Ferry Unit 3 and other SDV system events convinced the NRC staff that SDV systems in all BWRs should be modified to assure long-term SUV reliability.
Improvements were needed in three major SDV-IV hydraulic coupling, level i.tstrumentation, and system isolation.
areas To achieve these oojectives, an Office cf Nuclear Reactor Regulation (NRR) task force and a subgroup of the BWR Owners ficoup developed Revised Scram Discharge dd Franklin Research Center 4 c%.e N r n,
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b TER-C5506-63/65 System Design and Safety Criteria for use in establishing acceptable SDV systems modifications [9]. Also, an NRC letter dated October 1, 1980 requested all operating BWR licensees to reevaluate installed SDV systems and modify them as necessary to comply with the revised criteria.
In Reference 9, the SDV-IV hydraulic coupling at the Big Rock Point, Brunswick 1 & 2, Duane Arnold, and Hatch 1 & 2 BWRs was judged acceptable. The remaining BWRs will require modification to meet the revised SDV-IV hydraulic coupling criteria, and all operating BWRs may require modification to meet the revised instrumentation and isolation criteria. The changes in Technical Specifications associated with this effort will be carried out in two phases:
Phase 1 - Laprovements in surveillance for vent and drain valves and instrument volume level switches.
Phase 2 - Improvements required as a result of long-term modifications made to comply with revised design and performance criteria.
This TER is a review and evaluation of Technical Specifications changes proposed for Phase 1.
1.3 PLANT-SPECIFIC BACKGROUND The July 7,1980 NRC letter [2] not only requested all BWR licensees to amend their facilities' Technical Specifications with respect to control rod drive SDV capability, but enclosed the NRC staff's proposed Model Technical Specifications (see Appendix A of this TER) as a guide for the licensees in preparing the requested submitt'als and as a source of criteria for a technical evaluation of the submittals.
This TER is a review and evaluation of Technical Specifications changes for the Quad Cities Station Units 1 and 2 proposed by the Licensee, Commonwealth Edison (CE), in letters dated October 14, 1980 and October 22, 1981 (see Appendices B and C, respectively) in regard to "BWR Scram Discharge volume (SDV) Long-Term Modifications" and, specifically, the surveillance req 2irements for SDV vent and drain valves and the limiting conditien for operation (LCO)/ surveillance requirements for the reactor protection system and control rod withdrawal block SDV limit switches. The 4 NUd Franklin Research Center 4oi orn.r=.
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TER-C5506-63/65 adequacy with which the CE information documented compliance of the proposed Technical Specifications changes with the NRC staff's Model Technical Specifications is also assessed.
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- 2. REVIEW CRITERIA The criteria established by the NRC staff's Model Technical Specifications involving surveillance requirements of the main SDV components and instrumenta-I tion cover three areas of concern o surveillance requirements for SDV vent and drain valves o LCO/ surveillance requirements for reactor protection system SDV limit switches o Ico/ surveillance requirements for control rod block SDV limit switches.
l 2.1 SURVEILLANCE REQUIRDUDITS FOR SDV DRAIN AND VENT VALVES The surveillance criteria of the NRC staff's Model Technical Specification for SDV drain and vent valves are:
"4.1.3.1.1 - The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:
a.
Verifying each valve to be open* at least once per 31 days, and b.
Cycling each valve at least one complete cycle of full travel at least once per 92 days.
- These valves may be cicaed intermittently for testing under administrative controls."
The Model Technical Specifications require testing the drain and vent valves, checking at least once in every 31 days that each valve is fully open during normal operation, and cycling each valve at least one complete cycle of i
full travel under administrative controls at least once per 92 days.
Full opening of each valve during normal operation indicates that there is no degradation in the control air system and its components that control the air pressure to the pneumatic actuators of the drain and vent valves.
Cycling each valve checks whether the valve opens fully and whether its movement is smooth, jerky, or oscillatory.
During normal operation, the drain and vent valves stay in the open i
position for very long periods.
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j TER-C5506-63/65 and flakes, various fibers, lint, sand, and weld slag from the water or air may accumulate at moving parts of the valves and temporarily freeze them. A strong breakout force nay be needed to overcome this temporary freeze, producing a violent jerk which may induce a severe water hammer if it occurs during a scram or a scram resetting.
Periodic cycling of the drain and vent valves is the best method to clear the effects of particulate silting, thus promoting smooth op.ening and closing and more reliable valve operrtion. Also, in case of improper valve operation, cycling can indicate whether excessive pressure transients may be generated during and after a reactor scram which might damage the SDV piping system and cause a loss of system integrity or function.
2.2 LCO/ SURVEILLANCE REQUIREMENTS FOR REACTOR PROTECTION SYSTEM SDV LIMIT SWITCHES The paragraphs of the NBC staff's Model Technical Specifications pertinent to LCO/ surveillance requirements for reactor protection system SDV limit switches are:
"3.3.1 - As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PRCffECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.
s Table 3.3.1-1.
Reactor Protection System Instrumentation Applicable Minimum Operable
[
Functional Operational Channels Per Trip Unit Conditions System (a)
Action 8.
Acram Discharge l
Dlume Water 1
Level-High 1,2,5 (h) 2 4
Table 3.3.1-2.
Reactor Protection System Response Times l
Functional Response Time Un it (Seconds) 8.
Scram Discharge
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"4.3.1.1 - Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CRANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.
i Table 4.3.1.1-1.
Reactor Protection System Instrumentation Surveillance Requirements Operational Conditions channel in which Functional Channel Functional Channel Surveillance Unit check-Test Calibration-Required-8.
Scram i
Discharge Volume Water Level-High NA M
R 1,2,5 Notation (a) A channel may be placed in an inoperable status up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
(h)
With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2 Action 4:
In. OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In OPERATIONAL CONDITION 5, suspend all operations involving CORE ALTERATIONS
- and fully insert all insertable control rods within one hour.
- Except movement of IRM, SRM or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2."
i Paragraph 3.3.1 and Table 3.3.1-1 of the Model Technical Specifications require the functional unit of SDV water level-high to have at least 2 operable channels containing 2 limit switches per trip system, a total of 4 operable channels containing 4 limit switches per 2 trip systems for the reactor protection system which automatically initiates a scram. The technical objective of these requirements is to provide 1-out-of-2-taken-twice logic for nklin Research Center ADammenedThe hasuenwummme.
TER-C5506-63/65 the reactor protaction system.
The response time of the reactor protection system for the functional unit of SDV water level-high should be measured and kept available (it is not given in Table 3.3.1-2).
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Paragraph 4.3.1.1 and Table 4.3.1.1-1 give reactor protection system instrumentation surveillance requirements for the functional unit of SDV water level-high.
Each reac N r protection system instrumentation channel containing a limit switch should be shown to be operable by the Channel Functional Test monthly and Channel Calibration at each refueling outage.
- 2. 3 LCO/ SURVEILLANCE REQUIREMENTS FOR CONTROL ROD WITHDRAWAL BIDCK SDV LIMIT SWITCHES The NRC staff's Model Technical Specifications specify the following LCO/
surveillance requirements for control rod withdrawal block SDV limit switches:
2 "3.3.6 - The control rod withdrawal block instrumentation channel shown in Table 3.3.6-1 shall be OPERABLE with trip setpoints set consistent j
with the values shown in the Trip Setpoint column of Table 3.3.6-2.
Table 3.3.6-1. Control l Rod Withdrawal Blcck Instrumentation Minimum Operable Applicable Channels Per Trip Operational Trip Function Function Conditions Action 5.
Water level-high 2
1, 2, 5**
62 b.
Scram trip bypassed 1
(1, 2, 5**)
62 ACTION 62: With the number of OPERABLE channels less than required by the minimum OPERABLE channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.
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- With more than one control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
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TER-C5506-63/65 Table 3.3.6-2 Control Rod Withdrawal Block Instrumentation Setpoints
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Trip Function Trip Setpoint Allowable Value 5.
Water level-high To be specified NA b.
Scram trip bypassed NA NA" "4.3.6 - Each of the above control rod withdrawal block trip systems and
' instrumentation channels shall be demonstrated OPERABLE by the o
performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencles shown in Table 4.3.6-1.
Table 4.3.6-1. Control Rod Withdrawal Block Instrumentation Surveillance Requirements Operational Conditions Channel in Which Trip Channel Functional Channel Surveillance Function Check Test Calibration Required 5.
Water Level-NA Q
R 1, 3, 5**
High b.
Scram Trip NA M
NA (1, 2, 5**)
Bypassed
- With more than one control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2."
Paragraph 3.3.6 and Table 3.3.6-1 of the Model Technical Specifications require the control rod withdrawal block instrumentation to have at least 2 operable channels containing 2 limit switches for SDV water level-high and 1 operable channel containing 1 limit switch for SDV scram trip bypassed.
The technical objective of these requirements is to have at least one channel containing one limit switch available to monitor the SDV water level when the other channel with a limit switch is being tested or undergoing maintenance.
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.-.,..... 7 TER-C5506-63/65 SDV water level-high should be specified as indicated in Table 3.3.6-2.
The trip function prevents further withdrawal of any control rod when the control
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rod block SDV limit switches indicate water level-high.
Paragraph 4.3.6 and Table 4.3.6-1 require that each control rod withdrawal block instrumentation channel containing a limit switch be shown to be operable by the Channel Functional Test once per 3 months for SDV water level-high, by the Channel Functional Test once per month for SDV scram trip bypassed, and by Channel Calibration at each refueling outage for SDV water level-high.
The Surveillance Criteria of the BWR Owners Subgroup given in Appendix A, "Long-Term Evaluation of Scram Discharge System," of " Generic Safety Evaluation Report BWR Scram Discharge System," written by the NRC staff and issued on 4
December 1, 1980, ares 1.
Vent and drain valves shall be periodically tested.
2.
Verifying and level detection instrumentation shall be periodically tested in place.
3.
The operability of the entire system as an integrated whole shall be demonstrated periodically and during each operating cycle, by demonstrating scram instrument response and valve function at pressure and temperature at approximately 50% control rod density.
Analysis of the above criteria indicates that the NRC staff's Model Technical Specifications requirements, the acceptance criteria for the present TER, fully cover the BWR Owners Subgroup Surveillance Criteria 1 and 2 and partially cover Criterion 3.
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METHOD OF EVALUATION The CE submittal for the Quad Cities Station Units 1 and 2 was evaluated in two stages, initial and final.
During the initial evaluation, only the NRC staff's Model Technical Specifications requirements were used to determine if:
o the Licensee's submittal was responsive to the July 7,1980 NRC request for proposed Technical Specifications changes involving the surveillance requirements of the SDV vent and drain valves, LCO/ surveillance requirements for reactor protection system SDV limit switches, and LCO/ surveillance requirements for control rod block SDV limit switches o the submitted information was sufficient to permit a detailed technical evaluation.
During the final evaluation, in addition to the NRC staff's Model Technical Specifications requirements, background material in References 1 through 10, pertinent sections of " Commonwealth Edison Quad Cities Station Units 1 and 2 Safety Analysis Report," and Quad Cities Technical Specifications were studied to determine the technical bases for the design of SDV main components and instrumentation.
