IR 05000397/1998301
| ML17284A822 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/09/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17284A821 | List: |
| References | |
| 50-397-98-301, NUDOCS 9812160064 | |
| Download: ML17284A822 (30) | |
Text
ENCLOSURE U.S. NUCLEAR REGULATORYCOMMISSION
REGION IV
Docket No.:
License No.:
Report No.:
Licensee:
Facility:
Location:
Dates:
Inspector(s):,
Approved By:
50-397 NPF-21 50-397/98-301 Washington Public Power Supply System Washington Nuclear Project-2 Richland, Washington November 2-6, 1998 T. O. McKernon, Chief Examiner, Operations Branch R. E. Lantz, Examiner, Operations Branch M. E. Murphy, Senior Examiner, Operations Branch T. R. Meadows, Senior Examiner, Operations Branch John L. Pellet, Chief, Operations Branch Division of Reactor Safety ATTACHMENTS:
Attachment 1:
Attachment 2:
Supplemental Information Simulator Facility Report Attachment 3:
Facility Initial License Written Examination Comments and'Analysis Attachment 4:
Final Written Examinations,"Answer Keys, and Proctor's Comments 98i2i60064 98i209 PDR ADOCK 05000397" V
-2-EXECUTIVE SUMMARY Washington Nuclear Project-2 NRC Inspection Report 50-397/98-301 NRC examiners evaluated the competency of 4 reactor operator and 8 senior operator applicants for issuance of operating licenses at the Washington Nuclear Project-2 facility. The licensee developed the initial license examinations using NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Interim Revision 8. NRC examiners reviewed, approved, and administered the examinations.
The initial written examinations were administered to all 12 applicants on October 30, 1998, by facilityproctors in accordance with instructions provided by the chief examiner.
The NRC examiners administered the operating tests on November 2-5, 1998.
~Oerations
~
Good operator performance and communication practices were observed during the initial operator licensing examination (Section 04.2).
Allapplicants for reactor operator and senior operator licenses displayed the requisite knowledge and skills to satisfy the requirements of 10 CFR Part 55 and were issued the appropriate licenses (Sections 04.1, 04.2).
h The licensee initiallyfailed to submit an acceptable examination for administration to operator license applicants for the operating test portion of the examinations.
The final as-given examination met the requirements of NUREG-1021 and was considered good quality (Section 05.1.2).
-3-Re ort Details Summa of Plant Status The plant operated at essentialiy 100 percent for the duration of this inspection.
I. 0 erations
Operator Knowledge and Performance
.04.1 Initial Written Examination Ins ection Sco e
On October 30, 1998, the licensee proctored the administration of the written examination approved by the NRC to four individuals who had applied for initial reactor operator licenses and eight individuals who had applied for senior operator licenses.
The licensee graded the written examinations and its staff reviewed the results.
The licensee also performed a post-examination question analysis, which was reviewed by the examiners.
b.
Observations and Findin s The minimum passing score was 80 percent.
Allapplicants for a reactor operator and senior operator license passed the'written examination.
Scores for the reactor operator applicants ranged from 86.9 to 87.9 percent.
Allsenior operator license applicants passed with scores ranging from 90.9 to 96.9 percent.
The average score for reactor operator applicants was 87.6 percent and the average score for senior operator applicants was 93.4 percent.
The above grades reflected the results after examination changes recommended by the licensee as a result of post-examination question analysis were incorporated.
The examiners reviewed and accepted these recommendations based on the technical merits of each recommendation.
As a result of this analysis, two choices were accepted for one question (RO 30, SRO 30) because the wording in choice "b" was also a true statement for the conditions given in the question stem.
Also, Questions RO 9 and SRO 11 were deleted because no correct choice existed.
The submitted comments and analysis are included as part of Attachment 3 to this report.
Conclusions All reactor operator and senior operator license applicants passed the written examinatio.l
04.2 Initial 0 eratin Test The examiners administered the various portions of the operating test to the 12 applicants on November 2-5, 1998.
