ML19276H202
| ML19276H202 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/15/1976 |
| From: | Arnold R Metropolitan Edison Co |
| To: | Brunner E NRC Office of Inspection & Enforcement (IE Region I) |
| Shared Package | |
| ML19276H201 | List: |
| References | |
| GQL-0560, GQL-560, NUDOCS 7910150436 | |
| Download: ML19276H202 (3) | |
Text
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- fffMO n u a m us METROPOLITAN EDISON COMPANY sussioiany of cenenst rustic uritiries cospanarian POST OFFICE BOX 542 READING, PENNSYLVANI A 19603 TELEI HONE 215 - 929-3601 April 15, 1976 GQL 0560 l
2 8
Mr. d1 don J. Brunner, Chief Reactor Operations and Nuclear Support Branch U.S. Nuclear Regulatory Coz:: mission 631 Park Avenue King of Prussia, Pennsylvania 19h06
Dear Mr. Brunner:
1 Docket No. 50-289 Operation License No. DPR-50 Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Inspection Report No. 50-289/76-03 This letter and the attached enclosure are in response to your inspecticn letter of March 25, 1976, concerning Mr. R. Hurd's inspection of TAI-1 and the resultant finding of an apparent deficiency.
Sincerely, R. C.
old Vice President RCA:JJM:cas Inclosure 1413 269 191onoq g s
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ENCL 0SURE
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Metropolitan Edison Cc=pany (Met-Ed)
Three Mile Island Nuclear Station Unit 1 (TMI-1)
Docket No. 50-289 License No. DFR-50 Inspection No. 76-01 Response to Apparent Deficency Ancarent Deficiency Technical Specification 3 5 1.1 requires that the reactor shall not be in a startup code or in a critical state unless the requirements of Table 3 5-1 are met.
Table 3.5-1 requires that at least one source range instrument channel be operable, or that the reactor be shutdown, or that an Intermediate Range Channel be above 10-10 amps.
1 l
Contrary to the above, on October 19, 1975, the reactor was brought critical fro = hot shutdown without a source range instru=ent channel operable.
Restonse
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Descrintion x-On October 17, 1975, prior to reactor startup, pulse height discriminator gain was not set correctly due to a =athematical error made during the pre-startup test of a Source Renge Instrument.
The error was not reported and reviewed when discovered on November 2k, 1975 The inspection finding was promptly reviewed by the Plant Cperations Review Cc==ittee.
We feel the error =ade in the calibration did not cause or threaten to cause an unsafe condition in the plant for the following reasons:
a.
The error in the affected Source Range instrument was so small that it resulted in a negligible effect on the instrument reading.
b.
The instrument provides no auto =atic safety function, it is only for indication.
The exact count is uni =portant, only the change in count rate is relevant to safety.
c.
The acceptance criteria in many surveillance procedures is conservative and not necessarily the li=it of the instru=ent's performance requirement.
In this case, the instrument was technically operable.
1413 270
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d.
The other Source Range instru=ent passed its previous and subsequent surveillance without adjustment and it, therefore, was also operable during the period in question.
Technical Specifications require only one Source Range instrument be operable.
Preventative Action Appropriate Operations and =aintenance personnel vill be instructed to insure that ano=alies observed during testing and surveillance be reported pro =ptly to the appropriate manage =ent personnel for review.
The Unit Superintendent has issued a se=orandum to cognizant personnel emphasizing the necessity of thoroughly checking the data sheets for errors.
Additionally the data sheets vill be redesigned to facilitate checking for such errors.
They vill be completed by May 1, 1976.
r-a 1413 271
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UNITED STATES
/ ' ~ [C NUCLEAR REGULATORY COMMISSION I
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REGION I 5BM l
$31 PARK AVENUE KING OF PRUSSI A. PENNSYLVANI A 19406 Metropolitan Edison Company License No. DPR-50 Attention:
Mr. R. C. Arnold Inspection No. 76-03 Vice President Docket No. 50-289 P.O. Box 542 Reading, Pennsylvania 19603 Gentlemen:
This refers to the inspection conducted by Mr. R. Hurd of this office on 1
February 10-20, 1976 of activities authorized by NRC License No. DPR-50 and to the discussions of our findings held by Mr.
R. Hurd with Mr.
Colitz of your staff at the conclusion of the inspection.
i Areas examined during this inspection are described in the Office of Inspection and Enforcement Inspection Report which is enclosed with this letter.
Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observations by the inspector.
Based on the results of this inspection, it appears that one of your O
ectivities was not conducted in fu11 co=P 1ance with NRC resuirements.
1 as set forth in the Notice of Violation, enclosed herewith as Appendix i
A.
This item of noncompliance has been categorized into the levels as described in our correspondence to you dated December 31, 1974.
This notice is sent to you pursuant to the provisions of Section 2.201 of the NRC's " Rules of Practice", Part 2, Title 10, Code of Federal Regulations.
