ML20248E410
| ML20248E410 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/27/1970 |
| From: | Hendrie J US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Seaborg G US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20248A375 | List: |
| References | |
| NUDOCS 8910050222 | |
| Download: ML20248E410 (23) | |
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fwAs'HINGTON. D.C. 20545 - .
p" ' ~ , , January 27, 1970' w
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eHonorable Glenn T. Seaborg n Chdirman a >U.iS. Atomic Energy' Commission-B LWehhington, D. C. 20545 .
E _ Subje'ct: - REPORT ON JAMES A. FITZPATRICK NUCLEAR POWER PLANT-t , Dacr Dr.-Seaborg:
- -At a Special Meeting,. January 23-24,1970, the ' Advisory Comittee on 1
1Recctor Safeguards completed its review of the app l icati on of the Power Authority of the State of New York for authorization to con -
The project.was c "otruct= the -James A. FitzPatrick Nuclear Power Plant.
also: considered at a Subcommittee meeting on January 22, 1970.in Wash-ington,-D._'C. During its review, . the Committee had the- benefit of' dis-cussions with representative's: of 'the Power Authority of the : State of-Mew York, the Stone and Webster Engineering Corporation,: the General
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idlectric Company, the N2.agara Mohawk Power Corporation, the AEC Regu-letory Staff and their~ consultants. The Committee also had the benefit -
- of the documents listed.
b The applicant will own' the nuclear plant and will have sole responsibil-( ity for its construction. Under contract with the applicant, . Niagara Mohawk Power Corporation will operate the plant in conjunction with the z
- Nine Mile: Point plant and will participate .in the design. The applicant bas designed,- constructed and operated extensive hydro-electric genera-
. ting facilities and electrical distribution- systems cotaling 3200 MRe.
The FitzPatrick plant will be the' first steam stacion owned by the appli-
~ cent.;
The nuclear plant will use a single cycle, forced cireplation, boiling water reactor similar to the previously reviewed Ha:ch reactor (ACRS re-port of May 15,_'1969). The FitzPatrick reactor is dasigned to have a rated' output-of 2436 MWt with a maximum anticipated power level of 2550
~ PGiC . '8910050222 890921 PDR- ADOCK 050o0233 PDP G
arhe' site of the Fit: Patrick plant is located in Oswego County, New York, cn the' southeastern shore of Lake Ontario anc is about 3000 feet east the Nine Mile Point Nuclear Station of tr.a Niagara Mohawk Power Corpora-J tion'. Tbc minimum exclusion zone radius is 3200 feet ar.d che nearest
- The population i permanent residence is about 3700 feet from the plant. b t 1500.
within the four mile low population zone is es timated to be a ou
! '9swego, New York (population about 24,000) l i is seven f about 271,000, miles sol
_he site, and Syracuse, New York, with a popu at on o l j
l1 is located 36 miles southeast. j The meteorology, hydrology, geology and seismology of the site h i
investigated by the applicant and found acceptable.
j The applicant is studying design modifications to prevent h violat ths containment by pipe whipping and generation of missiles The Committee in t e un-likely event of a failure of the primary system piping.
believes that such design modifications as are practical Theshould Com- be plemented in a manner satisfactory to the Regulatory Staff.
mittee wishes to be kept informed about this matter.
The applicant proposes to locate motor-driven pumps fcr the emergen
'9 ore cooling system in a single The crescent-shaped Committee believes that the pump room adj toroidal suppression chamber.
rcom should be partitioned into water-tight compartments and the pum located so that flooding of one compartment will not impair the func-T tion of the pumps in the remaining compartment (s) .
be resolved in a manner satisfactory to the Regulatory Staff. <
The Committee recommends that the applicant i h give regard further to thecons to the design of the emergency on-site power system u t The Commit-need for synchronization of the diesel-driven generators. i l and in-tee also recommends that rigorous, realistic preoperat ona hich is service cdopted.
testing programs be undertaken on the dies these considerations.
A large number of instrument lines, approximately one inch t in penetrate the primary containment and terminate as clos pressure sensing devices.The isolation provisions for these instrument reactor primary system.
lines include a manual shutoff valve and a spring-loaded i h rimary exc check valve for each line, with both valves located containment.
factory without the inclusion of remotely operable isolation i va However, since these lines represent a potential source tion for be a p system and containment leak, it is essential that isolation, propercontrolatten d mini-given during design and construction stem i l
to questi is pres mization surized.
of possible mechanical damage while the
. l that are satisfactory to the Regulatory Staff. Also, the applicant should study and propose means to reduce the rate of possible leak-cge from instrument lines in the event of failure so that such leak-age would not damage the secondary containment. or bypass the building filters.
The applicant has stated that fuel handling proceduras and fuel pool design will be such that the dropping of the spent fuel cask will not jeopardize the continued availability of adequate cooling water for stored irradiated fuel. The applicant will review the final design opproach with the Regulatory Staff when this aspect of design has been completed.
Information on a number of items, identified in previous reports of the Committee, is to be provided by the applicant to the Regulatory Staff during construction. These include: ,
(a) A study of means of preventing common failure modes from negating scram action and of design features 'to make tolerable the consequence of failure to scram during an-ticipated transients. .
(b) Review of development of systems to control buildup of hydrogen in the containment following a loss-of-coolant accident.
(c) A study of the exfects of leakage through possible bypass paths, with particular emphasis on the main steam line isolation valves, and measures to deal with such leakage.
Other problems related to boiling water reactors have been identified by the Regulatory Staff and the ACRS and cited in previous ACKS re- -
ports. The Committee feels that resolucion of these items should apply equally to the FitzPatrick plant.
The Committee believes that the above items can be resolved during con-struction and that, if due consideration is given to these items, the James A. FitzPatrick Nuclear Power Plant can be constructe able assurance that it can be operated without undue risk to the healt' and safety of the public.
Sincerely yours, Jo eph M. hendrie
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Honorable Glenn T. Seaborg January 27, 197
= References' - James A. FitzPatrick Nuclear = Power Plant
- 1. Letter from LeBoeuf, Lamb, Leiby &-MacRae, dated _ December 31,19 License Application; Volumes I, II and III of the Preliminary Sa Analysis Report (PSAR) 2; Amendments Nos. 1-12 to License Application, including Supplemen Nos.1-11 to PSAR ,
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NUCLEAR HEGULATORY COMMIS$10N f " NR #'. U 2//
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- OFFAGES /a iono or paussia. ee n=svtvansa t uu
'....." 3 AUG 121985 i Docket No. 50-333 Power Authority of the State of New York James A. FitzPatrick Nuclear Power Plant ATIN:
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I Mr. Harold A. Glovier ' -
Resident Manager 1 P.D. Box 41 _
Lycoming, New York 13093 . -- '
Gentlemen: -
\
Subject:
Inspection 50-333/85-19 This refers to the routine, safety inspection conducted by Mr. L. Doerflein of this of fice on June 1, - July 14,1%5, at the J.A. FitzPatrick Nuclear Po Plant, wer to the discussions of our findings held by Mr. Doerflein with y other members of your staff at the conclusion of the inspection.