Each applicant participated in 2 or 3 dynamic simulator scenarios.
Each applicant also received a walkthrough test, which consisted of either 10 or 5 system tasks, depending on application type, with 2 followup questions for each task. The applicants also received an operating test administrative portion consisting of tasks or questions related to 5 subjects in 4 administrative areas.
b.
Observations and Findin s Allapplicants passed all sections of the operating test. The applicants generally performed well and used good communication practices and peer checks.
The applicants exhibited good plant knowledge and ability to find components when asked.
Some applicants exhibited slow control board awareness during the dynamic scenarios.
For example, some applicants were slow to recognize that a control rod drive pump had tripped off after a loss of power to the applicable electrical bus. While the slowness in recognizing the loss of the component did not exacerbate plant conditions, it did exhibit weakness on the part of some board operators in control board awareness.
The examiners observed good ownership, and application of principles for self verification and peer checks by the applicants throughout the examination.
Conclusions Allapplicants passed all sections of the operating test.
Overall, the license applicants demonstrated good performance and use of good communication practices and peer checks.
Operator Training and Qualification 05.1 Initial Licensin Examination Develo ment The licensee developed the initial licensing examination in accordance with guidance
provided in NUREG-1021.
05.1.1 Examination Outline a.
Ins ection Sco e
The licensee submitted the initial examination outline on July 1, 1998. The examiners reviewed the submittal against the requirements of NUREG-102 Ep
-5-Observations The chief examiner provided enhancement suggestions related to examination integrity and responsiveness to NUREG-1021 requirements, which were incorporated by the licensee into the job performance measure and dynamic scenarios outlines.
The submitted scenario outlines contained malfunctions associated with instrument malfunctions which lacked significant operator response to correct.
Additional enhancement suggestions were made to the scenarios in order to add balance to the malfunctions so that one particular applicant did not receive all the malfunctions.
Conclusions The licensee submitted final revised examination outlines prior to the final version of the examination submittal. The revised outlines satisfied the requirements of NUREG-1021.
Examination Packa e
Ins ection Sco e
The licensee submitted the initial examination package on September 1, 1998.
Because of extensive NRC comments on the initial operating test submittal, the licensee submitted a revised operating test package on October 29, 1998, following an onsite review by the examiners during the week of October 19, 1998. The chief examiner reviewed the submittal against the requirements of NUREG-1021.
Observations and Findin s The licensee submitted 130 draft written examination questions, of which 70 were designated as common questions to both the reactor operator and senior operator examinations.
The chief examiner provided comments or questions on 12 of the questions.
Additionally, the licensee performed internal audits, which identified other questions for revision. The majority of changes made to the questions were changes to question stems and answers for clarification. Although failure to make the above changes would not have invalidated the written examinations, it would have degraded their discriminatory value. The written examinations were adequate for administration.
The licensee submitted two sets of the administrative test, one intended for the reactor operator applicants and one for the initial senior operator initial and upgrade senior operator applicants.
The reactor operator and the senior operator administrative tests were similar but the senior operators test differed in the questions asked and in the emergency preparedness areas.
The examiners'eview indicated that the administrative portion of the examination was unacceptable for administration because a
number of the questions did not discriminate at the appropriate level ~
Similarly, the system walkthrough portion of the examination was inadequate for administration.
The majority of the job performance measure followup questions could be answered by direct look up and did not discriminate appropriately for open reference question The examiners reviewed the scenarios and found them of good quality. Some enhancement suggestions were made for balancing malfunctions between applicants.
As a result of the above comments, the licensee resubmitted the operating portion of the examination on October 29, 1998. The revised examination materials satisfied NUREG-1021 requirements and were of good quality.
C.
Conclustions Although the written examination was acceptable, the licensee initiallyfailed to submit an acceptable examination for administration to the operator license applicants for the operating portion of the examination.
The final as-given examination met the requirements of NUREG-1021 and was considered good quality.
