ML20093C533
| ML20093C533 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna, 05000000 |
| Issue date: | 03/30/1984 |
| From: | Bacanskas V, Jerome Murphy FRANKLIN INSTITUTE |
| To: | Bettenhausen L, Mccabe E NRC |
| Shared Package | |
| ML20093C531 | List: |
| References | |
| CON-NRC-03-81-130, CON-NRC-3-81-130 TER-C5506-524, NUDOCS 8404040098 | |
| Download: ML20093C533 (26) | |
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TECHNICAL EVALUATION REPORT r
h( AUDIT OF SUSQUEHANNA UNIT 2 TECHNICAL SPECIFICATIO PENNSYLVANI A P0h'ER a LICHT COMPANY f SUSQUEHANNA STEAM ELECTRIC STATION UNIT 2 3
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J NRC DOCKET NO. 50-388 FRC PROJECT C5506 FRC ASSIGNMENT 25
- I NRC CONTR ACT NO. NRC-03-81-130 FRC TASK 5 2!.
C'?:2 re: .*y ranklin Research Center Autnor: V. ?. Eacanskas 20tn anc Race Streets J. A. Murphy chilaceichia. PA 19102 FRC Grcup Leacer. :i. Ah=ed Prepared for
, Nuclear Regulatory Commission i wasnington, D.C. 20555 Lead NRC Engineer: L. Bettenhausen E. McCabe i
March 30, 1984 I
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This ' evert was preparec as an accosint cf work sponscrec Dy an agency of the Unitec States C-evernment. Neither the Unitec States Government nor any agency tnereof, or any of their ern: oyees, makes any e.arranty. expresses or im::liec. or assumes any iegal liability or responsibility for any thirc cartv s use or the results of such use. cf any information. apea-rates. precuct or process cisc:ose.: in tros recert c' epresents tnat its use os such thirc carty wourc not infringe arrvately ownec rignts
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l Prp ared by: Approved by:
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@M' Principal huthor Croug, Leader Gepartment Directo[ ('/
' Date: / ' Date: , Date-31 r.h'd 3 t i e:iE1 - } r-- 8" L.
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i egy y Gts #W % Fra n,s..an Research tenter yg _ _
20:h and Race Streets. Phde. Pa.19103 (215) 4451000
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TER-C5506-524 8
i CONTENTS action Title . ' Pace 1 INTRODUCTION . . . . . . . .
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1.1 Purpose of Audit . . . . . . . . . . . 1 1.2 Generic Background. . . . . . . . . . . 1 1.3 Plant-specific Eackground . . . . . . . . . 1 2 EVALUATION . . . . . . . . . . . . . . 4 2.1 Primary Containment Isolation Valves . . . . . . 4 2.2 Drysell to Suppression Pool vacuu: Ereakers . . . . 6 2.2 Automatic Depressurization System Valves . . . . . 8 2.4 Secondary Containment Isolation Dampers . . . . . 9 2.5 Suppression Pool Volume . . . . . . . . . 11 2,. 6 Diesel Generator Day Tank Levels . . . . . . . 12 2.7 Other Observations. . . . . . . . . . . 13 I
3 CONetuSIONS. . . . . . . . . . . . . . 14 4 ?.IFEP2NCES . . . . . . . . . . . . . . 15 9
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t FOREWORD This Technical Evaluation Report was prepared by Franklin Research.' Center
{ under a contract with the U.S. Nuclear Regulatory Commission (Of fice of * - ',.
Nuclear Reactor Regulation, Division of Operating Reactors) for techni, cal essistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by '
he NRC.
F.r. Joseph A. Murphy contributed to the technical preparation of tnis r e per through a subcontract with Schneider Consulting Engineers. -
M.r. L. Briggs, NRC Region ! Inspector, acco.mpanied the FRC perscnnel dering performance of the audit.
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- l. INTRODUCTION L1 PURPOSE OF AUDIT The . objective of the audit was to assist the Nuclear Regulatory Com=ission
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!ggaC) in determining whether the selected plant technical specifications are
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compatible with the as-built safety-related systems, structures, and components -
hf Susquehanna Unit 2. This technical evaluation report documents tdue results o f that audit. .
l.2 GENERIC BACKGROUND During the low-power testing phases at Grand Gulf Unit 1, it was found hat discrepancies existed between the technical specifications and the final ,
i j safety analysis report (FSAR), the NRC saf ety evaluatien repor (SER), and the plant's as-built condition. Many of these discrepancies have been climinated y amendments to the low-power license and by changing the technical specifi-ations. In order to gain additional assurance that the Grand Gulf technical spscifications were in agreement with the safety evaluations and the as-built condition, comparative audits were performed.
1 As 'a result of the problems found at the Grand Gulf plant, the NRC decided
. to conduct similar audits at the LaSalle plant, Washington Nuclear Plant 2, and Susquehanna Unit 2 to provide assurance that the plant technical specifications are compatible with the as-built plant.
1.3 PLANT-SPECIFIC BACKGRCCID On March 6, 1984, Franklin Research Center (FRC) was requested to assist NRC in performing an audit at Susquehanna Unit 2 to ensure that the plant technical specifications for selected safety-related systems are compatible with the as-built safety-related systems, structures, and components of the plant. The audit was to establish that hardware, its operating characteris-ties, and/or other conditions of the as-buil safety-related systems, struc-tures, and compenents are compatible with the parameters, descriptions, or
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l TER-C5506-524 other information set forth in the selected technical specifications. The ,
following scope of work for the audit was developed by NRC and discussed with tth2 FRC auditors at the Region I office on March 7,1984:
j 1. Primary containment isolation system (PCIS) valves: A sample of 25 PCIS valves were to be selected. The as-built condition of the
- valves and the surveillance procedures for the PCIS valves were t'o.be . _ _
r.eviewed to provide assurance that the as-bui'lt condition reflected - '.:
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the plant technical specification descriptions. .
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- 2. Drywell to suppression chamber vacuum breakers: Two vacuum breakers f were to be sele,cted at random. .The as-built condition of the vacuum breakers was to be verified to be in accordance with the plant tech-nical specification descriptions, and the existence of adequate surveillance procedures to address plant technical specification testing requirements was to be verified.
- 3. Automatic depressurization system (ADS) valves: The as-built condi- '
tien of -he ADS valves and associated surveillance procedures were to be reviewed to assure that plant technical specification requirements were adequately addressed.
- 4. Secondary' containment ventilation system automatic isolation dampers:
A physical verification was to be performed to determine if secondary containment ventilation system supply dampers were required to be addressed in Technical Specification Table 3.6.5.2-1.
- 5. . Suppression pool volume: Because a discrepancy existed between the technical specification maximum water level an3 the PSAR maximum level for thy suppression pool, the audit tean was to identify the reason for the discrepancy.
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- 6. Diesel cenerator day tank: The audit team was to examine an installed day tank to verify that sufficient volume existed to comply with plant technical specification requirements and that adequate surveillance procedures existed to provide assurance that the technical specifica-tien volume could be maintained.
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This scope considered the following:
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- 1. EG&G comparison of technical specifications with the FSAR; findings discussed by telephone with NRR representatives
- 2. NRC Region II Inspection Report 50-416/84-06 for the Grand Gulf plant; findings discussed with responsible Section Chief
- 3. previous Susquehanna problems (e.g. , PCIS valves)
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TER-C5506-524
- 4. dominant BWR plant. risk contributors; Susquehanna probabilistic risk analysis results (preliminary) e 5. those technical specifications which will be verified during start-up program inspections.
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- 2. EVALUATION Tnis section pr esents an item-by-item evaluation of compatibility of the plant technical specifications with the as-built condition of the plant for the primary containment isolation system (PCIS) valves, drywell to suppression
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pool vacuum breakers, automatic depressurization system (ADS) valves, secondaty ' ,
containment isolation dampers, suppression pool volume, and diesei generator dry tank level indication. .
2.1 PRIMARY CONTAINV.ENT ISCLATION VALVES j 2.1.1 Scoce The task required review of the technical specifications, plant drawings, and the as-built condition of 25 PCIS valves to assure that:
- a. the as-built ecndition reflects the description contained in the technical specifications
- b. the technical specification testing requirements are adegaately I addressed by surveillance procedures 1
j c. the electrical sche =atic drawings indicate that the isolation signals
- noted on Table 3.6.3-3 of the technical specifications are applied to actuate the valves.
, 2.1.2 Discussion i
Twenty-seven PCIS valves were physically inspected. Appendix A contains a list of these valves and the nameplate data" recorded.
1 Cne following documents supplied by the Licensee were reviewed:
i o Piping and instrumentation drawings (P& ids) for the following systems:
r I residual heat removal 1
reactor water clean-up
- high pressure coolant injection nuclear boiler - main steam reactor recirculation i
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I TER-C5506-524 containment atmosphere control reactor core isolation cooling reactor building chilled water reactor building component cooling water.
o Surveillance Procedure S0-259-011, "18 Month Manual Initiation of
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Dryvell Cooling Automatic Isolation System", .. .
o Surveillance Procedure SI-283-523, "lB Month Logic System Fungtional Test of Main Steam Line Isolation-Closure, Half Scram Channels Al, A2, B1, and B2"
! o Surveillance Procedure SI-283-501, " Main Steam Line Isolation Logic System Functional Test" o Surveillance Procedure 50-249-005, " Residual Beat Removal (RHR)
Division I and II Quarterly Valve Exercising."
I In addition, the data obtained by the Licensee from the performance of Surveillance Procedure 50-249-005 were reviewed for compliance with plant
.echnical specifications.
- l 2.1.3 Observations
[ o The reviewed surveillance procedures are in agreement with the plant I
technical specification requirements.
o The valve stroke times for 7 of the 27 valves reviewed were verified from the data recorded in Surveillance Procedure S0-249-005 and were
. found to be within the plant technical specification limits.
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i o tio discrepancies were identified for the 27 POIS valves reviewed.
2.1.4 Discrepancies None.
I 2.1.5 Re commendations I
During review of the P& ids by the auditors, several valves were deter-I mined to be first isolation valves outside the primary containment penetra-tions, but they were not listed as containment isolation valves. Discussions with the Licensee revealed tnat exemptions for thre "alves had been requested Mh TER-C5506-524 in the FSAR [1] and were granted in the plant SER [2, p. 6-33) . Review of the plant surveillance procedures indicated that these valves were subject to r
essentially the same testing requirements as valves listed as PCIS valves in the technical specifications with the exception of local leak rate testing.
Local leak rate testing could not be performed because no isolation valye exists inside primary containment. With these valves excluded from the techt _
nical specification PCIS valve listing, there is concern that the' surveillance requirements could be significantly modified or eliminated by the Licensee.
To obviate this concern, it is recommended that the valves be added to the technical specification PCIS valve table with the leak rate testing exception noted.