Areas examined during this inspection are described in the NRC Region I Inspection Repcrt which is enclosed with this letter.
Within these areas, the inspection consisted of selective examinations of procedures and representati v e
records, interviews with personnel, and observations by the inspector .
t Wi hin the scope of this inspection, no violations were observed.
No reply to this letter is required.
is appreciated. Your cooperation with us in this matter Sincerely,
?
IW Samuel J. C4111ns, Chief Projects Branch No. 2 Division of Reactor Projects inc'oscre W hp er. I inspection Report No. 50-333/85-19 .
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2 DETAILS q 3.
Persons Contacted R. Baker, Technical Services Superintendent
- J. Brons,
- R Converse,Senior Vice President Superintendent of Power Nuclear Generation M. Curling, Training Superintendent W. Fernandez, Operations Superintendent
'H. Glovier, Resident Manager H. Keith, Instrument and Control Superintendent D. Lindsey, Assistant Operations Superintendent R. Liseno, Maintenance Superintendent E.Superintendent Mulcahey, Radiological & Environmental Services R. Patch, Quality Assurance Superintendent T. Teifke, Security & Safety Superintendent The inspector also interviewed other licensee personnel during this inspection physics, including shif t supervisors, administrative security, ,
operations, health personnel. instrument and control, maintenance a,d contractor
- Denotes those present at the exit interview.
2.
Licensee Action on Previous Inspection Findings (Closed) Inspector Followup Item (333/81-06-05):
As part of the Analog Transmitter Trip System modification (no. F1-82-53) installed during t 1985 refueling outage, the Reactor Core Isolation Cooling System isolat subsystem thermocouple were connected to trip units.
therefore, the need to disconnect and reconnect and, eads, the which surveillance resulted in their frequent failure,<iuring routine monthly testing.
this item. The inspector had no further questions regarding (Closed) Devir. tion (333/81-07-01): 4The inspector noted that, during'
.1983, refueling, outage,thelicenseeinstalled:tsolationjive;slcapableof remrate manual operation from~the control room on the Reactor > Building' Closed the primaryLoop Cooling Water System influent and effluent lines penetr containment'.
The inspector reviewed the p'ackage forrthisM.
modification item. (no. F1-81-26) and had 'no further questions'iega;rding"thisW (Closed) Violation (333/83-01-01): The inspectcr reviewed procedure Containment Vacuum Re11of and Contair. ment D',f Revision 22, datt.d June 25, 1985, and ver11,ed that the licensee revi.-1 the procedure directly from thetotruck include a method of inert'ng the primary containment nitrogen fill connection. The inspector No %
further questions regarding this item.
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Il POWER AUTHORITY OF THE STATE OF NEW YORK j
{ JAMES A. FtT2PAtmicz NucLEAm Powen PLANY .I DSR #3 N k ~ ~
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- 0. h RAYMOND J. PASTERNAK it a...s. . s....., Po BOK 41 L yco,,, .. u yu. . i30s3 A
315 347 3840
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1 f~ August 24, 1981 SERIAL:
J. JAFP 81-0871 i-t t
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Boyce H. Grier. Director '
United States Nuclear Regulatory Comission Region 1 J
631 Park Avenue I. King of Prussia, PA. 19406
SUBJECT:
DOCKET NO. 50-333 - I&E INSPECTION NO. 81-07
Dear Mr. Grier:
'1 d
j With reference to the inspection conducted by Mr. J. C. Linville of your office on March 1-31, 1981, at the James A. FitzPatrick Nuclear y[ Power Plant, and in accordance with the provisions of Section 2.201 of Part 11 of Title 10 of the Code of Federal Regulations, we are submitting our response to Appendix A Notice of Deviation transmitted by your letter
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dated July 31, 1981 at eeieived by the undersigned on August 3,1981.
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APPENDIX A
.] NOTICE OF DEVIATION k Based on the results of an NRC inspection conducted during the period 1
March 1-31, 1981, it appears that one of your activities was not conducted
)
in conformance with your comitments to the Comission as indicated below:
7' A. Contrary to the licensee's comitments in Sections 5.2.3.5 and E
8 7.3.4.3 of the FSAR the five Reactor Building Closed Loop Cooling Water System containment effluent lines have only single manual d
h isolation valves outside containment instead of one valve which cicses automatically by process action (i.e. , reverse flow)'of' by f remote manual operation from the control room. s , o
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g Boyce H. Grier DiTctor August 24, 198; 1
United States Nuclear Regulatory Comission JAFP 81-0871
SUBJECT:
DOCKET NO. 50-333 I&E INSPECTION NO. 81-07 Page RESPONSE TO NOTICE OF DFVIATION
- i:
A. The FitzPatrick Plant has completed a review of the subject Inspection Report, applicable portions of 10CFR50-Appendix A, the Final Safety
- Analysis Report and other related documents. As a result of this review.
l the FitzPatrick Plant is in agreement that the installation of remote manual motor operated valves is desirable for operational considerations and for additienal safety margin. Accordingly, the Power Authority comits i '
to provide sucS valves for the rcactor building closed loop cooling system. Since an engineering evaluation and an assessment of equipment availability is necessary before the Authority can provide a firm modification schedule, the Authority plans to provide a schedule for completion of this modification by February 2, 1982
~
As a result of the review noted above, it has also been determined that the FSAR, in addition to providing general design criteria for contain-ment isolation as noted in the deviation, describes the isolation capabilit i of the containment penetrations in question in a manner which is essentiall.-
as-built. Specifically, FSAR Figure 9.5-1 and FSAR Table 5.2-2, as well as Techical Specification Table 3.7-1 accurately depict the as-built valve arrangement associated with reactor building closed loop cooling effluent
_: from primary containment.
Y It should also be noted that the Power Authority is currently reevaluating
, containment isolation design and capability in response to NUREG 0737
, and is also in the process of preparinc, an updated FSAR in accordance with 10CFR50.71(e).
As part of these efforts, the inconsistencies in the FSAR will be corrected and the Power Authority will submit a report with respect to containment isolation design as indicated in previous NUREG 0737 and 0578 related correspondence.