05.1.3 Licensin Conditions The chief examiner reviewed the final applications as submitted by the facilityfor the license applicants against the requirements of NUREG-1021, Interim Revision 8.
b.
Observations and Findin s The chief examiner verified that the facilitylicensee properly identified the required five significant reactivity manipulations on the applications.
The chief examiner also verified that the facilityhad properly documented these manipulations and that they were significant in accordance with NRC Information Notice 97-67.
c.
Conclusions The facility's program was adequate to ensure that initial license applicants satisfied the requirements for performance of significant reactivity manipulations 05.2 Simulation Facilit Performance The examiners observed simulator performance with regard to fidelityduring examination validation and administration..
t
-7-b.
Observations and Findin s The simulation facilitysupported examination administration well. However, minor unexpected simulator malfunctions and annunciations occurred during the dynamic scenarios.
Examples included:
(1) Annunciation of a high reactor vessel level alert alarm; (2) failure of the rod worth minimizer screen to update and insert control rod withdrawal block alarm annunciations; and (3) failure of the Division 3 diesel to tie-in to its safety bus. The simulator errors did not create a modeling problem with respect to mass or energy transfer and did not adversely affect the planned scenarios.
The simulator errors are further discussed in attachment 2.
Conclusions The simulator supported examination administration well with few minor exceptions.
V. Mana ement Meetin s X1 Exit Meeting Summary The inspectors presented the inspection results to members of the licensee management at the conclusion of the inspection on November 5, 1998. The licensee acknowledged the findings presented.
The licensee did not identify as proprietary any information or materials examined during this inspectio ATTACHMENT1 PARTIALLIST OF PERSONS CONTACTED Licensee P. Taylor, Operations Training Manager R. Guthrie, Operations Training W. Oxenford, Operations Manager J. McDonald, Engineering, General Manager D. Coleman, Regulatory Affairs Manager G. Smith, Plant General Manager C. Golightly, Simulator Group M. Westergren, Operations Training NRC J. Arildsen, Operator Licensing Branch, Office of Nuclear Reactor Regulation S. Boynton, Senior Resident Inspector
ATTACHMENT2 SIMULATIONFACILITYREPORT Facility Licensee: Washington Nuclear Project-2
'acility Docket: 50-397 Operating Examinations Administered at: Richland, WA.
Operating Examinations Administered on: November 2-6, 1998 These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility, other than to provide information, which may be used in future evaluations.
No licensee action is required in response to these observations.
During the performance of Scenario 5, Event 5, in which an earthquake caused a loss of coolant accident with a loss of Electrical Bus TR-S, the Division III HPCS diesel failed to tie onto the SM-4 bus. This was an unplanned and unexplained simulator error.
During the same scenario described above, a RPV high level alert alarm unexpectedly annunciated.
3.
In Scenario 1, during control rod withdrawal to critical the operators received an unexpected and repeated rod out block alarm and the rod worth minimizer display would not update dat ATTACHMENT3 Facility Initial Written Examination Comments and Analysis
%ritten Exam Performance Anal sis exam given at WNP-2 on Oct. 30, 1998 Our exam analysis revealed 11 questions for which greater than 25% ofthe students gave an incorrect response.
Ofthese 11, only 2 questions were determined to require 'modification that would alter the final grade.
These were:
1) question - SRO 11/RO 9/ex98010 and 2) question - SRO 30/RO30/ex98036.
For the remaining 9 questions.
No changes were made'to these questions.
However, training willbe provided to those students that missed these questions to upgrade their knowledge in the appropriate areas.
Question No.
Problem identified:
RO 9 Allapplicants missed this question.
9 of 12 answered A, 3 of 12 answered D.
ex98010 Recommendation:
Recommend deleting this question.
JustiTication:
The answer to this question was based on a previous revision to PPM 5.0.10.
The current revision (rev 5), no longer defines "MSIVoperation" as the basis.
Reference:
PPM 5.0.10 rev 5, pg 76, 77, 93.
