ML20203M857
| ML20203M857 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/29/1986 |
| From: | Lydon J BOSTON EDISON CO. |
| To: | Strosnider J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| #86-128, 86-128, CAL-86-10, NUDOCS 8609050027 | |
| Download: ML20203M857 (35) | |
Text
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.om--u-Executive Offices 800 Boylston street Boston, Massachusetts 02199 James M. Lydon Chief Operating Officer August 29, 1986 BECo Ltr. #86 128 Mr. J. Strosnider, Chief Reactor Project Section IB U.S. Nuclear Regulatory Commission 631 Park Avenue - Region i King of Prussia, PA 19406 Docket No. 50-293 License No. DPR-35
Subject:
Supplemental Response to NRC Confirmatory Action Letter #86-10
References:
(a) NRC CAL 86-10 Dated April 12, 1986 (b) BECo Response to CAL 86-10 Dated May 15, 1936 (c) NRC " Request for Additional Information" Letter Dated May 16, 19536 (d) BECo Response Dated June 16, 1986
Dear Mr. Strosnider:
This letter provides additional information requested by the NRC Region 1 through a list of comments and questions which we received on July 21, 1986. This list was a result of the NRC review of the BECo Supplemental Response Letter to CAL 86-10 dated June 16, 1986. A telephone conversation was held on July 31, 1986, between Mr. Strosnider, Dr. McBride and various Station management personnel to finalize the answers. Answers to the specific questions are included as Attachment I to this letter.
He trust the contents of this submittal combined with information provided in our preliminary response dated May 15, 1986, and the second response dated June 16, 1986, will provide information adequate to address the requirements of Confirmatory Action Letter 86-10.
8609050027 860829 3 DR ADOCK 0500 EE L5 i t
' BOSTON EDISON COMPANY Mr. J. Strosnider
-Page 2 Should you have any further questions concerning these matters, please do not
, hesitate to contact me.
Respectfully submitte .
James M. ydon JQ/ko Attachments: 1. Supplemental Response to CAL 86-10
- 2. Pressure Switch History Data
- 3. Calculation M-269
- 4. Procedure 3.M.3-8 (Control Room)
- 5. Procedure 3.M.3-8 (Cable Spreading Room) i e
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ATTACHMENT 1 SUPPLEMENTAL RESPONSE TO CAL 86-10 Boston Edison Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. OPR-35 NRC Question 1 Has BECo evaluated the merits of periodic. checks of the RHR system pressure and temperature after (1) startups, (2) any RHR system high pressure alarm or (3) after the leakoff system is initially placed in service? Also,.has BECo considered periodic checks of RHR system pressure in the absence of high pressure alarms?
Response
Although the alarm response procedure is keyed to the RHR discharge pressure high pressure alarm, action such as obtaining local pressure readings takes place after a second annunciation. The alarm response Procedure 2.3.2.1 will be clartfled prior to startup for action on any Panel C903-Left annunciator B-7 RHR high pressure alarm. This will cause local pressure and temperature data to be captured upon alarm annunciation. Additionally, a surveillance test for periodic data collection of RHR system pressure and temperature will be prepared prior to startup. Conceptually this surveillance test will record RHR system local pressures and temperatures from the temperature surveillance monitoring strips. A frequency of weekly surveillances during power operation is being considered, but frequency of the surveillance will be procedurally adjusted based upon the trend of the results. The resident inspectors will be appraised of the content of the proposed surveillance proceaure.
NRC Questien 2 The first paragraph of BECo's response to NRC question "e" describing the proposed leakoff method appears to differ from the method in Procedure TP86-85 and should be clarified. The paragraph infers that the bypass valve will be opened during a measurement step, closed, and reopened to establish the leakage path.
Response
BECo will follow the method described in TP86-85 for establishing the controlled leakoff. The response to NRC question "e" was presented as a concept of the method. The temporary procedures presents the actual method to be used.
Page 1 of 11
SUPPLEMENTAL RESPONSE TO CAL 86-10 (Cont.)
Boston Edison Company -Docket No. 50-293 Pilgrim Nuclear Power Station License No. OPR-35 NRC Question 3 Hill a reactor shutdown just be initiated or will the plant be shutdown (pending an engineering evaluation) if the 1.0 gpm limit is exceeded? Also, has BECo considered evaluating leakage rates above 1.0 gpm now; rather than waiting to conduct the evaluation until the leakage limit is exceeded?
Response
Yes, normal reactor shutdown will be initiated when the 1 gpm is reached with both valves shut. However we will be performing additional testing which may result in a change to the 1 gpm criteria.
Leakage greater than 1 gpm implies that the single closed MOV is at or beyond its local leak rate limit. Leakage rates greater than 1 gpm have been considered. The engineering evaluation has set a conservative limit of I gpm based upon measured Appendix J 1eakage, the Appendix J 1eakage limit, a conservative safety factor and high pressure water leakage. Steps required for additional engineering evaluation will involve retesting the boundary for Appendix J criteria and for high pressure water leakage which will be performed during this shut down.
NRC Question 4 The BECo response and the associated safety evaluation state that leakage in the RHR system will not be allowed to exceed 1.0 gpm. This is not strictly true since the measured parameter will be leakage through the bypass valve at 150 psig.
Changes in RHR check valve leakage (such as after a pump has been cperated) could make the leakoff measurements inaccurate and misleading. Has this been considered? Has BECo considered tracking the leakage rate into the torus over extended periods of time as a method of verifying stable RHR system in leakage?
The safety evaluation also states that all RHR pump flow will go into the RHR system. However, some flow could be diverted through the leakoff path if the "D" pump is idle (LPCI only requires 3 pumps operate). BECo should consider modifying the safety evaluation to address these two concerns.
Response
Leakage through the RHR pump discharge check valves was considered in BECo's evaluation of RHR system operation and isolation valve performance. The bypass valve addressed in S.E 1959 R1 was part of a three-pronged approach which included the additional temperature and pressure monitoring equipment. The temperature monitoring equipment in particular will provide notice of large leakage rates past the isolation valves independent of RHR check valve condition. Tests performed on the RHR system using a hydro test pump oa April 17, 1986, showed that the RHR discharge piping would become pressurized with Page 2 of 11
SUPPLEMENTAL RESPONSE TO CAL 86-10 (Cont.)