Section 2.201 requires you to submit to this office, within twenty (20) days of your receipt of this notice, a written statement or explanation in reply including:
(1) corrective steps which have been taken by you and the results achieved; (2) corrective steps which will be taken to avoid further items of noncompliance; and (3) the date when full com-pliance will be achieved.
In accordance with Section 2.790 of the NRC's " Rules of Practice", Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosures will be placed in the NRC's Public Document Room.
If this report contains any information that you (or your contractor) believe to be proprietary, it is necessary that you make a written application within 20 days to this office to withhold such information from public disclosure.
Any such application must include a full statement of the
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Metropolitan Edison Company
,s reasons on the basis of which it is claimed that the information is proprietary, and should be prepared so that proprietary information identified in the application is contained in a separate part of the document.
If we do not hear frem you in this regard within the speci-fied period, the report will be placed in the Public Document Room.
i Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely,
(%
1 n J. Brunner, Chief Rea tor Operations and Nuclear Support Branch
Enclosures:
l.
Appendix A, Notice of Violation
,_)
2.
IE Inspection Report No. 50-289/76-03 cc:
J. G. Herbein, Manager, Generation Operations - Nuclear R. W. Heward, Project Manager, GPUSC l
Miss Mary V. Southard, Chaircan, Citizens for a Safe Environment (Without Report-)
l 1
I bec:
l IE thil & Files (For Appropriate Distribution)
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TIC REG:I Reading Room Region Directors (II, III, IV) (Report Only)
State of Pennsylvania Miss Mary V. Southard, Chairman, Citizens for a Safe Environment IAl3 273
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License No. DPR-50 APPENDIX A NOTICE OF VIOLATION Based on the results of the NRC inspection conducted on February 10-20, 1976, it appears that one of your activities was not conducted in full compliance with the conditions of your NRC Facility License No. DPR-50 as indicated below.
This item is an infraction.
Technical Specification 3.5.1.1 requires that the reactor shall not be in a startup mode or in a critical state unless the requirements of Table 3.5-1 are met.
Table 3.5-1 requires that at least one source range instrument channel be operable, or that the reactor be shutdown, or that an Intermediate Range Channel be above 10-10 amps.
Contrary to the above, on October 19, 1975, the reactor was brought critical from hot shutdown without a source range instrument channel operabic.
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1413 2/4 llM '
r ' Form 12 (Jon 75) (P.cv)
U. S. liUCLF/.R REGULATORY CC:CIISSIO:1 0FFICE OF I::SPECTIO:7 A :D E:iFORCC!E::T REGIO:1 I 50-289/76-03 Docket :o:
50-289 IE Inspection Report 1:o:
Metr Politan Edi' son Company Lic ense' !!o :
DPR-50 Licensec:
i P.
O.
Box 542 Priority:
Reading, Pennsylvania J9603 C
Category:
Saiecuards Group
- Three Mile Island I, Middletown, PA Location:
e of Licensce:
PWR, 2535 MWt, B&W Routine, Unannounced Type of Inspection:
February 10-20, 1976 Dates of Inspection:
February 5-6, 1976 Dates of Previous Inspectien:
Reporting Inspector:
d.
<*;' Mot / / //
M27/7I UAIE R. O.'Hurd, Reacto[ Inspector Acconpanying Inspectors:
None DATE DATE 1413 275 DATE Other Acconpanying, Personnel:
DATs_"
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A.
B.
Davis, Section Chief, cactor P oject Section No. 1 M/.lbffi-
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ncvicved By:
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O* '
E.
C.'M6Cabe, Section dhief, Nuclear Support Section No. 1
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SUMMARY
OF FINDINGS i
Enforcement Action Items of Noncompliance Infraction Technical Specification 3.5.1.1 requires that the reactor shall not be in a startup mode or in a critical state unless the requirements of j
table 3.5-1 are met.
Table 3,5-1 requires that at least one source i
range instrument thannel be operable, or that the r or that an Intermediate Range Channel be above 10-1gaetor be in shutdown, 4
amps.
Contrary to the above, on October 19, 1975 the reactor was brought critical from hot shutdoi n without a source range instrument channel operable.
(Detail 2.a)
Licensee Action on Previously Identified Enforcement Items Corrective actions tiken in response to Region I Inspection Report 50-289/75-21 were ccto19ted.
(Detail 26)
Design Changes None identified.
l 1
Unusual Occurrences i
None identified.
Other Significant Findings A.
Current Findings 1.
Acceptable Areas Source Range Channel Operability, October 6 and December a.
9, 1975.
(Detail 2.c) b.
Shif t and Daily Checks.
(Detail 3) c.
Reactor Coolant Evaluation.
(Detail 5) 1413 276
i
) d.
Reactor Coolant System Pressure.
(Detail 6) e.
Restriction on Reduction of Bcron Concentration.
(Detail 7) f.
Reactor Coolant Chloride and Flouride Concentration.