Very truly yours, y
N RJP:V0:brp RAYMOND J. PASTERNAK Distribution:
O George b je t1 *C P. lisyne,"rry, PASNY, NY0PASNYT'RYOW \
( G. M. Wilverding, PASNY, NYO
- B. W. Deist, PASNY, NY0 f T. Dougherty, PASNY, NY0
- J. F. Davis, PASNY, NYO M. C. Cosgrove, PASNY, JAF 1 R. Baker, PASNY, JAF ei W. Fernandez, PASNY, JAF 1 NRC Resident Inspe; tor, JAF
~" -y NRCI 81-07 File
,;j Document Control Center ii la N*
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. 5.0 CONTAINMENT SYSTEMS 5.1 General- ,
e The containment syste=s include the primary containment which utilizes the pressure suppression concept and the secondary containment which is formed by the low-leakage reactor building surrounding the primary containment. The reactor building has an air recirculator system and a Standby Gas Treatment System ,
'(SGTS) to mix and filter primary containment leakage prior to its discharge to the environment via the main stack.
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5.2.1 Design
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The primary containment is a typical "lightbulb" pressure suppression system consisting of a drywell, pressure suppression
- chamber (torus), and a connecting vent system. The drywell has ,
. j-a steel spherical lower portion 65 feet in diameter, and a steel cylindrical upper portion 35 feet 7 inches in diameter. The overall height of the drywell is about 111 feet. The pressure suppression chamber is a steel torus located below and encircling .
the drywell, with a centerline diameter of approximately 108 feet and a cross-sectional diameter of 29.5 feet. Eight vent pipes lead from the dryvell to a header inside the torus, and 96 - 24
. inch diameter downcomer pipes project downward from the header I
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> vith Section III of the ASME Boiler and Pressure Vesse.1 Code, maximum.
dryvell pressures up to 62 psig are permissible for this design.- i; i
Combinations of live, dead, and seismic loads in conjunction with {
l thermal stresses have been considered in the design analysis'. The 8 l d :
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design also considered the jet forces that might act on the contain-
' i ment consequent to a pipe severance. Adequate strength has been >.
provided to prevent f ailure of the containment vall as a result of ,
direct jet impingement, and critical penetrations have been l-p P
provided with restraints and auxiliary stops to limit pipe movement ::.
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.and prevent fallure of the containment.
}] i The primary containment was designed to sustain the combination of -loads resulting from the design basis loss-of-coolant accident, l-earthquake, and the conventional live and dead loads within the
- stress limits defined in Subsection III B of the ASME Boiler Ii-and Pressure Vessel Code 1968 and applicable addenda in effect (
i as of June 1968. ';e find the design stress limits for the i
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primary containment system to be acceptable. t r
Containment piping penetrations satisfy the criteria of the j i t ASME Boiler and Pressure Vessel Code noted above. Lines connected !
to the reactor coolant system incorporate a sleeve to extend the f r
dryvell to the outer isolation valve, thus containing the effluent [
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g_z H."t in the event of'a pipe break. Hot lines which must sustain large i
. thermal-and mechanier.1 stresses are designed with combinations of penetration sleeves'and flued fittings. ,, j j.
!. Based on our review of the information contained in this j l
application and similar designs we . conclude that the primary ;
containment design basis is acceptable. j
~' 5.2.2 Missile and Pipe Whip Vrotection
- 4 Several' locations on the main steam lines and feedwater lines l.
are not restrained to prevent pipe whip in the event of pipe failure
'at these locations.'; The applicant has stated that the physical layout
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within the drywell- pr'ecludes restraints at these points. For all
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'other lines and locations, restraints have been provided where- ;
+ t a break could result in containment impact. . The. applicant has
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identified the unrestrained high stress areas in these lines where I breaks could result in pipe whip- such that the pipe could impact the primary containment vall. At those locations which are accessible !
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the applicant has provided 1-1/4 inch thick impact plates as sup- ;
. i pienencary protection for the drywell. In addition, he has agreed i i
i to perform augmented inservice inspection of these weld locations l o .,.- during each inspection period. At the remainder of, these identified l
areas the physical layout precludes installation of impact plates. ,
. . t Here, the applicant will perform augmented inservice inspection of yh n '
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, the welds during each inspection period. The requirements of this j Q. . .
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augmented inspection will be set forth in the Technical Specifications l '; ' t.. ; . .' *
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and will' call for 100% rather than 25:-inspection during each period.
The applicant has also considered thel effects of pipe whip on the iij ., .
. emergency
- core
.. ' soling systems. The systems are redundant and L';
- physically' separated such that a ruptured pipe could impact and
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affect only one of the redundant ECCS. The remaining ECCS components.
'- were shown to limit peak fuel clad temperature to 1370'F following the i.
most severe postulated break sequence.
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- , . The applicant has considered the effect of missiles ranging in
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size from nuts and bolts to valve bonnets, and concludes that no
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e In-addition, where possible, components are arranged so that the h-3
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direction of flight of potentia 1 missiles is away from the.contain-
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ment' wall.
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The effects of pipe whip and steam jet impingement on the shield
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and vessel support structure resulting from a LOCA occurring within o- .
the sacrificial shield area were analyzed and found to be acceptable.
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We conclude that the applicant has provided adequate measures to
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. 5.2.3 Containment Isniation The ability to isolate the primar containment provides the g.+
necessary integrity between the coolant system pressure beundary, r:3A 4 e
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of or the containment atmosphere, and the environs in the event Isolation is accidents or other non-nominal conditions. .
The numbers, types and locations h, -
' accomplished by means of valves.
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' of these valves- in the various ~ 1ines depend on the manner in which
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'the lines pene'.. rate' the reactor vessel and the containment.
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I necessary, the valves are equipped with operators and close b or fault
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-automatically when sensors detect certain accident *
- o conditions.
The consequene,s e of postulated pipe failures both inside and outside of the containment have been evaluated and are described in Section 10. - The isolation valves and their control systems hay ( been reviewed to assure that no single failure can result in a loss of containment integrity. An exception exists in the case of instrument lines connecting to the reactor coolant system which penetrate the containment and dead-end at instruments located in the reactor building. Such lines are provided with manually operated isolation valves and excess flow A break chsek valves, both of which are outside the containment.
in the line between the containment and the outer check valve would result in blowdown directly into thu reactor building.
The applicant has installed 1/4 inch diamator orifices in each I.
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L . :. ,. of.these lines inside the primary containment to prevent over-
, pressurization'of the reactor building and limit offsite doses -
', . to substantially below the' 10 CFR Part- 100 values in the event 4 of theLpostulated instrument lino break. Based on our review of
, .the design;we c6nclude that the provisions for instrument 1,ines penetrating the' primary containment are adequate and satisfy the-
- supplement to Safety Guide 11.
- Leakage through the closed main steam line isolation valves'
?[ follouing a postulated LOCA presently. relics on the low leakage characteristic of the vaives. The acceptability of present .
leekage limits and the need for an auxiliary scaling system
- are under study by the staf f. There is nothing in the existing
/ design which would preclude incorporation of an additional sealing *
' feature if such is determined to be necessary. The applicant
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will continue to study developments in this area.. .