Boston Edison Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. OPR-35 Response (Cont.)
only .33 gpm being added to the system. This test confirmed that the RHR pump discharge check valves are " tight". The valves have performed satisfactorily since plant startup and are expected to function satisfactorily in the future.
Valve degradation, when it occurs, is expected to be gradual and will be identified through a change in the flow measured during engineering tests at the bypass line and possibly an increase in piping temperature upstream of the RHR injection valves if there is a concurrent increase in isolation valve leakage. BECo believes these measures along with our scheduled Appendix J LLRT's are sufficient to monitor the performance of the RHR system containment isolation valves.
The intent of the bypass valve modification and subsequent Safety Evaluation (S.E. 1959 Rev. 1) was to address the effects of an incremental increase in leak rate caused by the addition of the bypass valve. Large leak rate increases were to be monitored by other means.
BECo has considered trending torus water level but considers this impractical given the accuracy of the present and available instrumentation ano the large volume in the torus.
The Safety Evaluation #1959 Rev. I did not address the reduction in RHR flow in the rare event during 3 pump LPCI injection concurrent with the condition when the fourth, idle pump, had the bypass line around the discharge check valve.
This was onitted as the incremental flow rate if belc,w the level of accuracy for punp curves or flow measuring devices (ie: less than 3 gpm out of a total flow of 14,400 gpm 'le: 3 pumps at 4800 gpm each).
BECo will revise SE #1959 Rev. I to clarify its intent and assumptions and forward it to you under separate cover by September 17, 1986.
NRC Question 5 The data sheet to procedure TP 86-85 requires that an RHR pump suction block valve (MO-1001-7D) be closed during the initiation of a controlled leakage path but not reopened (although the procedure does require this). Why has the latter step been left off the data sheet?
Response
The step to reopen the RHR suction valve M0 1001-7D was inadvertently deleted from signoff sheet A2. The procedure has been corrected to include this step.
Page 3 of 11 w%;
SUPPLEMENTAL RESPONSE TO CAL 86-10 (Cont.)
Boston Edison Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. DPR-35 NRC Question 6 The acceptance criteria for tests of the RHR injection check valves are not included in the response, as requested (NRC question "f"). The criteria should be determined and submitted to Region I.
Response
A 5 gpm leak rate criteria was selected for the following reasons:
- a. The injection check valves are designed to limit loss of primary fluid while the motor operated isolation valves close. As such, rapid operation is more important than duplicating the leak tight seating capability of the containment isolation valves. The 5 gpm leak criteria is well within the makeup capacity of the plant under normal and emergency conditions.
- b. 5 gpm leakage will not challenge the integrity of the low pressure RHR Diping. The low pressure RHR pipe is protected by two PSVs with a combined relieving capacity of over 100 gpm.
- c. 5 gpm leakage is well within the makeup capability of any of the CSCS Systems acting individually or in concert in the event of an accident event (i.e., HPCI makeup is nominally rated at over 4000 gpm).
- d. 5 gpm has reportedly been used and accepted by the NRC as an acceptable leak rate limit for swing check valves at several other operating nuclear power plants.
NRC Question 7 The RHR pressure gauge calibration frequency was stated to be "once per refueling outage" in the response to question "h". This frequency is not defined in the technical specifications. How is it defined and what is the justification fo'r the frequency?
Rasponse The once per refuel outage frequency is better stated as once per operating cycle which is defined by T.S. as "the interval between the end of one refueling outage and the end of the next subsequent refueling outage". The newly installed gauges in the RHR system are classified as supplemental instrumentation for which the owner (BECo) may determine the calibration frequency. BECo uses the once per cycle frequency for calibration on many instruments that support Technical Specification systems. Similar gauges are also located throughout the Station on various other systems and have not exhibited any major calibratton drifting with once per cycle testing.
Page 4 of 11 u :.
SUPPLEMENTAL RESPONSE TO CAL 86-10 (Cont.)
Boston Edison Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. DPR-35 NRC Question 8 The calibration histories of the RHR high system pressure alarm switches (PS1001-74A and B) should be submitted to Region 1. How is the proposed calibration frequency, "once per cycle", defined and justified?
Response
Calibration histories of PS1001-74A & B are included as Attachment 2. The once per cycle frequency is explained in the response to Item 7..
Additional information is needed to clarify a statement in the prior response to NRC question "i" concerning the setpoint for PS 1001-748. This statement indicated that the actuation setpoint was found tat low at 360 psig versus a setpoint of 392 psig. It should be clarified that 360 psig was obtained from the RHR pressurization tests and that the 360 psig was the point at which the RHR high pressure alarm was received in the control room. This reading was obtained from a test gauge mounted on the RHR system at approximate elevation
+3'6". The PS 1001-74B is mounted on rack 2206 at an approximate elevation of
-12'6". The pressure source used to obtain the 360 psig was a positive displacement pump which created pressure surges in the system. The 360 psig was not indicative of the actual pressure that actuated PS 1001-748.
NRC Question 9 Hill the 1001-28 and 29 valve on the "A" RHR loop be maintained normally closed?
Procedure TP86-84 Indicates that the 28A valve will be left closed, but the licensee has previously indicated that the 28A will be open and the 29A closed.
Response
The M0 1001-29A and B valves will be normally open and the MO 1001-28A and B valves will be normally closed. Procedure changes to establish tnis lineup were in progress but are being delayed because further safety evaluations are needed for the procedure change. If the valve lineup decision changes based on this pending safety evaluation, the resident inspector will be informed.
NRC Question 10 What is the accuracy and reliability of the temperature measuring markers? What specific change in temperature will require that additional measurements be taken with a portable measuring device? What is the accuracy and reliability of the portable temperature measuring device?
Page 5 of 11
SUPPLEMENTAL RESPONSE TO CAL 86-10 (Cont.)