3 (Detail 8) g.
Plant Heatup and Cooldown Rate.
(Detail 9) h.
Operability of ECCS components for Plant Startup.
(Detail 10) 1.
Operability of Instrument Channels.
(Detail 11) j.
RPS Channel Bypass Key.
(Detail 12) k.
RPS Shutdown Bypass.
(Detail 13)
- 1. _
Control Rod Operating Limits.
(Detail 14)
I m.
Maximum Rod Worth Test.
(Detail 15)
I n.
Control Rod Insertion Ti=e.
(Detail 16) l o.
Power Distribution Limits.
(Detail 17)
I p.
Reactor Building Pressure Trip Setpoints.
(Detail 18)
I q.
Reactor Coolant Pressure (LPI Actuation) Setpoints.
(Detail 19) t r.
Operability of Emergency Diesel Generators.
(Detail 20) s.
Operability of Unit Electrical Pcwer System.
(Detail 21) t.
Limits on Reactor Building Pressure while critical.
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(Detail 23) u.
Maintenance and Design Modifications.
(Detail 24)
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Reactor Coolant Pump Combinations Versus Power Level Limit.
(Detail 25) 2.
Unresolved Items 1
a.
Source Range Operability, Nove=ber 24, 1975.
(Detail 2.b)
I b.
Flux-Reactor Coolant Flow Comparator Channel.
(Detail 4.a)
Nuclear Overpower Trip Setting.
(Detail 4.b) c.
d.
Limits on Use of Reactor Building Polar Crane.
(Detail 22)
B.
Status of Previousiv Unresolved Items Not inspected.
Management Interview An exit interview was held onsite on February 20, 1976 at the conclusion of the inspecticn.
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, i 3-Licensee Attendees Metropolitan Edison Company Mr. J. J. Colitz, Unit No. 1 Superintendent Mr. D. L. Cood, Technical Analyst III Hr. C. E. Hartman, Electrical Engineer Mr. C. A. Kunder, Supervisor, Station Operations Mr. W. E. Potts, Supervisor, Quality Control The following items were discussed:
A.
Source Range Instrument Operability Requirements and Surveillance (Detail 2).
B.
Flux-Reactor Coolant Flow Comparator Surveillance Procedure (Detail 4.a).
C.
Nuclear Overpower Trip Setting (Detail 4.b).
l D.
Review of operation with Inoperable Rods (Detail 14).
E.
Documentation on operation of Polar Crane (Detail 22).
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DETAILS 1.
Persons Contacted Discussions were held with the following persons during the conduct of the inspcetion activities documented in this report.
Mr. M. L. Beers, Shift Supervisor Mr. N. R. Bennett, Instrument Technician Mr. D. J. Boltz, Shift Foreman Mr. T. L. Book, Shift Foreman s
Mr. K. P. Bryan, Shift Foreman Mr. J. J. Chwastyk, Shift Superiisor Mr. J. J. Colitz, Unit No. 1 Superintendent Mr. D. E. Cuny, Engineer I, Nuclear Engineering Department Mr. D. L. Good, Technical Analyst III Mr. D. Harper, Instrument Foreman Mr. L. D. Hydrick, Radwaste Foreman Mr. G. A. Kunder, Supervisor, Station Operations Mr. D. E. Lang, Instrument Technician Mr. H. Mitchell, Electrical Maintenance Supervisor Mr. L. G. Nell, Shift Foreman Mr. R. L. Parnell, Auxiliary Operator Mr. R. H. Porter, Shift Supervisor Mr. P. J. Schwartz, Control Room Operator (in-training)
Mr. D. M. Shove 11n, Maintenance Supervisor Mr. J. Wellace, Shift Supervisor Mr. H. Wilson, Instrument Foreman 2.
Source Ran e Channel Operability The inspector reviewed " Plant Precritical Checklist," Enclosure I of OP 1102-2, Plant Startup.
Item 22 of this checklist requires that one source range channel be operabic and that the operability be demonstrated by completion of SP 1303-7.2, Source Range Channel Surveillance Test.
The licensec stated that although two source range channels are provided, OP 1102-2 and SP 1303-7.2 meets the requirements of Technical Specifications (T.S.) Table 4.1-1 item 6 which requires that the Source Range Channel Surveillance test be performed prior to a startup and T.S. table 3.5-1 item 4, which requires at least one source channel operable prior to startup and until 1 of the 2 intermediate range channels is above 10-10 amps.
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The inspector reviewed the precritical checklists completed prior j
to reactor startups on October 6, 1975, October 19, 1975, November 24, 1975 and December 20, 1975 and verified that item 22 had been i
signed off indicating that one source range channel was available.
Review of the surveillance records for the source range channeJs indicated that the following channel tests (SP 1303-7.2) were performed.
Date Source Range Channel October 6, 1975 Channel 1 October 17, 1975 Channel 2 l
November 24, 1975 Channel 2 December 17, 1975 Channel 1 The inspector reviewed the data sheets used to document the comple-tion of the channel tests listed above.