Subsequently, the Licensee's response was
- compared directly to the requirements of the NRC staff's Model Technical Specifications. The findings of the final evaluation are presented in Section 4 of this report.
I The initial evaluation concluded that the Licensee's submittal was l
responsive to the NRC request of July 7,1980, but certain information was not l
available.
A Request for Additional Information (RAI) was sent to CE by the
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NRC on September 2, 1981. Thus, this TER is based on the initial submittal I
and the Licensee's response dated October 22,1981 (see Appendix C) to the RAI.
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TECHNICAL EVALUATION 4.1 SURVEILLANCE REQUIREMENTS FOR SDV DRAIN AND VENT VALVES NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS Paragraph 4.1.3.1.1 requires demonstrating that the SDV drain and vent valves are operable by:
verifying each valve to be open at least once per 31 days (valves may a.
be closed intermittently for testing under administrative controls) 4 b.
cycling each valve at least one complete cycle of full travel at least once per 92 days.
LICENSEE RESPONSE The Licensee proposed to revise page 3.3/4.3-3 of the Quad Cities Station Units 1 and 2 Technical Specifications by adding paragraph 6:
"6.
The scram discharge volume vent and drain valves shall be verified open at least once per 31 days. These valves may be closed intermittently for testing under administrative control. At least once each Refueling Outage, the scram discharge volume vent and drain valves will be demonstrated to:
a.
Close within 15 seconds after receipt of a signp1 for control rods to scram and I
b.
Open when the scram signal is reset."
and provided a revision of page 3.3/4.3-9 with this pertinent statement:
"6.
The operability of the Scram Discharge volume vant and drain valves assures the proper venting and draining of the volume, so that water accumulation in the volume does not occur. These specifications provide for the periodic verification that the valves are open, and for the testing of these valves under reactor scram conditions during each Refueling Outage. "
l The Licensee's answer to the RAI regarding cycling the drain and vent valves at least one complete cycle of full travel at least once per 31 days was as follows (see Appendix C):
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" CONCERN 1.
Commonwealth Edison's response in paragraph 3 does not contain the requirement of the Model Technical Specifications of paragraph 4.1.3.1.lb to cycle each valve at least one complete cycle of full travel at least once per 31 days.[*]
REQUEST 1.
Provide technical bases why the requested change is not applicable to Dresden Nuclear Power Station Unita 2 and 3.[**]
Response
The model Technical Specifications that were used as a reference to develop our submittal were attached to the July 7,1980 letter from D.
Eisenhut to all operating BWR licensees. The Franklin Research Center model Technical Specifications for Section 4.1.3.1.1 are not the same as the July 7,1980 model Technical Specifications.
In addition, the July 7, 1980, Model Technical Specifications were incorrect concerning the SDV vent / drain valve closure during individual CRD scram timing. The July 7, 1980, model is performed each refueling. A possible means of modifying our submittal would be to require verification of valve closure and subsequent re-opening during each acram, and take credit for that.
In summary it is our contention tt 1 our proposal is meaningful and provides a true test of the system."
The Licensee agreed to revise th'e proposed specification changes to require cycling each valve at least one complete cycle of full travel at least once per quarter..
EVALUATION The revised page 3.3/4.3-3 of the Quad Cities Station Units 1 and 2 Technical Specifications with the agreed-upon revision complies with the requirement of paragraphs 4.1.3.1.la and 4.1.3.1.lb of the NRC Staff's Model Technical Specifications which require verifying each valve to be open at least once per 31 days and cycling each valve at least one complete cycle of full travel at least once per quarter, respectively.
1
- Cn 10/22/81, the paragraph 4.1.3.1.lb was revised by the NRC:
"Once per 31 days" was replaced by "once per 92 days. "
- Appendix C is also applicable to Quad Cities Station Units 1 and 2. ddu Franklin Research Center aoen ner m rr.n n m
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TER-C5506-63/65 4.2 LCO/ SURVEILLANCE REQUIREMENTS FOR REACTOR PROTECTION SYSTEM SDV LLMIT SWITCHES NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS Paragraph 3.3.1 and Table 3.3.1-1 require the functional unit of SDV water level-high to have at least 2 operable channels containing 2 limit switches per trip system, a total of 4 operable channels containing 4 limit switches per 2 trip systems for the reactor protection system which automatically initiates scram.
s Paragraph 3.3.1 and Table 3.3.1-2 concern the response time of the reactor protection system for the functional unit of SDV water level-high which should be specified for each BWR (it is not specified in the table).
Paragraph
{
4.3.1.1 and Table 4.3.1.1-1 require that each reactor protection system instru-mentation channel containing a limit switch be shown to be operable for the functional unit of SDV water level-high by the Channel Functional Test monthly and Channel Calibration at each refueling outage. The applicable operational conditions for these requirements are startup, run, and refuel.
LICENSEE RESPONSE Pages 3.1/4.1-8, 3.1/4.1-9, 3.1/4.1-10, and 3.1/4.1-11 of the existing Quad Cities Statio3 Units 1 and 2 Technical Specifications address the NRC staff's Model Technical Specifications requirements of paragraph 3.3.1 and Table 3.3.1-1 by providing Table 3.1-1, " Reactor Protection System (Scram)
Instrumentation Requirements Refuel Mode," Table 3.1-2, " Reactor Protection l
System (Scram) Instrumentation Requirements Startup/ Hot Standby Mode," Table
(
3.1-3, " Reactor Protection System (Scram) Instrumentation Requirements Run Mode," and Table 3.1-4, notes for Tables 3.1-1, 3.1-2, and 3.1-3 with the following information for " Trip Function High-water level in scram discharge volume (4)":
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"1.
Minimum Number of Operable or Tripped Instrument Channels per Trip i
System (1) : 2 2.
Trip Level Setting: f 50 gallons l
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Action (2): A 4.
Action Required When Equipment Operability is Not Assured (1) : A
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NOTES:
1.
There shall be two operable trip systems or one operable and one tripped system for each function 2.
If the first column cannot be met for one of the trip systems, that trip system shall be tripped.
If the first column cannot be met for both trip systems, the appropriata actions listed below shall be taken:
A.
Initiate insertion of operable rods and complete insertion of all operable rods within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i 4.
Permissible to bypass, with control rod block for reactor protection system reset in refuel and shutdown positions of the reactor mode switch."
The requirements of paragraph 3.3.1 and Table 3.3.1-2 cf the NRC staff's Model Technical Specifications are cove'ed by page 3.3/4.3-10 of the Quad r
Cities Station Units 1 and 2 Technical Specifications which give the reactor protection system response time as follows:
"In the analytical treatment of the transients, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods.
This is adequate and conservative when compared to the typically observed time delay of about 210 milliseconds.
Approximately 90 milliseconds af ter neutron flux reaches the trip point, the pilot scram valve solenoid deenergizes and 120 milliseconds later the control rod motion is estimated to actually begin.
However, 200 milli-i seconds rather than 120 milliseconds is conservatively assumed for this time interval in the transient analyses and is also included in the allowable scram insertion times specified in Specification 3.3.C.
The scram times for all control rods will be determined at the time of each refueling outage. A representative sample of control rods will be tested following a shutdown.
Scram times of new drives are approximately 2.5 to 3 seconds; lower rates of change in scram times following initial plant operation at power are expected. The test schedule provides reasonable assurance of detection of slow drives before system deterior.ation beyond limits of Specification 3.3.C."
Pages 3.1/4.1-12, 3.1/4.1-13, and 3.1/4.1-14 of the present Quad Cities Station Units 1 and 2 Technical Specifications address the NRC staff's Model 4 Ndd Franklin Research Center Ac %.e w n.m n
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TER-C5506-63/65 Technical Specifications requirements of paragraph 4.3.1.1 and Table 4.3.1.1-1.
Pages 3.1/4.1-12 and 3.1/4.1-13 contain Table 4.1-1, " Scram Instrumentation
~~
and Icgic Systems Functional Tests Minimum Functional Test Frequencies for Safety Instrumentation, Iogic Systems, and Control Circuits," which provides the following information for " Instrument Channel High water level in scram discharge volume":
"1.
Group (3): A 2.
Functional Test (7): Trip channel and alarm 3.
Minimum Frequency (4): Every 3 months NOTESs 3.
A description of the three groups is included in the bases of this specification A.
On-off sensors that provide a scram trip function 4.
Functional tests are not required when the systems are not required to oe operable or are tripped.
If test are missed, they shall be performed prior to returning the systems to an operable status 7.
A functional test of the logic of each channel is performed as indicated. This coupled with placing the mode switch in shutdown each refueling outage constitutes a logic system functional test of the scram. system."
Page 3.1/4.1-14 contains Table 4.1-2, " Scram Instrument Calibration Minimum Calibration Frequencies for Reactor Protection Instrument Channels." The minimum calibration frequency for Instrument Channel High water level in scram discharge volume should be listed in this table.
It is not. The Licensee's response to the RAI regarding the reactor protection system SDV Channel Functional Test and Channel Calibration is given below (see Appendix C).
~
" REQUEST 2.
~
The technical Specifications for Dresden 2 and 3(*] state that each reactor protection system scram discharge volume water level-high instru-mentation channel containing a limit switch shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST once per 3 months, and A 5 nk!!n Research Cente Acw mr r
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4 TER-C5506-63/65 there are no tech specs for CHANNEL CALIBRATION.
Since the proposed frequency of the required surveillance for Dresden Nuclear Power Station Units 2 and 3 differs from the frequency requested by the Model Technical Specifications, provide technical bases for it.
RESPONSE
The proposed Technical Specifications regarding the SDV scram and rod block level switches are adequate.
A monthly functional test of the SDV scram bypass would require the reactor mode switch to be placed in either SHUTDOWN or REFUEL for the test, and this is unreasonable.
REQUEST 3.
Provide technical bases for not calibrating the scram discharge volume water level-high instrumentation channel.
Also provide technical basis for performing the scram trip bypassed jnstrumentation channel functional test once per refueling outage instead of once per, month as requested in the Model Technical Specifications.
l
RESPONSE
Magnetrol level switches are not, and cannot, be calibrated.
Therefore, calibration frequency in our submittal is designated as 'Not Applicable' for the scram discharge volume water level high cnannel."
l The Licensee is installing a second instrument volume containing four additional limit switches, for a total of eight limit switches for the reactor protection system.
In addition, the Licensee agreed to revise page 3.1/4.1-14 of the Quad Cities Station Units 1 and 2 Technical Specifications to incorpo-rate in Table 4.1-2 the Calibration Test each refueling for " Instrument t
Channel-SDV Water Level High. " The Calibration Test will consist of physical inspection and actuation of the level switches using water columns.
EVALUATION Pages 3.1/4.1-8 through 3.1/4.1-11 of the existing Quad Cities Station Units 1 and 2 meet the NRC staff's Model Technical Specifications requirements
- Appendix C is also applicable to Quad Cities Station Units 1 and 2.