Boston Edison Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. DPR-35
Response
The specific change in pipewall temperature will be determined after ambient operating conditions are established. This may require the frequency of temperature checks mentioned in response to question 1 to be increased to daily or once per shift to establish baseline conditions.
The specific change in RHR pipe wall temperature that will require additional measurements is expected to be an increase in 20*F from the normal ambient temperature. This ambient temperature will be established as part of the surveillance test referenced in response to Question No. 1.
The temperature measuring markers, reversable and nonreversable have an accuracy of t l'F. Portable temperature monitoring can be accomplished with a Rochester Instruments System "Supercal" in the thermocouple mode, which has an accuracy of 3 l'F. Each of these instruments based on performance history is considered to have good reliability.
NRC Question 11 Will the licensee verify seating of the RHR check valves after operating? If not, why not?
Response
The Nuclear Engineering Department (NED) is currently investigating methods for confirming RHR check valve operability. After researching the implications of operationally testing these valves and assessing such concerns as opening leakage paths between the reactor and low pressure systems, the investigation is now concentrating on check valve disk position monitoring. The valves were originally equipped with position monitoring instrumentation, however, operational problems resulted and the instrumentation was inoperable and finally removed. NEO is presently investigating new pcsition monitoring equipment.
If the investigation shows that the new valve position monitoring equipment represents an improvement over the original design and will operate reliably in a power plant environment, NED will recommend the installation of the equipment on the RHR check valves. This investigation is expected to be complete by November 1, 1986.
An alternative previously considered to verify check valve seating by pressurizing the low pressure system with the reactor at approximately 250 psia has been reviewed by NED and not recommended. This recommendation is contained in a NED response and states that there is a concern that such a pressure test (i.e., opening the in-series valves MO 1001-298, 288, 34B and 36B and creating a direct pressure path to the torus) could over pressurize several portions of the 150 psig design piping in the RHR System. The event of a single operator error or equipment malfunction could cause the loss of a significant portion of the reactor vessel coolant inventory.
The check valves ability to prevent reverse flow will however be tested once per refueling outage as previously committed.
Page 6 of 11
F' SUPPLEMENTAL RESPONSE TO CAL 86-10 (Cont.)
Boston Edison Company Doctet No. 50-293 Pilgrim Nuclear Power Station License No. DPR-35 NRC Question 12 Why will it take 9 months to submit a Technical Specification change to reduce compensatory surveillance testing in LCO's: considering that the issue was identified mid-1985 in connection with the on-line EQ modifications and also noted in the 1985 SALP report? Has the licensee considered contacting other facilities of a similar age to see if they have information that could be used to speed the evaluation and submittal process?
Response
As identified to the NRC in the NRC/BECo Management meeting held on June 12, 1986, one of the Management improvement items is the need to develop a Work Management Control Program which will identify and prioritize the entire NUORG workload. Until such time, it is not possible to accurately foreca;t completion dates that ultimately become commitments. The nine month time frame was chosen as a " Target" date only.
As for contacting other facilities, we contacted General Electric Company for background information on the Standard Technical Specification allowances. We had hope of being able to utilize whatever generic evaluations were avaliable and apply them to our specific situation. This avenue did not prove fruitful, as no generic studies were available. He have, however developed a preliminary justification for satisfying the safety significance consideration which will be presented soon to the appropriate levels of management. As new scheduling information becomes available, we will update the NRC Project Manager for Pilgrim Station in accordance with the ongoing Licensing Action Report iterations.
NRC Question 13 What is the justification for limiting RHR pipe temperature to no less than 15 degrees of saturation temperature? Is this temperature margin adequate, considering that pipe wall temperature (rather than interior water temperature) is the measured parameter?
Response
The purpose of applying a 15*F limit to the RHR pipe wall temperature is to guard against steam void formation. This margin is deemed sufficient based upon a bounding calculation for the uninsulated pipe based on heat conduction through the pipe wall. Considering the air is stagnant in the area of the uninsulated RHR piping (6 inches from the floor) the maximum temperature drop through the steel wall was 0.15'F. When the effects of air velocity (400 ft/ min) are considered, the additional temperature drop is approximately 0.5*F. Since the total temperature drop is less than l*F the 15*F limit on saturation temperature is considered conservative.
Page 7 of 11 1
SUPPLEMENTAL RESPONSE TO CAL 86-10 (Cont.)
l Boston Edison Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. DPR-35 NRC Question 14 Calculation M-269 should be submitted to Region I for review.
Response
Please refer to Attachment 3 of this letter.
NRC Question 15 The following comments concern the draft procedure TP86-81 which will control the test for spurious group I primary containment isolations during the next reactor startup:
15a. Will reactor level instrument vibration be monitored? If not, why.not?
Response
The reactor level instrument Racks C2205 and C2206 will be monitored for Vibration during the performance of TP86-81. A special Temporary Procedure TP86-82 has been written to control this activity (see attached copy). An outside vendor has been contracted by NED to perform the monitoring and analysis of the data taken. A change has been made to the startup procedure to include the vibration monitoring activity.
15b. Step VI.A indicates that the reactor mode switch will be placed in run for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Why?
Response
The Draft TP86-81 has been redefined and approved, by the ORC for implementation. The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> indicated in Step VI. a. of the draft has been lowered to 10-15 hours in the final version. This time period is required to stabilize the plant and should allow for any time dependent variables to be reproduced.
15c. Was a functional test conducted of the PCIS logic after GETARS modification was installed? What procedure was used fcr the functional test?
Response
A special Temporary Procedure (TP86-83) was written to satisfy the post installation testing cf the GETARS equipment as well as the functional testing of the PCIS logic. This Procedure was completed at the completion of GETARS installation.
Page 8 of li ,
SUPPLEMENTAL RESPONSE TO CAL 86-10 (Cont.)
Boston Edison Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. DPR-35 l
15d. Can th/ spurious isolation test be conducted at a power below the stated 30%?
Response
1 The isolation transient took place at a power level quite a bit lower than 30%
but, the purpose of this startup test is to as completely as possible, ceproduce that transient. Thirty percent is a comfortable power level for the
' plant to, operate at during the stabilizing period. In addition most of.the
,* major c;Jerating equipment will be inservice at the point.