The findings from this review are detailed below.
e^
a.
Startup of October 20, 1975
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The count rate amplifier gain value, Item 6.2.2.b, on the SP 1303-7.2 data sheets completed October 17, 1975 was incorrectly entered as 3.63.
The gain value is calculated by dividing the measured output by the input.
On the data sheets these values were entered as item 6.2.1 and 6.2.2a with the values of 2.0 and 0.6.
The actual value of item 6.2.2b is therefore 3.333.
The procedure lists the acceptance criteria for this value as 3.75 t 10% (3.375 to 4.125).
Therefore the Count Rate Amplifier gain was not within the acceptable range and the surveillance test for Source RanFe Channel No. 2 was not successfully I
completed prior to reactor startup of October 20, 1975.
Since the surveillance test on Source Range Channel no. I had not been performed during the previous week and T.S. 1.3 which defines " operable," states in part " tested periodically in accordance with specification 4," the reactor startup on October 19, 1975 was conducted without at least one source range channel operable.
This is contrary to T.S. Table 3.5-1 item 4 and is an Item of Noncospliance at the infraction level.
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Startup of November 29, 1975 The inspector noted that item 6.2.2.b of the data sheet for SP 1303-7.2 performed on Ncvember 24, 1975 had a value of 3.333 entered on the value of Count Rate Amplifier gain and item 6.2.3 the "as left" gain was entered as 3.73.
The procedural i
i step 6.2.3 of SP 130?-7.2 requires that steps 6.2.1, 6.2.2a and 6.2.2b be repeated to bring the Count Rate Amplifier gain within the acceptable value.
The inspector expressed concern that the data sheet did not provide documentation of the repeated steps.
This item is unresolved pending revision of the data sheet for SP 1303-7.2.
c.
Startup of October 6, 1975 and December 21, 1975 i
The Inspector review of the data sheets for SP 1303-7.2 performed on October 6, 1975 and December 19, 1975 identified no inadequacies.
3.
Shift and Daily Checks
-()
The inspector reviewed surveillance precedure SP 1301-1, Shift and Daily Checks.
This procedure is used to satisfy the requirements of TS Table 4.1-1 relating to shift checks on instrument channels.
The inspector varified that the procedure included the requirement for shift checks on Control Rod Relative Position and Reactor Coolant temperature channels.
The inspector reviewed the SP 1301-1 data sheets for each shift chack perfor=ed in November and December 1975.
This review verified that the reccrded values satisfied the stated acceptance criteria of " Readings are acceptable if within normal expected ranges for various plant conditions."
The inspector found no inadequacies in this area.
4.
Reactor Protection System Channel Tests The inspector reviewed surveillance procedure SP 1303-4.1, Reactor Protection System.
The purpose section of this procedure states that this procedure is used for compliance with the requirements of TS Table 4.1-1 relating to the mcnthly functional test of the RPS instrument chcnnels.
The inspector specifically reviewed the sections of the procedure pertaining to the Flux-Reactor Coolant Flow Comparator Channel, Pump Flux Comparator Channel. Power Range Channel, High Reccter Ccolant Pressure Channel and Low Reactor Coolant Pressure Channel.
The inspector had the following findings.
1413 281
!,J Flux-Reactc' joolant Flow Comparator Channel a.
The inspector determined that procedure SP 1303-4.1, section 6.5.9, checks the Flux Imbalance / Flow Trip bistable at seven different power / flux imbalance / flow trip points.
The inspector expressed concern that procedure did not in-clude verification of the trip bistable at the extremes of the flow range described in TS Figure 2.3-2.
The procedure only checked trips at flow reductions to maximum of 8%.
TS Figure 2.3-2 indicates that the trip point is reduced by flow decreases from rated of up to 51%, during power opera-tion with only two pumps in service.
s The inspector discussed the above with the licensee and l
reviewed the appropriate electrical schematics of trip circuits. The flux imbalance / flow trip point in each RPS circuit is generated from a linear amplifier which receives the reactor flow signal from the individual loop flow elements via differential pressure transmitters, square rooters and a summer.
The licensee stated thct verification over the x,j small flow range confirmed the overall circuit linearity.
The inspector also noted that power operation with less than rated flow had not occurred and therefore trips at the lower flow condition had not been required.
This item is unresolved pending procedural changes or further technical support for not performing trip point verification for reduced flows.
b.
Nuclear Overpower Trio Setting In addition to section 6.5.8 of SP 1301-1 which specifics the setting of the cierpower trip bistable to within TS Table 4.2-1 item 1 requirements, the inspector reviewed the procedure and documentation relating to reducing of the overpower trip setting to below 5" when the RPS channels are in Shutdown Bypass. Plant cooldown procedure OP 1102-11 and plant heatup
. procedure OP 1102-1 both require proper action; however these procedures refer to RPS operating procedure OP-1105 vith
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' subsequent referral to the surveillance procedurc.