I i 2.2 DRYaTL* TO SUPPRISSION POOL VACUUM SREAKERS 2.2.1 Scoce l The task required selection of two vacuum breakers to assure that they are tested to meet technical specification requirements and that their I I it;stallation agrees with the technical specification description.
l 2.2.2 . Discussion It was not possible to physically inspect the vacuum breakers installed batween the drywell area and the suppression p>ol because the wetwell air space had been inerted and current atmospheric samples (oxygen, nitrogen, and other gases) were not available to allow authorization for entry. Documenta-tion for the vacuum breakers was reviewed to assure that the actions required for proper " Vacuum breaker operation are being performed by the Licensee.
The following documents supplied by the Licensee were reviewed:
. o FSAR Section 6.2.1.1.3.2 o surveillance Procedure 50-259-002, " Operability Check of Suppression Char 6er Drywell Vacuum Relief Breaker Valves
- o Surveillance Procedure SM-259-002, "18 Month vacuu= Relief Breaker Valve Set Pressure Test" o Technical Manual for Suppression Pool Drywell Vacuu: Relief Br ea<er valves - 10M 166 ,
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I TER-C5506-524 o Surveillance Procedure for Containment Exit o Technical Specifications for the Suppression Cha=ber Drywell Vacuum
. Breaker Relief Valves.
In addition, Surveillance Procedure 50-259-002 had been performed on February 14, 1984, and the results of the completed procedure were rev[ewed .
for compliance with plant technical specification requirements. .: .
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2.2.3 Observations o The set pressure and set pressure tolerance for the vacuum breakers.
required by the technical specifications are verified by surveillance procedure SM-259-002 at 16-month intervals. Tne vendor technical I
manual (10M 166) data are also in agreement with technical specifi-cation requirements and Surveillance Procedures SM-259-002 and l S0-259-002 requirements. -
I o Operability testing performed under Surveillance Procedure S0-259-002 is in accordance with plant technical specifications and the vendor technical =anual instructions, i
I o verification that the covers for the vacuum breakers are in the proper position is established in the surveillance requirements contained in
- the containment exit procedure and is in accordance with technical
- specification requirements.
2.2.4 Discrepancies o The technical specifications do not contain any specific requirements j' for setting and calibration of the limit switches for the vacuum l breakers.
t o The vendor technical manual indicates that the limit switches should indicate that the valve is fully closed or fully open.
i o th'e surveillance procedures require verification that the limit i switches are operable and properly calibrated. However , the procedures do not contain any information on how to calibrate the limit switches or on what are considered acceptable data.
o The FSAR, on page 6.2-5, states:
"Each of the inboard vacuum breakers is connected to a common alarm which indicates when any valve is not fully closed. Each of the outboard vacuum breakers is connected to a common alarm which indicates when any valve is not fully closed. There is individual gA i u - -
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TER-C5506-524 vacuum breaker position indication in the main control room for each valve.
The normally closed switches are held open when the valve is fully closed. The switches have a hysteresis or differential travel of 0.025". The switch hysteresis is multiplied through the mechanical
' linkage so that when the valve is opening under differential pressure the disk of the inboard valve is 0.32" off the seat before the "not *
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fully closed" light comes on. The outboard . valve can be 0.2" of f 'thi.:
seat under similar conditions. When the valve is closing under %
dif ferential pressure or when the valve is opening or closing ty the actuator, the mechanical linkage assures that the "not fully closed" light is on unless the disk is on the seat. "
There are no similar requirements in the plant technical specifica-tions. The discrepancy was brought to the attention of the Licensee and the NRO resident inspectors.
2.2.5 Ree: mendations s
I The Licensee should determine the reason for the discrepancy between the i rechnical specifications and the FSAR regarding limit switch setting and alibration for the suppression chamber drywell vacuum relief breaker valves.
The surveillance procedures should be revised as required to ensure that the limit switches are properly calibrated in accordance with the FSAR description.
2.3 AUTOMATIC DEPRESSURIZATION SYSTEM VALVES
- 2. 3 .1 Scooe The task required verification that the installed condition of the ADS
, valves and the plant surveillance procedures reflect the plant technical
, specification requirements.
2.3.2 Discussion Two of the ADS valves were physically inspected for proper installation; nameplate data were recorded and are included in Appendix B.
! The following documents supplied by the Licensee were reviewed:
I l 1 o FSAR Section 6.3.2.2.2 o Flant Technical Specifications Sec.rien 3.4.5.1, Table 3.3.1-1
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- l o Surveillance Procedures 50-283-001, 50-283-002, SI-280-303, SI-283-321, and SI-283-322.
I 2.3.3 Cbservations
? o.. The installed condition of the valves is represented by the plant drawings.
- o The reviewed surveillance procedures establish testing requirements in -
accordance with the plant technical specification requirements.
2.3.4 Discrepancies -
None.
2.3.5 Re commendations -
None.
2.4 SECCCARY COWAINMEN"' ISOLATION DAMPERS
. 2.4.1 Scooe Tne task was to assure that Technical Specification Table 3.6.5.2-1
- contains supply dampers for the secondary containment ventilation if necessary.
l In addition, discussions at Region I indicated that the standby gas treatment systems (SGTS) dampers required specific review.
2.4.2 Discussion i
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Technical Specification Table 3.6.5.2-1 identifies those valves and dampers ,that are part of the secondary containment (reactor building) l l boundary. No dampers associated with the SGTS are listed on the referenced table. Discussions with the Licensee and review of the PEIDs revealed that the SGTS dampers open on a seconcary containment isolation and are not used j for isolation purposes.
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. A subsequent review of the normal EVAC supply and exhaust dampers revealed that many of the secondary containment isolation dampers were not
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included in the draf t Technical Specification Table 3.6.5.2-1 for Unit 2 or the issued Unit 1 Technical Specifications.
However, discussions with the NRC
' licensing Project Manager for Susquehanna Unit 2 revealed that a revised Table 3.6.5.2-1 had recently been issued in NUREG-1042 (Susquehanna Unit 2 Technical Specifica tions) .
A copy of this table was provided to the auditors for review.
The revised Table 3.6.5.2-1 addresses all secondary containme t'. .- - -
ventilation system isolation dampers identified on ,the flow diagrams. .
2.4.3 Observations -
o Tne revised Table 3.6.5.2-1 contained in NUREG-1042 adequately addresses all secondary ventilation system isolation dampers. -
o Tne SSTS dampers do not provide any secondary containment isolation f unction; isolation is provided by the normal HVAC dampers. ~
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2.4.4 Discrecancies Ncne.
2.4.5 Re co=cenda tions -
b'ith the present configuration of the SGTS, the supply dampers (i.e.,
crossover) between Zone I (Unit 1 reactor building) and the recirculation plenum are not required to provide a secondary containment isolation function in the event of a Unit 2 secondary containment ventilation system isolation.
Should additional modifications be made to the SGTS to return the system to its original design' (two-:ene operation on a secondary containment ventilation system isolation), the supply dampers to the recirculation rystem plenum for Zone I (Unit 1 reactor building) should be added to the Unit 2 Technical Specifications as secondary containment ventilation system isolation dampers; similarly, the tone II supply dampers to the recirculation system plenum should be included as secondary containment ventilation system isolation k
dampers to the Unit 1 Technical Specifications.
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l r TER-C5506-524 2.5- SUPPRESSION POOL VOLUME 2.5.1 Scooe The task was to identify the re.ason for a discrepancy in suppression chamber water volume between th'e' technical specifications and the PSAR.
2.5.2 Discussion -
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Table 6.2.1 in the FSAR lists the following information concerning the suppression chamber water volume. ' '
Minimum, ft3 122,410 Maximum, ft3 131,550 Pool depth (normal) , ft 23 i
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Section 3/4.6.2 of th? technical specifications, " Depr e ssur iza tion
[t Systems," states ,the following:
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"Suoeression Chamber i L* .
Limitina Condition for Doeration 3.6.2.1 The suppression chamber shall be OPERA 3LE with:
- a. The pool water:
- 1. Volume between 133,540 f t3 and 122,410 ft3, equivalent to a level between 24'0" and 22'0"."
The minimum volumes identified in the technical specifi ions and the FSAR are in agreement. The normal pool depth maintained by th s Licensee is 23 f t and is in agreement with the FSAR and .the technical specifications.
2.5.3 Observations o The Licensee stated that the suppression pool high-level alarm is set at 23 ft 9 in. (This level corresponds to a suppression pool volume of 132,14 7 f t3.) The Licensee also stated that when the worst-case instrument error is considered, the high-level alarm point (23 ft 9 in) ensures that the maximum pool level of 24 ft, corresponding to a volume of 133,540 ft3, is not exceeded.
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TER-C5506-524 2.5.4 Discrepancies The information available from the Licensee does not provide a technical explanation for the discrepancy in the maximum suppression pool water volume identified in the FSAR and the technical specifications. (The technical cpecification value is 1990 f t greater than the PSAR.)
- 2.5.5 Recommendations .-
The Licensee should determine the reason for-the discrepancy in maximum suppression pool water volume and whether this discrepancy has any effect on suppression chamber performance. If the value in the PSAR is correct and the value in the technical specifications requires revision to confor= to the FSAR, the high-level alarm setpoint will require readjustment to a lower '
value. Further, if this instrument error argument is valid, the low-level clarm setpoint should also be reevaluated.
2.6 DIESEL GENE'RATOR DAY TANK LEVELS 2.6.1 Scooe T.$e scope of work included verification of the diesel generator day tank volume and surveillance requirements to assure that a minimum acceptable level j is maintained in th,e day tank.
l 2.6.2 Discussion l
Tne "A" diesel generator day tank was inspected to ensure that a method for verification of day tank level is provided and surveillance procedures are established.
2.6.2 Observations
, o The diesel generator day tank is provided with level instrumentation l to indicate the tank level and alarm at the low-level setpoint. plant i
surveillance procedures require verification of the tank level in accordance with technical specification requirements.
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1 TER-C5506-524 2.6.3 Discreoancies x
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2.7 ' OTER OBSERVATIONS -*
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9 During the physical inspection of the PCIS valves, several observations
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concerning the as-installed condition of the valves were made. They are as
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'I o Namco limit switches: Several Namco limit switches located inside ,
primary containment were identified as models environmentally qualified for outside primary containment use only. Discussions with the I
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y Licensee revealed that replacement of these limit switches was in progress. In addition, Namco limit switches for use inside contain-ment were environmentally qualified with the electrical connection
- provided with a sealant. The limit switches observed at the plant used a standard strain-relief connection with no sealant evident.
o Junction boxes: The junction boxes inspected inside primary contain-ment did not have pressure equalization (weep) heles necessary to ensure post-accident environmental qualification for high pressure.
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- 3. CONCLUSIONS ,
The audit confirmed that a discrepancy exists between the maximum allowable suppression chamber volume in the FSAR and plant technical specifications. The audit also revealed a lack of quantified acceptan,ce criteria in calibration procedures for the suppression chamber to drywel'1 ." ~ .
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vacuum breakers. No further discrepancies were identified anong the FSAR, the '
technical specifications, and the as-built conditions for the equipment evaluated. --
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- 4. REFERENCES
- 1. Susquehanna Steam Electric Station Units 1 and 2 Final Safety Analysis Report, Pennsylvania Power and Light Company September 1978
- 2. hUREG-0776, including Supplements 1 through 5, " Safety Evaluation,Repopt -
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related to the operation of Susquehanna Steam Electric Station, Units 1:
and 2,* USNRC, April 1981 '
3.