This will, help us to more accurately reproduce the shutdown sequence that led to the spurious
, isolation transient.
Y[ '
, f NRCQuestionifI Does BECo plan'to conduct a sampling review of other systems, given the large number of' drawing / loose wire problems discovered during work on the reactor mode o, switch? -'
Response
An investigation was conducted on other Control Room and Cable spreading room panels. no compression connections similar to the type found loose in C915, 917 and 916 were identified. Therefore there is no further effort planned on that matter. The drawing discrepancies have all been dispositioned as electrical equivalents, duplication of. wires, or typographical errors, with two' exceptions. The first exception was a ' relay contact not shown on PCIS logic elementary diagram. This is considered an isolated case and measures have been taken to correct the drawing. The second exception was the miswiring of a reactor manual control circuit at the mode switch itself.
Investigation showed the logic to be more restrictive to plant operations in
- its mf,swired form than it would be wired ' correctly. Regular surveillance testing would have identified a miswiring that would have resulted in a less restrictive mode of operations. This was the only case of a circuit
- performing other than was indicated by the elementary diagrams, therefore, no walk down of other elementary diagrams is planned. The miswiring was corrected by returning the circuit to its design conditi,on.
The panel wiring diagram discrepancies dispositioned as electrical equivalents dealt with the neutral connections of relay coils. There is a chance such deviations exist in other panels but those discrepancies would, most likely, also be limited to neutral connections cr typographical errors. Active relay contacts and the relay coil supply leads were portrayed accurately on the wiring diagrams. Drawing revisions / upgrades are a continuous process. For example, the NED Plant Design Change (PDC) closeout program consists of, but is not limited to, updating electrical wire diagrams to reflect changes made due to PDC's. Beyond that program, the Technical Group is included in the E-203 walk down program. This program involves the walkdown of the AC and DC power distribution. systems. Print discrepancies found during this program are bein~g corrected through the E-203 procedure developed by the' Technical Group (reference Attachments 4 & 5).
Page 9 of 11 1
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SUPPLEMENTAL RESPONSE TO CAL 86-10 (Cont.)
Boston Edison Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. DPR-35 NRC Question 17 Why is local venting of the RHR system needed in addition to the keep fill system?
Response
Local venting in RHR is being used for assurance that there is no formation of air pockets in local high points. The venting Procedure TP86-84 schedules the venting to be performed once per week for four weeks (after maintenance is completed). This schedule is intended to detect the presence of air in local pipe runs and to be used as an input to regular venting practice. The results and locations will be incorporated into normal RHR System surveillance tests.
NRC Question 18 Is the RHR system always filled and vented after the 1001-34 and -36 valves are opened to depressurize the system? ,
Response
Since the keepfill system is continuously in service, the RHR system is always filled after the use of the 36 and 34 valves, but the system was not vented after it was used to depressurize the system. The results of the venting procedure discussed in Question 17 are intended to assess the presence of air and the severity of a potential problem in this area.
NRC Question 19 Has BECo considered the personnel safety aspects of the leakoff measurement process? At what location will system pressure be measured and what will be the expected water pressure at the measuring point? Will the measurement equipment withstand this water pressure?
Page 10 of 11
SUPPLEMENTAL RESPONSE TO CAL 86-10 (Cont.)
Boston Edisen Company Docket No 50-293 Pilgrim Nuclear Power Station License No. DPR-35
Response
The leakoff measurement process has not yet been tested in the field. Prior to actual use, a walkthrough and practice use of the procedure is planned to validate the method. This is to be performed prior to unit startup so that the pressure source would be controlled. This walkthrough has not yet been conducted because of the redireccion of resources to Local Leak Rate Testing and the RHR pump wear ring inspection. Industrial safety concerns and appropriateness of equipment use would be additionally verified as part of this walkthrough. However, procedural steps to "slightly opening valve M0 1001-100D" and the fact that the vent bottle is vented (vent is equipped witn a particulate filter for radiological concerns) reduces the possibility of over pressurizing the equipment. The system pressure of 150 psig is monitored in the RHR quadrant at PI 1001-808. This gauge is at approximately the same elevation as the test connection (within 3' elev.). The components used for measurement equipment are Tygon tubing or rubber hose and a "Nalgene" reagent container (vent bottle) are judged to be sufficient to withstand 150 psig water pressure even if not vented.
NRC Question 20 Has BECo considered testing the leakage of injection caeck valves in ECCS systems other than LPCI?
Response
In June, 1985, the NRC designated Generic Issue No. 105 " Interfacing Systems LOCA at BWR's" as a high priority issue. Shortly afterward in September, 1985, the NRC's Office for Analysis and Evaluation of Operational Data (AE00) issued a case study report that based on operational events indicated the likelihood of interfacing LOCA was higher than previously assessed and that this represents a trend with serious safety implications. As a result of the AEOD Report, a BHR Owners Group Committee of which BECo is a member was formed to assess the significance of this issue. This committee has worked closely with INP0 and with the NRC Task Manager on this generic issue.
Among the possible fixes to address the root causes for these events is testing of isolation check valves and removal of the air operator on testable check valves. PNPS has removed the air operators from all ECCS injection check valves. The air operator was the major contributor to the events described in the AE00 report. At the present tine, the significance and need for further corrective actions are being studied by the NRC, INPO and the BWR Owners Group. A decision on testing of these check valves will be based on the findings of the BWR Owners Group Committee.
Paga 11 of 11 '
ATTACHMENT 2 SUPPLEMENTAL RESPONSE TO CAL 86-10 Boston Edison Company Docket No. 50-293
- Pilgrim Nuclear Power Station License No. DPR-35
- Pressure Switch History Data Conclusion
, PS 1001-74A 1974 399 Set Point 1985 392 Set Point 12 Year Span Lost 7 PSI 1
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, ., .;N MODEL NO. SERI AL NO. LOC A TSON 'M'
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c,la N #, G E. OR BE GN TE L D AT A Se9EE T NO. SURVEILLANCE FREQUENCY g .,a- .d M241 REV.