The inspector expressed concern that this caused an inconsistency in the type of documentation provided for the overpower trip setting. The licensee stated that OP 1102-1 and OP 1102-11 would be revised to specify the performance of the specific steps in SP 1303-4.1 relating to the overpower trip settings.
This item is unresolved pending review of the revised proce-dure.
c.
RPS Surveillance October, November, December, 1975 The inspector reviewed the data sheets completed for SP 1303-4.1 during October, November and December 1973.
The inspector i
verified that the setpoints for Flux-Reactor Coolant Flow Com-parator, Nuclear Overpower, Low Reactor Coolant Pressure, High Reactor Coolant Pressure and Pump Flux comparator were within the requirements of TS table 2.3-3 The inspector found no inadequacies in this area.
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5.
Reactor Coolant Leakage Evaluation The inspector reviewed SP 1303-1.1, Surveillance Procedure for RC System Leakage.
The purpose section of thic report states, "To evaluate reactor coolant system leakage in accordance with TS tabic 4.11-2 item 7."
The inspector reviewed the data sheets for SP 1303-1.1, performed during the December 1975 on a twice daily basis when the reactor average tecperature was greater than 5250F.
All of the evaluations indicated leakage was within the Technical Specification 3.1.6 requirements.
The inspector also reviewed the minutes of PORC Meeting No. 313 December 15, 1975 concerning the leakage of the bonnet gasket on MU-V49.
The inspector had the following findings.
The leak was detected by the control room operator's observa-a.
tion that the Makeup Tank level decrease was greater than normal, increases in indication of airborne activity in the Reactor Building and increases in Reactor Building Sump fill rate.
The above observations over a period of 12/12 to 12/14/75 resulted in an inspection of the reactor building where the leakage from the gasket on MU-V94, a high pressure injection check valve, was observed.
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The minutes of the PORC meeting held December 15, 1975 in-dicated that this leak did not present a threat to public safety since all sources of release to the environment vere through controlled and monitored systems.
Airborne releases are through the RB Purge system and liquid releases from the RB sump are through the Liquid Waste Disposal system.
I c.
The PORC recommended that MU-V94 be repaired as soon as possible because continued operation would increase the amount of leakage and make repairs more difficult.
Subsc-quently the plant was shutdown on December 16, 1975 and the valve leakage repaired.
The inspector found no inadequacies in the above evaluations and actions.
j 6.
Reactor Coolant System Pressure j
The inspector reviewed the Control Room Operator's Log and Shift l
and Daily Check (SP 1301-1) data sheets for November and December 1975.
The inspector found no indication that the Reactor Coolant System' pressure exceeded the 2750 psig limit established in TS 2.2.1.
7.
Restriction on Reduction of Boron Concentration Technical Specification 3.1.1.lb states, "The boron concentration in the reactor coolant system shall not be reduced unless at least ons reactor coolant pump or one decay heat removal pump is circu-l lating reactor coolant." The inspector reviewed SP 1301-1 Shift and Daily Checks and determined that all of the Reactor Coolant Pumps were shutdown from 10/1/75 to 10/5/75, 10/16/75 to 10/18/75, 11/13/75 to 11/23/75 and 12/17/75 to 12/20/75.
By review of the Shift Foreman's log the inspector determined that Deca, Heat Removal System Loop B was placed in service at the time the last Reactor Coolant Pump was shutdown and that loop B continued to operate until an RCP was placed in operation.
The inspector found no inadequacies in this area.
8.
Reactor Coolant Chloride and Flouride Concentration The inspector reviewed the data sheets 1or the Surveillance Proce-dure for Reactor Coolant System Chemistrf (SP 1301-3) for the 1413 284
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' period of October, November and December, 1975. The data sheets list the results of a daily determination of Flourides and Chlorides in the reactor coolaat sample.
The inspector determined that the T3 3.1.5.2 limit of 0.15 ppm for chloride and the TS 3.1.5.3 limit.
of 0.10 ppm for Fluorides were not exceeded.
9.
Plant Heatup and Cooldown Rate Technical Specification 3.1.2.1 requires limits on Reactor Coolant system temperature, pressure and the rate of change of temperature during plant cooldown and heatup.
The inspector reviewed Operating Procedures 1102-1, Plant Heatup, and 1102-11, Plant Cooldown and verified that the Technical Specification requirements were included in the procedure.
The procedures alue required that, during each heatup and cooldown, the reactor coolant temperature and pressure be recorded by a computer trend recorder.
These charts are main-tained as documentation of compliance with the pressure and tempera-ture limits.
The inspector reviewed the pressure / temperature charts for the plant heatups which occurred on 10/5/75, 10/17/75, 11/27/75 and 12/20/75, and the plant cooldowns which occurred on s
- (J 10/16/75, 11/12/75 and 12/17/75.