Based on our review we conclude that the primary contain-i
. ment isolation provisions are adequate, 5.2.4 Leakage Testing Program Leakage testing of the reactor primary containment and associated systems is intended to provide initial and periodic 11 .
verification of the leaktight integrity of the containment.
9 The applicant has stated in Amendment Nos. 4 and 5 that the primary reactor containment and its components have been designed
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_...- 5-8 so that periodic integrated Icakage rate testing can be performed at the calculated peak pressure. Penetrations, including ,
personnel and equipment hatches and airlocks, and isolation valves, have been designed with the capability of being individually
. leak tested at calculated peak pressure. I We conclude that the containment system will permit contain-s.
ment leakage rate testing in compliance with the AEC proposed l l
" Reactor Containment Leakage Testing for Water Cooled Power j
. Reactors," 10 CFR 5 50.54(o). Appendix J, and therefore is acceptable.
/~ In addition to agreeing to meet the requirements of proposed ,
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' Appendix J PASNY has agreed to perform a leak test of drywell to suppression chamber piping, headers, downcomers and vacuum breaker valves at each refueling outage. They will also determine receptable bypass leakage limits and other test criteria and will be ,
- l required to perf orm frequent surveillance testing of the vacuum l breakers. We have not completed our review of the details of the test and surveillance program. However, the applicant has indicated I
his intention to base it on the recently approved Browns Ferry leak check program. We find this com=itment acceptable pending completion of our review.
5.3 Secondary Containment i . . '
- The reactor building, together with the Standby Gas Treatment System (SGTS) and the main staca, form the secondary u..) .
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wuct AR REGULATORY comenssion =
- 15 memon # OF PAGES C ses rans avsmus EINe or enuar:4. Pewnsyt.v4= 4 temos JUL 3 1 381 Docket No. 50-333 Power Authority of the State of New York James A. FitzPatrick Nuclear Power Plant ATTN: Mr. R. J. Pasternak
' Resident Manager P. O. Box 41 Lycoming. New York 13093 Gentlemen:
Subject:
Inspection No. 50-333/81-07 This refers to the routine, safety inspection conducted by Mr. J. C. Liny 111e of this office on March 1-31, 1981 of activities authorized by MRC License No.
DPR-59 and to the discussions of our findings held by Mr. Linville with Messrs.
Bayne. Klausman. Diest, yourself and other members of your staff at the conclusion of the inspection..
Areas examined 'during this inspection are described in the Office of Inspection
-. and Enforcement Inspection Report which is enclosed with this letter. Within v
these areas the inspection consisted of selective examinations of procedures and representative records.' interviews with personnel, and observations by the inspector.
Our inspector also verified the steps you have taken to correct the item of noncompliance brought to your attention in the enclosure to our letter dated February 11,1981. We have no further questions regarding the si!aps you took to correct item A.
Based on the results of this inspection, it appears that certain of your activities were not conducted in full compliance with NRC requirements. Enforcement action pertaining to the replacement of G safety relief valve will be provided under separate correspondence to J. P. Bayne, Senior Vice President. Nuclear Generction, Power Authority of the State of New York.
Another activity appears to be a deviation from connitments made in your FSAR regarding containment isolation provisions as set forth in the Notice of Deviation enclosed herewith as Appendix A. In your reply, include your cownents concerning this item, a description of any steps that have been or will be taken to prevent recurrence and the date all corrective actions or preventive measures were or will be taken.
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Power Authority of the State of New York 2 JUL 311981 3 l In accordance with 10 CFR 2.790 of the Comission's regulations, a copy of this letter and the enclosures will be'placed in the NRC's Public Document Room. If this report contains any information that you (or your contractors) believe to be exempt from disclosure under 10 CFR 9.5(a)(4), it is necessary that you (a) notify this office by telephone within ten (10) days from the date of this letter of your intention to file a request for withholding; and (b) submit within 25 days from the date of this letter a written application to this office to withhold such information. Consistent with section 2.790(b)(1), any such application must be accompanied by an affidavit executed by the owner of the infonnation which identifies the doctment or part sought to be withheld, and which contains a full statement of the reasons on the basis which it is claimed that the information should be withheld from public disclosure. This section further requires the statement to address with specificity the considerations listed in 10 CFR 2.790(b)(4). The information sought to be withheld shall be incorporated as far as possible into a separste part of the affidavit. If we do not hear from you in this regard within the specified periods noted above the report will be placed in the Public Docement Room. The telephone notification of your intent to request withholding, oc any request for an extension of the 10 day period which you believe necessary,'.ihould be made to the Supervisor Files, Mail and Records, USNRC Region I at (215) 337-5223.
s Should you have any questions concerning this inspection, we will be pleai,ed to discuss them with you.
Sincerely.
Bo e H. Grier Director
Enclosures:
- 1. Appendix A, Notice of Deviation
- 2. Office of Inspection and Enforcement Inspection Report Number 50-333/81-07 cc w/encls:
George T. Berry, President and Chief Operating Officer J. P. Bayne, Senior Vice President-Nuclear Generation A. Klausmann, Director. Quality Assurance M. C. Cosgrove, Site Quality Assurance Engineer J. F. Davis, Chainnan, Safety Review Comittee C. M. Pratt, Assistant General Counsel ,
G. M. Wilverding. Manager-Nuclear Licensing } l Public Document Room (PDR) t Local Public Document Room (LPDR)
Nuclear Safety Infomation Center (NSIC)
State of New York ,
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( EQ O e APPENDIX A NOTICE OF DEVIATION
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Power Authority of the State of New York Docket No. 50-333 I James A. FitzPatrick Nuclear Power Station License No. OPR-59 Based on the results of an NRC inspection conducted during the period March 1-31,1981, it appears that one of your activities was not conducted in confonnance with your commitments to the Comission as indicated below:
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A. Contrary to the licensee's comitments in Sections 5.2.3.5 and 7.3.4.3 of the FSAR the five Reactor Building Closed Loop Cooling Water System containment effluent lines have only single manual isolation valves outside containmr.nt instead of one valve which closes automatically by process action (i.e., reverse flow) or by remote manual operation from the control room.
l The Power Authority of the State of .New York is hereby requested to submit to this office within twenty-five days of the date of this Notice, a written statement or explanation in reply including the corrective steps which have been taken and results achieved (or corrective steps that are planned), and the date when corrective action will be completed.