4F 225A5750 DS53 REW.
0 x % '4 4, s 34NUF , k . F.DWG NO. eNSTRUM ENT a0 ANU AL NO. ROUTINE C AL4BR ATION FREQUENCY ,,
72 OO IN , TRUMENT R ANGE RED BLUE 25-600 psi COMY..OL iND AL ARM SETT629GS
~g' '
4 a 28 Contret #1 closes @ 400 psi (ine) pres. (reset diff 25 psi) 5,EciAL n EcAunONS (si %
Alrrm dischare:e headers hinh prens. and shutdown suction headers. *
- c,.Ll_ . TION eRocEDURE NO. wsNoOR G.E. SPEC. NO.
3C Mercoid aav. S (U N E.TED TEST INSTRUM EN T4
REASON FOR A L A R M PT. TRIP PT. A L A R M PT. TRIP PT. SET CATO BY/APPR'O NI LO NI LO NE LO NI LO P.B. RESET RATE POINT p; ROUT.
12 M' 9/12 /7h DFC/PFK X MAL.
375 399 ., ,, ~
mm 'ahn 360 .
TEST INSTRUMENT USED REMARKS 't5 3tl 77J iS 4:
ROUT.
2/25 RPL/BJL Trip 392 PSI siv ! es No Adj. '
86 M A L. **
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C ALIBR ATION MAINTENANCE PERFORMED aa oio.
DATE BT cat cal. P. M . PARTS REPLACED DESCRIBE M ALFUNCTION nw/* -
9/12/7h DFC X Checked setpoint and adjusted RPL 2/25/86 BJL X NONE Checked Setpoint Per MR86-140 4,
l mm . e 9
t l I J-m . . . .
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ATTACHMENT 2 (Cont.) i SUPPLEMENTAL RESPONSE TO CAL 86-10 Pressure Switch History Data Conclusion 4
- PS 1001-74B
^
1974 398 Set Point 1986 394 Set Point 12 Year Span Lost 4 PSI l
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Contact #1 closes at'265 psi inc 'Contset"clopesjz( $y %,g, %;n m y
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,.'.mQ REMARKS
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p 5 Y (12/17/78 h' I
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Jischcrac "cader Precs Mercoid g @g%gS w g)<3
.u m,% y.f ;rg Lj woonono sanias No. Locarion - p s.,
E.L. 17L 6 Rack.. ..2262 f , %j %MyH L 23 304 3D749312 '
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w a,.ur. ma r o*o. no. mst maavat no. INSTALLATION X \
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72 PIPING ' ,
}
g, y-gg]g n -.~ .
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Contact ,71 close at 400 psig ;(inc) pres Rose _
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- gw a era ,w;d Mh, ,
,*-* y@ie E,MMpweewm1mm q a% aggh/,; tonthetf,: closes lh @n OD.<aEM~ n m o .c ,q}pp% Contact #1 closes at 265 psi ine i .. ,v w.e.v Contact #1 closes at 241 psy dcc (Contact l;o ~ 'yQ: h;p.&x;i ,;cg ' .z:;u d y Q fnMd.t B & %ld % y4 $ G, -n ,47 e.g.5 %o .a': e% .Q. . ~>_m; p s;c.,;:e , dao.m G.%&!Q h x .p.m &.gasw p .aw p, qz . _~ _r-- = v .,. ..h m aynw u n(p.m m , ,p ag4;,..vg g ;aqq a.y , 3 ' 9:AQW g WW ., , aq(. , y; .6., M, y m . n WW< . q p.m pMp@:myg ng s p ,,. ~ f . , 4 .3 4 r.,(; q g>. . $g > y<y> i,[.; W .[.(.g$-lJ. M y 7f W, 4, rg ;Q( p,; g~ ~j. . . . w. y . ? , - e. 4 n m is%,w ~.. g u x w & g,V 2 Q.g%y}f33 gsag ; . aa, m;W6g:fq;. sty t ty y s . f g qup, 7 gtg+r3:g.g h, .:& 3 p 3.M ,.wpg m.n u,%, y,u% k9 e c. c, si NQRq.. Q:q:, mg..p> 4;$a y C g_ ,' q u .; 'a ez \w vg.ni.q
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- f;
9
- N.,TRuM Ny R ANGE OG
.- <- ~ RED SLUE CsNf wL ANo ALARM s E T Y IN G s ' nn.n~1 .:',or. n a t M 731, };ac) pres. .;eset diff 25 psi a 38 erECsAL WECAurioNs
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+ 0 0 '. . '. (. ...~<.O!O $; ( , {rCG3 C nd ';'lu t do ,7 suC O t{gn 'j,cp,dCIS C,.Ls u.ATioN PRoCEoumE No. 4 30 vENooR .. s.E. SPEC. No.
- ,g
-s ..t. 16C2CO10 , , . tEv. tum.E .v En TEs t sNsTRuMENis S 32 7 yt. 7, 3 RECALIBRATION AND MALFUNCTION RECORD * ** ,,,,,, AS FOUND AS LEFT CONTROLLERS i W FOR ALARM PT, TRIP PT. ALARM PT, TRIP PT. gg to 38 D*,TE RECAL* BY/APPR'O NI Lo HI Lo Hf Lo NI LO P.B. RESET RATE NINT ROUT, ,4' ,' ~ ~ ' ", 12 asi 9/12/74 PFW/DFC X 385 398 inc , , L. .. Ts ,.w. WHD 362 370 reset ' TEST INmTRUMENT USED REMARKS , ggt gg- .__ 77.T ROUT. is 42 RPL/JRP Trip 394 PSI 2/ W Nc Adj. E
- 86 "- "
D0'L x Resci. 350 PSI Nec. TE.T IN,TRUMENT USED as,, .se REMARKS I-600H * ** ROUT. as 48 MAL. }gg' gg TE T 8N ,TRUMENT USED REMARKS 24 SO
- T ROUT.