The inspector determined that the i
limits specified in TS 3.1.2.1 were not exceeded during the plant i
heatup or cooldown listed above.
10.
Operability of ECCS Components for Plant Startup The inspector reviewed the " Plant Precritical Checklist" Enclosure I of OP 1102-2, Plant Startup, and verified that followirc specific conditions were included.
These conditions are required to be satisfied prior to the reactor being made critical per TS 3.3.1:
Two makeup (HPI) pumps are operable.
(TS 3.3.1.lb) a.
b.
Two decay heat removal pumps are operable.
(TS 3.3.1.lc) c.
Core flood tank discharge valves are open and the electric circuit breakers are open.
(TS 3.3.1.2c) d.
Two reactor building spray pumps and associated nozzles, fans and coolers are operable.
(TS 3.3.1.3a) e.
All manual valves in the discharge lines of the sodium thio-sulfate and sodium hydroxide tanks shall be locked open.
(TS 3.3.1.3c)
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Two decay heat closed cycle cooling water pumps must-be operable.
(TS 3.3.1.4c)
The inspector determined that items 5, 9, 20, 13, and 16 re-spectively in the Precritical Checklist correspond to the condi-tions listed above.
The ins ector reviewed the Precritical checklist for the Plant Startups on 10/6/75, 10/19/75, 11/24/75 and 12/20/75 and observed that the items listed above were signed off in an appropriate manner to indicate the conditions of TS 3.3.1 were satisfied prior to criticality.
11.
Operability of Instrument Channels The inspector reviewed " Plant Precritical Checklist," Enclosure I i
of OP 1102-2, Plant Startup and verified that the minimum number of operable channels for the following instruments were included in this checklist.
The instrument channel operability is a requirement of TS 3.5.1.1, which must be satisfied prior to criticality and while in a critical state.
i
)
Flux / imbalance / flow channel, 2 channels operable.
(TS tcble a.
3.5-1, item 7) l b.
High Reactor Coolant Pressure Channel, 2 channels operable.
(TS table 3.5-1, item 8a)
HPI Reactor Coolant Pressure Channel, 2 channels operable.
c.
(TS table 3.5-1, item la (ESF))
d.
LPI Reactor Coolant Pressure Decay H' at Valve Interlock e
bistable, 1 channel operable. ' (TS table 3.5-1 item 2a (ESF))
The inspector determined items 22A, 22B and 46A respectively in the Precritical Checklist correspond to the instrument operability requirements listed above.
The inspector reviewed the Precritical Checklist for the Plant Startups on 10/6/75, 10/19/75, 11/24/75, and 12/20/75 and observed that the items listed above were signed off in an appropriate manner to indicate the operability conditions of TS 3.5.1.1 were satisfied prior to criticality.
The inspector reviewed the SP 1301-1, Shif t and Daily Checks data sheets for the period of October, November and December.
This review indicated that the channel operability requirements listed above were satisfied during the time the reactor was critical f
during that period.
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RPS Channel Bypass Key The inspector determined, by discussion with the licensee, that only one cuannel bypass key was in the control room in accordance with Technical Specification 3.5.1.2.
Plant keys are kept in a key locker adjacent to the Shift Foreman's desk.
This key locker is normally incked.
When any key is removed, the person charged with the key is entered on the key log.
The inspector reviewed this log to verify that only one RPS Channel Bypass Key was checked out at any given time. The inspector also observed that only one such key was available in the key locker.
By inspection and discussion, the inspector determined that the licensee scheduled RPS surveillance i
testing on a one channel per week basis and therefore the surveil-lance schedule aids in preventing any attempt to bypass two channels j
simultaneously.
l The inspector found no inadequacies in this area.
13.
RPS Shutdown Bypass Key Technical Specification 3.5.1.4 states that the shutdown bypass key
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switch shall not be used during reactor operation.
The inspector determined, by discussion with the licensee, review of OP 1102-1, Plant Heatup and SP 1303-4.1, RPS Surveillance that the RPS t
channels are removed from Shutdown Bypass before Reactor Coolant pressure exceeds 1700 psig and that the only time the Shutdown Bypass key switch is used during reactor operation is during the channel test when the channel is bypassed.
The inspector reviewed the OP 1102-1 documentation for plant heatups on 10/6/75, 10/18/75, 11/24/75 and 12/20/75 and verified that each RPS channel had been removed from shutdown bypass prior to power operation.
The inspector found no inadequacies in this area.
14.
Control Rod Operating Limits The inspector reviewed the Shift Foreman's Log, Control Room Operator's and t'.e data sheets for SP 1301-1, Shift and Daily Checks to determine if the following technical specification limits on control rod operation had been exceeded during October, November and December.
a.
Control Rod Insertion-Withdrawal Limits.
(TS 3.5.2.5b) b.
Operating rod group overlap.
(TS 3.5.2.Sa) c.