NUREG-0831, " Technical Specifications, Susquehanna Steam Electric Station, Unit 2,* USNRC, September 1983 C
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APPENDIX A
- r PRIMARY CONTAINMENT ISOLATION VALVES REVIEWED .
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l The information contained within this appendix was compiled from the 1 I
controlled as-built piping and instrumentation diagrams at the Susquehanna Unit 2 site and from the nameplates on the primary containment isolation vcivca.
Where information is noted as "not accessible," physical obstructions (insulation, etc.) prevented the' recording of data.
valva Number: EV-241F028B Function: Main Steam Isolation valve Type:
Atwood & Morril 26-in globe valve with electrohydraulic actuator Location: Penetration X-7B inboard; Line No. --
Purchase Order: 8856M1; Specification No. 21A9257, Rev. 3 Sarial Number: SNil221 Acce:scries: Solenoid valves, Gould Allied Control Serial No. SV24123C, Cl, C2 .
Limit switch, Namco, EA740, EA700 Valve Number: HV-241F022C Function: Main Steam Isolation Valve Type: Atwood & Morril 26-in globe valve with electrohydraulic actuator Loca tion: Penetration X-7C inboard; Line No. 26-G001 -
Purchase Order: 8656M1, Specification No. 21A9257, Rev. 3 S2 rial Number: Not accessible Acesseories: Solenoid valves, Gould Allied Control Limit switches, Namco, EA740, EA700 Valve Nucher: EV-241F016 Punction: Main Steam Line Drain Valve Type: 3-in gate valve with Limitorque motor operator Location: Penetration X-83 inboard; Line No. 3-DBA-208 Purchase Order: Not accessible S2 rial Number: Not accessible Accessories: Limitorque motor operator, size SMB-00 Serial No. 21657
, . Reliance electric motor, Class RH insulation t-Z_b .
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Valve Number: HV-255F002 Function: HPCI Steam Supply +
Type: 10-in gate valve Location: Penetration X-ll inboard; Line No.10-DBA-202 Purchase Order: Not accessible Serial Nurber: Not accessible Accessories: Limitorque motor operator, Size SMB-1 Serial No. 218058 Reliance electric motor, Class RH insulation Namco EA170 limit switches (2) _
Valve Number: HV-255F100. -,'
Punction: HPCI Steam Supply '
Type: 8-in Masoneilan globe valve with pneumatic actuator - '
Location: Penetration X-ll inboard; Line No. S-DBA-202 Purchase Order: 8856-J065BAC
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, Serial Number: N00186-5-2 Accessories: Air operator, Masoneilan Model No. 38-2076L-9-9
! Solenoid valve, Asco NPKX8321 ale Limit switches (2) , Namco EA180 Valve Number: NV-255F006 "
Functica: HPCI Injection Type: 14-in Anchor Darling gate valve with Limitorque motor operator Location: Outboard; Line No.14-DSS-220 Purchase Order: 8856-P-10A Serial Number: E-5853-49-1 Accessories: Limitorque motor operator, Size SMS-3 Serial No. 343670 Reliance electric motor, Class H insulation Valve Number: EV-255F04 2 Function: HPCI Suction Type: 4-in Anchor Darling with Limitorque motor operator Location: Penetration X-94 outboard Purchase Order: Not accessible -
Serial Number: E-5853-70-1 Accessories: Limitorque motor operator, Size SMB-000
, Reliance electric motor, Class B insulation
, Valve Number: HV-251F009 Function: RHR Shutdown Cooling Suction Type: 20-in globe valve with Limitorque motor operator Location: Penetration X-12 inboard; Line No. 20-DCA-208 Purchase Order: 8856-P-17A Serial Number: Not accessible
. Accessories: Limitorque motor operator, Size SMB-1 Reliance electric motor, Class RH insulation
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Valve Number:
b HV-241F104 I
Function: RWCU Return h Type: 4-in gate valve with Limitorque motor operator j Location: Penetration X-94 outboard Purchase Order: Not accessible Serial Number: Not accessible
' Accessories: Limitorque motor operator Reliance electric motor, Class B insulation Valve Number: HV-244F004 ,
Function: RWCU Suction , , . _ _
Zrpe: 6-in gate valve with Limitorque motor operator -: _
Location: Penetration X-14 outboard; Line No. 6-DBC-201 - '
Purchase Order: Not accessible
- Serial Number: Not accessible Accessories: Limitorque motor operator, Size"SMB-00 Serial No. 213611 Reliance electric motor, Class RH insulation Limit switch, Namco EA170 Valve Number: HV-244F001 Function: RWCU Suction Type : 6-in gate valve with Limitorque motor operator Location: Penetration X-14 outboard: Line No. 6-DBC-201 Purchase Order: 8856-P-10A Serial Numoer: Not accessible Accessories: Limitorque motor operator, Sire SMS-00 Serial No. 213453 Peliance electric motor, Class H insulation -
Limit switch, Namco EA180 Valve, Number: SV-22605 Function: Containment. lnstrument Gas Type: 2-in Target Rock Model 75KK-285 solenoid operated globe valve Location: Penetration X-80C outboard; Line No. 2-HCB-221 Purchase Order: 8856-J-70 Serial Number: SV12605 Accessories: None Valve Number: SV-25752B Function: Containment Atmosphere Sample Type: ,, 2-in Target Rock Model 75KK-211 solenoid operated globe valve Location: Penetration X-80C outboard Purchase Order: 8856-J-70 Serial Number: Not accessible Accessories: None M_ _ . , .
A-4'
j' .
-Valve. Number: HV-21345
- Function: Reactor Building Component Cooling Water Type 4-in gate valve with Limitorque motor operator Location: Penetration X-24 inboard; Line No. 4-HBD-230 Purchase Order: 8856-P-12-A Serial Number: E9052-1-3 Accessories: Limitorque motor . operator, Size SMB-00 Peerless electric motor, Class H insulation valve Nu-ber: EV-25713 -
Function: Containment Purge
- Type: 24-in Henry Pratt butterfly valve,with pneumatic actuitor .: "" ,
~
Location: Penetration X-26 outboard; Line No. 24-MBB-217.- '
Purchase Order: 8856-P-31-AC .~
Serial Number: D-0026-1-2 Accessories: Pneumatic operator , Bettis Modei~ 1416-SR-3-M3 Solenoid valve, Circle Seal Controls Model SN-315-9101-3-B Limit switch (2), Namco EA740 Valve Number: EV-243F019 Fanction: Reactor Coolant Sample
. 7ype: 3/4-in Masoneilan globe valve with pneumatic operator L: :a tion: Pene ration X-60S inboard; Line No. 3/4-DCA-243 Purchase Order: 8856-J65-BAC Serial Nammer: N00186-14-2 Accessories: Pneumatic operator , Masoneilan Model 38-200761-9-9 Solenoid valve, Asco NPKX8321A1E Limit switch (2), Namco EA180 Valve Number: EV-243F020 Function: Reactor Coolant Sample Type: 3/4-in Masoneilan globe valve with pneumatic operator Location: Penetration.X-60B outboard; Line No. 3/4-DCA-243 Purchase Order: 8856-J-69B-AC Serial Number: N00186-15-2 Accessories: Pneumatic operator , Masoneilan Model 38-20 Solenoid valve, Circle Seal Controls Model SN-315-9101-1-B Limit switch (2), Namco EAlB0 d
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t APPENDIX B r .
AU'IOMATIC DEPRESSURIZATION SYSTEM VALVES ' ,
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- The following information was obtained from the nameplate data from two ,
of the inboard automatic depressurization system valves. It should be noted
- that the valve tag numbers could not be located.
Valve 1: Crosby Electromatic Actuated Relief Valve Direct Acting Safety Relief ASME III Class l' ~
Body and Bonnet ASME SA-105 * ' ._ _
Inlet Hydrostatic Pressure, 2370 psig .
Outlet Hydrostatic Pressure, 975 psig -
Drawing No. 08A63790 ..
Solenoid valve Serial Nos. S66274-279, S66274-273, S66274-272 Valve 2: Crosby Electromatic Actuated Relief Valve Direct Acting Safety Relief ASME III Class 1 Body and Bonnet ASME SA-105 Inlet Hydrostatic Pressure, 2370 psig Outlet Hydrostatic Pressure, 975 psig .
Crosby Tag No. h*i65BP-lN GE Specification No. GE 22A6441 Solenoid valve Serial Nos. 366274-289, S66274-305, S66274-285
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P.O. BOX 1625,loAHO FALLS,loAHO 83415 March 27,1984 Mr. F. L. Sims, Director Reactor Research and Technology Division Idaho Operations Office - DOE .
Idaho Falls, ID 83401 TRANSMITTAL OF SUSQUEHANNA, UNIT 2, REPORT A6816 - LPL-109-84 Ref: J. M. Fehringer and J. C. Stachew, Audit of Nuclear Plant Technical Specifications Susquehanna Steam Electric Station, Unit 2, Docket No. 50-388, EGG-EA-6541, March 1984
~~
Dear Mr. Sims:
Enclosed is the referenced final report. This report determined that there are inconsistencies between eight Technical Specification Sections, the Final Safety Analysis Report and the Safety Evaluation Report for Susque-hanna Steam Electric Station, Unit 2. This report issued under FIN A6816 completes Node 106-D1 on the FY1984 NRC Support Milestone Chart.
Very truly yours,
---.-C/ L L. C'r i
L. P. Leach, Manager Reactor Evaluation Programs JMF:jh
Enclosure:
As Stated cc: J. N. Donohew, NRC/DL (5)
G. C. Meyer, NRC/DL J. O. Zane, EG&G Idaho (w/o Enc.)
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. . 3 EGG-EA-6541 March 1984 AUDIT OF NUCLEAR PLANT TECHNICAL SPECIFICATI0t:S SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2
. DOCKET NO. 50-388 J. M. Fehringer J. C. Stachew
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This is an informal report intended for use as a preliminary or working document
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Prepared for the Q U.S. NUCLEAR REGULATORY COMISSION M EGE6'd ha DOE ContractM-AC07_-ID0157L __ n. v, w, 21.. a;ww., e.<mw heiArE%i .
1 TABLE I. (Continued)
SECTION CONSISTENT / INCONSISTENT 3/4.6.5 SECONDARY CONTAINMENT Secondary Containment Automatic Inconsistent Isolation Dampers -
Standby Gas Treatment System '
Consistent 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Drywell and Suppression Chamber Consistent Hydrogen Recombiner Systems Drywell and Air Flow Systems Consistent Drywell and Suppression Chamber Consistent Oxygen Concentration 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. Sources-Operating Inconsistent 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS Distribution - Operating Inconsistent i D.C. Sources - Operating Inconsistent Primary Containment Penetration Inconsistent Conductor Overcurrent Protective Devices 1
'e a
6
The FSAR does not identify specific designations for the 480VAC buses, the 125VDC and 250VDC fuse boxes. Therefore, the completeness of the T/S 3.8.3.1 cannot be verified.
