ML20137C059
| ML20137C059 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 08/16/1985 |
| From: | Frisch R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| GL-83-28, NUDOCS 8508220198 | |
| Download: ML20137C059 (24) | |
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0 Consumers Power Company General offices: 1945 West Parnell Road, Jack son, MI 49201 e (517) 7se-0660 August 16, 1985 Director.
Nuclear Reactor Regulation US Nuclear Reguintory Connaission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT -
RESPONSE TO REQUESTS FOR ADDITIOANL INFORMATION ON GENERIC LETTER 83-28 Nuclear Regulatory Commission (NRC) letters dated May 16, 1985 and July 1, 1985 requested additional information on Consumers Power Company responses to Generic Letter 83-28. Our response to the May 16, 1985 request is in the enclosed attachment I and our response to the July 1, 1985 requent is in the enclosed attachment II.
For your reference, NRC Region III has conducted a plant specific implementation inspection of Generic Letter 83-28 requirements. The results of this inspection are documented in Inspection Report 85013 dated August 7, 1985.
h AA40 Ralph R Frisch Senior Licensing Analyst CC Administrator, Region III, USNRC NRC Resident luspectors - Big Rock Point Plant Attachment 0500220198 850016 PDR ADOCK 05000155 P
PDR i
9 OC0885-00048-NLO2
I ATTACHMENT I Consumers Power Company Big Rock Point Plant Docket 50-155 Responses to May 16, 1985 Request for Additional Information on Generic Letter 83-28 August 16, 1985
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10 Pages 0C0885-00045-NLO2
1 RESPONSES TO MAY 16, 1985 REQUEST FOR ADDITIONAL INFORMATION ON GENERIC LETTER 83-28 Item 2.1.2 Licensee needs to submit detailed information describing its vendor interface program for reactor trip system components.
Information supplied should state how the program assures that vendor technical information is kept complete, current and controlled throughout the life of the plant and should also indicate how the program will be implemented at Big Rock Point.
CPCo Response The Big Rock Point Plant vendor interface program for reactor trip system components coincides with the program for safety-related equipment.
Presently the Big Rock Point Maintenance Department vendor files are being reviewed for the purpose of discarding out of date manuals and categorizing the retained vendor manuals as either information or controlled. Other departmental vendor files are being maintained as information copies leaving only those responsible for performing the work i.e. the Maintenance Department Instrument and Control group (I&C) and Mechanical and Electrical (M&E) group with sets of controlled vendor manuals, to be used for controlled work activities.
The review of the files is about 95% complete for the I&C group and about 60%
complete for the M&E group. During this review effort a user index is being generated which provides a cross-referencing system that includes vendor, component identification number, system, and maintenance procedure number that specifically tie to the particular piece of vendor information whether it be a manual, print, bulletin or other information.
A new plant vendor information control procedure to be implemented this year will contain the following aspects. As new components are being ordered vendor equipment technical information and applicable revisions will be requested. Such information will be routed to the Document Control Center t
(DCC). From DCC one copy will be routed for technical review and approval, Following approval and return to DCC the vendor document will be sent to Engineering Records Center (ERC) for control and issue. Revisions to documents initiated by the plant will also be reviewed, approved, and sent to ERC for control and issue.
Where vendors have gone out of business our experience has been.that the i
product line has usually been taken over by others. As a result there has been good experience in procuring parts and obtaining information from the new vendor.
Also most of the equipment at Big Rock Point is original plant equipment.
Over its 20-year-plus operating history the plant has had to rely to a considerable degree on its own expertise and experience in maintaining the equipment in good working ccndition.
Equipment specific maintenance
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CLO885-0004C-NLO2 1
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2 procedures have been developed for most of the complex safety-related
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equipment and these procedures have been revised to reflect plant operating and maintenance experience over the years.
Item 2.2.2 Licensee needs to present its evaluation of the NUTAC program and describe how it will be implemented at Big Rock Point. The staff found the NUTAC program fails to address the concern about establishing and maintaining an interface netween all vendors of safety-related equipment and the utility. Accordingly, the licensee will need to supplement its response to address this concern.