SS SS MAL. TEST INSTRUMENT USED REMARKS ROUT. MAL. T E.T IN,TRUMEN T USED REMARKS ROUT. M A L. TF T IN'81RUMENT USED REMARKS ROUT. MAL. l TE.T INiTRUMENT USED REMARKS g 3 O l "' m ' ' \'istreu syuem, v. 'O D D D f se. A su E . ,N * - 1 A'3 e .u . aan f-- N , e -. [9-O
- C at 444fioN poocEDywE NO 6 & ~vect m ,
15 ( * > nU.O eNgf4WwtNT MANUAL me o. j. wo2EL No -
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i .# 1. . , 9 '~ eme tauutNT m a ae ce r p e ep No. 4 EW. ., CONT-at AND aba#w SE f f rNG S G E 04 BEcMTEL DAT A SHE E T NO. s .. !) - . _v 22 spE au m e a ution., ' '" ' ' '~~ r~~~ J ': * .?:.:O lion I:00 t} o _ s C ALIBR ATION MAINTEN ANCE PERFORMED aNa oso DATE BY cat. c a L, P.M. P ARTS REPLACED DESCRIBE M ALFUNCTION FFW/ , 9/12/74 DFC X ! Checked setpoint and ad.1usted RFL 2/25/86 JRP X j None Checked Setpoint Per MR86-141 4 l l i ! l l l l . l I i ! _ L_ - f I i 1 j - I
- f. _ _ '
I i I l ______: l l l 3 l i l 1 i i > l l . I t i l ' l I I _ _ 1 l ! l-6 ! i _.4 __i. - t m m . ,. a e no-a < '" O Ul.sf cO rewrd i~, .asur - ~, c) p q q' ("T p, (] - g-} _ .s ' .__p .%w# a, ATTACHMEtlT 3 ' CAPITAL CALCULATION COVER SHEET Aursor:iz ATios so. '/M Rs ""dIon E C80MBY PILGRIM U NIT 1 Sheet 1 of <h CALC !!O. N-259 REV. 6 File flo. SR [3 Subject //4t ria ic Alc A cd ;Q4 T-r NO lCO/ - ZRS o A tlSR C f% f o,o f_ pc7 y Preliminary Calc. Finalization Discipline Group Leader /vS8MM/ due date Approval /s/ ffd[d**uu Date 6/7/4 / Independent Verifier 8 / te / Statement Attached @ By uE WrASA Chk'd 7.) //gM/) kll /SWhf $ IV Vb /s/ $bhizi 1%l Shh{ / bdlOS[ ' % RTEeMisuE h) . Accc/49 Oc c Cf
- TER 1 A FOR 'uA TEAL l'cilgN;c pp sf 77/E Mo /oo!-22 B e2 /40 / co /-9c)B 4,tp / AifD 17/C Rf/2 S*fSTEf4 Minor revision (s) made on page(s) of this calculation. See next revision.
Replaces calc = _ Voided by calc ( or attached memo O PAGE 1.;of 4 . F/ c26V PRELIMINARY CALCULATK)N SHEET CAPlT^i'lZ U, HOR ATION NO. REV DATE PREPA".ED BY DATE FINAL @h CHECKED BY DATE REV O DATE f V, - Edison ceana APPVD BY <- DATE
SUBJECT:
SHEET E OF f (OI VGiq .- SR b NSR
(?c%./N3D sy$rcht [kCG fiaOQr6 d L A ~f~
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' /50-?c qs)i 3 /e '7. 99 s r m flin,t o. Cp;. (Scc Et& IGct(cmcurs) ren a rr>
LR (tal W e d X (4 P i [ LR, = 33gpiv, AP, - 6 lo pd LRt(Led T.4 dm (a P& LRz= ? 4 P2. " 9 8 o es i A.
'h LR 1 =
LRe(9 % f ., .33 (280f , ,6<3 pp a sea.faJ fea
@ P, ) *- (fo I 0)~ &MfsiA.
ASsuPrriotas: Skl/a,isud FAN 6ddw we realed(ANfa;r e
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..oigf,- e ssoi rs,1 NED -ICO, B.E.Co. FORM 3930
CAPlT^L b PRELIMINARY CALCULATION SHEET HORIZATION .e REV OATE P~.EPARED BY '/ DATE E4 h FINAL gh CHECKED BY ' DATE #/
c ;
REV O DATE>f2d
, Edison APPVD BY -
DATE !d
~
SUBJECT:
SHEET 7 OF M
('A t C u L A _r2ct11&-
SR Q We e vi&led N /er' /eal e. b 8e. //,sAy f J~
NSR D accef 'ewc e C's e r? e
/r .
as J /ar- (enz) = /. o nu d er ei Ap J,, Ja cces '<-w. eri l-em ,7.F9st-efkrpy 7,-uu ~ J ~en u ga/
y N M </d DJ k W =gt) u fi
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f = J~44y lAlin fle te k C% fv le-re- y'fe f c2?B **b'C xas th/#ny*ms$ h f;dL a 9tB =%
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Mto -loot B E.Co. FORM 3930
4 r
e Attachment I - Ind:psnd:nt Verification Statement /f//- g Celculation ##- , Revision # d has been independently verified by the following method (s), as noted below:
kat G
Yol Y Design Review. 7 including verification that:
'V Design inputs were correctly selected and included in the calculation.
'd Assumptions.are adequately described and are reasonable.
. Input or assumptions requiring confirmation are identified, and if any exist, the cal-culation has been identified as " Preliminary" and a " Finalization Due Date" has been specified.
- / Design requirements from applicable codes, standards and regulatory documents are identified and reflected in the design.
- ' Applicable construction and operating experience was considered in the design.
- ' The calculation number has been properly obtained and entered.
V An appropriate design method or computer code was used.
/ A mathematical check has been performed. .
V The output is reasonable compared to the input.
Alternate Calculation Q including verification of asterisked items noted above.'-'
The alternate calculation ( pages) is attached.
Qualification Testing O for design feature includ-ing verification of asterisked items noted above and the following:
The test was performed in accordance with written test procedures.
Most adverse design conditions were used in the test.
Scaling laws were established and verified and ttror analyses were performed, if applicable.
Test acceptance criteria were clearly related to the design calculation.