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TER-CS506-63/65 of paragraph 3.3.1 and Table 3.3.1-1 in regard to the minimum number of operable instrument channels per trip system and number of trip systems, and are acceptable. The Quad Cities Station Units 1 and 2 reactor protection system SDV water level-high instrumentation consists of 2 operable channels containing 2 limit switches per trip system, for a total of 4 operable channels containing 4 limit switches per 2 trip systems, making 1-out-of-2-taken-twice logic. The specified trip level setting of < 50 gallons for scram initiation and operating modes of refuel, startup/ hot standby, and run are also acceptable.
The reactor protection system response time of 290 milliseconds specified on original page 3.3/4.3-10 of the Quad Cities Station Units 1 and 2 Technical Specifications is acceptable and addresses the requirements of paragraph 3.3.1 and Table 3.3.1-2.
The provision of the present Quad Cities Station Unita 1 and 2 Technical Specifications given on page 3.1/4.1-12, Table 4.1-1, for a reactor protection system SDV water level-high Channel Functional Test once per 3 months does not meet the NRC staff's Model Technical Specifications requirement of paragraph 4.3.1.1 and Table 4.3.1.1-1 for the Channel Functional Test to be performed monthly. However, the Licensee is installing a second instrument volume containing four additional limit switches, for a total of eight limit switches for the reactor protection system. This increases significantly the reliability of the system and provides technical bases for acceptance of the proposed surveillance requirweents to perform Channel Functional Test quar terly.
The Licensee agreed to revise page 3.1/4.1-14 of the Quad Cities Station Units 1 and 2 Technical Specifications to incorporate in Table 4.1-2 the Calibration Test "each refueling" for " Instrument Channel - SDV water level l
high."
The Calibration Test will consist of physical inspection. and actuation l
of the level switches using water columns. This meets the NRC staff's Model Technical Specifications requirements of paragraph 4.3.1.1 and Table 4.3.1.1-1 and is acceptable.
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TER-C5506-63/65 4.3 LCO/SURVEILLANC3 REQUIREMENTS FOR CONTROL ROD WITHDRAWAL BLOCK SDV LIMIT SWITCHES NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS Paragraph 3.3.6 anel Table 3.3.6-1 require the control rod withdrawal block instrumentation to have at least 2 operable channels containing 2 limit switches for SDV water level-high and 1 operable channel containing 1 limit switch for SCV trip bypassed.
Paragraph 3.3.6 also requires specifying the trip setpoint for control rod withdrawal block instrumentation monitoring SDV water level-high as indicated in Table 3.3.6-2.
Paragraph 4.3.6 and Table 4.3.6-1 require each control rod withdrawal block instrumentation channel containing a limit switch to be shown to be operable by the Channel Functional Test once per 3 months for SDV water level-high, by the Channel Functional Test once per month for SDV scram trip bypassed, and by Channel Calibration at each refueling outage for SDV water level-high.
LICENSEE RESPONSE The Licensee proposed to revise pages 3.2/4.2-14 and 3.2/4.2-16 of the
~ Quad Cities Stations Units 1 and 2 Technical Specifications. On page 3.2/4.2-14, Table 3.2-3, " Instrumentation that Initiates aod Block," provides the following information:
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TER-C5506-63/65
" Table 3.2-3 Minimum Nuncer of Operable Trip
~~
or Tripped Instrument Instrument Level Channels per Trio S/ stem (1)
Setting i
High water level in scram
< 25 gallons discharge volume (SDV) 1 SDV high water level scram NA*
trip bypassed NOTE:
1.
For the Startup/ Hot Standby and Run positions of the reactor mode selector switch, there shall be two operable or tripped trip systems for each function except the SRM rod blocks.
IRM upscale and IRM downscale need not,be operable in the Run position, APRM downscale, APRM upscale (flow biased), and RBM downscale need not be operable in the Startup/ Hot Standby mode. The RBM upscale need not be operable at less than 30% rated tneraal power. One channel may be bypassed above 30% rated thermal power provided that a limiting control rod pattern does not exist. For systems with more than one channel per trip system, if the first column cannot be met for one of the two trip systems, this condition may exist fer up to 7 days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than 7 days the system shall be tripped. If the first column cannot be met for both trip systems, the system shall be tripped."
on page 3.2/4.2-16, Table 4.2-1, " Minimum Test and Calibration Frequency for Core and Containment Coolihg Systems Instrumentation, Rod Blocks, and Isolations," line 12 was added as shown in the following tables
- This line was added in the revised edition.
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TER-C5506-63/65-
" Table 4.2-1 Instrument Instrument Instrument Channel Functional Test (2)
Calibration (2)
Check (2)
Rod Blocks 11.
High water level in scram discharge Once/3 months Not applicable
.None volume (SDV) 12.
SDV high level trip Refueling outage Not applicable None bypassed NOTES:
2.
Functional test, calibrations, and instrument checks are not required when those instruments are not required to be operable or are tripped."
The Licensee's response to the RAI, Requests 2 and 3, given in Section 4.2 of this report, is also applicable to this section.,
In addition, the Licensee agreed tot 1.
delete " Instrument Channel-SDV high water level scram trip bypassed" from Table 3.2-3 (revised page 3.2/4.2-14) and Taole 4.2-1 (revised page 3.2/4.2-16) 2.
revise the first sentence of Note 1, Table 3.2-3 on page 3.2/4.2-14, to state the following or its equivalent:
"For the Startup/ Hot Standby and Run positions of the reactor mode selector switch, there shall be two operable or tripped trip systems for each function except the SRM rod blocks and ' Instrument-Rod Block, SDV high water level'."
3.
incorporate into Table 4.2-1 (revised page 3.2/4.2-16).the Calibration Test "each refueling" instead of "Not applicable" for
" Instrument-Rod Block, High water level in scram discharge volume."
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EVALUATION l
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The existing Quad Cities Station Units 1 and 2 scram discharge system has six. level switches on the scram discharge volume (see FSAR page 3'.5-5 and Section 10.6) set at three different water levels to guard against operation of the reactor without sufficient free volume present in the scram discharge headers to receive the scram discharge water in the event of a scram. At the first (lowest) level with a setpoint of 3 gallons (see FSAR Table 7.7.1, A. N ranklin Research C
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TER-C5506-63/65
" Typical Protection Systems Setpoints"), one level switch initiates an alarm for operator action. At the second level with a setpoint of < 25 gallons f
(see the present Quad Cities Technical Specifications, page 3.2/4. 2-14, Table f
3.2-3), one level switch
- initiates a rod withdrawal block to prevent further withdrawal of any control rod.
At the third (highest) level with a setpoint I
of 50 gallons (see FSAR, Table 7.7.1), the four level switches (two for each l
reactor protection system trip system) initiate a scram to shut down the reactor while sufficient free volume is available to receive the scram dis-charge water.
Reference 9, page 50, defines Design Criterion 9 -("Instrumenta-tion shall be provided to aid the operator in the detection of water accumula-tion in the instrumented volume (s) prior to scram initiation"), gives the technical basis for "Long-Term Evaluation of Scram Discharge System," and defines acceptable compliance ("The present alarm and rod block instrumentation meets tais criterion given adequate hydraulic coupling with the SDV headers").
Thus, if the Quad Cities Station Units 1 and 2 scram discharge system is 1
modified (long term) so that the hydraulic coupling between scram discharge j
headers and instrumented volume is adequate and acceptable, then the present alarm and rod block instrumentation consisting of one trip system with one instrument channel containing one limit switch is also acceptable.
when the reactor of Quad Cities Station Units 1 and 2 is in operational l
conditions of startup and run, " Scram Discharge Volume Scram Trip" cannot be bypassed, and operational condition " refuel with more than one control rod withdrawn" is not applicable (see FSAR, page 3.5-5:
" Interlocks are provided whicn prevent the inadvertent withdrawal of more than one control rod with the mode switch in the refuel position"). Thus, the NRC staff 's Model Technical Specifications requirements of paragraph 3.3.6, Table 3.3.6-1, paragraph 4.3.6, and Taole 4.3.6-1 for " Trip Function 5. b. Scram Discharge Volume Scram Trip
- The existing SDV system in regard to control rod withdrawal block SDV limit switches does not comply with the present Quad Cities Technical Specifications which require two trip systems with two level switches (see page 3.2/4.2-14, Table 3.2-3, and Note 1).
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TER-C5506-63/65 Bypassed" are not applicable to Quad Cities Station Units 1 and 2 for the specified operational conditions. Therefore, the Licensee agreed to delete
" Instrument Channel-SW high water level scram trip bypassed" from the proposed revision of page 3.2/4.2-14, Table 3.2-3, and page 3.2/4.2-16, Table 4.2-1.
Since the existing system has only one trip system with one instrument channel containing one control rod withdrawal block SW limit switch and is acceptable, to reflect this, the Licensee agreed to revise the first sentence of Note 1 in Table 3.2-3 on page 3.2/4.2-14 as follows or its equivalent:
"For the Startup/ Hot Standby and Run positions of the reactor mode selector switch, there shall be two operable or tripped trip systems for each function except the SRM rod blocks and ' Instrument-Rod Block SW high water level'."
The specified Trip Level Setting of < 25 gallons in Table 3.2.3, page 3.2/4.2-14, for control rod withdrawal block instrumentation, monitoring SW water level-high, is acceptable.
It meets the NRC staff's Model Technical Specifications requirements of paragraph 3.3.6 and Table 3.3.6-2.
Since the Licensee agreed to incorporate into Table 4.2-1, on revised page 3.2/4.2-16, the Calibration Test "each refueling" instead of "Not' applicable" for " Instrument-Rod Block, high water level in scram discharge volume," the revised Table 4.2-1 will comply with the NRC staff 's Model Technical Specifications requirements of paragraph 4.3.6 and Table 4.3.6-1 for the Channel Functional Test once per 3 months and Channel Calibration each refueling for control rod withdrawal block SW water level-high.
The Channel Calibration will consist of physical inspection and actuation of the level switch using water column.
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CONCLUSIONS Table 5-1 summarizes the results of the final review and evaluation of the Quad Cities Station Units 1 and 2 Phase 1 proposed Technical Specifications changes for SDV long-term modification in regard to surveillance requirements for SDV vent and drain valves and LCO/ surveillance requirements for reactor protection system and control rod block SDV limit switches.
The folicwing conclusions were mader o
The revised page 3.3/4.3-3, with the Licensee's agreement to incorporate a revision into the proposed specifications changes that requires cycling each valve at least one complete cycle of full travel at least quarterly, complies with the NRC staff's Model Technical Specifications requirements of paragraphs 4.1.3.1.la and 4.1.3.1.lb.
o The original pages 3.1/4.1-12 and 3.1/4.1-13, Table 4.1-1, of the Quad Cities Station Units 1 and 2 Technical Specifications, which provide for the reactor protection system SDV limit switches water level-high Channel Functional Test to be performed once per 3 months, do not meet the surveillance requirement (paragraph 4.3.1.1, Table 4.3.1.1-1, of the NRC staff's Model Technical Specifications) for the test to be performed monthly. However, the Licensee is installing a second instrument volume containing four additional limit switches, for a total of eight limit switches, for the reactor protection t
system. This increases significantly the reliability of the system and provides technical bases for acceptance of the proposed -
surveilladce requirements to perform the Channel Functional Test quarterly.