This additional information should describe how current procedures will be modified and new ones initiated to meet each element of item 2.2.2 concern.
The staff notes that the licensee's October 2, 1984 submittal incorporated Generic Letter 83-28 Item 2.2.2 into the Big Rock Point Living Schedule as Issue 100 with a submittal scheduled for December 31, 1985. The staff also notes that the written interpretation of the issue for Living Schedule Item 100 is not sufficiently comprehensive; thus, additional discussions with the NRC staff are recommended so that the response will be more complete when received.
CPCo Response As described in the previous section, Big Rock Point has to rely to a considerable degree on its own expertise and experience in maintaining the plant. Except for some cases (recent modifications or equipment replacements), vendors no longer have the involvement nor expertise base necessary to provide a useful " tool" in maintaining the older equipment installed at Big Rock Point. However, to maximize the benefit of vendor support the following actions will be taken at Big Rock Point.
A listing of vendor contact personnel will be developed. Plant engineers and maintenance supervisors will furnish their personal lists of vendor contacts currently used.
These will be reviewed and summarized by vendor and equipment type into one document available to all personnel. Having a coordinated document will permit timely revision (additions, deletions, or changes) and provide plant personnel with ready access to appropriate vendor contacts.
This listing will be completed consistent with our Living Schedule Issue 100 currently scheduled for December 31, 1985.
Big Rock Point's review of the NUTAC program has concluded that not all I
portions of the program are prudent for implementation at this facility, primarily due to its age.
Operating Experience Responses are now developed for all of the following documents by the plant l
staff. The Plant Review Committee reviews, evaluates and approves actions l
taken in these responses.
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GLO885-0004C-NLO2
l 3
A.
NRC Documents
. I&E Bulletins
. I&E Information Notices
. Generic Letters B.
INPO Correspondence (SEE-IN)
. Significant Event Reports
. Significant Operating Experience Reports i
All vendor correspondence, primarily General Electric Service Information Letters (SILs), are also evaluated for applicability and action.
t Although a high percentage of these reviews conclude non-applicability to Big l
Rock Point, the plant still supports continuation of these reviews.
Nuclear Plant Reliability Data System (NPRDS) f When NPRDS was started in the mid-70s, the industry made the decision that the older vintage plants need not participate in the program. The decision was based on the minimal worth gained from exchange for equipment failure information generated and received by the older units. Due to the age difference in components of the older units, information exchange was concluded to have little or no benefit. Big Rock Point is still in agreement with the decision not to participate.
1 i
Certain equipment failures are reported undet NRC reporting requirements (10CFR50. 73).
Furthermore, our preventive maintenance program incorporates operating experience identified as a result of equipment failure reporting not only from 10CFR50.73 but from our internal corrective action program as well as our operating experience review program described above and manufacturers recommendations.
Item 3.1.1 - and Item 3.2.3 Results of review of test and maintenance programs shall identify any post-maintenance testing that may degrade rather than enhance safety and shall describe actions to be taken including submission of Technical Specification changes.
CPCo Response i
Review of test and maintenance programs have not identified any post-maintenance testing required by the Plant Technical Specifications that degrades rather than enhances safety. No Technical Specification changes are necessary.
Item 4.5.2 Licensee needs to submit procedure T 30-01 referred to in previous response.
CPCo Response A copy of T30-01 is attached.
i GLO885-0004C-NLO2
ntFORid AT_IO, N COPY RECEIVED Form No. Admin 1.3.A.7.4 Consumers sn MAY 81985 Power CSES3%
BRP PROCEDURE APPROVAL AND AUTHORIZATION NUCLEAR LLCENSING Procedurc No.
T90-01 Rev No.