Exceeding 241'F WW at 600 psig in the RPV exceeds HCTL, which exceeds otentiall PCP QUESTION ¹ 10.
ex98010 WNP-2 WRITTENEXAMINATION EXAMKEY 8/12/98 The plant is in an ATWS. Suppression Pool level is normal, Suppression pool temperature is 241'F, and reactor pressure is 600 psig.
Which ONE ofthe following is correct concerning these conditions?
A. ALOCA may cause a Wetwell/Drywell interface failure.
B.
A reactor depressurization may cause the MSIVs to become inoperable.
C.
Spraying the Wetwell may cause a containment failure due to low internal pressure.
D.
An emergency depressurization may cause SRV Tailpipe damage.
ANSWER:
B QUESTION TYPE:
SRO/RO PPM 5.0.10, rev 4, page 101 KA¹ &KAVALUE:
295013AK3.02 3.6/3.8 REFERENCE:
SOURCE:
NEW QUESTION. SRO Tl, Gl, ¹10 RO Tl, G2, ¹9 LO:
8303 RATING:
ATTACHMENT:
JUSTIFICATION:
L2 PPM 5.2.1 A is incorrect because a LOCAwould have no effect on the drywell fioor/seal.
C is incorrect because spraying the wetwell airspace would have no effect under these conditions. D is incorrect because there would be no damage with wetwell.level in the normal range during an ED.
COMMENTS:
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The Heat Capacity Temperature Limit (HCTL) is the highest wetwell temperature at which initiation of RPV depressurization will not result in exceeding the PCPL before the rate of energy transfer to the containment is within the capacity of the containment vent.
(Refer to discussion of the Primary Containment Pressure Limit.)
The HCTL is used to preclude failure of the containment or equipment necessary for safe shutdown by assuring that RPV blowdown does not cause containment pressure to exceed the PCPL.
For RPV pressures below 60 psig. the rate of energy transfer to the containment with 5 SRVs open is within the capacity of the containment vent.
Therefore when the energy discharged from the RPV to the wetwell is equal to or less than the energy discharged outside the primary containment tnrough the open vent, primary containment pressure will not exceed the PCPL.
Five open SRVs is the Minimum Number of SRVs Required for Emergency Depressurization.
(Refer to the discussion of'the Minimum Number of SRVs Required for Emergency Depressurization in sectio'n 7.8) 60 psig is the Minimum RPV Flooding Pressure.
(Refer to the discussion ot'he Minimum RPV Flooding Pressure in Section 7.9)
For RPV pressures above
psig, wetwell heatup during RPV depressurization is proportional'o RPV pressure.
Thus this segment of the HCTL decreases with increasing RPV pressure.
lfdepressurization is performed with more than 5 open SRVs, the integrated energy addition to the v,e!well is less than that resulting from fewer SRVs open because the rate ot depressurization of the RPV is faster.
The HCTL also assumes all heat removal capability from the containment is lost, and the airspace and water in the wetwell are in thermal equilibrium.
Operation a-: pressure above the minimum pressure in the RPV at which an SRV is set to lift (109l psig) is not expected to occur. therefore, the HCTL is not defined above this pressure.
An override in the RPI'ressure control flowpath of PPM 5.1.1 and PPM 5.1.2 direct the operator to control F,'PV pressure below HCTL. The wetwell temperature control.flowpath ot the PPM 5.2.1 requires emergency RPV depressurization when RPV pressure and wetwell temperature can not be maintained below HCTL.
Two graphs are giv.n. LE 39 ft. and GT 39 ft..
The unsafe area of the HCTL curve, is above and to the ric'.>t or'he level curve.
The HCTL is reter.need in PPM 5.1.1, PPM 5.1.2, and PPM 5.2.1 with the following identifier:
Pk(k'EDURE.'4l:hiHER 5.0.10 kEVIRI<)Y PACiE 77 of 312
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Question No.