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U.S. NUCLEAR REGULATORY COM1!SSION h DCS NUMBERS T6333-810117
- OFFICE OF INSPECTION AND ENFORCEMENT 50333-810209 50333-810210 Region ! 00333-810213 81-07
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- 50333-810219 ,
Report No. 50333-810220 50333-810225 { i Docket No. 50-333 /%/ l License No. OPR-59 Priority
-- g%Cate
Facility Name: James A. FitzPatrick Nuclear Power Station Inspection at: Scriba, New York Inspection conducted: arch 1,1981 - March 31,1981 Inspectors: (. h LinWile Ras M
ht Inspector f /
date st%ned rs L.Doerf1 pin,ResidentInspector Clt(/El
( date signed
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date sig ed Approved by:
H' B. KisterMhtef Reactor vrojects
"/ / /
date signed Section 1C t
Inspection Sumary:
Inspection on March 1,1981-March 31,1981 (Report No. 50-333/81-07)
Areas Inspected: Routine inspection by the Restcent Inspectors l108 hours) of licensee action on previous inspection items; review of Licensee Event Reports; licensee action on TMI Task Action Plan Items; licensee acti,n on IE Bulletins and Circulars; control room inspections; Log and Record review; plant tours; witnessing surveillance tests
, and observation of Maintenance activities.
Results: Of the nine areas inspected no items of noncompliance were noted in eight areas. tour items of noncompliance were identified in one area. (Failure to get prior NRC approval when making a change involving Technical Specifications, Failure to develo a written nuclear safety evaluation as required by 10CFR50.59(b), Failure of PORC to review a change in a plant system affecting nuclear safety, Failure to document the basis for accepting a condition adverse to quality). One Deviation from a licensee commitment was identified in one area (Failure to install an isolation valve capable of remote manual isolation from the control room on the RBCLCW system effluent lines from containment).
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b DETAILS
- 1. Persons Contacted
- J. Bayne, Senior Vice President, Nuclear Generation
- A. Klausmann Vice President. Quality Assurance
- B.. Diest, Manager-Nuclear Operations
- R. Baker,' Superintendent of Power N. Brosee. Maintenance Superintendent
- R. Burns, Assistant to Superintendent of Power
- V. Childs Assistant to Resident Manager
- R. Converse. Operations Superintendent
- M. Cosgrove, Site Quality Assurance Engineer W. Fernandez. Technical Services Superintendent H. Kieth Instrument and Control Superintendent E. Mulcahey, Radiological and Environmental Services Superintendent
., C. Orogvany Reactor Analyst Supeivisor .,
- R. Pasternak. ' Resident Manager -
D. Tall, Training Coordinator ,
The inspectors also interviewed other licenses personnel during this inspection including Shift Supervisors, Administrative. Operators, Health Physics, Security, Instrument and Control, Maintenance and Contractor Personnel. ,
- Denotes those present at' an exit interview.
., 2. Licensee Action On Previous Inspection Findings V (Closed) ' UNRESOLVED ITEM (333/80-O'8-02) In response to NUREG 0737 Item I.C.6, the licensee issued Work Activities Control Procedurs 10.1.2, Equipment and Personnel Protective Tagging, Revision 3 dated December 18, 1980. This requires that tags be second checked and should provide added assurance that tagouts are adequate for the scope of work.
- (Closed) UNRESOLVED ITEM (333/80-12-01) The licensee issued Operations Department Standing Order No. 4. Shift Relief and Log Keeping, Reviston 7, dated October 28, 1980, to change the title of the Equipment Status Log to the Auxiliary Operator's Log.
l (Closed) UNRESOLVED ITEM (333/80-15-05) The licensee submitted a followup 1
report to LER 80-050 which provided the general results of the Type B and C tests conducted during the 1980 refueling outage. The inspector reviewed N -
the results of the Type B and C tests and had no further questions.
(Closed) UNRESOLVED ITEM (333/80-21-03) Since the licensee transferred 5
the seventh STA who did net hold a bachelor's degree to another department,
, it is no longer necessary to define the equivalency.
(Closed) UNRESOLVED ITEM (333/80-21-04) The licensee issued Operations 7 Department Standing Order No. 4. Shift Relief and Log Keeping, Revision 7 dated October 28, 1980, which changed the title of the Equipment Status Log to the Auxiliary Operator's Log. In addition, the licensee issued
- Instrument and Control Department Standing Order No. 6 Instrument and Control Work Activities Control Log Book, Revision 0, dated December 1980,
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and Radiological and Environmental Services Department Standing Order No. 2 Maintenance of Radiation Protection Records, Revision 1, dated l
, December 19, 1980, which provide departmental equipment status logs for shift turnover purposes. -
1 (Closed) UNRESOLVEDITEM(333/80-21-05) The licensee issued Revision 5 l' to Operations Department Standing Order (00$0) No.1, Operating Staf' Responsibilities and Authorities, dated December 9,19M deleting tne ambiguity regarding the authority of the person in the control room during emergency conditions.
(Closed) UNRESOLVEDITEM(333/80-21-06) The licensee revised the Emergency Plan and Procedures, Revision 6, dated December 12, 1980 and issued procedure no. CLI-20 MSA Portable Combustible Gas Indicator, Revision 0, dated December 19. 1980 to make interim provisions for drawing a containment atmosphere sample during accident conditions assuming that the secondary containment is accessible for some period of time.
, (Closed) Severity Level V folation (Supplement 1)(333/80-21-07) The licensee revised Preop 02A, Revision 0, dated December 10,1980, deleting the requirement for a test of the SRV acoustic monitors at 100 percent power. In addition, the licenses will revise the Work Activities Control Procedure for the Control of Modifications by April 1,1981. 1 (Closed) UNRESOLVEDITEM(333/80-21-09) The licensee made previsions for mdving essential personnel to the control room if the technical
/y support center becomes . uninhabitable during accident conditions. This change is documented.in the Emergency Plan and Procedures, Revision 6, dated December 12,.1980.
l (Closed) UNRESOLVED ITEM (333/80-21-101 The licensee issued Emergency Plan and Procedures, Revision 6 dated December 12,1980, which describes the Operational Support Center and its use. ,
(0 pen) UNRESOLVED ITEM (333/80-21-11) The licensee is making provisions to perfom a helium leak test of the sample system and the standby gas treatment system during the October 1981' refueling outage. The licensee r issued work request 10516 and purchase requisition 22412 to provide for 2- the completion of this work. This item will remain open pending the result
- $ of these tests and their subsequent incorporation into the leak reduction program.
(0 pen) UNRESOLVED ITEM (333/80-BC-01) This item which was identified in inspection report 80-15 remains open pending completion of modification of V the containment vent and purge valve Icgic such that at least one of the r automatic safety injection actuation signals is uninhibited and operable to initiate valve closure when any other isolation signal may be blocked, reset or overridden. The licensee stated in their January 9,1981 letter to NRR that this modification will be completed by the end of the next refueling outage scheduled for October 1981.