7 Procedure Title unmn.y REACTOR PROTFCTION SYSTEM TEST AT POWER ORIGINAL ISSUE I
ORIGINATOR DATE l
METHOD OF REVIEW (SPECIFY ONE)
QA-05#
CURRENT REVISION STATUS AUTHOR DATE METHOD OF REVIEW (SPECIFY ONE)
DDHerboldsheimer 10/01/84 QA-05#
REISSUE HISTORY REV REV FOR APPLICABILITY DATE METHOD OF REVIEW (SPECIFY ONE) 6 EMcNamara 3/01/8h QA-05#
Prepared by:
LCP/9 N Technical Review:
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/#/4 /8%
Department Approval:
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/0[9 8[
Management Approval M
ID i Distribution:
Authorized Period of Use:
MASTER COPY:
DOCUMENT CONTROL March 6. 198h - March 6. 1986 CONTROLLED COPIES:
SS SURVFITlJNCE TFST BTNDED NOTE: Authorized period shall not exceed two years from date of last Review and Approval for Applicability.
INFORMATION COPIES:
QUALITY CONTROL PROCEDURE IMPLEMENTATION HISTORY Reviewed for System or Component Operability Performed by Completed / Reviewed by Method of Verification O Functional Test Titte Tale O Physical Inspection O Administrative Review Date Time Date Time Maintenance Order No. (if applicable)
Revision 7 Page 1 of 6 Initials
-T30-01 MONTHLY REACTOR PROTECTION SYSTEM TEST AT POWER 1.0 PURPOSE To confirm and document the operability of each sensor and component of the safety system and, thereby, its input to a trip function.
2.0 PRECAUTIONS AND LIMITATIONS 2.1 Strict adherence to the alphabetical and numerical sequence of the procedure is essential.
2.2 Trip tests are to be performed on one logic unit at a time followed by immediate reset and verification of reset of the test switch involved before. proceeding to the next step.
3.0 PREREQUISITES 3.1 Shift Supervisor approval of test obtained.
(SS) 3.2 Power Controller permission to perform test obtained.
(0P) 3.3 All logic unit test switches checked in normal (up) position, events recorder on and inking, both logic units
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reset.
(OP)
4.0 REFERENCES
AND ATTACHMENTS 4.1 Technical Specifications 6.1.5a.
4.2 Operating Procedures SOP-38 4.3 IE Bulletin Number 84-02, Date, March 12, 1984 OP1084-0125A-BA02
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Revision 7 Page 2 of 6 T30-01 Initials 5.0 PROCEDURE INSTRUMENT & CONTROL GROUP PIC0 AMMETER TRIP UNIT OUTPUT VOLTAGE TEST 5.1 Test equipment available:
Calibration Due Date Accuracy (TECH) l 5.2 Measure picoammeter trip voltages at test jacks on rear of each picoammeter (minimum voltage = 11.0 volts).
- 1
- 2
- 3 (TECH) 5.3 Voltage tests satisfactory to continue procedure.
(I&,C SUPV) 5.4 Measure picoammeter trip voltages at test jacks during Step 5.30.
OPERATIONS GROUP LOGIC UNITS (RE-03A AND RE-03B ON C02-5)
Insert channel trip signals for all sensors (one at a time on one reset of power switches on C01)y reset of test switch on logic and logic unit at a time followed b by manually operating test switches A extinguished) and operations recorder (pen offset) peration (lamps and B for each trip function.
Check power switch o for each sensor switch.
OP1084-0125A-BA02
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Revision 7 Page 3 of 6 T30-01 Initials Channel 1 Channel 2 Power Switch A (all lamps out)
Power Switch B (all lamps out)
- Power Switch B (output power lamp out)
Operational Record (pen offset)
NOTE: X indicates that no operation should occur.
3/vv V 5.5 Neutron Flux Hi Level Channel 1 only X
X X
X X
X X
X 5.6 Neutron Flux Hi Level Channel 2 only X_
X X
X_
X X
X X
5.7 Neutron Flux Hi Level Channel 3 only X
X X
X X
X X
X Proceed to Step 5.8 only if no channel trips or operations recorder offset occur in Steps 5.5, 5.6 and 5.7.
NOTE:
Prior to resetting system after doing Step 5.11 (Channel 1) and Step 5.19 (Channel 2) perform relay verification.
5.8 Neutron Flux Hi Level Channel 1-2
-X X
X X
5.9 Neutron Flux Hi Level Channel 2-3 X
X X
X i*
5.10 Neutron Flux Hi Level Channel 1-3 X
X X
~X 5.11 High Enclost re Pressure A 5.12 Visually verify that the following relay contacts change state when the relay coil is de-energized.