- 8. T/S Section 3/4.8.4.1 (Primary Containment Penetration Conductor .
Overcurrent Protective Devices)
T/S Table 3.8.4.1-1 (Primary Containment Penetrition Conductor Ov4rcurrent Protective Devices) identifies the overcurrent protective devices required to determine electrical equipment operability.
The FSAR does not identify any of the overcurrent protective devices listed in T/S Table 3.8.4.1-1. Therefore, the completeness of T/S Table 3.8./4.1-1 cannot be verified.
Table I contains a summary of the Susquehanna-2 T/S sections reviewed; consistencies and inconsistencies with the FSAR and/or the SER are shown.
1 i
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- 3. DISCUSSION The following inconsistencies were identified:
- 1. T/S Section 3/4.3.2 (Isolation Actuation Instrumentation)
The completness of T/S Table 3/4.3.3.2 (Isolation Actuation ._ _ .
Instrumentation) cannot be verified by the FSAR Table 7.3-5 (Containment and Reactor Vessel Control System Instrumentation Specifications).. A total listing / discussion of all instrument channels identified in T/S Table 3/4.3.3.2 are not addressed in FSAR Table 7.3-5. .
- 2. T/S Section 3/4.6.2.1 (Suppression Chamber)
The FSAR Section 6.2 page 6.2.1-92 identifies a maximum allowable water volume of 131,550 ft3in the suppression chamber. The T/S Limitng Conditions for Operation (LCO) 3.6.2.1 identifies a maximum allowable water volume of 133,540 ft3in the suppression chamber.
- 3. T/S Section 3/4.6.3 (Primary Containment Isolation Valves)
T/S Table 3.6.3-1 (Primary Containment Isolation Valves), identifies isolation valve data (isolation timing and input signals) that cannot be matched with the isolation valve data in the FSAR Table 6.2-12 (Primary Containment Isolation Valve Summary). There is no correlation between the valve designations identified in 'he FSAR and
, in the T/S.
I
- 4. T/S Section 3/4.6.5.2 (Secondary Containment Automatic Isolation
! ;. Dampers) i l
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ABSTRACT This report documents the review of the Susquehanna Steam Electric Station Unit 2 (Susquehanna-2) Technical Specifications (T/S) to determine if selected sections of the T/S are consistent with the Susquehanna Final Safety Analysis Report (FSAR) as amended and the Susquehanna Safety Evaluation Report (SER) as supplemented. Inconsistencies are listed in _ _
this report but no further ev'aluation was conducted to determine if the inconsistency was an indication of an error in any of the subject documents.
FOREWARD This report is supplied as part of the " Audit of Nuclear Plant Technical Specifications" being conducted for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Licensing, by EG&G Idaho, Inc., NRC Licensing Support Section.
The U.S. Nuclear Regulatory Commission funded the work under authorization B&R 20 19 10 11 1 FIN No. A6816.
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lL.)fiQ Pennsylvania Powei & Light Company O- Two Nonh Ninth Street
- Allentown. PA 18101 + 215 s 770 5151
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!'!M*.1. 1.. -y:h*4 Nc* man W. Curtis V;:e President Engineering & Construction Nuclear 215mC-7501 Mr. Harold R. Denton, Director -
Office cf Nuclear Reactor Regulation '
U.S. Nuclear Regulatory Co= mission * -
Vashington, D.C. 20555 ,
SUSC:UEEANNA STEAli II.ICTRIC STATION CERTIFICATION OF I.7IT 2 TICENICAI. SPECIFICATIONS ER 100508 FILE 841-8
?LA-211L Decke: No. 50-3c5
- :.:: Mr. Den:en:
In respense to Mr. Eisenhu:'s letter dated March 3, ice , a::a:hed are Fennsylrania Fever & l.igh: Ccepany's prepesed 7:1: : Ta:. ni:a* Spa :1fic::1 ens , .
f An amend:en: to our license applica: ion has been subri::ed which revises
-( See:icn 16.2 of the Final Safety Analysis Report (FSAR)- to reference this N- letter as containing our preposed Technical Specificatiens fer Susquehanna SIS Uni: 2. -
We have reviewed the draft Unit 2 Technical Specificatient, provided to us on Februarv 17, 1984 including revisions received from :he NRC staff on March 10 and 20, 1984 Based on that review, I certify : hat, to :he best of my kncviedge, the Technical Specifications for Suscuehanna SES Uni: 2 as proposed in this letter
- accurately reflect the plant, the FSAR and supplementary correspendence, and the SIR analysis with the exception of the lack of an es:ablished limit on the =easurement of secondary con:ain=ent bypass leakage
- hrough the feedvater penetration.
The lack cf this li it is not a safety concern since we are ceasuring the leakage and keeping :he :etal leakage frem this seurce and the MS' Drains te vi:hin f.0 scfh, consistent vi:h our analysis. A change to :he a::isting Susque'nanna SES Unit 1 Technical Specifications is in our internal review process. Prior to exceeding five percent power, we vill submit a request to .
revise the Technical Specifications for both units to incorporate this limit.
The operating license for Suscuehanna 1 was issued on July 17, ICS2, and the s:artup of this unit, in our,judgenent, was very successful resul:ing in this unit being the first BWR since the TMI incident :c achieve ce==ercial rera:ien. The enperience vi:h Uni 1 led to :he identifica:icn cf reiz:ively l
g minor clarifications of language that have been incorporated into the Unit 2 l > Technical Specifications. Additional changes fre: the Unit I docu=ent have
\- been made te reflec: the plan: configuration of :ve units, :e incerporate the offe:: ef NEC staff resolutier f indus: y gener.: ssues, and :e incerperate a nu ber cf =inor ad inis:ra:1te chances, ~ vh
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(5 30 !.1 ER 100450 File 841-8 P.r. Harold R. Denton Our overall as sessment is that the Unit 1 Technical Specifications are sour 3 and constitute a good basis for plant operation and =enitoring ce=pliance.
The changes ir.orporated into the Unit 2 Technical Specifications, beyond those requirec co reflect system configuration, are relatively minor, but should contribute to elimination of misinterpretation in carrying out surveillances and operating this unit. Prior to exceeding 5% powcr on Un,it A -
PP&L expects to request changes in the Unit 1 specifications to mak2 them. -
comparable to Unit 2.
If you have any co=ments or questions please contact us.
Very truly yours,
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N. ~.'. Curtis Vice President-Engineering & Cctstruction-Suclear '
cc: K. L. Perch - SEC R. E. Jacebs - NRC
- t. R. Hoff=an - NRC ~
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Tvin: QO-Docket No. 50-374 Commonwealth Edison Company ATTN: Mr. Cordell Reed ~~ ~'
Vice President .
Post Office Box 767 Chicago, IL 60690 Gentlemen:
This refers to the special inspection conducted by Messrs. A. L. Madison, S. Stasek, S. Guthrie and D. Evans of this office on March 6 through 9, 1984, of activities at LaSalle County Station, Unit 2, authorized *by NRC Operating License NPF-18, and to the discussion of ,our findings with Mr. R. D. Bishop at the conclusion of the inspection.
The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel.
No items of noncompli'ance with NRC requirements were identified during the course of this inspection.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure (s) will be placed in the NRC Public Document Room unless you notify this' office, by telephone, within ten days of the date of this letter and submit written application to withhold information contained therein within thirty days of the date of this letter. Such application must be consistent with the re-quirements of 2.790(b)(1). If we do not hear from you in this regard within the specified periods noted above, a copy of this letter and the enclosed inspection report will be placed in the Public Document Room.
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We will gladly discuss any questions you have concerning this inspection.
Sincerely, Ndfigis,37 Sirr ' .* .its"
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C. E. Norelius, Director -- --
Division of Project and Resident Programs
Enclosure:
Inspection Report -
No. 50-373/84-07(DPRP) cc w/ encl:
D. L. Farrar, Director ,
of Nuclear Licensing G. J. Diederich, Station Superintendent R. H. Holyoak, Project Manager DMB/ Document Control Desk (RIDS)
Resident Inspector, RIII Phyllis Dunton, Attorney General's Office, Environmental Control Division l
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lius Walker /1d Raf 03/16/84 V n/
. . 1 U. S. NUCLEAR REGULATORY COMMISSION -
REGION III Report No. 50-374/84-07(DPRP)
Docket No. 50-374 License No. NPF-18 Licensee: Commonwealth Edison Company Post Office Box 767 ,
Chicago, IL 60690 Facility Name: LaSalle County Nuclear Station, Unit 2 ,.
Inspection At: LaSalle Site, Marseilles, Illinois Inspection Conducted: March 6 through 9, 1984 t<.,. I s p- , ", j Inspectors: X. Madison Date M n s t&&L /k ~-
- ~M S. St'asek
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({fg f!k.t S. Guthrie # *W
, Date c-M P " . 4 ,{ l ' , y
- 7. xc_s 2 E. Evans Date A 1 - . , e Approved By: R. D. Walker, Chief 7' ' ~7
Projects Section 2C Date Inspection Summary .
Inspection on March 6-9, 1984 (Report No. 50-374/84-07(DPRP))
Areas Inspected: Special unannounced safety inspection to verify Technical Specification conformance to as-built plant configuration; review surveillance procedures and surveillance program implementation. The inspection involved a total of 80 inspector-hours by four inspectors.
Results: Of the two areas inspected, no items of noncompliance or deviations were identified.
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- DETAILS *
- 1. Persons Contacted R. D. Bishop, Administrative and Support Services Assistant Superintendent J. C. Renwick, Technical Staff Supervisor The inspectors also talked with and interviewed various members of
"" ~~
the Operations and Technical Staff. .
- 2. Technical Specification (TS) Review At the request of the Office of Nuclear Reictor Regulation (NRR),
Region III assigned four inspectors to review two sections of the Unit 2 Technical Specificaticos for technical adequacy and conformance to actual plant design:
- a. Section 3/4.6.3 Primary Containment Isolation Valves
- b. Section 3/4.8.2 Electrical Distribution The inrpectors reviewed the applicable sections of the Final Safety Analysis Report (FSAR) to ensure Technical Specification conformance.
The inspectors reviewed as-built drawings and performed in plant walk-downs to' verify that equipment in place matched ~that described ic the Technical Specifications. The inspectors reviewed the Technical Specifications action statements for technical adequacy including verifying adequate electrical power for performance of all Emergency Core Cooling Systems (ECCS). The inspectors also reviewed the licensee's surveillance program to ensure compliance with Technical Specifi* cation requirements.
(1) Section 3/4.6.3 Primary Containment Isolation Valves The inspectors' review of this section revealed two minor discrepancies:
(a) Table 3.6.3-1 lists the primary containment isolation valves.
However, under Automatic Isolation Valves, a. 14., only one Tip Guide Tube Valve Ball Valve; 2C51-J004 is listed. There are actually five valves; 2C51-J004A, B, C, D and E.
The licensee's surveillance program and" procedures recognize the existence.of these five valves and the required testing and surveillance 'has been performed.