To meet the NRC staff's Model Technical Specifications requirements o
of paragraph 4.3.1.1 and Table 4.3.1.1-1, the Licensee agreed to revise page 3.1/4.1-14 of the present Quad Cities Station Units 1 and 2 Technical Specifications to incorporate into Table 4.1-2 the Calibration Test "each refueling" for " Instrument Channel - SDV Water Level High."
The Calibration Test will consist of physical
~
l inspection and actuation of the level switches using water columns.
o The NRC staff's Model Technical Specifications surveillance requirements in paragraph 3.3.6, Table 3.3.6-1, paragraph 4.3.6, and Table 4.3.6-1 for control rod block SDV scram trip bypassed are not applicable to the operational conditions of startup, run, and refuel with more than one control rod withdrawn. Therefore, the Licensee agreed to delete " Instrument Channel-SDV high water level scram trip i
bypassed" from the proposed revision of page 3.2/4.2-14, Table 3.2-3, and page 3.2/4.2-16, Table 4.2-1.
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TER-C5506-63/65 o
The existing SDV system has only one trip system with one instrument channel containing one control rod withdrawal block SDV limit switch and meets NRC criteria.
To reflect this, the Licensee agreed to revise the first sentence of Note 1, in Table 3.2-3 on original page 3.2/4.2-14.
o To meet the NRC staff's Model Technical Specifications requirements of paragraph 4.3.6 and Table 4.3.6-1, the Licensee agreed to incorporate into Table 4.2-1 on revised page 3.2/4.2-16 the Calibration Test "each refueling" instead of "Not applicable" for instrument Channel-Rod Blocks, high water level in scram discharge volume." Channel Calibration with the Magnetrol level switch will consist of physical inspection and actuation of the switch using a water column.
o The remaining surveillance requirements are met by revised pages 3.2/4. 2-14, 3.2/4. 2-16, 3.3/4.3-3, 3.3/4.3-9, and original, unrevised pages 3.1/4.1-12, 3.1/4.1-13, 3.1/4.1-14, and 3.3/4.3-10 of the Quad Cities Station Units 1 and 2 Technical Specifications.
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Table 5-1 Evaluation of Phase 1 Proposed Technical Specifications Changes
[5 for Scram Discharge Volume Iang-Tern Modifications a
Quad Cities Station Units 1 and 2 i
N's N
Technical Specifications 3
g NRC Staff Model Proposed by Surveillance Requirements (Paragraph)
Licensee Evaluation l
C SDV DRAIM AND VENT VALVES l
{
verify each valve opea Once per 31 days Once per 31 days Acceptable (4.1.3.1.la)
(pp. 3.3/4.3-3 and 3.3/4.3-9 revised)
Cycle each valve one Once per 92 days Once per quarter Acceptable f.
E complete cycle (4.1.3.1.1b)
(p. 3.3/4.3-3, second f
revision)
I i
REACTOR PROTECTION SYSTEM SDV LIMIT SWITCNES l
Minimum operable channels 2
2 Acceptable i
per trip system (3.3.1, Table 3.3.1-1)
(pp. 3.1/4.1-8 to 3.1/4.1-11) i SDV water level-high NA 0.290 seo max.
Acceptable j
response time (3.3.1, Table 3.3.1-2)
(p. 3. 3/4. 3-10)' ' '
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SDV water level-high Channel functional test Monthly Every 3 months Acceptable (4.3.1.1, Table 4.3.1.1-1)
(pp. 3.1/4.1-12 and (see p. 20 3.1/4.1-13) of this TER) m E
Channel calibration Each refueling.
Each refueling Acceptable os (4.3.1.1, Table 4.3.1.1-1)
(p. 3.1/4.1-14, 1
Table 4.1-2, to be 5
revised)
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Technical Specifications a5 NRC Staff Model Proposed by N
Surveillance Requirements (Paragraph)
Licensee Evaluation N
CDNTROL ROD BIDCK SDV LIMIT SWITCHES
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Minimum operable channels
,g per trip function SDV water level-high 2
1 Acceptable *
(3.3.6, Table 3.3.6-1)
(p. 3.2/4.2-14, Table 3.2-3, revised)
Not applicable AccQptable*
(3.3.6, Table 3.3.5-1)
(p. 3.2/4.2-14, h
Table 3.2-3, second revision) y SOV water level-high Trip setpoint NA
. < 25 gallons Acceptable (3.3.6, Table 3.3.6-2)
(p. 3.2/4.2-14, Table 3.2-3, revised)
Channel functional test Quarterly Once per 3 months Acceptable (4.3.6, Table 4.3.6-1)
(p. 3.2/4.2-16, Table 4.2-1, revised)
Channel calibration Each refueling Each refueling Acceptable (4.3.6, Table 4.3.6-1)
(p. 3.2/4.2-16, Table 4.2-1, second revision)
N Channel functional test Monthly R>t applicable Acceptable
- f (4.3.6, Table 4.3.6-1)
(p. 3.2/4.2-16, in Table 4.2-1, second revision) as
- See Reference 9, p. 50, and pp. 21 to 25 of this TER.
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TER-C5506-63/65 6.
REFERENCES 1.
" Degradation of BWR Scram Discharge Volume Capability" NRC, Office of Inspection and Enforcement, June 12, 1980 IE Bulletin 80-14 2.
D. G. Eisenhut (NRR)
I4tter "To All Operating Boiling Water Reactors (BWRs) " with enclosure, "Model Technical Specifications" July 7, 1980 3.
" Failure of 76 of 185 Control Rods to Fully Insert During a Scram at
~
a BWR" NRC, Office of Inspection and Enforcement, July 3, 1980 IE Bulletin 80-17 1
4.
Supplement 1, " Failure of 76 of 185 Control Rods to Fully Insert During a Scram at a BWR" NRC, Office of Inspection and Enforcement, July 18, 1980 IE Bulletin 80-17 5.
Supplement 2, " Failures Revealed by Testing Subsequent to Failure of Control Rods to Insert During a Scram at a BWR" NRC, Office of Inspection and Enforcement, July 22, 1980 IE Bulletin 80-17 6.
Supplement 3, " Failure of Control Rods to Insert During a Scram at a BWR" NRC, Office of Inspection and Enforcement, August 22, 1980 IE Bulletin 80-17 7.
Supplement 4, " Failure of Control Rods to Insert During a Scram at a BWR" NRC, Office of Inspection and Enforcement, December 18, 1980 IE Bulletin 80-17 8.
Supplement 5, " Failure of Control Rods to Insert During a Scram at a BWR" i
NRC, Office of Inspection and Enforcement, February 13, 1981 i
9.
P. S. Check (NRR)
Memorandum with enclosure, " Generic Safety Evaluation Esport BWR Scram Discharge Systas" December 1,1980 i
10.
P. S. Check (NRR) i Memorandum with enclosure, " Staff Report and Evaluation of Supplement 4 to IE Bulletin 80-17" June 10, 1981 C_.s.
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i APPENDIX A NRC STAFF'S MODEL TECHNICAL SPECIFICATIONS i
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- Note: Applicable changes are marked by vertical lines in the margins.
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.__...na TER-C5257-63/65 7
REACTIVTTY CCNTROL SYSTEMS LIMITING CCNDITICN FOR OPERATICN (Continued)
ACTICN (Continued) 2.
If the inoperable control rod (s) is inserted, within one hour disarm the associated. directional control valves either:
a)
Electrically, or D)
Hydraulically by closing the drive water and exhaust water isolation valves.
3.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
c.
With mere than 8 control rods inoperable, be in at least H'OT SHUTD0hlt within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:
a.
Verifying each valve to be open* at least once per 31 days and b.
Cycling each valve through at least one complete cycle of full travel at least once per 92 days.
4.1.3.1.2 When above the preset power level of the RWM and RSCS, all withdrawn control rods net required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:
a.
At least once per 7 days, and o.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when any control rod is immovable as a result of excessive friction or mechanical interference.
4.1.3.1.3 All control rods shall be demonstrated CPERA8LE by performance of Surveillance Requirements 4.1.3.2, 4.1.3.4,.4.1.3.5, 4.1.3.6 and 4.1.3.7.
"These valves may De closed intermittently for testing under ac:nihistrative Controls.
ia GE-STS 3[4 1-'
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nklin Research Center A chassen of r:t Fransen anomasse
TER-C5257-63/65 REACTIVITY C0 CRCL SYSTEuS CNTROL 700 FAX! MUM SCRAM INSERTION TIMES LIMITING CONDITION FOR CPERATION 2.1. 2. 2 The =tximus scre insertion ti== of each c:r. trol red from the fully withdrawn position to notch position (5), based on de-energization of the scram pilot valve solencids as time :ero, shall not exceed (7.0) seconds.
APPLICASILITY: OPERATIONAL'f.ONDITICHS 1 and 2.
ACTTCN:
Vith the maximurs scrara insertion time of one or, ore control rods excaeding-(7.0) seconds:
Declare the control rod (s) with the slow insertion time inoperable, a.
and b.
Perform the Surveillance Requirements of Specification 4.1.3.2.c at least once per 50 days when operation is edntinued with three or more control ' reds with maximum scram insertion times in excass of (7.0) seconds, or c.
Ee in at least HOT SHUTDCVN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE RECUIREWENTS 4.1. 3. 2 T'ha maximum scram insertion time of the control rods shall be dem:n-strated th-ugn measure ent with reacter c:alant pressurt greater can or
'acual to 850 psig and, durin; single c:ntr:1 rod scram tice tests, the emntrol red drive pumps isolatad from the ace.culat:rs:
For all centrol rods prior to THERMAL PCVER exceeding 4G% of RAf!D a.
THERMAL PCVER following C3RE ALTERATIONS or after a reactor shutdown that is greatar than 120 days,
~
b.
For specifically affected indivic 11 c:ntrai reds following.aintenance on or modification to the centrol ted =r control red drive system which could affect the scram inser". Ton time =f thosa specific con.rs) i rods, and i
c.
For IC% of the control reds, on a rotating tasis, at leas. once per 120 days of c;eration.
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2/4.3 INS ILHENTATICN 3/4.3.1 REACTCR PROTICTTCH SYSTEM INSTEL'u!NTATICH L!wI IN3 CCNDITTCH FCR CPEEATICN
- 3. 3.1 As a =inf u=, the react:r pr: taction rysta: inst:.:::.ntation channais sa:en in Tule 3.3.1-1 shall be CPERAELE with the REACTOR PE3TECTICH S'fSTT4 7.13?CNSE IME as shown in Table 3.3.1-2.
17:LI: ABILITY: As shown in Tabit 3.3.1-1.
/~~iCN:
Vith. the numbe'r of CPERABLE channels less than required by the Minimum a.
CPE?ABLE Channels per Trip System requiracent for one trip systam, place at least one inopertale channel in the tripped esndition within one hour.
Vith the number of CPERASLE channels less than required by the Ninimus
- PE?A!LE Channels per Trip System requirement for hc2 tris systams, placs at least one inoperable channel in at least one trio systas" in the tri::ed c:ndition wicin one hcur and taka taa ACTICH required by Tule 3.3.1-1.
c.