Isolation Valve Control Auxiliary Relay 1K4A (OP)
Isolation Valve Control Auxiliary Relay 1K4B (OP)
Vent Trip Auxiliary Relay 1KSA (0P)
Vent Trip Auxiliary Relay 1KSB (OP) 5.13 High Enclosure Pressure B f
5.14 Reactor low Water A i
5.15 Reactor Low Water B t
X lX-l 5.16 Steam Drum Low Water A Y
!Xj 5.17 Steam Drum Low Water B 5.18 Reactor High Pressure A T
X-OP1084-0125A-BA02
Revision 7 Page 4 of 6 T30-01 Initials 5.19 Visually verify that the following relay contacts change state when the relay coil is de-energized.
l Isolation Valve Control Auxiliary Relay 2K4A l
(0P) i Isolation Valve Control Auxiliary Relay 2K4B i
(0P)
{
Vent Trip Auxiliary Relay 2K5A (0P)
Vent Trip Auxiliary Relay 2K5B (0P) 5.20 Reactor High Pressure B X
X 5.21 Recirculation Valves Closed A X
X 5.22 Recirculation Valves Closed B X
X 5.23 Steam Isolation Valves Closed A X
X 5.24 Steam Isolation Valves Closed B X
X 5.25 Condenser Low Vacuum A X
X 5.26 Condenser Low Vacuum B X
X 5.27 Dump Tank High Level A X
X 5.28 Dump Tank High Level B X
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5.29 Manual Scram
,X iX' Neutron Monitoring Power Range Alarm and Trip Test NOTE: This test item serves to document the quarterly isolation valve simulated instrument test as outlined in the last paragraph of 3.7.b of Technical Specifications. Also see Check Sheet C-3, Item 8C.
f Complete Steps 5.30 through 5.33 for each picosaseter.
j 5.30 Test one picoammeter for proper operation of upscale alarm j
and trip points when neither of the other two units are in an upscale or downscale trip condition.
5.31 Put the range switch in trip test position and depress the trip test button and rotate the adjust potentiometer clockwise and verify proper alarm indication at 10512 and scram trip operation at 120 1 2.
Set Points Channel 1 Alarm Trip (0P)
Channel 2 Alarm Trip (0P)
Channel 3 Alarm Trip (OP)
OP1084-0125A-BA02
Revision 7 Page 5 of 6 T30-01 Initials INSTRUMENT AND CONTROL GROUP 5.32 Measure picosameter trip voltages at test jacks and record results in trip condition (maximum voltage = 0.5 V)
- 1
- 2
- 3 (TECH)
OPERATIONS GROUP t
5.33 Return the potentiometer to full CCW position, reset trip and return picoammeter operation to normal (trip output voltage > 11.0 volts) before proceeding to next unit.
NEUTRON HONITORING INTERNEDIATE RANCE PERIOD ALARM AND TRIP TEST COMPLETE STEP 5.32 FOR EACH LOG "N" 5.34 When the reactor protection period trip function is bypassed, rotate the " infinity set and trip test" poten-tiometer on the Log-N period amplifier CW to verify proper alarm operation to 15 1 2 seconds and trip operation at 10 1 2 seconds. Return potentiometer to normal.
I Set Points Channel 4 Alarm Trip (0P)
Channel 5 Alarm Trip (0P) 5.35 Check both logic units reset, all test trip switches reset to normal, both operations recorders indicating normal and all alarms cleared.
(OP) i i
5.36 Report completion time and results of test to Shift i
Supervisor and Region Power Controller and enter same in Control Room Log Book.