Test results (documented in ) were reviewed by the calculation Preparer or other cognizant engineer.
Independent Verifier Comments:
/7 .
See NED Procedure 3.05, Sec. 7.1.1 /s/ /d/ e #M Independent Verifier SM[M ~
Dd te' Preparer concurrence with
>~ M R5 findings and comment resolu- <f
/s/
r
- - g tion Preparer or other / Date
BOSiON
~ ~
EDISON -
0 b' Nb f[$J7}
l NUCLEAR OPERATIONS DEPARTMENT PILGRIM NUCLEAR POWER STATION Procedure ,No. 3.M.3-8
, INSPECTION / TROUBLE SHOOTING - ELECTRICAL CIRCUITS 1
List of Effective Pages 3.M 3-8-1
' 3.M.3-8-2 3.M.3-8-3 Attachments .
3.M.3-8A-1 '
3.M.3-8A-2 Approved Mf .
ORC Cfiairman Date d We r /4 /9/*/
3.M.3-8-1 Rev. 6 ATTACHMENT 4
. ~
I. PURPOSE To prcvids guidalinas fcr personnel inspecting and tr'oubleshooting electrical circuits and to document that investigation and its results.
II. DISCUSSION Inspecting / Troubleshooting may be performed due to any malfunction of an electrical circuit switch results in abnormal operation. The malfunction may take the form of electrical grounds, equipment failure, short circuits, open circuits, sequencing problems, or loose wires as examples.
Troubleshooting electrical circuits may require equipment or circuit isolation. The momentary opening of circuits under controlled -
circumstances and with the Watch Engineer's understanding of those components affected and his permission to perform does not constitute making the systems inoperable.
III. REFERENCES -
A. PNPS Single Line Diagram E-13: Single Line Relay & Meter Diagram 125V & 250V DC Systems -
B. PNPS Single Line Diagram E-14: Single Line Diagram 120V Instrument AC, Vital & Reactor Protection AC Systems & 124 VDC Power System C. Special Order 81-02 dated 10/8'/81 4 IV. PREREQUISITES 1
A. Plant Conditions
, 1. Any condition permissible within Technical Specification Requirements. .
B. System Conditions i
- 1. Any condition permissible within Technical Specification Requirements.
C. System Isolation
- 1. Proper tagging procedure is to be observed.
s D. Radiation Work Permit (RWP)
- 1. Check with Health Physics on conditions and requirements; initiate an RWP if required.
E. Signed Maintenance Request if required.
F. Any of the following in order to accomplish the return to normal operation.
- 1. Bechtel Prints .
I. tenJor Prints
- 3. Vendor Manuals
- 4. BECo Prints
- 3.M.3-8-2 Rev 6
. _ . _ . _ _ -._. -. ~ _ . _ _ - - _ -
V. CPPORSTUS_
. 1. Cny test equipment necessary to, perform the Inspection / Troubleshooting
~ ~
required.
VI. PRECAUTIONS A. The approach and areas of investigation must be thoroughly discussed with the Watch Engineer and documented on the checklist.
B. Maintenance to be performed as a result of investigation shall be documented on a signed Maintenance Request.
C. Special note must be taken in tnose areas where redundant equipmen't cross interlocks are involved. There is a potential especially in AC distribution Panels Y1, Y2, Y3, and Y4 and DC distribution Panels D4, DS, and 06 breakers for enabling redundant safety related systems Isolation of these loads other than momentary shall be on a component only basis, unless reviewed and approved by ORC.
VII. PROCEDURE A. The procedure will comprise those steps as discussed and agreed to by the Watch Engineer and documented on Attachment A.
VIII. ACCEPTANCE CRITERIA -
A. Inspection / Troubleshooting - Electrical Circuits shall be considered complete when Attachment A is signed off with no unexplained i discrepancies and with second verification of Systems Returned to Normal.
IX. ATTACHMINTS A. Inspection / Troubleshooting - Electrical Circuits Check List 3.M.3-8-3 Rev 6
- ~
ATTACHMENT A -
, INSPECTION / TROUBLESHOOTING - DATE: U ~/ 7 Y [>
ELECTRICAL CIRCUITS CHECK LIST' MR# //['/
/
- 1. State Problems V b L 'A).b ,Y l H G r'Et' I~ A ].L W E L'r 2 A ) , .4 N n (Cl?s u P D RgM;($
$ 12 1 1 i 9 4 PPLI(' 4 il/ h' 1 e1 F' Cc P1PRK6S]t.4.s l i o rCG lg 4 A f,
/
[ 4 N I~ L C l n! fj, _ Le 1~L' t L R f6 Pl A n !) ^ ^ ^ ^-^* ^ " ~ ' ^ ^ ' u _kA Der mW v
- 2. State Trouble Shooting Technique Potential Effects VineA L1 s/ i r~u < PKe r Ate n/ht<C-_
- es/VK
~~
/ Arv;// K h ,
i
/f. 3.5HiDNEFL 7 Maintenance Techniciand' N 6 d/1 E Watch Engineer fu l f .
- 3. Cause rY 4 Corrective Action 4 r
- 5. Associated Documentation (if applicable) MR# A Jumper Loq Entry # //,'[' ,'l
- 6. As Left Condition A 3 b c' # O
- a. Systems Returned to Normal ./I4- _ , _ _ _
/b ef Vofified Signed
- b. Other 4 s-/
3.M.3-0A-1 Rev. 6
Discrepancies: . Af4N$ -
5 ['[- L ( (- Y M '
-] l 6' O fl.', C'. '[ ] I ~ E N l'.) { .. c; f( 0 - 910 -C5 des- cK C -901 - c K 6-of- c K c-qq-og 6e3 - eg -
CC/ - og l- 931 ^ c) k Co).~ g g
~
b~ Nh"0 N Cyes og C-/ 7p eK C903 - CK. C -17 i- e l<
d7J.7-c(
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Oc (c cg
')7l- Cli. w :EA 't? ck t'.. j 7c _ (W p.5'O. N' - c;Q C.-c4 -c X Date performed [-/7~ d Plant Status
~
dd/A68 The Acceptance Criteria specified in Section VIII was met except as noted in discrepancies.