The : revisions of Specification 3.0.3 are not a;plicable in CPERATICHAL C*,h0ITICN 5.
.c.'IVE*LLuCE RECUIREkENTS
' 4.3.1.1 Each rtectar pr.tection system instrumentation channel shall be
- an:..strated CFE?ASLE by the per'ar=anca cf the CHANNEL CiECX, C4ANNEL FUNCTICNAL T257 and CFANNEL CALI3 RATION :perations for the CPE*ATICNAL CNCITIONS and at tae frequencies shown in Tule 4.3.1.1-1.
4.3.1.2 LOGIC SYSTEM FUNCTICNAL TESTS and sia'ulated automatic operati=n of a.11 enannels shalf he performed at least onca per 18 months.
l 4.3.1.3 The PSCTOR PROTECT *CN.SY3 TEM RESPCMSE TIME of each reacter trip "un:tica shewn in Table 3.3.1-2 shall be deconstrated to be within its 11eit at least enca ;er is sonths. Esca test shall include at least c9e logic train su:n tea, esta logic trains a n tasted at least ::: per 35==nt 5 and sne enannel ;er fuction such that all channels are tas ad at least e... e ever N ti=ss 13 :ntas where N is the total nuc er of radundant channels in a,.y 4;e:ific sact:r trip function.
. :: n :nanneis art incperable in one tri; rystam, select at least one
^
in::ercie c.annt) in that trip system ts :laca in the tripsed c:ndition, t.x: :: wasn this -culd cause the Trip Function ta ec:ur.
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TER-C5257-63/65 T:SLE 3.3.1-1 (Continued)
- Et:~0:7 FOTECTION SYSTEM INSTRUMENTATION ACTICN 10TI:N1 In C?ERATICdG'. CONDITICH 2, be in at least HOT SMUTDCW within 6 ho:rs.
In CFERATICN!L CCHDITICN 5, suspend all operations involving CCP.E ALTIRATION5* and fully insert all insertable control rods within o.e hour.
1CTION 2 Lock the mec:ce mode swit:h in the Shutdown position within one tour.
AOU:N3 Se is at le.ast STARTUP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
ACU:N4 In 0:U.A IGNE CCNDITION 1 or 2, be in at least HOT SHUTC%N within 6 heurs.
In 0?FX'IONE CONDITICH 5, suspend all opeiations involving CORE ALTIRATI*N5" and fully insert all insertable control rods within ore hoar.
1* TION 5 Se i: at least HOT SHUTDCW within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
CI",N 5 5e 1: ST;ATU? vith the main staas line isolation valves closed Wthin 2 heu:s or in at least HOT SHUTDL'N within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
A; TION 7 Initfa a a escucitan in THER.".;L PC4ER within 15 minutas and redu:e t:-::ine first stage pressure to < (250) psig, equivalent i
to T.iE*ML PC4' ER less than (30)% of FATED TriERFAL PC'4ER, within 2 ho:rs..
A:U:NS In C?PAUCNA*. CONDITION 1 or 2, be in at least. HDT SHUTCaw wi.hin 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
In CFE?. CONE CON 0!TICH 3 or 4, verify all insertable c:ntrol rods t.s te fully inserted 4th,in one hour.
I In 0?E*fiCNA* CONDITION 5, suspend all eperations involving CCRI ALTERATION 5" and fully insert all insertable' control rods within are hoar.
l A TION 9 In-C?:.?A70MA:. CONDITION 1 or 2, he in at least H3T SHUTUGH within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
^
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t In C7E3,ATICNR CONDITICN 3 or 4, lock the reactor mode switch in the Sh.: do.n p=sition within one hour.
In C?E:.CONA*. CONDITICN 5. suspend all c;erations involving I
CORI ALU?.AU:HS* and fully insert all insarta.ble c:ntrol rods within c e n::r.
I
- '.z:::: ovecent of 1.M SRM cr special c:vable datectors, or replacement of
? ?.d strin;s ;r:vicec 5??. i:stru entati:n is C?!?JELE per Specificati:n 3.3.2.
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TER-C5257-63/65 TA3LE 3.?.1-1 (Continued)
REA TOR P".CTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channo) may be placed in an inopeiable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillar.:e without placing the trip systes in the tripped condition provided at least one CPERASLE channel in the same trip system is sonitoring that ; ara =eter.
b)
The " shorting links' shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn
- and shutdown =argin damonstrations pa.-formed per Specification 3.10.3.
(c) An APRM channel is incparable if there are less than 2 LPRM inputs per level or less tha.n (11) LPRM inputs to an APRM channel.
(d) These functions are not required to be OPERA 3LE when the reactor pressurs vessal head is ur.2 cited or removed per Specification 3.10.1.
(e) This function shall be automatically bypassed when the reactor = cts switch is not in the Run position.
(f) This function is not required to be CPERABLE when PRI. WARY CONTAINMENT INTEGRITY is not re uired.
(g) Also actuates the stancy gas treat.:nent system.
(h) With any control r.d withdrawn. Not applicable to c:ntrol rods removed per Specification 3.S.10.1 or 3.9.10.2.
(i) These fun *c-fens a. e auto atically by;assed <nen tu-bine first stage
- ressure is < (250) psig, equivalent to THER"AL P%'ER less than (30)%
of PATED TMEPuAI. FC'ER.
4 (j) Also actuates the E3C-RPT system.
" Net requirac fer co.t ci rods removed per Specification 3.9.10.1 or 3.9.10.2.
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- 57tVVENTATIC.4 2 't. 2. 6 CCfGOL RCD VIT-!CRAVAL ELCCX TNST3LHENTATICN L*M**!NG CCNGITICN FOR C:ERATION
- 1. 3. 5.
The contn' red withdrawal block instrumentatica channels shown in Tcle 3.3.5-1 shall be CPEFJ3LE whh their trip setpcints set csnsistant with
-. values sh:vn in the Trip ietpoint coluca of T c le 3.3.5-2.
A:?LICA!ILITY: As shcwn in'Tahle 3.3.5-1.
A"T*CN:
With a control red withdrawal block instr::entation channel trip a.
setpoint less conservative than the value shown in the A11cwahle Values eslumn of Tule 3.3.5-2, declare the channel inoperable until' the channel is restored to CPERABLE status with its trip setpoint adjusted consistant with the Trip 5etpoint value.
b.
With the number of CPERAELE channel.3_ Jess than required by the Minimu= OP!EABLE Channels ;er Trip Func-ica, requirement, taka the ACTICH required by Table 2.3.5-b The provisions of Specification 3.0.3 tre net a,pticahia in OPERA-e.
TICNAL CCNDITICN 5.
!*.'iVIILLANCE REOUI19 DENTS 4.3.5 Each of the above required control r:d withdrawal block trip systacs and instru=entation enahnels shall be da ce.stratte CPEEAELE by the perf:::ance
- f t.e CMANNil CMECK, CMANNEL FUNCTICNAL TE3T anc CMANNEL CALIEPATICN :pera-l
-i:..s f r the' CFERATICHAL CCNDITICNS and at the frequencias sn:wn in Tacle 4.3.5-1.
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61 8
2 5
61
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Upscale (c) 3 2
61 2
5 61 Inuperative(C)
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2, 5 61 Inoperatig) c.
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~
a.
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1 2, 58*
62 g
h ' Scram trip Dypasseil I
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62 g
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1 62 y
c.
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62
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TER-C5257-63/65 TAELE 3.3.5-1 (Continued)
CCNTROL 700 VITHOP.AVal ! LOCK INST?M'.INTATION ACTION AZIN 50 T ke the ACTICN ttquired by 5;ecificati:n 3.1.4.3.
A;T~:N 51 With the nu=ter of CPERABLE Channels:
Cne less than required by the Mini::.. 07EFJELE Channels a.
per Trip function requirement, restore the inoperable channel to C?EPABLE status within 7 days er place the ineparable channel in the tripped c:ndition within the next hour.
b.
Two or scre less than re:;uired by the Mini =um CPE?a!LE Cha.nels per Trip Function recuirement, piaca at least one inoperable channel in the tripped, condition within one hour.
AZ:M $2 With the nu=ter of CPEPA3LI :hannels less' than requi ed by the Mini =um CPERABLE Channels ;'er Tri; Funcion equirement, place the in:perable channel in the tri;;ed c:nditica within ene hour.
NOTES
~
Wita TdEF#AL PCVER 1 (20)% of RATED TdE7?AL PCKER.
With =cre than ene e:ntrol red withcrawn. Not ap;11camle to cente:1 rods rec:ved per 5;eciffcation 3.9.10.1 cr 3.9.10.2.
Ths REM shall b's aut==atically bypassed wnen a peripheral c:ntrol red is '
t.
selected.
This func-icn shall be automatically bydassed if datactar esunt rata is t.
> 100 c;s or the !?Ji channels are on range (2) er higher.
This funni:n shall te automatically bypassed wnen the associated IFJi
- nannels are en range 8 or higher.
This fun: tion shall to automatically bypassed when the IRM channels art t.
- n ? nge 3 or higner.
Tais func-icn shall be aut matica11y by;assed when the IPJi channeIs ire a.
- n range 1.
j
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TER-C5257-63/65 ET,*.E 4.3.5-1 (Continued)
C0hiROL 203 VITHORAVAL ?t0CX INSTRUwENTATION !URVEILLt.NCE RECUIREMEhTs NOTES:
a.
Neut.:n detact:rs cay be excluded fr:m CHANNEL CALIE.uTICN.
b.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior is startup, if not perfor=ed within the previous 7 days.
When making an unscheduled change fr: CPERATIONAL CONDITION i to c.
CPE.MTICSAL CCSDITION 2, perfor= the recuired surveillanca within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after antaring OPERATIONAL CONDITICM 2.
Wits THERMAL PCWER 1 (20)% of RATED THERFAL PCWE2.
Vith any centrol red withdrawn. Not a:plicable to control rods removed per Specification 3.3.10.1 or 3.9.10.2.
6 e
I 25-i 3 I/4 2-55 nklin Resear
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=~:-
e APPENDIX B COMMONWEALTH EDISON LETTER OF OCTOBER 14, 1980 AND SUBMITTAL WITH PROPOSED TECHNICAL SPECIFICATIONS CHANGES FOR QUAD CITIES STATION UNITS 1 AND 2 4
d0 Franklin Research Center 4 w.a n. re en =
~
~
^ - - ---~~~~~-- -
TER-C5506-59/62 f
kEGULATORY INFORMATION DISTRIBUTION SYSTEM CRIDS) j ACCESSION NBR: 6010240203 OOC.DATE: 80/10/14 NOTARIZED: YES 00CXET s
~~
FACIL50-237 Dresden Nuclear Power Station, Unit 2, commonwealen E 05000237 50-249 Oreseen Nuclear Pawer Station, Unit 3, Commonwealth E 05000249 50-254 Quao-Cities Station, Unit 1, commonwealth Edison Co.
05000254 50-265 Quac-Cities Station, Unit 2, Commonwealth Eatson Co.
05000265 AUTH.NAME AUTHOR AFFILIATION JANECEK,R.F.
Commonwealth Eoison Co.