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OP1084 0125A-BA02
Revision 7 Page 6 of 6 T30-01 Initials 5.37 List all corrective actions (eg, H0s, DRs, ERs, etc) necessary to complete this test:
None Required (SS)
Test Completed by: Operator Date Operator Date i
Technician Date 6.0 REVIEWS Reviewed by:
Shift Supervisor Date OTA Date OPSUPT/SUPV Date OP1084-0125A-BA02
ATTACHMENT II Consumers Power Company I
Big Rock Point Plant Docket 50-155 l
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Response to July 1, 1985 Request for Additional Information on Generic Letter 83-28 August 16, 1985 11 Pages 0C0885-00045-NLO2
RESPONSES TO JULY 1, 1985 REQUEST FOR ADDITIONAL INFORMATION ON GENERIC LETTER 83-28 Item 1.1 (1)
The licensee needs to submit information clearly defining the responsibilities and authorities of the personnel who will perform the post-trip review and analysis. We recommend that the post-trip review team include a member of plant management at the Shift Supervisor level or above who holds or has held a license at the Senior Operator level on the plant (the Shift Supervisor who was on shift at the time of the event would be acceptable) and who has the responsibility and authority to obtain all necessary personnel and data to ensure a thorough cnd complete post-trip review.
In addition, the post-trip review team should include an STA or an engineer who has had special transient analysis training. These two people should have a joint responsibility to concur on a decision / recommendation to restart the plant.
CPCo Response As described in the attached Reactor Trip Report procedure, the post-trip review team consists of a Shift Supervisor and the On-Call Technical Advisor (OTA), which is Big Rock Point's shift Technical Advisor, plus the assistance of operating personnel on shift at the time of the trip event. These two individuals provide a recommendation for start-up to the Plant Superintendent who authorizes the restart.
Item 1.1 (2)
The licensee needs to addrese tue methods and criteria comparing the event information with known or expected plant behavior. We recommend that the pertinent data obtained during the post-trip review be compared to the applicable data provided in the FHSR to verify proper operation of the systems or equipment. Where possible, comparisons with previous similar events should be made.
CPCo Response The OTAs and Shif t Supervisors are kept current with known or expected plant behavior and transient analyses. Following the event, copies of all charts that can be useful for evaluation are made by Operations personnel (Step 2.9).
Using these and the guidance of Section 3 of the procedure, the OTA performs a review of the event and identifies any deficiencies.
Item 1.1 (3)
The licensee has indicated that if the cause of the trip cannot be determined, the O&M Superintendent or the On-Duty Superintendent prescribes which portions of the precritical check list must be performed prior to requesting that the Plant Superintendent obtain General Office authorization for restart. We find that this action to be taken by the licensee is not sufficient to ensure safe plant operation. We recommend that if any of the restart criteria listed below are not met, an independent assessment of the event should be performed by the Plant Review Committee or a group with similar authority and experience.
In addition, the licensee should have procedures to ensure that all physical evidence necessary for an independent assessment is preserved.
NUO885-0427A-NLO4
2 I
l Recommend restart criteria:
a.
The post-trip review team has determined the root cause and sequence of events resulting in the plant trip.
l b.
Near term corrective actions have been taken to remedy the cause of the
- trip, c.
The post-trip review team has performed an analysis and determined that the major safety systems responded to the event within specified limits.
d.
The post-trip review has not resulted in the discovery of a potential safety concern (eg, the root cause of the event occurs with a frequency significantly larger than expected).
CPCo Respcuse If the cause of trip is unknown, the Plant Review Committee performs a review of the situation and recommends a course of action. All information associat-ed with the event and review will be preserved with the trip report.
(See Step 3.7) The Plant Superintendent has the responsibility to review the actions of the event and subsequent corrective actions prior to authorizing a restart (Step 3.6).
Item 1.1 (4) l The licensee has not provided, for our review, a systematic safety assessment program to assess unscheduled reactor trips. We recommend that the licensee develop a systematic safety assessment program in accordance with the guide-lines provided below.
Recommended guidelines:
a.
The criteria for determining the acceptability of a restart of the plant.
b.
The qualifications, responsibilities, and authorities of key personnel in the post-trip review process.
1 c.
The methods and criteria for determining whether plant parameters and system responses were within the limits provided in the FHSR.
d.
The criteria for determining the need for an independent review.
CPCo Response The Reactor Trip Report was developed to provide, in it self, the guidelines for performing a systematic safety assessment of all unscheduled reactor trips. This procedure incorporates the guidelines provided to insure that response was proper and that anomalies are corrected prior returning the reactor to power operation.