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(0 pen) UNRESOLVED ITEM (333/81-02-04) The licensee completed modification revents the outside containment isolation valves for the no. F1-81-02 drywell floor andwhich eq p'uipment drain sump Ifnes from automatically ope reset of a containment isolation signal if their control switches are not in the closed position. In addition, the licensee revised his response to '
I NUREG 0578, itap 2.1.4 to indicate that it is his position that it is not I
necessary that automatic isolation be provided for non-essential systems which are closed within the reactor containment like the Reactor Building Closed Loop Cooling Water System. 'This item remains unresolved pending NRR's rapponse to the licensee's letterz,
- 3. In Office Review Of Licensee Event Reports (LER's) l The inspector reviewed LER's to verify that the details of the events were l
clear.ly reported. The inspector determined that reporting requirements had been met, the report was adequate to assess the event, the cause appeared accurate and was supported by details, corrective actions appeared appropriate to correct the cause, the form was complete and generic applicability to other plants was not in question.
LER's81-018, 81-019*, 81-020*,81-021, 81-022, and 81-023 were reviewed.
- Reports selected for onsite followup.
No items of noncompliance wem identified.
- 4. Licensee' Action On TMI Task Action plan Items Item I. A.1.3 'STA Trained Per LL CAT 8 The licensee response to these requirements dated January 8,1981 indicates that the STA's received the training required by the NRR staff's. October 30, 1979 letter. The program correlates well with the proposed INPO program found ,in Appendix C, paragraph 6.2 to 6.4 and 6.6 to 6.8 to NUREG 0737.
However, the Ifeensee's response also indicated that the STA's had completed 200 hourt of programmed on the job training. Through interviews with four STA's the inspector determined that two STA's almost completed and the other four STA's had just started when the )rogram was curtailed because of outage demands during the summer of 1980. Tw licensee had already identified this i error and had taken action to revise his response in a letter dated March 13, 1981 to indicate that this training was only partially completed.
Item I.A.1.3.1 Shift Mannino i License Plant Standing Order No. 26. Revision 1, Overtime Policy, Correlates well with the criteria in NUREG 0737, pages 3-6 and 3-7. Revision 0 of the procedure was issued on October 22, 1980 compared to an implementation date of November 1,1080.
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Item I.A.2.1.4 Innediate Upgrading of RO and SR0 Training and Qualifications The licensee issued Revision 2 of Indoctrination and Training Procedure No. 4, !
Licensed Operator / Senior Operator Replacement Training, dated August 8,1980.
The procedure includes the subject areas listed in paragraph A.2.C of Enclosure.1 to Harold Denton's letter dated March 28, 1980. The procedure was approved September 10,1980 compared to the 1eptementation date of August 1, 1980. To date, the licensee has not developed the lesson plans on the use of installed plant s stems to mitigate accidents involvin a
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several damaged core which sho d have been lated by January 1, 981 accord ng to item !!.B.4.1 of NURES 0737. The icensee requested a waiver i of this training requirement for the last R0 and SRO which received licenses !
in February 1981, and stated in that request that the training would be '
completed by February 1981. In a letter dated March 13, 1981 the licensee requested that development of this training be delayed until April 1,1981 and that im lamentation be delayed until af.ter completion of the current -
simulator se of the requalification training in late April 1981. This item will rtmain unresolved until the licenses has developed his training program on'this subject area.
Item I.C.5 ' Feedback Of Operatino Experience Licensee Plant Standing Order No. 28. Revision 0, Operating Experience Feedback, identifies responsibilities for review, feedback, and incorporation into training programs of bperating experience information. It also requires
-n audits of the program by the Quality Assurance department. The procedure was dated December 29, 1980 compared to an implementation date.of January 1,1981 Item I.C.6 Verify Correct Performance Of Operating Activities Licensee Work Activities Control Procedure No.10.1.2, Revision 3 Equipment and Personnel Protective Tagging. provides the recuired assurance,that equipment is properly released fo,r maintenance anc maintenance. However, the licensee did not addressreturned-to-service the requirement for from providing similar verification for surveillance testing or jumper / bypass control. Consequently, the licensee prepared, issued and implemented Operations
{- Department Standing Order No.16. Verification of Correct Performance of Operating Activities, Revision 0, dated March 26,1981 to cover these areas.
Item'II.K.3.22.A RCIC Suction, Licensee Operating Procedure 19. Revision 5. Reactor Core Isolation Cooling System (RCIC), provides clear guidance for manual shift of RCIC suction from the condensate storage tank to the suppression pool. The procedure was dated December 18, 1980 compared to an implementation date of January 1,1981.
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, 5. Licensee Action On IE Bulletins and Circulars f
- a. The inspector verified that for the IE Bulletins listed below the i Itcensee's written response was provided within the time period stated I in the Bulletin, included the information required to be reported, !
included adequate corrective action comitments based on infonnation Presented in the Bulletin and was accurate. The inspector further verified that any corrective action taken by the licensee was as i described in the response. '
The following Bulletins were closed out:
i IES 80-02. Inadequate Quality Assurance For Nuclear Supplied )
Equipment l l
IES 80-13. Cracking In Core Spray Spargers The following Bulletins will remain open for the reasons indicated below:
IEB 80-08. Examination of Containment Liner Penetration Welds The licensee committed to update his response dated July 3, 1980 within thirty days. This submittal was not made. The licensee stated that the update will be submitted by May 1, 1981.
IEB 80-25, Operating Problems With Target Rock Safety - Relief
Valves at BWR's This Bulletin remains open pending completion of tha modification
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to the pneumatic supply system durin comitted in the licensee response. g the 1981 refueling outage as IEB 81-01. Surveillance of Mechanical Snubbers This Bulletin will remain open pending completion of the inspection program for the 15 Pacific Scientific mechanical snubbers during the 1981 refueling outage and submission of the report within 60 days of its completion.
- 6. Reactor Building Closed Loop Cooling Water System Containment Isolation Yalves During followup on IE Bulletin 80-24. Prevention of Damage Due to Water Leakage Inside Containment (October 17, 1980 Indian Point 2 Event) in January 1981 the inspector noted that the five Reactor Building Closed Loop Cooling Water System containment effluent lines have only single manual isolation valves outside containment. Since this is not consistent with the requirements of General Design Criterion 57 which requires that containment penetrations of this type have a single isolation valve outside containment capable of remote manual isolation, the inspector reviewed the licensee's FSAR commitments on this matter.
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Paragraph 5.2.3.5, Primary Containment Isolation, on page 5.2-9 of Supplement 5 to the FSAR states in part, " Valves on lines that penetrate the primary containment but do not courunicate with the reactor vessel, with the primary containment free space, or with the environs require only)one flow or byvalve remote which closes manual automatically operation from theby process control action room." (i.e.,
Page reverse 5.2-10 cf' Supplement 5 further states, " Lines, such as those of the Reactor Building Closed Loop Cooling Water 5) stem, which do not connect to the Reactor Coolant Pressure Boundary or open into the primary containment air space are provided with at least one a-c powered valve on the effluent line and a check valve on the influent line." j l
Paragraph 7.3.4.3, Physical Arrangement, on pages 73-75 of Supplement 13 states in part " Process lines that penetrate the primary containment but do not comunicate directly with the reactor vessel, the primary containment frere space or the environs have at least one Group C isolation valve located outside the primary containment which may close either by process action (reverse flow) or by remote manual operation.