SRO 30 Problem identified:
RO30 9 of 12 applicants missed this question; 6 SROs, 3 ROs. All9 answered B.
ex98036 Recommendation:
Accept either answer B or D.
Justification:
Answer B would be correct iftemperature got high enough to receive the
"HI-HF'larms. However, the students are not required to memorize the input logic to these alarms, and would expect an immediate isolation of RWCU-V-1 and 4 Refere'nce:
PPM 4.601.A3 pg 27, 28 PPM 4.601.A2 pg 33, 34.
PPM 4.601.A12 pg 28 PPM 4.601.A2 pg 15 PPM 4.601.A3
QUESTION ¹ 36 ri ex9S036 WNP-2 WRITTENEXAMINATION EXAMKEY 8/12/98 The plant is operating at 99% power, when the following alarms are received:
LEAKDET RWCU CH A DIFF FLOW HIGH/FLOWHIGH (59.5 GPM)
LEAKDET RWCU/RCIC PIPE AREATEMP HIGH Which ONEofthe following describes the expected plant response?-
In addition to the operating RWCU pump trip,....
A.
RWCU-V-4 closes immediately.
B.
RWCU-V-4 and V-1 close immediately.
C.
after a 45 second time delay, RWCU-V-4 only closes.
D.
after a 45 second time delay, RWCU-V-4 and V-1 close.
ANSWER:
D QUESTION TYPE:
SRO/RO KA¹ &KAVALUE:
295032EA1.05 3.7/3.9 REFERENCE:
SOURCE:
LO:
PPM 4.601.A3 drop 3-4, rev 11 PPM 4.601.A12 drop 6-2, rev 16 NEW QUESTION SROT1, G2, ¹10 ROT1, G3, ¹3 5035 RATING:
ATTACHMENT:
H3 NONE JUSTIFICATION:
D is correct because both RWCU-V-1 and 4 close after a 45 second time delay.
COMMENTS:
3-4 LEAKDET RWCU DIFFEIKNTIALFLOW HI/CH A 3-4 WINDOW LEAKDET RWCU CH A DIFF FLOW HU FLOW HI SOURCE LD-FS-605A (58.5 GPM d,flow and 45 second time delay)
(LD-RLY-K7A)
LD-FS-15 (253.5 gpm blowdown flow and 1.6 second time delay)
(MS-RLY-K27)
AUTOMATICACTIONS
~ RWCU-V-4 Closes
~NTE:
Delta flow indication is from U3-Fl-620 on H13-P602.
Total blowdown flow is indicated on RWCU-FI-602 at H13-P602.
1.
Ifdifferential fiow is GE 58.5 gpm and automatic isolation is imminent, perform all of the following:
fP-60468}
'.
Consider throttling open RWCU-V-104, Cleanup System Bypass, to preclude lifting RWCU HXR relief valves.
b.
Stop RWCU-P-lA(1B).
c.
Close RWCU-V-1.
d.
Close RWCU-V-4.
P e.
Ifnecessary, close RWCU-V-40 and secure RWCU CRD purge.
f.
Ensure RWCU-FCV-33 is closed.
2.
Ifisolation has occurred:
a.
Ensure RWCU-Y-4 closed.
b.
Consider throttling open RWCU-Y-I04, Cleanup System Bypass, to preclude lifting RWCU HXR relief valves.
'c.
Secure RWCU pump CRD purge if RWCU-V-I or V-4 and RWCU-V-40 are closed.
d.
Ensure RWCU-FCV-33 closed, locally ifpossible (RB 501, SW).
Ensure RWCU Pump tripped.
(CONTINUED ON NEXT PAGE)
PROCEDURE NllMBER 4. 601.A3 REVIS I0 Id PAGE 27 of 78
LEAKDET RWCU DIFF FLOW HIICH A (CONTINUED FROM PREVIOUS PAGE)
3.
Ifsystem restart is not anticipated in the near future, consider closing RWCU-V-34 or RWCU-V-3S to stop any leakage through RWCU-FCV-33.
4.
5.