NUO885-0427A-NLO4
3 Additionally, during the week of July 15, 1985, Region III Inspector, Tom Taylor performed a refiew of our actions resulting from Generic Letter 83-28. He reviewed our Reactor Trip Report procedure and reviewed its use on past events and found that our actions were in compliance with the requirements of Generic Letter 83-28.
l The Reactor Trip Report procedure has been reviewed against the requirements of the Generic Letter and in our opinion adequately addresses the concerns of the Generic Letter.
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VOLUME 1:
ADMINISTRATIVE PROCEDURES CHAPTER 4 ATTACHMENT C REACTOR TRIP REPORT Trip Number from DCC I
Trip Date/ Time
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Purpose Evaluation and review of each unscheduled reactor trip involving blade motion is required to determine that response was proper and that anomalies are l
corrected prior to returning the reactor to power operation. The cause of the trip must be determined, the proper operation of safety related equipment that has been challenged must be verified and assurance established that the trip event did not have any other detrimental effect on the plant in terms of nuclear safety. The trip review will provide the data for the required l
Licensee Event Report (IIR).
References 1
1.
NRC Generic Letter 83-28, dated July 8, 1983.
2.
Responses to NRC Generic letter 83-28, dated September 6, 1983, November 7, 1983 and June 1, 1984.
3.
INPO Good Practice OP-211, September, 1983, draft.
Responsibility
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(
Operations Department Personnel (usually SS) will perform the trip review for events as described in Section 1 and 2.
Technical Department Personnel (usually an OTA) will perform the trip review with the assistance of the operating personnel on duty at the time of the trip event as described in Section 3.
The Control Room log and/or the Shift Supervisor log should contain a chronological record'of observations during the trip event. Written statements may be requested by the OTA or others to assist in the
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investigation process. The trip review shall in no way interfere with operations or maintenance activity in progress to secure a safe condition.
Event Reports (ER) are required for all Reactor Trip Eve.nts and are subsequently reviewed by the Plant Review Committee.
07/01/85 4 - 40 Rev 331 AD0584-2515A-BA01-BA05
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VOLUME 1:
ADMINISTRATIVE PROCEDURES CHAPTER 4
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1 Procedure NOTE:
Control Room Operator. or Shift Supervisor shall complete Section 1.
i 1.0 List Steady State Plant Conditions Prior To Trip 1
1.
Reactor Power Level MWt 2.
Primary System Temperature
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3.
Primary System Pressure psig 4.
Feed Pumps operating (circle)
One or Both 5.
Reactor Recirculation Pumps on (circle)
One or Both 6.
Turbine By Pass Valve (circle)
Auto Manual 7.
Turbine By Pass Valve (circle)
Closed Open 8.
Pico-Ammeter readings 1.
2.
3.
9.
Turbine rolling (circle)
Yes No 10.
Generator synchronized (circle)
Yes No 11.
List any other off normal equipment status.
Completed by Date Time 2.0 The Shift Supervisor shall complete Section 2 through Step 2.9 for all unscheduled reactor trip events, involving blade motion.
l NOTE:
If reactor trip occurs with primary system temperature below l
212*F and the cause is identified and corrections made, obtain approval l
for startup from Plant Superintendent or his designate. Evaluation by Technical Department Personnel in Section 3 is not required.
07/01/85 4 - 41 Rev 331 l
AD0584-2515A-BA01-BA05
VOLIME 1:
ADMINISTRATIVE PROCEDURES CHAPTER 4 2.1 Operation in progress at time of trip f
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l 2.2 Chronological events I
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l 2.3 Verify that rods fully inserted without delay 2.4 Verify that all four containment ventilation isolation valves closed.
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l 2.5 Trip caused by (list sensor and root cause)
NOTE: The operations recorder units (two) for a protection channel go from li inches per hour to 1\\ inches per minute when the first input sensor pen offsets and remains on "high speed" until an adjustable time period expires or sooner if all input sensors reset.