Test Equipment Used (Serial No./ Calibration DATE DUE .4 Maintenance Technician saw ee%jD te I -/7 - 7d Maintenance Engineer > Date _8 ' / 9- Ob ,_
Watch Engineer Nt
(< r
/ ,M Date __ si <t
/'184-j Gv-3.M.3-80-2 Rev. 6
BOSiON
~
EDISON (1,-9LME S'i'RE4DikC- R66!l T
l NUCLEAR OPERATIONS DEPARTMENT PILGRIM NUCLEAR POWER STATION Procedure ,No. 3.M.3-8
. INSPECTION / TROUBLE SHOOTING - ELECTRICAL CIRCUITS I
a List of Effective Pages
+ 3.M.3-8-1 3.M.3-8-2 3.M.3-8-3 Attachments i
3.M.3-8A-1 3.M.3-8A-2 Approved 48 .
ORC Cfairman Date $ for be /k /9Pt 3.M.3-8-1 Rev 6 l
i ATTACHMENT 5
_I. PURPOSE To pr:vida guidelines for personnel inspecting and troubleshooting electrical circuits and to document that investigation and its results.
II. DISCUSSION Inspecting / Troubleshooting may be performed due to any malfunction of an electrical circuit switch results in abnormal operation. The malfunction may take the form of electrical grounds, equipment failure, short circuits, open circuits, sequencing problems, or loose wires as examples.
Troubleshooting isolation. The momentary electrical circuits may require equipment or circuit opening of circuits under controlled circumstances and with the Watch Engineer's understanding of those components affected and his permission to perform does not constitute making the systems inoperable.
III. REFERENCES -
A.
PNPS Single Line Diagram E-13: Single Line Relay & Meter Diagram 125V & 250V DC Systems B
PNPS Single Line Diagram E-14: Single Line Diagram 120V Instrument '
AC, Vital & Reactor Protection AC Systems & f24 VDC Power System C. Special Order 81-02 dated 10/8'/81 IV. PREREQUISITES 8
A. Plant Conditions
- 1. Any condition permissible within Technical Specification Requirements.
P B. System Conditions
- 1. Any condition permissible within Technical Specification Requirements.
C. System Isolation
- 1. Proper tagging procedure is to be observed.
D. Radiation Work Permit (RWP)
- 1. Check with Health Physics on conditions and requirements; initiate an RWP if reqJired.
E. Signed Maintenance Request if required.
F. Any of the following in order to accomplish the return to normal operation.
- 1. Bechtel Prints .
- 2. tenJoe Prints
- 3. Vendor Manuals
- 4. DECO Prints a 3.M.3-8-2 Rev. 6
, w
. .V. APPARATUS
- 1. Any test equipment necessary to, perform the Insbection/ Troubleshoot.ing required.
VI. PRECAUTIONS A. The approach and areas of investigation must be thoroughly discussed with the Watch Engineer and documented on the checklist.
B. Maintenance to be performed as a result of investigation shall be documented on a signed Maintenance Request.
C. Special note must be taken in those areas where redundant equipment cross interlocks are involved. There is a potential especially in AC distribution Panels Y1, Y2, Y3, and Y4 and DC distribution panels,D4, DS, and D6 breakers for enabling redundant safety related systems.
Isolation of these loads other than momentary shall be on a component only basis, unless reviewed and approved by ORC.
VII. PROCEDURE A. The procedure will comprise those steps as discussed and agreed to by the Watch Engineer and documented on Attachment A.
VIII. ACCEPTANCE CRITERIA -
l
- A. Inspection / Troubleshooting - Electrical Circuits shall be considered
! complete when Attachment A is signed off with no unexplained
! discrepancies and with second verification of Systems Returned to Normal.
IX. ATTACHMENTS A. Inspection / Troubleshooting - Electrical Circuits Check List r
l l
l 3.M.3-8-3 Rev 6 I
I l
l
t',TTACHMENT A
~
INSPECTION / TROUBLESHOOTING - DATE: 6 - / 7 ELECTRICAL CIRCUITS -
CHECK LIST MR# A//4
- 1. Sy te Problems l?t AL'N /$ ,V 2N Gt'I~Y i~ All NEUR 2AL ann (;2en.ch lTi.'s S E C /dP 111sh APPJ,i6A Th At sF r2 n Pcp g gls.A/
L U6-N, N All. f4 NGL 6 ld AARLF KPREA DEN 4 RM1v1)
EL L9 YBMn 6
- 2. State Trouble Shooting Technique Potential Effects Vi$t'AAAf lH&PE6[ No rMiM6- No N2 75or'//Kh, t
Maintenance Techniciad6//EE;9AA/5///DREM. Watch Engineer n b, 7 f w Q/E)
- 3. Cause #4
/
- 4. Corrective Action A l'
s
- 5. Associated Documentation (if applicable) MR# #[Af f Jumper Loq Entry # //
- 6. As left Condition I$ Id////O
- a. Systems Returned to Normal .# _ #A . , , , ,
Signed Verified
- b. Other 2f/N 3.M. 3-8 A-1 Rev. 6 i
4
1
( . .
Discrepancies: el /
- l'E
, , s
- O i
l C -ll - C/' l 6-y7-(K ;s
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_ 4 '/7-[b Date Performed Plant Status "..4f I .
I
- /.'?. . .: ;
i -
. ,: s '- -
The di Acceptance s::repancies . Criteria specified in Sectior VIII was me g/c.m%e .t .
-~~
+.. ,.
- res A .. .._.s e i r.
e- y ~~
spy Test Eq'uiprrent Used (Set ial No./ Calibration DATE DUE - E cp i
, : .2 f ty
.u.
\
Maintenance Technicia ihmsfo ,/ E d @ ate -
Maintenance Engineer ' Jd Ce-r/ A--- Date E~-[ -- bC "
Watch Engineer .# } O- Date BID'IM U.::
Q
' h>.g.
n .,.y 3.M.3-8A-2 Rev. 6
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