The failure to provide isolation valves" capable of remote manu'al operation
' from the control rocm on the Reactor Building Closed' Loop Cooling Water System effluent lines from the containment is considered to be a deviation
, from the licensee's comitments to the NRC in the FSAR (333/81-07-01).
- 7. Replacement of.G ' Safety Relief Valve (SRV)
After reviewing.the licensee's. response to IE. Bulletin 80-25, the inspector asked the licensee why the response did not address the replacement of G SRY !
which occurred after it failed to lift in response to a manual open signal following a scram on January 17,1981. The licensee stated that this was not {;
necessary since he determined that the failure was caused by the excessive I use of Loc-tite on the solenoid assembly as described in.!E Inspection Report I 50-333/81-02. In order to verify that there were no other contributors to the failure of G SRV the inspector reviewed work request package no.12301 associateo with the failure of the G SRV.. During this review the inspector {
determined that the licensee had removed SRY serial no.1012 set at 1140 I l
Work Tracking Form psig)and (WTF No. installed SRVControl 4 and Quality serial Inspection no.1013 set at 1090 Report (QCIR psig).
No. F 81-0021, step 2.6.5. clearly indicated that this had been done. However, step 2.F.2 of ,
- QCIR F 81-0021 indicated that the required setpoint for the G SRV was 1140 psig. Although the QCIR gave no indication of the reason for this discrepancy the Quality Control Supervisor stated that it was his understanding that the discrepancy had been reviewed and approved by the Plant Operations Review
.mittee (PORC). The inspector was unable to find any reference to this in the 1981 PORC minutes. Paragraph 6.3.(E)4 of Technical Specification states that it is a responsibility of the PORC to " Review proposed changes or modifications to plant systems or equipment that affect nuclear safety."
Failure to review the G SRV setpoint change is an item of noncompliance.
(333/81-07-02).
The inspector reviewed the Target Rock Quality Control Data for SRV's with serial no's.1012 and 1013 and determined that their setpoints were 11a0 psig and 1090 psig respectively.
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8 Paragraph 2.2.8 of Technical Specifications states that " Reactor Coolant System safety / relief valve nominal settings shall be as follows:
Safety / Relief' Valves 2 valves at 1090 ps g 2 valves at 1105 ps g .
7 valves at 1140 ps g The allowable setpoint error for each safety / relief valve shall be i 1 percent." By replacing G SRV which had been set at 1140 psig with an SRV set at 1090,psig, the'Itcensee was no longer in compliance with the combination required by this specified Limiting Safety System settipg.
In subsequent discussions the licensee stated that they considered that this change was safe because the lower setpoint was more conservative.
Since the licenses had purchased only one spare SRV they chose the lowest pressure setting under.the impression that it could be used conservatively in place of a valve with.any of the three settings. Consequently, the SRV was replaced without any formal safety evaluation. A review of the safety evaluation forJmendment 43 to the license indicated in paragraph 4 that the specified combination of SRV setpoints "will assure that the analysis.
of the containment structure for the effects of multiple consecutive relief valve actuations satisfies.the structural. acceptance criteria set forth in the Mark I Short Tenn Program," 10CFR50.59 (a) (1) states in part, "The holder of a license authorizing operation of a production or utilization facility may (t) make changes in the facility as described in the safety analysis report, ... without prior Commission approval unless the proposed change, test or experiment involves a change in the technical specifications incorporated in the license or an unreviewed safety question." Based on
- the discussion 'above, it appears that the change in the setpoint.for G SRV 1 involved a chan This is an item of g noncompliance33/81-07-03). (ge in the technical specifications.
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Although the licensee stated that the change in setpoint of G SRV was discussed in meetings identified as outage or PORC or department head meetings, there is no written record of these discussions. Licensee p
c Quality Control personnel stated that these undocumented discussions described above were the basis for accepting the change in G SRY setpoint
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- without question. They also did not doct. ment this basis on QCIR F1-81-0021.
10CFR50, Appendix B, Criterion XVI, states in part "The identification of i
^ the significant condition adverse to 'tuality, the cause of the condition, and the corrective action taken shall be documented and reported to f appropriate levels.of management." The Quality Assurance Program, Section 16, paragraph 2.5, states in part, " Records shall be maintained to show objective evidence that when conditions adverse to quality have been
- identified, action has been taken to correct the existing conditions ...
These records shall document the instructions for corrective action and the verification that corrective measures have been satisfactorily accomplished." Quality Assurance Procedure 10.1 states in part:
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"6.7 If a deficiency is found anytime during an activity, the QC
- inspector shall fill out a Deficiency and Corrective Action j Report and submit it to the QCS or his designee for approval M and processing... the Deficiency and corrective Action Report L
shall be referenced in the applicable QCIR.,
!' 6.8 When a process on a' safety-related system fails to meet established criteria due to noncompliance with specifications, procedures or drawings, unsatisfactory workmanship or deviation from operational
[ - standards, the QC inspector shall imediately notify the QCS or ,
, his designes.
6.9 The QCS or his designes shall notify the supervisor responsible for
- the activity being perfomed. If the deficiency cannot be resolved ;
} at this level the SQAE or his designee shall initiate stop work action...."
[ Failure to initiate a Deficiency and Corrective Action Report and failure i of the Quality Control Supervisor to doctanent his basis for accepting this condition adverse to quality on QCIR No. F81-0021 is an item of j noncompliance (333/81-07-04).
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L In addition, the licensee maintenance supervisor stated that the undocumented
' discussions discussed above were the basis for his conclusion which he documented on WTF No. 4 that the change in 6 SRY setpoint was not a
[ modification and therefore required no safety evaluation.
,' After discussions betweenihe inspector, NRC IE Region I Management, and the Licensee on March 25,1981, the License's performed a safety evaluation i
of the change in G SRV setpoint and submitted a proposed change in the technical specification SRY setpoint combination. Based on the completion status of ' torus modifications, the Licensee action and NRC NRR concurrence, 1
NRC IE Management concluded that continued operation of the facility was
! justified. -
- 8. Control Room Observations a.
Using a plant specific checklist, the inspectors verified selected plant i parameters and equipment availability to ensure compliance.with the i limiting conditions of operations of the plant Technical Specifications.
l Items checked included:.