(b) Table 6.2-21 in the FSAR requires a valve closure time of 140 see for valves 2E12-F008 and 2E12-F009 (RHR shutdown cooling suction). However, T.S. Table 3.6.3-1 lists 141 secs for valve closure time and the licensee's surveillance procedures comply with the Technical Specifications.
2
- j. .. ; .
y The actual closure times as verified by recent testing is <35 -
see for these valves. *
. Resolution and correction of these apparent discrepancies will be tracked as an unresolved item (374/84-07-01). No licensee action is required at this time.
During this review the inspectors also found several discrepancies in the licensee's surveillance mat'rix and procedures related to this Technical Specification section. These are discussed in Part 3 of this report. ,
(2) Section 3/4.8.2 Electrical Distribution This section describes requirements ihr A.C. and D.C. electrical distribution for both operating and shutdown conditions. No discrepancies in the technical specifications were found.
However, the inspectors did find five deficiencies in the licensee's electrical drawings and labeling of breakers.
The licensee has committed to correcting these deficiencies and their action will be tracked as an open item (374/84-07-02).
Space A-2, MCC 235X-1 (2AP71E) and Space AA-4, MCC 235X-2 (2AP72E) appeared to have been additions to the motor control centers.
Further review revealed no electrical loading concerns; however, -
the inspector questioned the effect these additions had on the seismic qualifications of the affected motor control centers. The licensee has committed to provide an analysis concerning the seismic qualification of those motor control centers performed by Sargent and Lundy, the architect-engineer. Resolution of this condern will be tracked as an open item (374/84-07-03).
- 3. Surveillance Matrix and Program i-l As part of the Technical Specification review, the inspectors reviewed portions of the licensee's surveillance matrix that were applicable to the sections of the Technical Specifications under review. The matrix is designed such that a direct written correlation between Technical Specification required surveillance, specific components and specific procedures should exist. The purpose of the review was to ensure th'at all components and surveillance for those components addressed in the Technical Specifications were listed in the matrix and that a procedure to perform that surveillance for each component existed. The inspectors also reviewed some procedures on a spot check basis to ensure that the procedure actually performed the required surveillance for the specific component and that the procedures were technically adequate. -
- a. Technical Specification 4.6.3.1 requires that each valve listed in Table 3.6.3-1 be tested to verify full travel and operability following maintenance. However, the matrix does not list an '
applicable procedure to fulfill this tequirement for several valves.
3 i_ . _ _ ..__ ____ _ __ _ _
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Whether or not the required surveillance is actually being performed could not be determiBed without further information 4
from the licensee. The licensee has agreed to provide the required information. Resolution of this matter will be tracked as an open item (374/84-07-04).
- b. Valve 2E51-F069 was not listed in the matrix. However, further investigation confirmed that procedures existed to perform the
~
required surveillance. ,
The matrix referred to LIS-NB-15 and 16,for the required sur-veillance on Excess Flow Check Valves, whereas the procedures were actually LIS-NB-115, 215,116 and 216.
~
These and other minor discrepancies w'ere noted and referred to the licensee, They will be corrected as part of an ongoing review by the surveillance group. This surveillance group was established February 1,1982 and is charged with the responsibility of coordinating surveillance at LaSalle Station. The inspectors feel that this is a positive step and will enhance the licensee's performance in the surveillance area.
- c. In the review of procedures it was noted that calibrated stopwatches were not required to perform closure time measurements. However, further investigation revealed that calibration stopwatches were l
actually being used. ANSI 18.7 (1976) Administrative Controls and Quality Assurance requires that procedures for- tests -and maintenance-i specify any special equipment to be used. The licensee has agreed to revise applicable surveillance procedures to require a calibrated stopwatch for measuring valve closure times. Completion of this will be tracked as an open item (374/34-07-05). ,
- d. The inspectors also reviewed maintenance work requests to ensure that required surveillances were being performed following valve maint'e nance. No violations of requirements were identified.
However, a potential source of confusion was identified in that the specific test requirement was not noted on the work request in all cases. Identifying the specific test requirements on the work request not only ensures that the desired tests are performed, but also allows Quality Control, Quality Assurance, and othe'r reviewers the opportunity to verify that Technical Specification requirements are met. The licensee agreed that specific test requirements should be listed on the work request.
No items of noncompliance were identified.
- 4. Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection and summarized the scope and findings of the inspection activities. The licensee acknowledged those findings.
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- - _ . _ . _ . ._. _ . . . . ... .. .a .. _.
,s C:mmonwrith Edison .
1 One First Natenal Plata. CNcago lis. noes Accres: Reply to. Post Othee Box 767 l CNcago. litinois 60690 ,
Ma rch 21, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
LaSalle County Station Unit 2
- Technical Specification Certification NRC Docket No. 50-374 Reference (a): D. G. Eisenhut letter to Cordell Reed dated March 8, 1984.
(b): C.W. Schroeder letter to H. R. Denton dated January 13, 1984.
Dear Mr. Denton:
The purpose of this letter is to respond to Reference (a).
The interaction of LaSalle County Station personnel with the Standard Technical Specifications dates back to approximately 1974 when G. J. Diederich, then Assistant Superintendent for Operations, was a member of the BWR Standard Technical Specifications Committee. Mr.
Diederich, who is now Station Superintendent at LaSalle County Station, thus gained first hand knowledge of the development philosophy, and NRC staff positions as they were incorporated into the original BWR Standard Technical Specifications ~(STS).
In 1918, Commonwealth Edison Company prepared the original draft of the LaSalle County Station Technical Specifications. This preparation included reviews by individual system test engineers, departmental reviews and a series of meetings with the entire operating staff to review the tech specs in detail. Further reviews were performed on a chapter by chapter basis by the NSSS vendor (GE) and the A/E (Sargent and Lundy).
Following submittal of FSAR Chapter 16 (Amendment 39, October 1978), the NRC requested that future versions be submitted as marked up copies of
Two years prior to the Unit i license issue, the Tech Specs for Unit 1 were thoroughly reviewed by Commonwealth Edison Engineering, Station Staff, Nuclear Licensing, and Nuclear Safety for accuracy. These l reviews included providing each page and all subsequent changes to the applicable system test engineers and other " experts" for review and comment. These comments were reviewed and many discussions were held within the Company and with NRR (Messrs. Bottimore, Bournia and ~ ~ ,
reviewers).
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PDR ADOCK 05000374 0j l A PDR 9 r 6 i
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H. R. Denton - 2* - March 21, 1984 NRR issued many changes during this period (several dozen) to incorporate staff requirements, design changes and CECO requests. These changes also received multiple reviews by cognizant individuals.
h During the almost two full years sihce the Unit 1 License N#E-ll
( was issued, it has been our experience that the Unit 1 Technical Specifications accurately reflect the plant and the FSAR. Certain g
!g specifications were found to have minor discrepancies that were either 1 corrected by license amendments or were determined to be adequately l controlled and id'entified in the Unit 1 Technical Specification upgrade p to match Unit 2 (Reference b). The Unit 2 Tech Spec preparation started t
with the current Unit 1 Tech Spec at the time as the draft document and changes were made where differences existed. This wAs submitted to NRR as a draft. Additional changes were made and submitted in May, 1983 to D. Hoffman (NRR). These changes included improvements over Unit 1, clarifications, relaxations and new revised staff requirements where necessary. Such changes were held (at NRC request) for review and issuance at Unit 2 licensing, with the intention to then promptly backfit on Unit 1. The proof and review copy was received in August, 1983 and again was reviewed on site for accuracy by system test engineers and other " experts". Subsequently discussions were-held with the staffst reviewers including, a meeting at the Bethesda offices on September 20, 1983. Since the license condition identified in SSER Supplement 5 item 1.10(7)(1) on reactor containment electrical penetrations' redundant i
fault current devices was not issued due to installation of subject devices, a clarifying upgrade to the Unit 2 Technical Specification 3.8.3.2 will be submitted as an administrative change to note the backup i devices.
The status of Unit 1 and Unit 2 Technical Specifications has been discussed on several occassions between the NRC Staff and
.. Commonwealth Edison Company. During the Unit 2 operational readiness review meeting in Bethesda, Commonwealth Edison Company again stated our intention to upgrade the Unit 1 Technical Specifications to match the Unit 2 Technical Specifications. This-action.was agreed to by NRR management. On January 13, 1984, Reference (b) was submitted to fulfill our commitment. These changes were justified based on the fact that the NRC hdd'just issued the exact same specifications cn Unit 2 less than a month before (12/16/83). The NRC rejected this Unit 1 Technical Specification amendment request. Commonwealth Edison Company is in the process of reformatting our request and expect resubmittal in the near future.
It is.our understanding that the NRC Region III recently concluded an extensive onsite review of the Technical Specifications for Containment Isolation and AC/DC power. We understand that review, which will be documented in an inspection report to be issued in the near future, concluded that those specifications are technically adequate.
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H. R. Denten - 7* - March 21, 1984 Based upon the detailed, iterat'ive process utilized to prepare the Unit 1 Technical Specifications, the positive two year operating experience with the Unit 1 Technical' Specification.s, the use of the Unit 1 Technical Specifications as the basis for the Unit 2 Technical Specifica-tions, and the positive.three month experience since the operating license ~
~~
was issued with the Unit 2' Technical Specifications, I conclude and certify that the Unit 2 Technical' Specifications do accurately reflect the plant and the FSAR. Furthermore, I am satisfied that, because of
. these factors, no further adequacy reviews.are warranted by Commonwealth Edison Company at this time.
Certain issues as to the interpretations of specifications and overly restrictive action statements that have been previously identified by Commonwealth Edison Company, owners groups, and NRR generic letters will continue to be pursued. Commonwealth Edison Company is also participating in the BWR Owners Group Technical Specification Improve nents Committee and expects substantial changes in Techneial Specifications to result from that effort. Finally, we are endouraged by-the work that the NRC is initiating (NUREG-1024) to provide an overall upgrade of Technical Specifications.
To the best of my knowledge and belief the statements contained herein are true and correct. In some respects these statements are not based on my personal knowledge but upon information furnished by other Commonwealth Edison and contractor employees. Such information has been reviewed in accordance with Company pr&ctice and I believe it to be i reliable.
- Enclosed for your use are one signed original and thirty-nine (39) copies of this letter.
Very truly yours, C .' .c.% . '.O. \ i 6 2../-
Cordell Reed Vice President Im -
cc: Dr. A. Bournia - Telecopy NRC Resident Inspector - LSCS SUBSCRIBED and SWORN to befoRe me this #/a e day '
of % heA; , 1984
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EGG-EA-6539 March 1984 AUDIT OF NUCLEAR PLANT TECHNICAL SPECIFICATIONS LA SALLE COUNTY STATION, UNIT 2 o DOCKET NO. 50-374 t
- P J. M. Fehringer .
D. M. Beahm ,
Idaho National Engineering Laboratory Operated by the U.S. Department of Energy
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. 1 g, This is an informal report intended for use as a preliminary or working document
-s N o W W7thfIl0 6W Prepared for the
- p 2 'U.S. NUCLEAR REGULATORY COMISSION g;,g E G E G ia.no i ,Under DOE Contract No.-DE-AC07-ID01570 .