07/01/85 4 - 42 Rev 331 AD0584-2515A-BA01-BA05 i
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VOLUME 1:
ADMINISTRATIVE PROCEDURES CHAPTER 4 2.6 Circumstances of manual scram (if it occurred) 2.7 Based on available information (primarily from charts), determine if any Reactor Protection System input or output appears to have been challenged without proof of operation.
If so, initiate NO and/or DR to verify proper operation prior to startup of plant.
NOTE: See note at 2.5.
List any corrective action document by number and short narrative title.
2.8 Based on available information, determine if the Core Spray System or the Containment Isolation System were challenged without proof of operation.
If so, initiate ER to evaluate prior to startup of plant.
2.9 If the reactor trip occurs with the primary system temperature above 212*F prior to or during the event and blade motion has occurred, make copies of charts that can be useful for evaluation. The evaluation must be completed in Part 3 of this procedure by Technical Department personnel and the remaining steps in Part 2 (this part) shall not be completed.
2.10 Corrective actions (DR and ER dispositioned):
Completed and reviewed by SS Date Time Reviewed by Plant Superintendent or his designate and approval given for startup.
Date Time NOTE: On Call Technical Advisor shall complete Section 3.
3.0 Complete only if reactor trips with temperature >212' and blade motion has occurred.
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ADMINISTRATIVE PROCEDURES CHAPTER 4 3.1 Verify trip with sensor operation that caused trip by use of the operations recorder chart. Make copy of charts to be a part of this trip report. List sensors that tripped.
NOTE: See note at 2.5.
3.2 Using the neutron monitoring charts and feed water charts and others verify that sensor operation was proper for initial trip and subsequent trip sensor operation.
(ie, did all three pico channels trip and if not, why not, etc? Same for drum level and high reactor pressure and high dump tank level etc.)
If any reactor protection sensors did not operate that appear to have been challenged, initiate maintenance orders to have them tested prior to plant startup.
List H0 number and initiate DR for each verified failure.
NOTE: See note at 2.5.
3.3 Through interviews, log records, chart review verify that Engineered Safety Features (ESF) and other equipment needed for accident mitigation or safe shutdown operated properly if challenged.
ESF listed below were unchallenged or operated properly.
YES or NO Core Spray Containment Spray i
Emergency Condenser Containment Isolation Containment Vacuum Relief Reactor Depressurizing System Steam Drum Safety / Relief Valves 07/01/85 4 - 44 Rev 331 AD0584-2515A-BA01-BA05 l
VOLUME 1:
ADMINISTRATIVE PROCEDURES CHAPTER 4 Other accident and transient mitigation equipment.
Liquid Poison HELB Ventilation Damper CV-4190 Degraded Voltage Trip System Core Spray Recirculation System Off Gas Isolation Valves and Monitors Fire Pumps Diesel Generator (s)
Fuel Storage Area Monitors Containment Water Level Monitors Transfer to 46 Kv line Rod Drive Catcher If "no" on any of the above items, list Deviation Reports or Event Reports (ESF failures and spurious operations and other as outlined in Volume 1A, Chapter 15, may be reportable to the NRC.)
3.4 Through interviews, log records and chart review, verify that no other safety related detrimental effect occurred to the plant and equipment because of the trip; (ie, rapid primary system temperature transient water chemistry problems, water hammer).
List any DR items or other comments generated from this review.
Comments or DR references:
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ADMINISTRATIVE PROCEDURES CHAPTER 4 3.5 Complete narrative description of event listing probable sequence of trip events, root cause(s) and unresolved problems if any.
OTA (or equivalent) signature Date/ Time
/
3.6 Review of trip report and disposition of all safety related corrective action documents completed and reactor startup authorized by Plant Superintendent or his designate.
Initial Date/ Time
/
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ADMINISTRATIVE PROCEDURES CHAPTER 4 e
3.7 If the cause of trip is unknown, the PRC shall review the situation as it is known and recommend a course of action.
If cause of trip remains unknown, responsible Corporate Office approval for startup must be obtained. Attach documentation from PRC review and other pertinent data.
Startup Approval from Corporate Office
. NOD Vice President or Alternate Date/ Time l
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