HECIP.NAME 4ECIPIENT AFFILIATION 6
Office of Nuclear Reactor Regulation, Director SuSJECT: Forwards proposed Tech Spec changes,per NRC 800707 request.
Cnenges provice acci operational & surveillance requirements.
COPIES RECEIVED:LTR k ENCL SIZE: N k
DISTRIBUTION CODE: A0013 TITLE: General Distrioution for after Issuance af coeroting Lice'nse NOTES:1 copy:SEP Sect. Lor.
05000237 RECIPIENT COPI E"3 RECIPIENT COPIES 10 CODE /NAME LTTR ENCL 10 CODE /NAME LTTR ENCL ACTION:
CRUTCMFIELO. 04 13 13 IPPOLITO,T.
04,
13 13 8 MTER t. A L : D/DIR,$uM FACOS 1
1 11E 06 2
2 NRC POR 02 1
1 OELD 11 1
0 GR. ASSESS SR 10 1
0 REG FILE 01 1
1 EXTERNAL: ACR3 09 16 16 LPOR 03 1
1 N3IC 05 1
1 kJ/shest; h
$40 'O O l
i
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$)
i TCTAL Ne oER OF COPIES RE4JIRED: LTTR JT* ENCL s**
m f-A B-1 h* *
$0d Franklin Research Center 4a aw n.rrmann==.am.
~~
e s
l TER-C5506-59/62 Commonwealth Edison one new usuone raza. enieseo. mino s Accress Booty to: Post Cthce Sox 767 Chacago. !!Imoes 60690 -
October 14, 1980 6
Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission washington, DC 20555 Subfect: Oresden Station Units 2 and 3 Quad Cities Station Units 1 and 2 Proposed Amendment to Appendix A, Technical Specifications, to Operating Licenses OPR-19, 25, 29, and 30 NRC Oceket Nos. 50-237/249 and 50-25A/265 Reference (a):
D. C. Eisenhut letter to All Operating Boiling Water Reactors (BWR's) dated suly 7.
1980 l
Dear Sir:
In accordance with the request in Reference (a), and cursuant to 10 CFR 50.59, Commonwealth Edison proposes to amend Appendix A, Tecnnical Specifications, to Operating Licenses OPR-19, OPR-25, OPR-29, and OPR-30 for Drescen Units 2, 3 and Quad Cities Units 1, 2, respectively.
The proposed amencments would add surveillance requirements for the scram disenarge volume (SOV) vent l
l and drain valves and LCO/ surveillance requirements for RPS and control red block SOV limit switenes.
The proposed changes provide additional operational and surveillance requirements and thereby strengthen the provisions for assuring continued operability of the control rod drive system i
during reactor operation.
As such, the proposed enanges do not present an unreviewed safety concern nor do they present any additional hazard to the health and safety of the puolic.
The oroDosed changes were prepared in accordan'ce with the guidance provided in Reference (a) and have received On-Site and Off-Site review and approval.
The changes are included in Attachment.s 1, 2, 3, and 4 for Dresden 2, Dresden 3, Quad Cities 1, and Quad Cities 2, respectively.
Pursuant to 10 CFR 170, Commonwealth Edison has reviewed the procosed changes and determined tnem to ce two (2) Class III and two (2) Cisss I Amendments.
As such, a fee remittance in the amount of 58,800.00 nas been provided.
l 8010240q Q }
4 B-2 dbu Franklin Research Center a h ono m r,n.en m t..-.-.-.-.n-,
~
- - ~
~ ~ ~ ~ ~ ~
, ^ = ~ ~ ~.
TER-C5506-59/62
.- Plasse address any questioits you may have concerning this matter to this of fice.
Three (3) signed originals and fift nis transmittal are provided for your use. y-nine (59) cooles of Very truly yours, b-Robert F.
anecek Nuclear Licensing Administrator Boiling Water Reactors SUBSCRISED and SWORN to before me,this
, # _/
day of
/ TE c 1980
/ '. !
hw Notary Puolic cc:
RIII Inspector, Dresden RIII Inspector, Quad Cities 9
7283A 4
B-3 000 Franklin Research Center a % w w rnmenv nm.
s.,
a l
l TER-C5506-59/62 1
l 1
I l
l ATTACHMENT 3 Quad Cities Station Unit 1 Procesed T.echnical Specification Changes Revised Pages:
3.2/4.2 l 3.2/4.2 16 i
3.3/4.3-3 3.3/4.3-9 t
4 B-4
'Jbd Franklin Research Center a on men of The Frannen humans
- ~
- ~ *
-- ~
l TER-C5506-59/62 l
(
QUAD-CITIES CPR-29
=
TABLE 1 H DEIR58tENDHON THATUlmATE!200 SLAG mmmme ammr at censur er fHegue imummme chumussw iets W
!asmus lhe Lesus serase 2
AP!54 essee Clow tusP 53.551W
- G8 2
APIRI assus Olshal at 2atte/ Hot s12/US M scue Stumby asest 2
Aries esmessmen auM M isse 1
And Mssa nmaar wasse (*e emeF sG.5EDN. 4W t
Ibd uses mmmer somucera a.1/US W seus 3
350 dsumuse 8
- 23/ M M same 3
JRd unumys g;gg/m 'as :sse 28 558 detssar net a Stares commed
'22 W bene cae cameur.
ha 3
384 detsser aar a 2ane postura al 'ast bouw cae casas.
~
3se SIM esse s138 cases / ss P
DIN dessnesse8 1108 lmmes/sss t
1 Higt water 'evel a scrse tacaerge venue (%*l7) 52 gates 1
- SN high 14ter level St=3Dt R
t=ip bypassed I
==.
L Fe mesummesmes Summpme Aus enemme of tpe remer amo summer sums tsee ene to fueensamee ineens sie namm em one tamms =
os see me emma meammme me due es-emme amme ame :n eesses a cue den omanas Miru enemmmmm aMes soumme dem emeses. Asw em eune asummene sus seemame e as storments :tamos,uman J se arit emumm suma se see er emees meeso tne ammet me espeanu mme uma is se eJ eso ensame sus emme samt fue me enemme numme e fammmes, numme sameneuse me same emmesi 4 theexamme umas amen mum i amosesumusume to seems 3 se ese samemums se ear er een we - se summme see as sensa.
L vt _ _
ane amo e seems Tne mee sumsg u a same af mas seeG!112WE 1 usemummam, se exame amm at a um se summe rumes 4 meammmmee-enemmemee 2ToeCPe.
1 toe esee!Beimmeamtesemom L the:Es enemmemp eeemumme a me meme su cessme Iraque 4,1 mr te m as due ammme ne tame e sammma F
ma mamma teeeememmemmeamammmt as same smus men as aumanse summmmeeunger som rummmes as same same est e ammme 3 mut.
L Sesuetummammeuma esammer meme suas e a tse Amman e m ammer susum 1 these eenemmeuma se sme e susy summt e
4 12/42-14 h.Ve 30 4
B-5 MJ Franklin Research Center 4 cem,an er ne r= man ir.ames
)
_m.___
a TER-C5506-59/62 1
~
QUAD-CITIES CPR-29 TABLE 4.3-1 IIDummi TEIT 20 CAUSEATION REll:DCT FOR CNE MO CONTamMENT 120LME 2TD15 MSTIrJMDTATEM.
100 SL012L MO 13DLA110NSA husamme hugumme Pusemat hsemummt tammes Tasca tuihumm;Il casra gastesessmann 1.
Asasar W weist insi (D
Gun /3 musee Cum / day L DTned het preusse (D
Qum/3 muse ltme 3.
Easter les pressse (D
Qun/3 mamas ase 4 Castamust sys, mesimus
- a. 2/3 ase huge (D
Qun/3 ansee ame
- n. Cammammst summe IU Gaus/3 maae
.we 5.
(as emmme aus mest suas (D
Qun/3 asses Nee diusage L (kunneange w smanal Ashams outsp Asmasst ames,
.we aus aims 1.
APRI dummande GI 5 Qun/3 suses Mme 2.
APatsu vemWe (D (3 Ksemag amas Mme 3.
asamen s (n 5 (a nue 4.
III daunuses S Q3 (S (D Muse 5.
250aume (D (3 Ashout estage Name L Amt duumscae G103 Osm/3summe ame 7.
Es uansen S Ol (S (31 Mass L SIE desser net a stams (5 O!
(B
.ime puutan 3.
me deemer.mt nr stumm peneium S
(G Mme
- 10. 2X deusesWr (S Q)
(S QI Man
- 11. lile near 'eust a a:me once/3 treetths Mut sommen Name
. disasse vessee (SC7)
- 12. SCV high level trip Ref::alirsg
!t:2. T14 %
tTene -
bypassed otatage iium sammans tussen L
M $ tueranse
% outage Ashefst outsp Mme t
2.
nense mes b
. tu g,,f 3,,,,
e,,,j,,,
3.
2ammen W presse al om,f3,,,,,
3,,
4 2sexue hee =
(y (q Adment anage Cnce/ day 5.
Ansar to hw user evel al om,f3,,,,
g,,,f,,,,
ans amasam-1.
Samufme hip msw Qua/3 meses Cace/3 meses Mene 2.
Tiseme ne nip tensesam 4sfuset outage hhsms :mrage sane 3.
tan m :nsam once/3 mesms Onca/3 mouns Name I
12/42-16 4
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o TER-C5506-59/62 (UAD CITI!':S CPR-29
- 3. The conerni rod drive housing support
- 3. The corre:mes of the omtrol em!
system shall he in place during rea6 or withdrawal sequence input to the fetwer opera:po and when the reactor RWNI osmputer shall tw ventied dier undans sy sene is prewurried ah...r f i.adant the v. perwe.
asnimphern pecsuse ws Is fuel us the reestor vesel. unies all soneroe ri=Js p
g g
g g
gg; g
are fully inserted and SpecAcasion 3
I" ***'
iry of the rod worth minimiaer in property fWNll its function shall he Control rod *ithdrawal scquenen vended by the following checis:
a.
shad M etaMiM so that man.
4.
The RWM computer online diag
- imum resetsvity that could he added by dropuut of any incre.
nostic test shall be successfully ment af any sme control blade W,
would not niake the enre more h.
Proper annunciation of the sefec*
than 0.013 ak superrnsical tion error of one uus-of acqwnce h, Whenever the reactor is in the
$sattup/ Hot 54andhv or Run e.
The rod binck function of the-mode heiow 20% rated thermal l
RWM shall he venaied by with-power, the rod worth minimizer drawing the Srst rod as an out-shall he operable. A second opera.
of-sequence control rod no rnoce for or qualined technical person than'to the bkick point.
may he uned as.a suhssitute for.in inoperable rod worth minimiter which faals after withdrawal of as fasas 12 control rods to the fully withdrawn poselaofL Ther ruh worth minimiecr may alwa he hypewd for row poner phis.o testing to demonstrate the snus.
Jown m.irgin requirements oe Specarication.t3 A :( a nucle.ar engensur sa preent and vendo the maepJvy-esep rod movements of the test powedure.