Operation with inoperable rods..(TS 3.5.2.2)
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The inspector review indicated that the only time one of the above was exceeded was on 11/12/75 when two rods were inoperable.
This exceeded the recuirement of TS. 3.5.2.2a which states, " Opera-tion with more than one inoperable rod as defined in TS 4.7.1 and 4.7.2.3 in the safety or regulating rod banks shall not be per-mitted." The plant was shutdown immediately and the rods repaired.
The licensee reported this item of noncompliance as A0 75-39 and this item was reviewed in NRC Inspection Report 76-01 as detail 3.b(1).
The inspector had no further questions in this area.
- 15. Maximum Rod Worth Test The inspector reviewed Enclosure 5 of SOP 199 dated 5/17/75,
" Ejected Control Rod Worth Measurements" this test is performed to verify compliance with TS 3.5.2.3 which states in part "the worth of a single inserted control rod shall not exceed 0.65 percent ak/k at rated power or 1.0 percent ak/k at hot zero power." The data from SOP 199 performed 6/22/75 indicated the
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worth of control rod H-14 was 0.58 percent Ak/k at zero power hot condition.
The licensee stated that H-14 had been deter-mined to be.the strongest rod from computer code analysis and that the.od would be worth less at rated power.
The licensee stated the evaluation and measurement was still valid since no core or control rod alteration had taken place since the i
time of the test.
The inspector had no further questions in this area.
16.
Control Rod Insertion Time The inspector reviewed Startup Test TP 330/5, CRD Trip Test, perforr-ed during the initial Startup program to satisfy the require-ments of TS 4.7.1.1.
The test measured the rod drop time for each control rod.
The maxicum time for a rod drop with the reactor coolant being hot and no flow was 1.12 sec versus the TS limit of 1.7 sec, and foe the coolant hot and at rated flow the maximum time was 1.3o3 versus the TS limit of 1.4 sec.
The inspecter found no inadequauies in this evaluation.
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( 17. Power Distribution Limits Technical Specification 3.5.2.4d and 3.5.2.5d requires determination j
of quadrant tilt every two hours when the reactor power is above 15%
and.of core imbalance every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when reactor power is above j
40%.
The inspector reviewed the data sheets for SP 1301-1 for October, November and December 1975.
This review verified that the determinations had been performed when required by the technical specifications.
The Quadrant Tilt did not exceed the limit of TS 3.5.2.4a (4%).
The Core Imbalance did not exceed the limit specified by TS Figure 3.5-2E.
The inspector found no inadequacies in the determination listed above, s
18.
Reactor Building Pressure Trip Setpoints Technical Specification table 2.3-1 item 8 specifies that the Reactor Protection System trip setpoint for Reactor Building Pressure be less than or equal to 4 psig.
The inspector reviewed SP 1303-4.13, Reactor Building Emergency Cooling and Isolation System Analog Channel Test, Sections 6.1.7,'6.2.7 and 6.3.7 which are used to
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check / adjust the Reactor Building High Pressure Trip bistables.
I The inspector determined that the procedures required that the bistable trip be within 2.4 psig to 2.6 psig or adjusted to within that range.
The bistable trip point is determined by using an g
analog test input voltage to the bistable which corresponds to the analog input from the pressure trans=itter.
The inspector reviewed the surveillance records for SP 1303-4.13 performed on 10/3/75, 11/7/75 and 12/3/75 and verified that the "as found" setpoints were within the. requirements of SP 1303-4.13 and therefore the l
requirements of TS table 2.3-1 item 8 were satisfied.
19.
Reactor Coolant Pressure (LPI Actuation) Setooint Technical Specicification 3.5.3.1 specifies that LPI actuation setpoints for low reactor coolant pressure are set greater than or equal to 1500 psig and 500 psig.
The inspector reviewed SP 1303-4.11, High and Low Pressure Injection Analog Channel Test.
The inspector determined that this procedure checked / adjusted the setpoints to a value of 1540 to 1550 psig and 540 to 550 psig.
~,aa trip points are determined by using an analog test input voltage to the bistable which corresponds to the analog input from the pressure transmitters.
The inspector reviewed I
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the surveillance records for SP 1303-4.11 performed on 10/6/75, 11/6/75 and 12/2/75 and verified that the "as found" setpoints were within the requirements of SP 1303-4.11 and therefore the i
requirements of TS 3.5.3.1 were satisfied.
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?O. Operability of Emergency Diesel Generators Technical Specification 3.7.le requires that the Diesel Generators be operable and at least 25,000 gallons of fuel oil be available in the storage tank prior to making the reactor critic 1.
The inspector reviewed " Plant Precritical Checklist", Enclosure I of OP 1102-2, Plant Startup Procedure.
Checklist item 28 specifies the requirements of TS 3.7.le.
The inspector reviewed the " Plant Precritical Checklist" completed prior to plant startup on 10/6/75, 10/19/75, 11/29/75 and 12/20/75.