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EGG-EA-6539 LA SALLE COUNTY STATION, UNIT 2 AUDIT OF NUCLEAR PLANT TECHNICAL SPECIFICATIONS Docket No. 50-374 TAC No. 54184 .
Published March 1984 l
i J. M. Fehringer D. M. Beahm I
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l j EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Responsible NRC Individual and Division:
C. Meyer/ Division of Licensing
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l Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6816
e CONTENTS
- 1. INTRODUCTION ..................................................... I
- 2. REVIEW CRITERIA .................................................. 1
. 3. DISCUSSION ....................................................... 2
- 4. CONCLUSIONS ...................................................... 5
- 5. REFERENCES .................................'...................... 5 TABLE I. LaSa11e-2 Technical Specification /FSAR/SER Consistency Summary .............................................. 3 1
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AUDIT OF NUCLEAR PLANT TECHNICAL SPECIFICATIONS I. INTRODUCTION The LaSalle County Station, Unit 2 (LaSalle-2) is a boiling water reactor (BWR) plant. It has been selected for an audit to determine if the LaSalle Technical Specifications (T/S)1 , are consistent with the LaSalle Final Safety Analysis Report (FSAR)2 as amended, hnd the LaSalle Safety Evaluation Report (SER)3 as supplemented. The specific sections of the T/S selected for audit and summary results are listed in Table I.
Inconsistencies between these sections of the T/S and the FSAR and SER were identified but no further evaluation was conducted to determine if the inconsistencies were indications of error in any of the subject documents.
- 2. REVIEW. CRITERIA
+
The T/S Limiting Conditions for Operation (LCOs) and Action Statements for each technical specification listed in Table I (Section 3) were compared with the FSAR and SER to determine if the T/S are consistent to ,
the FSAR and SER. Emphasis was on the T/S Operational Mode 1, power operation, with exceptions noted in this report. Setpoints and Itsts of valves, instruments, overcurrent protective devices and electrical buses in the T/S were checked against tables in the FSAR and SER.
The SER was reviewed to ensure that requirements in the SER were addressed in the T/S.
The T/S bases and surveillance requirements were not reviewed in this audit cf the T/1.
An explanation of each inconsistency between the T/S and the FSAR and SER is included in this report.
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TABLE I. LASALLE-2 TECHNICAL SPECIFICATIONS /FSAR/SER CONSISTENCY
SUMMARY
l SECTION CONSISTENT / INCONSISTENT 3/4.3 INSTRUMENTATION 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION Consistent 3/4.3.3 EMERGENCY CORE COOLING SYSTEM Consistent ACTUATION INSTRUMENTATION 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING Consistent 3/4.5.3 SUPPRESSION CHAMBER Consistent 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity Consistent Primary Containment Leakage Consistent i
Primary Containment Air Locks Consistent MSIV Leakage Control System Consistent Primary Containment Structural Consistent Integrity Drywell and Suppression Chamber Consistent Internal Pressure Drywell and Suppression Chamber Consistent Purge System 3/4.6.2 DEPRESSURIZATION SYSTEMS l
, Suppression Chamber Consistent l Suppression Pool Spray Consistent Suppression Pool Cooling Consistent 3/4.6.3 CONTAINMENT ISOLATION VALVES Co'nsistent l
e 3
- 4. CONCLUSION As shown in Table I, 24 technical specification sections were compared with information in the FSAR and SER for LaSalle Unit 2. Inconsistencies were identified in two sections of the technical specifications shown in Table I. This review did not determine the significance of the ;
1
, inconsistency or which of. the documents was in error. l
- 5. REFERENCES -
- 1. LaSalle County Station, Unit 2, Technical Specifications Rev. December 1983
- 2. LaSalle County Station, Unit 2, FSAR up to Amendment No. 63
- 3. LaSalle Co':nty Station, Unit 2, SER up to Supplement No. 7 9
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NRCPomu 2 ' '
U.S. NUCLEAR RECULATORY COMMISCON j 095 413 BIBUOGRAPHIC DATA SHEET EGG-EA-6539
- 3. TITLE AND SUBTITLE 2. (Leere bee *J Audit of Nuclear Plant Technical Specifications LaSalle County Station, Unit 2 s. RECIPIENT S ACCESS oN NO.
- 7. AUTHORISI 5. DATE REPORT COMPLETED uomTw l vE AR
. March 1984
- 9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (lactuae Isa Coors DATE REPORT ISSUED uoNTM lYEAm e .
March 1984 EG8G Idaho, Inc. s. rt,,,, ,,,,
Idaho Falls, ID 83415
- 8. (Leave Nmkl
- 12. SPONSORING ORGAN 6Z ATION NAME AND MAILING ADDRESS Itacevor Isa Cocal P UI WTM Division of Licensino Office of Nuclear Reacto- Reculation 11. FIN NO.
U.S. Nuclear Regulatory Commission Washington, DC 20555 A6816
- 13. TYPE OF REPORT Pr esco Cove ag o tractusere asest Technical Evaluation Report (TER) February 13, 1984 to March 12, 1984
- 15. SUPPLEMENTARY NOTES 14. (Leave swm*1
. I6. ABSTR ACT C00.* ores or deu)
This report documents the review of the LaSalle County Station, Unit 2 (LaSalle-2)
Technical Specifications (T/S) to determine if selected sections of the T/S are consistent with the LaSalle-2 Final Safety Analysis Report (FSAR) as amended, and the LaSalle-2 Safety Evaluation Report (SER) as supplemented. Inconsistencies are listed in this report but no further evaluation was conducted to determine if the inconsis-tency was an indication of an error in any of the subject documents.
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- 17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS e
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17b IDENTIFIE RS.' OPE N-EN DE D TE RMS
- 18. AV AILABILITY STATEMENT 19. SECURITY CLASS (Ta s recorrs 21 NO OF P AGES Unclassified 22 Pa CE Unlimited 2a g g v C yq S$ gp a,,os.es s _
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UNITED STATES ,
- [gs* * * &,,, c,
- NUCLEAR' REGULATORY COMMISSION IV *"* *
- nEGioN ii
- ! . $. i 101 M ARIETTA STREET.N.W.
's,h. , f ATLANTA. GE oRGI A 30203 MAR 13 E Mississippi Power and Light Company
- ATTN: Mr. J. B. Richard l
. :* Sdnior Vice President, Nuclear P. O. Box 1640 '
Jackson, MS ,39205 G:ntlemen:
SUBJECT:
REPORT NO. 5.0-416/84-06 On February 21-24, 1984, NRC inspected activities authorized by NRC Operating License No. NPF-13 for your Grand Gulf facility. At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the enclosed inspection report.
l l Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress.
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Within the scope of the inspection, no violations or deviations were identified.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosures will be placed in NRC's Public Document Room unless you notify this office by j telephone within ten days of the date of this letter and submit written application to withhold information contained therein within thirty days of the date of the letter. Such application must be consistent with the requirements of 2.790(b)(1).
Should you have any questions concerning this letter, please contact us.
Sincerely, af l ca ~ _
vid M. errelli, Chief roject ranch 1 .
Division of Project and Resident Programs '
Enclosure:
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Inspection Report No. 50-416/84-06 7 cc w/ enc 1: , i 1 o A Jn i V J. E. Cross, Plant Manager [ M 4' 7 (#
- V W s ' AR_ , ,
Ralph T. .Lally, Manager of Quality l
Middle South Services, Inc.
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pa cah UNITED STATES .
/ o NUCLEAR REGULATORY COMMIS$10N REGION ii .
U-101 MARIETTA STREET, N.W.
g . . [E ATLANTA, G EORGIA 30303,
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Repo.rt No.: 50-416/84-06 Licensee: Mississippi Power and Light Company Jpckson, MS 39205 , , _ _
Docket No.: 50-416 License No.: NPF-13 1
Facility Name: Grand Gulf 1 -
1 Inspection at Grand Gulf site near Port Gibson, Mississippi Inspectors: C-S. Butler u sb b- '
'Date Signed
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Approved by: G '/<' / U 2 t^ '
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SUMMARY
Inspection on February 21-24, 1984 *
- Area's Ihspected This special announced inspection ir}volved 234 inspector-hours on site in the , _
area of verification of the accuracy 'of the Technical Specification.
Results e Of the areas inspected, no violations or deviations were identified.
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REPORT DETAILS l
- 1. Persons Contacted Licensee Employees
- J. E. Cross, Plant Manager
- R. F. Rogers, Assistrnt Plant Manager - Operations
- C. R. Hutchinson, Assistant Plant Manager -~ Maintenance "J. W. Yelverton, Assistant Plant Manager - Support
- J. C. Roberts, Technical Support Staff
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- F. M. Walch, Maintenance Superintendent
- G. A. Zinke, Technical Engineering Supervisor "L. F. Daughtery, Compliance Superintendent
- J. D. Bailey, Compliance Coordinator Other licensee employees contacted included numerous engineers, operators, mechanics, security force members, and office personnel.
Other Organizations "M. G. Farschon, General Electric. Site Operations Manager
- Attended exit interview
- 2. Exit Interview The inspection scope and findings were summarized on February 24, 1984, with those persons indicated in paragraph 1 above. The Technical Specifica-tion.(TS) discrepancies were described to plant management by the inspectors.
NRC representatives stated that the problems found are indicative of the need for another review of Technical Specifications to find and correct any errors.
- 3. Licensee Action on Previous Enforcement.. Matters Not inspected.
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- 4. Unresolved Items .
Unresolved items were not identified during this inspection.
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. 2 ,
- 5. Suppression Pool and Containment Spray The inspectors compared applicable sections of the Final Safety Analysis Report (FSAR), as-built drawings, surveillance, and operating procedures a'nd actual plant systems to Technical Sp'ecifications (TS) associated with the suppression pool and containment spray. The following are discrepancies that were identified: .
- a. FSAR 5ection 6.2.7.5 indicates that the suppression pool level indica-tion system is made up of four level detector channels, (two detector l channels per division). It also indicafes that each of these channels l provides a .high-water-level alarm, low-water-level alarm, low-low- !
water-level alarm, as well as a signal to open suppression pool makeup valves.
In actuality, there are three active level detector channels per division. Two channels are wide range and one channel is narrow range.
There is also one additional channel per division which is only used for indication at the remote shutdown panel. Each wide range channel supplies input to their respective division's suppression pool makeup system 'in one out of two logic as well as providing a low-low-level alarm at 16'10". The narrow range channel in each division provides the divisional low-level-and - high-level alarms -(18'53 " sand 18' 9" , - ;
i respectively). TS are written to conform to the FSAR but are not in clear agreement with the actual plant design.
- b. TS 3.5.3 (ECCS), 3.6.3.1 (Depressurization Systems), 3.3.7.5 (Accident Monitoring Instrumentation), and 3.6.3.4 (Suppression pooi Makeup) all
, relate to required- operability of suppression pool and level instru- -
mentation. They do not recognize the difference between narrow and
. wide ranges and therefore do not identify what level detector channel is to be used to meet the TS operability requirement. As a result, I divisional operability is left to the interpretation of the reader in l the action statements as well as in the surveillance requirements.