Monitor RPV level, Reactor Building sumps, RWCU piping and equipment area temperatures, and Reactor Building ARMs for indication of RWCU leak.
Monitor Reactor mater conductivity at R%CU-CR-601 per PPM 1.13.1, Chemical Process Management and Control.
Ifinstrument operability is in doubt, refer to Technical Specification 3.3.6.1 in Modes 1, 2; aild 3,.
7.
Ifapplicable, refer to PPM 4.12.4.1A, High Energy Line Break.
8.
Ifapplicable, refer to PPM 4.11.2.1, Liquid Radioactive Spills.
REFERENCES:
N CVI 02E31-05,7,9 (GE 807E171TC) (Leak Det.)
EWD-4E-0017, EWD-4E-0023 EWD-19E-0006, EWD-19E-0028, EWD-19E-0038 LER 92-039-00-05 fP-60468)
PROCEDURE NUMBER 4.601. A3 REVISION Ph CiE 28 of 78
4-2 LEAKDET RWCU DIFF FLOW HI/CH B 4-2 WINDOW SOURCE AUTOMATICACTIONS LEAKDET RWCU CH B DIFF FLOW HI/
FLOW HI LD-FS-605B (58.5 gpm hfiow and 45 second time delay)
(LD-RLY-K7B)
RWCU-V-1 Closes LD-FS-16 (253.5 gpm blovPdown flow and 1.6 second time delay)
(MS-RLY-.K26)
5QTI.,: Delta fiow indication is from LD-FI-620 on H13-P602.
Total blowdown fiow is indicated on RWCU-FI-16 at H13-P613.
1.
Ifdifferential flow is GE 58.5 gpm and automatic isolation is imminent, perform all of the following:
g-60468)
a.
Consider Throttling open RWCU-V-104, Cleanup System Bypass, to preclude lifting RWCU HXR relief valves.
b.
Stop RWCU-P-lA(1B).
c.
Close RWCU-V-I.
d.
Close RWCU-V-4.
e.
Ifnecessary, close RWCU-V-40 and secure RWCU CRD purge.
f.
Ensure RWCU-FCV-33 is closed.
2.
Ifisolation has occurred:
a.
b.
Ensure RWCU-V-1 closed.
Consider throttling open RWCU-V-104, Cleanup System Bypass, to preclude lifting RWCU HXR relief valves.
C.
d.
e.
Secure RWCU pump CRD purge ifRWCU-V-1 or 4 and RWCU-V-40 is closed.
Ensure RWCU-FCV-33 closed, locally ifpossible (RB 501, SW).
Ensure RWCU pump tripped.
(CONTINUED ON NEXT PAGE)
PROCEDURE NUMBER 4.601.A2 REVISION PAGE 33 of 87
4-2 LEAKDET RWCU DIFF FLOW HIICHB (CONTINUED FROM PREVIOUS PAGE)
3.
Ifsystem restart is not ar,icipated in the near future, consider closing RWCU-V-34 or RWCU-V-35 to stop any leakage through RWCU-FCV-33.
4.
Monitor RPV level, Reactor Building sumps, RWCU piping and equipment area temperatures, and Reactor Building ARMs for indication of RWCU leak.
5.
Monitor Reactor vrater conductivity at RWCU-CR-601 per PPM 1.13.1, Chemical Process Management and Control.
6.
Refer to Technical Specification 3.3.6.1 and 3.3.6.2 in Modes 1, 2, and 3.
7.
Ifapplicable, refer to PPM 4.12.4.1A, High Energy Line Break.
8.
Ifapplicable, refer to PPM 4.11.2.1, Liquid Radioactive Spills.
REFERENCES:
EWD-1E-0060 EWD-4E-0016 EWD-19E-0029 EWD-19E-0041 CVI 02E31-05, 7, 11 (Leak Det.)