A Conerne rods shall not be withdrawn 4 Prior to contrnt nid withdrawal for for startup or refueling unless at leaw maattup or during refueling. weniy thar
=o source range channels have an at leans two wiurce range chanazis observed ununt rate equal in n' Fre: iter have an ohnerved count rate of at least than three counts per wennd and thee three counts per sciannit SAM's are fuHy inwrted.
3.
During operation with limiting con.
- 3. When a limiting 6mtrol rod pattern trol rod putterns. an determined by the saists, an instrument fianctional tes of nuclear engineer. casher-the R8M shall he performed prior to I
- I a.
both RBM channeh shall he operahie.
and daily thereafter.
hw control rod withdrawal shall he
- s. ww ser==.etsmurw.
me seen..s sists = wetram erm x esea me n.w s.
hiocked or
'h eu.am we, awinierrae-c=em,.i. 'ae' im'me.
eame si ave.uu
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. ssr me manse was ame.
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s =st rue om erot mis to wem, mas 3.
Chun imot efw Jerm 1e' vial is treet e I
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branklin Research Center a cm on ce n. nw.en in nme
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l TER-C5506-59/62 QUAD-CITIES DPR-19
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h, an end.or-cycle daisyed nestmn fraction o(0.005
- c. a 5-7' ; cife Doppler reassivity feedback.
i
- d. the red scuam innernos case shown in S; -ia"= 3.3.C.
- e. the i-pommible rod drop veioctry of 3.11 fps.
C the demi a aondent and scram reacdvity shape fiancnoa, and l
g themoderssor tempersnare as which criticahry occurs, t
In most cases the wordt o(insequence rods or rod segmena will be alwsmanialty less than 0.013.1k.
Perihar. the addhion o(0.013.1k worth of rescamsy as a result of a rod drop and in conjuncnon wish the actual value of the other important =mdaae analysis parameters described above, would saast likefy ruumit in a peak fbei enthalpy wher=aanity less than 130 cal /g design Ihmit. However. the GAI3.1k limit is applied in order to allow rocar for ihmse reload changes and ease o(vertAcanon wedsens repensive tenhancal s; --'d--"
changes.
i Should a control drop acadent result in a peak thei energy content of 30 cal /3. (swer thaa 660 (7
- 7) Anni rods ase conservauvely==ha='d o perforate.This would resultin sa oshine dose weil heiow t
due guadelinevalue of 10 CFR 100. For 3 x 3 thei. fewer than 350 rods are conservatively estimated to perfoenam.wish nearly the sasne consequences as for the 7 s 7 thei case because of the rod power dadhsense.
The rod words minimiar provides automatic suoervisnoa to assure that out o(sequence control rods wel aos be withdrawn or inserted: i.e it limbs operaser deviamens (rota planned withdrawai sequemann (reference SAR Sect.on 7.9). It serves as a beidsso to,,.e.1 comuoi o(control rod westh. la theevent that the rod worth minsmazer is out o(service when requared. a haansed operator or other 7mnw hm, I employee can manually fuldll the control rod paaern conformance thacnos of the rod worth mimmi=. In this case, the normal procedural controis are backed up by procedusal connois to assure conformance.
- 4. The sousua range moatsor(SRM) sysissa performs no assoimanc safety system thaenaa 1.e. it has se summa flancnon. It does provide the operator with a visual indicados o(neutros tevel. This is seeded for knowledgeshie and er5cient reactor starmy at low neocon tevels. The conses,uences of russmty aandants are thacnons of the initial nemuon flus.The requirement of at least 3 couais per samed assures that any tr*an-u should it occur, begins at or above the initial value of 10' of rated power used in the aindyses of treaments from cold conditions. One operable SRM channel would be l
adagesse to monitor the sooroach to criticality asing Ec -- paaerns of scattered control rod wahdrawuL A minunum of two operable SRM's is provided as as added conservansa.
- 3. The ed block rnonuor (RBM) is designed to assornaasally prevent thei damage in the event of enumenea red withdrawal from locanons of high power density durms high power operauca. Two chnassis ase provided, and one of these may be bypassed from the conscie for main'aa=aca andsor it sing. Trippin5 of one of the ehmaank will hiock stroneous red withdrawal soon enough to prevent (L damaga This sysaem backs up the operssor, who withdraws control rods according to a writtaa sequemma The sp=m6d restnctions with ot s channel out of service conservatively assure that thei i
l dassage will not ocast due to rod withdrawal errors when this condanos exists. During reactor opueden wink cartain limishas castrol rod patisms, the withdrawal o(a d== gamed angle control red could rundt is one of me e fbst rods with MCFR*s tous than 127. Dursag use of ack ponerns it is judged that !
I tammig of the REM symmen to ammue its operability poor no wahdrawei of such rods will assure that ha-psoper wedsdrawet does aos assur. It is the resposabdity of the Nuctsar Enyneer to identify these Ihnidag 4
peamms and the designased rods either winan the pattsms ase attany===hWhad or as they develop due to the occurrea:e o(Inoperahis control tods la other than Ilmstas patterms.
- s. m.,emuner er 2. sesem otamuses unamme=== mas armas vahme amanes en woner +=esar w dreansme as enwaiams,me ame =ner is tae venema amme ame amme, m.m. --. - wende car inn pursassa,arttemessi emas tr..== Ave are es. ans car ene tammans oc e=====&ame woor esammus auras
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l TER-C5506-59/62 ATTACHMENT 4 Quad Cities Station Unit 2 Prcoosed Technical Specification Changes Revised Pages:
3.2/4.2-14 3.2/4.2-16 3.3/4.3 3 3.3/4.3-9 l
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as mustaises 1
Ifish sets ined a seen (sman mee. (SI:71 s25 stus 1
SD7 hich water level scran FA trip 2.,f. M mt
- 1. For the Stee".upAtst Standby and Run poettions of the resetor made telecter switch, there s..a12 3e two ope-aale or tripped tris systems ter ca::: 0.mc.
tien exeopt the 33F. rod 32neus. T!DC upscale and U.M d=ve. scale runed not bc operaale in the Burt ps.tition. AP!tM dowrtacale, AF??. upscale (f10w tittred),
and R1L*( deemetale need not be operable in the Startuar?.st Staruf ay as4a.
i l
'Dba MIT: wpecale hoed rant be operable at losa than 3C% Pated the: a1 power.
One chtanol may te Bypassed aseve 307. ratest thetual power provided tast a 11mittag centrol rad pet *etit does not exist. Far systcas vita r.tre thaa 1
ene thennel per tras system, if the first colussi cannot te met far one of i
the tars trip syisees, tais conditir:3 esy asist int us to 7 days provided that durinC that time *.no soernala systes is functianally tested i==
a mediately and daily thereaf tert if this condition lasts lonCtr thats 7 days the syssam shall be tripped. If the first esinan cannot De met for bota trip systems, the systems shall De tripped.
- 2. W is tae parecat of drive now rottvired to produce a rated core now of e
9s r h In/ur..r13 levet settias is in percent of rated pow.c l
(2511 x8t).
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TER-C5506-59/62 QUAD-CITIES CPR-30 TABLE 431 IIDuMIBI TM AND CAURIAHON REI5 DICT RR Cl:RE AND CONTAINMDIT C:0UNG 3T5TDRIDISTRIDADITATION. '
308 ILOCII, ANO 130LAftolt34
'N M
kasammt husamme kusument Cammme fasta Calibraemp3 Chenra IEEE hum m mense L Reser be.Isw water level (D
.Que/3 austas (kes/dsy L bywed but :sesase (D
Ome/3 masens Mme 3.
Russter Iss :sesase (D
Once/3 mouns
.we 4 Catasumut seray miennes
- a. 3/3 can tager (D
Osse /3 mouns Mene
- n. Casamment :susure (D
Oma/3 means Name 5.
'- - cae ensus puse (D
One/3 mulas imme deusses 6.
theensitage 4.W enantal asfusing outage Ammung antage -
Nue aus amas L APRI duumacale (D W Qua/3 mouns Rue L AMBA !se vanson W (D Renment antage Mene e'
3.
358uomme (S (3 (S (3 lesse 4 DRE dammeste (S (D 5 (B Neu 5.
ilms escue (D (D P=*=== outage None 4.
IIBM demucae (11 ( 3 Qua/3 masens Muse 7.
3M8uncWe (S (3 (S (D None L 554 detscar not a stamm (S (D E
Mme P=*=
9.
Det detectar est a status postus 5
5 Mme
- 10. Sital damasse (S (D (S (D None
- 11. Hisp natur hvet 'a snet once/3 :ter:tt.s Mat asemame name
. damage vehme (SIN)
- 12. SCW high level trip Refuell.W
!k:t acpi%h1.
bytassed cutage inus smsmem mese.
L 3 Reels taunt hgt tamparattue tsfumag cutap Refaumg anage
,ge L Stumme hgiso" (D
Qua/3 monens Ones/ day 3.
Stassins ion pressse tu g,,j3,,,,,
3,,
4 Stamune Inge ratenes (D (g gg gg 5.
Asmar km tw =sur eva (D
oim/s anuni sac masen L 318m"*e hien b Onne/3 manns cnes/3 sentns Mme L Twom aren 4 :tmeerinre ReNeing curate Re*.ent curan son,
3-18" resc=r :resswe ance/3 mones oncu3 3,,eg 12/42-16 4
B-ll 0000 Franklin Research Center A 08vieson of The Fransen sneemme
4 TER-C5506-59/62 QUAD-CITICS DI*R-30
- 3. T*ne centroi rod drive housing support
- 3. The torrectnces of the contret red system shall he ta pl. ace ductag reactor withdrawal sequence input to the power operation and when the reactor RWM computer shau be venAed aftse
}
oestant system is ptessurized above leading the sequence.
asmospheric pressure with fuel in the reassor vessel, unless all constal rods g
,.g, e
l are fully inserted and Spee:Acanon drawal towards criticality. the capabil.
3.3.A.I is met.
i'I of the rod worth minimizer to properly fulfil its functiost sinall be
- a. Control rod withdrawal sequences vers 6ed by the following checks:
shall be establahed so that rnas.
imum reactivity that could be
- a. The RWM computer online daag.
added by dropout of any inere.
g,gg meat of any one control blade g
would be cuch that the rod drop accident
- b. Proper annunciation of the sette.
design limiit of 280 cal /cm..is.not c.scoeded.
tim error or me out4uomence
- b. Whenever the reactor is in the senssol red shall be venaed.
(
SaartspeHot Standby or Run
- s. The rod block function of the mode below W. rated thermal l RWM shat! be ver:Asd by wita.
power, the rod worth mantmszer drawing the erst roJ ns an out.
shan be operab!e. A second opera.
. of.nequerrte control rod no enere ser or qualsAed techas:al person than to the block posas.
may be used as a suasutate for an inoperable rod worth mansmiser whach fasis aAer withdrawal of at least 12 control rods to the fully withdrawn possuon. The rod wenh snintmiart may also be bypassed for low power physics notisrg to demonstrate the shut.
down margin reevirements of Speciacation 3.3.A if a nuclear espneeris present and verines the sosp.by.sae,s rod movementa of the seu procedure.
A Centrol rods shall not be withdrawn
( prior to control rod withdras.,I fc.?
for saartup or refuehng unless as ! cast startup or during refueling, verdy that two noerce taye channels have an at least two source range channels observed count rate equal to or yteater have an observed count rate of at least than three counts per second and these three counts per second.