- c. In none of the above listed TS is the level. instrumentation required to initiate automatic suppression pool makeup addressed as a requirement for suppression pool operability. This is an accident mitigation Yunction and should have an associated surveillance. It would logically follow that at least TS 3.6.3.4 (Suppression Pool Makeup) -
should include the wide range level instrumentation as part of its operability requirement and a surveillance should be included. In TS 3.3.7.5 (. Accident Monitoring Instrumentation) only two suppression pool level detectors are required, and a seven day Action Statement applies .if only one is available. In reality, it appears this should read that two wide range level channels per division are required, minimum channels operable per division is one, and if only one division is operable, then the 7-day Action Statement applies.
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- d. By annotating what level detectors are required in the daily operating log and surveillance procedures, the license has made an effort to compensate for these unclear technical spe'ci fications . In spite of this, some problems were observed. The daily operating log indicates that for operability statement "a" of TS 3.5.3, the narrow range level detectors are to be used to verify that suppression pool level is "b", the 3 18'4 3/4" (Condition 1, 2, or 3); for operability statement - -
wide range level detectors are to be used 'to verify that suppression pool level is 312'8" (Condition 4 or 5) since narrow range indication does not go down this far. However, wide range is calibrated for post accident temperature (170*F), thereby indicating approximately 3" higher supp: ession pool level than what is actually present under normal conditions. The licensee has agreed to resolve this temperature calibration issue. This will be identified as Inspector Follow-up Item (IFI) 416/84-06-01.
- e. Furthermore, since the . wide range indication is utilized by the licensee in conditions 4 or 5, a channel calibration per surveillance requirement 4.5.3.1.b.3 (ECCS) is required. A channel calibration is performed by surveillance procedure 06-IC-IE30-R-0001, but only for surveillance requirements 4.3.7.5 (Accident Monitoring Instrumentation) and 4.6.3.4.c (Suppression Pool Makeup). The fact that this surveil-lance procedure does not recognize surveil. lance _ requirement 4.5.3.1.b.3 (ECCS) further demonstrates the need for individual level instrumenta- It is also tion identification in these associated suppression pool TS.
important that all TS relating to the suppression pool cross reference each other. As it stands now, only 3.6.3.1 and 3.5.3 reference one another. The FSAR states that the level sensors are spaced 90 degrees apart around the pool. Actually, the two groups of sensors are spaced 180 degrees apart.
- f. TS 4.5.3.1.a.2 contains an apparent typographical error. The licensee noted that a change request has been prepared to surveillance require-ment 4.5.3.1.a.2 (ECCS) to indicate suppression pool level as 12'8", in lieu of 12'5" (IFI 416/84-06-02).
- g. TS 3.6.3.1 (Depressurization Systems) and 3.3.7.5 (Accident Monitoring
, Jnstrumentation) specify suppression pool temperature requirements.
There are actually installed 24 temperature detectors / alarms. is(2 divisions with 6 pairs per division). The suppression pool -
azimuthally divided into six sectors, with two pairs (one pair per division) in each sector. By licensee designation, 12 of these detectors are used to meet TS 3.6.3.1 and the other 12 are used to meet TS'3.3.7.5. Consequently, only 12 channels undergo the channel requirement 4.6.3.1.c functional test required by surveillance Instru- -
(Depressurization Systems). 4.3.7.5 (Accident Monitoring Neither TS indicates mentation) does not require a functional test.
4 which temperature channels are to be used; therefore, leaving divisional /
sector operability to the interpretation of the reader in the Action Statements as well as in the surveillance requirements. In fact, surycillance requirement 4.6.3.1.c implies that you can use any 12 temperature channels as long as there are two channels in each sector.
Since cnly 12 channels receive functional testing, only these 12 should be credited by TS. _ _
TS Table 3.3.7.5-1 apparently should state as " required number of {
channels" .12,2/ sector rather than the present 6,1/ sector. Then the !
present statement of 6,1/ sector for " minimum channels operable" would allow operation for up to 7 days in an Action Statement.
It was further observed in TS 3.6.3.1 that the combining of action statements on suppression pool level and temperature instrumentation with the use of "and/or" was very ambiguous.
- h. TS 3.6.3.2 (Containment Spray) contains an error and the licensee stated that a change has been prepared to operability statement 3.6.3.2.b I
to indicate the use of a "RHR" heat exchanger, in lieu of a "SSW" heat exchanger (IFI 416/84-06-03). Another inconsistency in surveillance l
{'3 l requirement 4.6.3.2.b was pointed out by the inspectors.
In order to demonstrate operability of containment spray, this surveil-lance requires verification that each RHR pump develops a flow of at least 5650 GPM while recirculating water through the RHR heat exchanger to the suppression pool. This is accomplished by surveillance proce-dure 06-OP-1E12-0-0023, where the same recirculation flow verification is used to determine LPCI and suppression pool cooling operability.
t However, surveillance requirements 4.5.1.b.2 (LPCI) and 4.6.3.3.b l
(Suppression Pool Cooling) specify a recirculation flow of at least 7450 GPM. The FSAR indicates that containment spray flow emitting from i the spray nozzles into the containment is 5650 GPM. This would imply that a RHR pump flow capability of 7450 GpM is reduced to 5650 GPM after passing through the piping and containment spray nozzles.
Therefore, one would suspect that surveillance requirement 4.6.3.2.b (Containment Spray) should also require a RHR recirculation flow acceptance criteria of.at least 7450 GPM. At the time of the inspec-tion, the licensee was unable to provide their spray flow analysis to
" justify the lower RHR recirculation flow of 5650 GOM. .
In regards to an inoperable train, Action statements of TS 3.5.1 (LPCI) and 3.6.3.3 (Suppression Pool Cooling) allow for seven day continued operation when only one train is available. Since containment spray is n more important, i .e. , 5as less redundancy, action statement 3.6.3.2.a g- j - (Containment Spray) only allows a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action period when one train l ' .. is inoperable. A review of the RHR Pump operability data sheets in l
surveillance procedure 06-OP-1E12-0-0023 revealed an allowance of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to analyze test results. In essence, this allows an additional
3 - * '
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96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> possible delay to the Action periods discussed above.
Licensee representatives agreed to review this matter and make appropriate changes to the surveillance pro'cedure (IFI 416/84-06-04). -
- 6. a. TS 3/4.6.6.2 - Secondary Containment Automatic Isolation Dampers / Valves The inspector compared the TS list of automatic valves and dampers to- -
the licensee's surveillance procedures. The completed results of the most recent surveillance on secondary containment isolation were reviewed to see that the valve lists and identification are compatible.
Approximately 5% of the valves and dampers were examined in the plant by the inspector. No discrepancies were identified.
, The inspector asked licensee representatives if an actual plant walk-down had been conducted by the licensee to verify the accuracy of the TS lists of primary, drywell, and secondary valves. Licensee repre- l l sentatives stated that walkdowns were done at various times to resolve l specific questions, but no comprehensive effort could be identified which had as its objective the verification of the TS tables. The inspector stated that, although this is not a regulatory requirement, it would seem to be a prudent action to confirm TS accuracy. Licensee representatives agreed to consider further action (IFI 416/84-06-05). )
- 6. TS 3/4.6.6.3 - Standby Gas Treatment The inspector rev'iewed the surveillance procedure for SGTS. The TS surveillance requirements and the implementing procedures appear adequate to ensure SGTS reliability. The inspector walked down the majority of the SGTS hardware in the plant to ensure that the hardware is compatible with the TS. No discrepancies were observed.
- c. TS 3/4.6.7.1 - Hydrogen Recombiner The inspector examined the two hydrogen recombiner systems installed in the plant to ensure compatibil.ity with the TS. The completed results of the preoperational tests of this equipment were reviewed to ensure that the recombiners are capable of performance described in the
, surveillance section of the TS. No discrepancies were observed.
, 7. TS 3/4.8 - Emergency Power Supplies -
The inspectors selected several sections of the TS and the corresponding j surveillance procedures for exan:ination to veri fy the. adequacy of the l procedures and the TS as they relate to the existing equipment. The 4
following TS and surveillance procedures were examined and evaluated. -
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6 TS Sections 3/4.8.1 AC Sources - Operating 3/4.8.2 DC Sources - Operating 3/4.8.3 Onsite Power Distribution Systems (Operating) 3/4.8.4 Electrical Equipment Protective Devices Primary Containment Penetration , Conductor Protective Devices -- -
Motor Operated Valve Thermal Overload Protection l
Reactor Protection System Electric Power Monitoring
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Surveillance Procedures
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j 06-OP-1R20-W-0001 ' Plant AC and DC Electrical Power Distribution Weekly Lineup 06-EL-1L51-R-0001 125V Battery Charger Capability Test 06-IC-1C71-SA-1001 RPS Electrical Protection Assembly Channel Functional Test 06-EL-1C71-R-0012 RPS Electrical Protection Assembly Calibration
. 06-EL-1L11-0-0001 125V Battery Capacity Discharge Test 06-EL-1R65-Q-1001 MOV Thermal Overload Protection Device 06-EL-1R65-R-0001 MOV Thermal Overload Protection Device As a result of this review, the following discrepancies were-identified: -
- a. Surve'111ance procedure 06-EL-1L51-R-0001 appears to be inadequate in that the battery chargers are never tested at the equalizing voltage (140 VDC 2 1 volt). The chargers are only tested at 105 volts at 400 amperes for two (2) hours. In. addition, in the battery discharge test, there is no time limitation specified for when the batteries must be recharged to full capacity (IFI 416/84-06-06).
- b. An apparent typographical error was found in TS Table 3.8.4.2-1. The B designation was omitted from valve number QSP4151898. Licensee representatives have since stated informally that the TS is correct.
This will be confirmed during a future inspection.
- c. TS requirement 3.8.4.3 appears to be inappropriate for the way the RPS electrical power monitoring assemblies (EPAs) are designed. Two EPAs
'are in series which means both units must be operable to supply power to the RPS bus. The Action statement in the TS requiring only one (1) .
EPA unit to be restored to service when two are inoperative does not seem appropriate for the circumstance. ) There is no provision for manual bypass of the individual EPA units.
Surveillance procedure 06-IC-IC71-SA-1001 appears inadeouate in that it only requires testing of the EPAs that are not providing power to the Reactor Protection System (RPS) bus. The procedure does not assure that the EPAs associated with the normal power supply (MG sets) will
. be tested during the six month surveillance test as required by TS Section 4.8.4.3.a (IFI 416/84-06-07).
7 -
The NRC inspectors also performed walkdowns of the systems icentified above to randomly verify that equipment described in the TS was actually installed in the plant. All equipment examined in the plant was found to be properly identified in the TS with the exception of the items discussed above.
- 8. ECCS Systems and Actuation Instrumentation ,
! a. TS 3/4.5.1 ECCS - Operating The requirements for ADS operability c6ntained in paragraphs 3.5.1.a.3 and 3.5.1.b.2 were . reviewed. The TS paragraphs require "at least 7 operable ADS valves". This number appears to be incorrect. The Safety Evaluation Report page 6-22 states that the ADS employs eight of 20 i
SRVs. The action statement paragraph e.1 allows the operation up to 14
- days with only six ADS valves operable, and up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with five or l 1ess ADS valves. This appears to be an unacceptable TS (IFI 416/84-l 06-08).