For each startup item 28 was signed off in an appropriate manner to indicate that the requirements of TS 3.7.le were satisfied prior to startup.
21.
Operability of Unit Electrical Power System l
Technical Specification 3.7.2 specifies that the electrical systems and components listed below must be operable when the reactor is critical.
Two 230 kv lines to Unit 1 (TS 3.7.2a).
a.
b.
Both 230/4.16 kv unit auxiliary transformers (TS 3.7.2b).
Both emergency diesel generators (TS.3.7.2c).
c.
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d.
Engineered safeguards electrical buses, switchgear, load shedding and automatic diesel start (TS 3.7.2f)
Station batteries (TS 3.7.2g).
e.
The licensee stated that the operability of the systems and components-listed above are not documented by a specific document but that, if any of these systems or components were out of service or had not successfully completed a surveillance test, an entry would be made in the Shift Foreman's log and appropriate action taken.
The inspector reviewed the Shift Foreman's Log for October, November and December and the surveillance tests completed on this system or component during that same period.
The inspector found no indication that any of the
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j The inspector had no further questions in this area.
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- 22. Limits on Use of Reactor Building Polar Crane Technical Specification 3.12.3 states "During the period when the reactor coolant system is pressurized above 300 psig, and is above 200 F, and fuel is in the core, the reactor building polar crane hoists shall not be operated over the steam generator compartments."
The licensee stated that documentation concerning crane operation was not maintained.
That the limitation of TS 3.12.3 was included in the Polar Crane operating procedure (RP 1507.1) and in any Job Order which required use of the crane.
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The inspector interviewed several cognizant licensee personnel to determine the operating history of the Polar Crane during October
,3 November and December 1975.
The inspector determined that based
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on the interviews the polar crane was not operated when the reactor
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coolant system was above 300 psig and 200 F during this period of time.
i The inspector expressed concern that without specific crane operating records or logs the documentation and verification of compliance with TS 3.12.3 was not possible.
The licensee stated that this item would be reviewed.
This item is unresolved.
23.
Limits on Reactor Building Pressure While Critical Technical Specification 3.6.4 states "The reactor shall not be critical when the reactor building internal pressure exceeds 2.0 psig or 1.0 psi vacuum." The inspector reviewed SP 1301-1 " Shift and Daily Checks" for October, November and December, 1975.
The inspector verified that the reactor building pressure was recorded every shif t and it did not approach the limits specified in TS 3.6.4.
The inspector also reviewed OP 1102-2 " Plant Startup Procedure" and verified that the procedure " Precritical Checklist" (Irem 23) required verification of Reactor Building pressure within TS limit prior to criticality.
The inspector found no inadequacies in this area.
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- 24. Maintenance and Design Modifications The inspector asked for the documentation concerning preventive l
maintenance of emergency core cooling systems during December 1975.
The licensee stated that all maine ance work is performed under control of a work request and that e.o work requests for preventive maintenance on emergency core cooling systems were knplemented during December 1975.
l The inspector asked for the QC inspection and test records for i
design changes implemented during November and December 1975 on j
the Containment Structure and Emergency Power Systems.
The licensee stated that no design modifications had been performed on these systems during November or December 1975.
The inspector had no further questions in this area.
25.
Reactor Coolan't Pump Combinations Versus Power Level Limit 1
)
Technical Specification 3.1.1.la limits reactor power levels when i
operating with less than four reactor coolant pumps in operation.
i I
The inspector determined by review of the Shift Foreman's Log, Control Room Operator's Log and SP 1031-1, " Shift and Daily Checks,"
data sheets for October November and December 1975 that, whenever the reactor was in pouer operation, all Reactor Coolant Pumps were in operation. Therefore during this time period reduced power i
level operation was not required to satisfy TS 3.1.1.a.
4 4
The inspector had no further questions on this item.
26.
Corrective Action For Previous 1v Identified Items of Noncompliance I
i Region I Inspection Report 50-289/75-21 and the licensea's response dated November 10, 1975, a.
EI 75-07 (1) Operations personnel were instructed by memorandum, dated November 21, 1975 on the details of the occurrence and the specific areas of interest contained in the Environmental Technical Specification.
(2) A memorandum dated August 28, 1975, instructed operation personnel on the correct monitor for delta T.
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i (3) The inspector reviewed the operating logs and noted that the logs contained indication of supervisory j
review on a daily basis.
b.
EI 75-04 2
(1)
The inspector reviewed the procedure revision which requires analysis results from a waste neutralizing tank release to be analyzed within eight hours of the commencement of the release.
(2) The inspector reviewed Chemistry Training Memo #1 dated August 20, 1975 and the signoffs by the chemistry technician which indicated the.r had read the memorandum.
The memorandum emphasnad the importance of immediately notifying the Shift Supervisor tpon recognition of a violation.
The licensee's corrective actions for this item of noncompliance O
vere fe==d te he cemg1ete.
The inevector hed eo ferther see tiees on this item.
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