SRM's are fully inserted.
i
- 3. During operation with limiting con.
- 3. When a IImiting control rod patters i
trol red patterns, as determined ay the exists. an instrument functional test or nuclear engineer, esiner:
the RBM shall be performed prior to 8**
- 585nated Ms)
- a. both RBM channels shall be l
eperahic.
and daily there Aer.
- s. n wr= 4:enuma =w n.e.wnen tinns t e stu.e.
as,..
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IL ronsrol rod withdrswal shall be blocked, or
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TER-C5506-59/62 QUAD CIT *Te DPR b. the delayed neutron fraction chosen for the bounding reactivity curve
{
- c. a beginning-of-life Doppler reactivity feedback
- d. scran times slower than the Technical Spech.fication rod scram insertion rate (sece.on 3 3.c.t>
- s. the etaximum possible red drop velocity of 311 fp's
- f. the design accident and scram resetivity st. ape function, and
- g. the moderator.ta=perature' at which criticality occurs In.nost cases the worth of insequence rods or red seg=ents in conjunction l
with the actual values of the other i=portant accider.t analysis pars: stars described acove, would nest likely result in a peak fuel enrhalpy suo-stantially isss than 263 cal /g design 'is t.
j Should a control drop assident ruult la a peak fuel ase gycontent of 230 csug, femtr than 660 (7 a *
- 7) that sods are -.
..h estimated to perforata This would result in an ossite dose well below the gudelinevalueof to CFR 100.For 3 a 8 fusi.fawertaan 150 rods are conservativety estima:sd e periorssa,with nearly the same consequences as for the 7 a 7 fusi ease because of the rod power i
ur -
5 The red worth minimizar provides assomatic senervison to assure that out of sequence control rods will see be withdrawn or inserted: La It !imus operator deviat.ons from planned withdrawal asquences (referene SAR Section 7.91. It serves as a bachp to preesdural control o(control rod worth. la he event that the rod worth minimiaeris out of service when required, a lacensed operator or other quala5ed techancal employee esa manually fula!! the control red pattern conform.4 ace ihastion of the rod worth rainimiaer. In this case. the normal procedural controts are backed up by sadependent procedural controis to assure conformanos.
- 4. Thersource range monitor (SRM) system performs no automatic safety system ihnetton i.e it has as serass fkocnon. h does provide the operstor with a visual indication of neutron level This is needed for knowledgeable and c.cient reactor startup at low =eutron levels. The consequences of a
numenvisy acudents are ihneuens of the initial neutron *as. The res,uarement of at least 3 counts par sessed assures that any transient, should it occur, begins at or above tne initial value of 104 of rated t
power used in the analyse of transients from cold conditions. Ona operable SRM c%aael would be adequate to monitor the approach to criticality using homogeneous partsras of scanered control r:d wishdeswaL A miairaum of two opershie SRM's is provided as an added conservatism.
- 5. The red blacit monuar (R3M) is designed to automatically prevent that damage in the event of erroneous red withdrawal from locanons of high power density during high power operation. Two abannels ase provided, and one of these may be bypassed them the consoie for maintenance and/or sessing. Tripping of os e of the channels will block erroneous rod withdraws! soon enough to prevent ibet damage.This sys
- backs up the operator, who withdraws control rods accordin5 88 4 *cttten 6
g sequenor. The specified restrictions with one chaansi est of sarwice conservatively assure t=at fuci damage will not occur due to rod witadrawal errors wisen this cond:nos asists. Dursag reactor operation with certain 14** ting control. rod patterns, the withdrawal of a designated single centrol red could result in one or more fuel rnds with MCPh's less than the MCP.S. fuel cladding inte,rit7 safety limit.During use of suchpat: erns is isjudged that testing of the R3M system to assure its cper:bility prior to witad::wal of such rods will assure that im,r'reae-withdrawal does not c::ur. It is the respor. sit:lity of ths Nuc! car Eagines a identify these lirnating patterns and ttte designated r::s cuher when the patterns are initiM!y assablished or as they develop due to the occurrence ofincperah!c
- control rods in other than ilmiting paaerna.
- s. -,.
suir oc u. s== ossaur est = *.===tv amer== tr= w=,se,mune are e.snine
'--=g ta the Unh.
mas fut amnar. "hume suuttficatamm grarsee for the og taw unham, a tang w _
pg3 gas vertfjetamt that the vm&ves are 8,W5, afd fx tpe teting af thee Tatves wing gemager Sean Whitame GBripe GEM *3 "* osEEe8*
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APPENDIX C COMMONWEALTH EDISON LETTER.OF OCTOBER 22, 1981 WITH ANSWER TO RFI FOR QUAD CITIES STATION UNITS 1 AND 2 e
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1 TER-C5506-59/62 REGJL&T'4Y INF0ad4 TION 3IST9IBJTION SYSTEM (PIDS)
J 40CE33I34 13R:A11329011J 33C.DATE: 91/10/22 NOTARIZED: NO 30:<ET s FACI.150-237 3eeseen laclear ponec Station, Unit 2, :ommonweal tn E 35300237 53-249 3*esden Naclear Po,ec Station,
' nit 3, :enetawesItw E 35300249 J
53-234 1aas-Cities station, Jnit 1, Commonwealen Esi som :o.
3530025a 5 3-235 lano-Ci ti es Stati on, Jnit 2, Oossonwealte Esiso1.:o.
35300265 A UT +. N 4 M E AUTHJM &PFILIATION 4403:8,T.J.
Coenoiweal tn Edison Co.
RECI8.N4*it RE:Is!ENT AFFILIATION IPP06IT3,T.A.
Joarating 9eactors 3renen 2 SUSJECT: Formaeds response to 12C 511902 eeouest for andt into ee neososes Teen Soec cnsnges for seras oisenarme vol re7e 1 3esin valves.
COPIES RECEIVE 3 LTR.[ ENCL..[ SIZES. f.
3IST9I5 JT! 3N C30E 43013 TITLE: 3eneral 31strication for after Issuance of Goerating...i:ense 10TES:1 cosy:SEP Sect. Lor.
35300237
- lE I8IENT COPIES RECI8IENT 008tE3
!) 033E/N44Ei LTTR E.NCL I3 00DE/NAME.
LTT4 ENCL ACTION:
3Rd 85 BC 01 13 13 3R8 s2 BC 01 13 13 i
INTERNAL: EL) 1 0
IAE 06 2
2 144/3Hs3 DEPY03 1
1 199/DL DIR 1
1 NR4/DL/09A3 1
0 NRR/DSI/ RAS 1
1 RE3 PILE 04 1
1 EXTERN 4L ACiS 07-to 16 LPO4 03 1
1 NR: POV 02' 1
1 MSI:
35 1
1 NTIS 1
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CoNIfnonwoneth Edisost one mest Nationes P' ara, CNeago. iii ros Accress Aeory to: Pos; c!flCe Sox 767
~
cnicago muos 60600 October 22, 1981 g i8 C
/
Mr. T. A.
Ippolito, Chief s
Operating Reactors - Branch 2 4
3
.D OCT2 31981= {s Division of Operating Reactors U.S. Nuclear Regulatory Commission C u.a.,emas wasAsus %
wasnington, DC 20555 g
g
+
Suoject: Dresden Station Units 2 and 3 pp' Quad Cities Station Units 1 and 2 Response to Request for Information Concerning SOV Vent and Orain Valve Tecnnical Specifications NRC Docket Nos. 50-237/249 and 50-254/265 References (a):
T. A. Ippolito letter to L. O. De1 George dated September 2, 1981 (b):
R. F. Janecek letter to Director of NRR dated October 14, 1980
Dear Mr. Ippolito:
Commonwealtn Edison has received your Reference (a) request for additional information concerning our Reference (b) proposed Technical Specification changes pertaining to Scram Disenarge Volume (SOV) vent and Drain Valves.
Although you only requested information for Dresden Units 2 and 3, our response includes Quad Cities Units 1 and 2 as well, since the questions are directly applicable.
- Your requests and our responses are provided in the attachment to this letter.
Please address any further questons you may have in tnis regard to tnis office.
[
One (1) signed original and fifty-nine (59) copies of this trtnsmittal are provided for your use, s
very truly yours, s '= = l 4 --<.
l Thomas J. Rausch i
Nuclear Licensing Administrator Boiling Water Reactors Attachment cc: RIII Inspector, Oresden RIII Inspector, Quad Cities 273CN 9110290ttoiit'C22' DR ACCCX 05000237 l
p PDM nklin Research Center A Chaemon of The Fransen buemme
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l TER-C5506-59/62 Attachment Commonwealtn Ecison Company Oresden Units 2 and 3 Quad Cities Units 1 and 2 Response to Request for Information Concering SOV Vent and Drain valve Technical Specifications CONCERN 1.
Commonwealth Edison's response in paragraph 3 does not contain the requirement of the Model Tecnnical Soecifications of paragraon a.l.3.1.lb to cycle eacn valve at least one complete cycle of full travel at least oncs per 31 days.
REQUEST 1.
Provide technical bases why the requested change is not applicable to Dresden Nucles: Power Station Units 2 and 3.
Response
The~model Technical Specifications that were used as a reference to develop our submittal were attached to the July 7, 1980 letter from D. Eisenhut to all operating SWR licensees.
The Franklin Research Center model Tennical Specifications for Section a.1.3.1.1 are not the same.as the July 7, 1980 mooel Tecnnical Specifications.
In addition, the July 7, 1980, model Technical Specifications were incorrect concerning the SOV vent / drain valve closure during individual CR0 scram timing.
The July 7, 1980, model Technical Specifications were modified to ensure a meaningful test is performed each refueling. A possible means of modifying our suomittal would be to require verification of valve closure and suDsequent re-opening during each' scram, and take credit for that.
In summary it is our contention that our proposal is meaningful and provides a true test of the system.
REQUEST 2.
The Technical Specifications for Dresden 2 and 3 state that each reactor protection system scram discharge volume water level-nigh instrumentation channel containing a limit switch shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST once per 3 months, and tnere are not tech specs for CHANNEL CALIBRATION.
Since the proposed frequency of the required surveillance for Oresden Nuclear Power Station Units 2 and 3 ciffers from the frequency requested by the Mccel Technical Soecifications, provide technical bases for it, ranidin Research Center e ~-
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TER-CS506-59/62
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6 -
9
RESPONSE
The proposed Technical Specifications regarding the 50V scram and rod block level switches are adequate.
A monthly functional test of the SDV scram bypass would require the reactor mode switch to be placed in either SHUTDOWN or REFUEL for tne test, and this unreasonable.
REQUEST 3.
Provide technical basis for not Calibrating the scram discharge volume water level-nigh instrumentation enannel. Also provide technical basis for performing tne scram trip byoassed instrumentation channel functional test once per refueling cutage instead of once per month as requested in tne Model Technical Specifications.
RESPONSE
Magnetrol level switches are not, and cannot, be calibrated. Therefore, calibration frequency in cur submittal is cesignated as "Not Applicable" for the scram discharge volume water l
level high channel.
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