LER 92-039-00-05 (P-60468)
PROCEDURE NUMBER 4.601.A2 REVISION PAGE 34 of 87
6-2 LEAK DET RWCU/RCIC PIPE AREA TEMP HIGH 6-2 WINDOW SOURCE AUTOMATICACTIONS LEAKDET RWCU/RCIC PIPE AREA TEMP HIGH LD-TRS-624
~
point 2 - 118'
point 3 - 130'
point 4 - 128'
point 5 - 125'
point 6 -140'one l.
Identify the source of the high temperature.
l 2.
Compare recorder reading with temperature points on H13-P632 and H13-P642:
LD-TRS-624 Point 2 Point 3 Point 4 Point 5 Point 6 LD-MON-1A A3-3 (LD-TE-24A)
A1-4 (LD-TE-24C)
A2-4 (LD-TE-24E)
A3-4 (LD-TE-24G)
A1-5 (LD-TE-24J)
LD-MON-1B A3-3 (LD-TE-24B)
Al-4 (LD-TE-24D)
A2-4 (LD-TE-24F)
A3-4 (LD-TE-24H)
A1'-5 (LD-TE-24K)
3.
Ensure adequate ventilation to area.
4.
Check the RWCU/RCIC PIPE AREA for leakage and isolate ifpossible.
5.
Refer to PPM 5.3.1, Secondary Containment Control.
6. Ifa high energy release is confirmed:
a.
Refer to.PPM 4.12.4.1A, High Energy Line Break.
b.
Refer to PPM 13.1.1, Classifying the Emergency.
REFERENCES:
EWD-19E-0008, 0026, 0030 PROCEDURE NUMBER 4. 601. A12 REVlSlON
PAGE
,28 of 30
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2-5
. LEAK DET RWCU ROOMS TEMP HI-HI 2-5 WINDOW SOURCE AUTOMATICACTlONS LEAKDET RWCU ROOMS TEMP HI-HI Any of the following on LD-MON-1A, after a 1.0 second time delay:
A1-2 A2-2 A3-2 A3-3 Al-4 A2-4 A3-4 (LD-MON-1AFK6)
or Al-5 (LD-MON-1AFK8)
(160'F)
(160'F)
('150'F)
(160'F)
(160'F)
(16O'F)
(160'F)
(160'F)
~
RWCU-V4 Closes
~
RWCU-Y-4 Closes
~
. RCIC-V-8 Closes 5CITE: The high alarm setpoint-is set 20'F (minimum} LT the HI-HI(Isolation) setpoint.
1.
Identify the alarmed temperature point(s) on LD-MON-IA(H13-P632) and compare with reading of adjacent sensors as indicated below:
LD-MON-IA(H13-P632)
LD-TRS-608 LOCATION LD-MON-1B (H13-P642)
A1-2 (LD-TE-3A)
A2-2 (LD-TE-3C).
, A3-2 (LD-TE-3E)
Point Io Point I I Point I2 RWCU-P-IA Rm RWCU-P-IB Rm RM,'CU-HX-Rm A 1.-2 (LD-TE-3B)
A2-2 (LD-TE-3D)
A3-2 (LD-TE-3F)
LD-MON-IA (H13-F632) LD-TRS-624 L'OCATION LD-MON-IB (H13-P642)
A3-3 (LD-TE-24A)
AI -4 (LD-TE-24C}
A2-4 (LD-TE-24E)
A3-4 (LD-TE-24G)
AI-5 (LD-TE-24J)
Point 2 Point 3 Point 4 Point 5 Point 6 54S'ORTH PIPE CHASE 5's'ot'TII PIPE CHASE 5" RwCU PIPE CHASE 5" RwCU PUMP RM MEZZANINE 5QI 'WCUIRCIC MEZZANINE A3-3 (LD-TE-24B)
A I-4 (LD-TE-24D)
A2-4 (LD-TE-24F)
A3-4 (LD-TE-24H)
A1-5 (LD-TE-24K)
(CONTINUED ON NEXT PAGE}
PROCEDURE NlJMEER 4.60I.A3 REVISION PAGE 20 of 78
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