A review of the paragraph 4.5.1.b pump testing criteria was conducted.
Significant inconsistencies were noted in the pump flow characteristics for the following pumps. The TS for High Pressure Core Spary requires at least 7115 gpm with 182 psid, while -the SER page- 6 states - -
l 7115 gpm with S40 psid, and the FSAR Figure 6.3-2 lists 7115 gpm with
- approximately 387 psid. The TS for Low Pressure Core Spray' requires at
, least 7115 gpm with 261 psid, while the SER page 6-22 states 7115 epm
! with 340 psid and the FSAR lists 7115 gpm with approximately I 311.6 psid. The TS required flows. appear considerably less conserva-tive than either the SER or FSAR (IFI 416/84-06-09).
- b. JTS 3/4.3.3 Emergency Core Cooling System Actuation Isolation The inspector verified the incorporation of the following instrument surveillances of TS Tables 3.3.3-1, 3.3.3-2, and 4.3.3.1-1 into the plant's surveillance program.
LPCI Pump A Start Time Delay Relay ADS Times Drywell Pressure High Reactor Vessel Water Level - Low, Low, Level 2 -
The following surveillance procedures were reviewed to ensure the required TS frequencies, trip setpoints, and allowable values were correctly incorporated.
06-IC-1921-R-0012, Rev. 22, Reactor Vessel Water Level Calibration 06-IC-1521-M-1010, Rev. 21, TCN9 Reactor Vessel Water Level (HPCS) 06-EL-1821-M-0001, Rev. 21, TCN3 ADS Times Functional Test and Cali-I bration l
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06-OP-1000-D-0001, Rev. 20, TCN27 Daily Operating Log (Items 64 & 16) 06-IC-1521-R-0009, Rev. 21, TCN3 Drywell High Pressure Calibration
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(ECCS) 06-IC-1821-M-1011, P.ev. 20, TCN4 Drywell High Pressure (HPCS)
Functional Test 06-EL-1E12-M-0001, Rev. 22, RHR Pump Start Time Delay Relay Functional
- 06-EL-1E12-M-0001, Rev. 21, RHR Pump Start Time-Delay Relay Calibration. -
t The following Drywell Pressure ' Transmitters and Reactor Vessel Level Transmitters were reviewed for proper field installation in accordance with the as-built drawing and the piping and instrumentation diagrams.
Logi.c diagrams were reviewed for actuation of the appropriate equip-
! ment. No defic'iencies were noted. ,
l l Division I PT N094A Division I PT N094E Division II' PT N0948 Division II PT N094F Division III PT N067C Division III PT N067G Division III PT N067L Division III PT N067R i
Division I LT N091A Division I LT N091E Division II LT N0918 Division II LT N091F Division III LT N073C Division III LT N073G c Division III LT N073L Division III LT N073R The licensee has previously identified a problem on instruments which utilize atmospheric pressure as one side of a differential pressure detector instrument. Due to a possible low atmospheric pressure condition around the plant, certain detectors may be as much as .5 psig nonconservative. This includes drywell pressure and containment spray.
The licensee has not yet submitted all the appropriate changes at this
. time (IFI 416/84-06-10).
- 9. Drywell and Primary Containment Integ-ity An inspection was performed of the following sections of the Grand Gulf TS:
SECTION SUBJECT pAGES .
3/4 6.1.1 Primary Containment Integrity 3/4 6-1 3/4 6.1.2 Containment Leakage Rates 3/4 6-2, 3, 4 ;
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a
. 9 3/4 6.1.3 Containment Air Locks 3/4 6-5, 6 3/4 6.1.4 MSIV Leakage Control System ' 3/4 6-7 -
j 3/4 6.1.6 Containment Structura1 Integrity ,3/4 6-9 3/4 6.1.7 Containment Internal Pressure 3/4 6-10 -- -
3/4 6.1.9 Containment Purge System 3/4 6-12 Emphasis was placed on the following specifics:
(1) Literal correspondence between the TS, and the installed hardware configuration.
(2) Adequacy and completeness of the surveillance requirements.
(3) Review of associated surveillance procedures and results generated by their execution.
(4) Adequacy and completeness of the Action Statements.
(5) Familiarity of licensee personnel with the TS. and.. the associated._._
hardware systems and testing requirements. i The following discrepancies were identified:
- a. Surveillance Requirement '4.6.1.4 concerns the operability of each MSIV leakage control subsystem. Item C address the functional testing of
! the subsystem heaters but does not acknowledge that there are no heaters on the outboard subsystem; whereas there are four on the l inboard subsystem. Section 6.7 of the FSAR reveals that no heaters are l required in the outboard system, which is not identical to the inboard I system in a number of aspects. Additionally, licensee surveillance l procedures accurately reflect the existing hardware configuration.
Clarification of the wording of the TS will resolve the ambiguity (IFI 416/84-06-11),
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- b. ' Surveillance Requirements 4.6.1.1.a requires a leak rate retest of the equipment hatch seals every time each penetration subject to a Type B test, except the containment air *ocks, is reclosed.
This is not what was intended as it would require a retest of the equipment hatch seals following the opening of Type B pentration areas such as:
(1) electrical penetrations (2) ECCS test return line orifice plate (3) fuel transfer tube j
5 10 Surveillance Requirement 4.6.1.1.b is vague as to what must be secured in position, and how. Correction of the wording in the TS will resolve these ambiguities (IFI 416/84-06-12).
- 10. TS 3/4.3.2 Isolation Actuation Instrumentation TS 3/4.3.2 Isolation Actuation Instrumentation was reviewed to determine if._ _ ,
the requirements entailed therein are clear,1f the LCOs are realistic, if the channels and trip systems appear technically adequate, if there are procedures for performing the surveillances, and if those requirements can be performed. -
Seventeen applic'able sur've111ance procedures were analyzed for technical adequacy and incorporation of TS acceptance criteria. Ten channels and associated trip systems were analyzed through examination of electrical prints, logic diagrams, and system descriptions for technical adequacy.
The review revealed that TS 3/4.3.2 appeared technically adequate, the requirements realistic, and there were procedures for performing those
, selected requirements reviewed.
l
! The procedures reviewed, with those exceptions to be detailed, appeared adequate. The procedures reviewed included but were not limited to:
06-IC-1821-M-1004 06-IC-1821-M-1010 06-IC-1C71-M-0001 06-1C-1821-M-1004
^
06-IC-1E31-M-0003 06-OP-1000-D-0001 06-IC-1E31-M-1001
. D6-Dp-1G33-M-0002 06-IC-1E31-M-0023 06-IC-1321-M-1003 l
l l
11 t
Of those procedures, some discrepancies were identified in operation surveillance procedure 06-Op-1000-D-0001, which entails the operations' semi daily surveillance (channel . checks) as required by TS Table 4.3.2.1-1. The item numbers, detailed below refer to the TS table specific requirement, by line number. The discrepancies are as follows:
- a. With regard to items 1.b through 1.g and 5.m it was observed that the- -
procedure line item (as referred to by t'he TS cross reference index) does not conform to the TS requirement. This appears to be an error in the TS cross reference document. . . .
- b. Line. item 63 of procedure 06-OP-1000-D-0001 stipulates a channel check of slave trip ' unit B21-LS-693 to level instrument B21-LIS-N691 at
< 53.5". The slave trip unit referenced on the semi-daily surveillance sheet has undergone a station modification such that it no longer is
. slave unit to instrument B21-LIS-N691, as the procedure indicates, but
! now is slave to instrument B21-LIS-N695. There are three problems associated with this issue, of serious concern.
- (1) The station modification was completed, according to the licensee, in December 1983; however, the procedure has yet to be changed to reflect the modification.
(2) The operations staff, if they were to be following the procedure, would be noting in mode 4 that slave unit B21-LS-693 is not in alarm as it should be since the master trip unit B21-LIS-N691 is tripped and is the procedure referenced master trip for that slave. They are not noting this fact.
(3) The operations staff has been ignoring the procedural requirements simply because they know the modification, discussed above, is installed and as such that slave trip unit should not be in alarm.
The operations staff however, has not in the period since December initiated a procedure change to reflect the station modification.
- c. Further, TS item 4.3.2.1-1.f requires that a semi-daily channel check be performed on high drywell pressure, ECCS, Division 3. The TS cross reference index refers to line item 15 of procedure 06-Op-1000-0-0001
. -for the performance of that required surveillance. That line item number appears to test reactor vessel level 2 and 8. -
d~. Line items 15 and 63 (reactor vessel level it) require a semi-daily .
channel check of those applicable instruments at < 53.5". However, the instruments referred to in the prccedure for those channel checks are calibrated for elevated temperatures, and in modes 4 and 5, are pegged
~high such that the required cnannel check can never be satisfactorily performed. Here again, although operations staff has known of this inadequacy, no procedure change has been implemented to resolve the inadequacy, i.e., refer to an instrument which reads correctly in modes 4 and 5.
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12 In summary, the TS (3/4.3.2) apoears to be adequate although there are some inadequacies in the implementation of the requirements. Licensee representa-tives agreed to review and correct as necessary the semi-daily log sheet and the TS cross reference document (IFI 416/83-06-13).
- 11. TS 3.6.4 - Containment and Drywell Isolat' ion Valves
~
The inspector reviewed'TS 3.6.4 " Containment and Drywell Isolation valves"~
and~ Tables 3.6.4-1 to verify that the licensee has adequately identified all primary containment and drywell penetrations and associated isolation valves and included them in the TS. In addition,. the inspector reviewed the limiting condition for operation and surveil' lance requirement for TS 3.6.4
'to ensure that they'were appropriate and being properly' implemented by the licensee. -
The inspector reviewed Final Safety Analysis Report (FSAR) Table 6.2-44
" Containment Isolation Valve Information" and Figures 6.2-76 thru 80
" Containment Leak Rate Test System" and compared them to TS Table 3.6.4-1, plant surveillance procedures and selected "as built" penetrations observed in the plant. No discrepancies were identified. The inspector reviewed the following plant surveillance procedures:
l 06-OP-1M10-M-0001 " Containment and Drywell Penetration Isolation Monthly Check" 06-OP-1M10-C-10C1 " Containment and Drywell Penetration Isolation Cold Shutdown Check" 06-OP-1B21-R-0006 " Containment, Drywell and Auxiliary Building Isolation Valves Functional Test" 06-OP-1821-C-0003 " Nuclear Boiler Valve Operability" The procedures were reviewed to determine if they adequately fulfilled the associated surveillance requirements and corresponded to the TS Table 3.6.4-1 of isolation valves. Some minor discrepancies were identified and resolved.
No valves were identified that were not.. covered by procedure or TS.
The inspector toured the Unit 1 Containment and Drywell and randomly selected approximately 45 isolation valves and verified that they were .
included in TS Table 3.6.4-1 and plant surveillance procedures. No discrepancies were identified. The conments of paragraph 6.a and the IFI identified there also include this large group of valves.
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