Power distribution limits Availability and proper valve lineup of ESF systems Availability and proper alignment of onsite and offsite emergency power sources Reactor Control panel indications Primary Containment temperature and pressure t
Drywell to suppression chamber differential pressure
-- Standby Liquid Control Tank level and concentration Stack monitor recorder traces
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- 5. The inspectors directly observed the following plant operations to ensure adherence to approved procedures:
Routine power operations 1 Issuance of.RWP's and Work Re uest/ Event / Deficiency forw Nf7f ;
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Surveillance Tests .
Maintenance Activities
- c. Selected lit annunciators were discussed with control room operators to verify that the reasons for them were understood and corrective action, if required, was being taken.
The inspector had discussions with the licenses on reducing the number I of lit annunciators in the control room. On a day to day basis, there are approximately twenty lit annunciators in the control room. Some of these alarms are nomal for the existing plant conditions and therefore, i srovide little infonnation or have no meaning. Although the licensee 1as shown rome improvement in this area, the inspector will continue to review and evaluate the licensee's efforts during future inspections.
- d. Shift turnovers were observed weekly to ensure proper control room and shift manning on both day and back shifts. Shift turnover checklists
, and log review by the oncoming and offgoing shifts were also observed by the inspector. .
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- e. Shift Logs and Operating Records (1) Selected shift logs and operating records were reviewed to:
Obtain infomation on plant problems and' operations Detect changes and trends in performance
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Detect possible conflicts with Technical Specifications g or regulatory requirements Determine that records are being maintained and reviewed as required Assess the effectiveness of the communications provided by the logs
. (2) The following logs and records were reviewed:
Shift Supervisor Log Nuclear Control Operator Log
-- Night Orders -
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Shift Turnover Check Sheet f Protective Tag Record Log
'k' -- Daily Instrument Checks
-- Daily Core Surveillance Checks Liquid Radwaste Discharge Log Gaseous and Particulate Sample Logs Weekly Chemistry Status Log No items of noncompliance were identified.
- 9. Plant Tours (a) During the inspection period, the inspector made observations and conducted tours of plant areas including the following:
Control Room Relay Room
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L ecw annum mesu mum.
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-- Reactor Building Turbine Building Diesel Generator Rooms
- Electric Bays "
-- Pumphouse - Screenwell Standby Gas Treatment Building
-- Battery Rooms Radwaste Building 4
-- Stack Cable Spreading Room Refuel Floor
- Crescent Rooms Torus Room Cabis Tunnel
- Drywell Entrance Area (b) During the plant tours.the inspector conducted a visual inspection of selected piping between containment and the isolation valves for leakage or leakage paths. This included verification that manual valves were shut, capped and locked when required and that motor operated or air operated valves were not mechanically blocked.
Other items verified during the plant tours included:
Proper completion and use of selected radiation work pemits Proper use of protective clothing and respirators Proper personnel monitoring practices Proper control of ignition sources and flamnable material Equipment tag outs in conformance with controls for removal of equipment from service ;
Normal security practices are being followed
- Piar.t housekeeping and cleanliness practices are in conformance with approved licensee programs No items of noncompliance were identified.
- 10. Surveillance Observations ,
1 The inspector observed portions of the surveillance procedures listed below to verify that the test instrumentation was properly calibrated, the redundant system or component was available for service, approved procedures were used, and the work was performed by qualified personnel.
-- F-ST-34B, Reactor Building Logic Systeur, Functional Test, Revision 2, !
dated February 28, 1980, conducted on March 5, 1981
-- F-ST-4C, HPCI Pump Operability Test, Revision 7. dated March 11, 1980, conducted on March 13, 1981 s
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-- F-ISP-77, Reactor Shroud Level Calibration, Revision 8, dated July 1979, y conducted on March 31,1981
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No items of noncompliance were identified.
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- 11. Maintenance observations The inspector observed portions of the maintenance activities Itsted below to verify that the redundant train was available for service, approved procedures were used, and the work, was performed by qualified personnel. l WR 11-821. Repair HPCI Steam Trap T-3 and RCIC Steam Trip T-6 perfonned on March 24,1981 WR 03-8907 Replacement of Select Switch For Rod 2603 perfonned March 31, 1981
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The i'nspector detemined that the HPCI and RCIC systems were capable of perfoming their intended functions in the intended manner while these traps were removed from service for repairs. However, the inspector .
questioned the licensee's basis for classifying them as QA Category II. j Licensee QC Personnel produced system equipment listed with a handwritten <
column identifying the QA classification and system drawings with handwritten markings identifying the QA classification boundaries. . , ,
In addition, they produced component data sheets prepared by General Physics Corporation identifying the QA classification. However, the equipment list for the HPCI and RCIC system did not list the traps in question and the boundary markings on the system drawing were unclesr.
Since the List of Safety-Related Structures Systees and Components in QAP2.1 , Quality Assurance Program Scope, Revision 5 dated March 28, 1980 indicates that the HPCI and RCIC systems are safety-related and the equipment list does not s cify otherwise for the traps in question, ,
j it' appears that the traps sho 1d have been classified as QA Category I.
In addition, paragraph 5.5.2 of QAP 2.1 requires that technical reviews using the Equipment List and the original specification be used to detemine the classification of components in safety-r 14ted structures and systems. l The licensee's practice of classifying components in safety-related systems '
and of doctanenting this classification is unresolved pending further review by the inspector (333/81-07-07).
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- 12. Unresolved Items Unresolved items are matters about which more infomation is required in
-o6er to ascertain whether they a're acceptable items, items of non'ecmpliance,
'or deviations. The unresolved item identified during this inspection is discussed in paragraph 11.
- 13. Exit Interview ,
At periodic intervals during the course of thi,s inspection, meetings were held with senior facility management to discuss inspection scope and findings. ;
On March 17, 1981, and April 6, 198.1, the inspectors met with licensee l representatives (denoted in paragraph 1) and summarized the scope and findings of the inspection as they are detailed in this report. During the latter meeting the unresolved item was discussed.
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ENCLOSURE-(2) TO JAFP-89-0689 Response to Notice of' Violation 2.0 NOTICE OF VIOLATION As n' result-of the inspection-conducted on May 1 through May 26, 1989, and in'accordance with the " General Statement of Policy and Procedure for NRC Enforcement Action," 10 CFR Part 2, Appendix C, 53. Fed. Reg. 40019 (October 13, 1988),
the following violation was identified:
2.0 10CFR 50 Appendix B Criterion V requires that activities affecting quality shall be prescribed by procedures appropriate to the circumstances.
Contrary to the above:
(2.1) On May 22, 1989 the Emergency Diesa! Generator Day tank level calibration procedure F-IMP-93.5 Revision 2, dated October 2, 1985 and entitled, Fuel Oil Day Tank Level Functional Test did not appropriately address the calibration of the level switches to a standard.
(2.2) On May 22, 1989 procedure No. F-OP-22, Revision 16, dated December 12, 1988, Diesel Generator Emergency Power was not prescribed by procedures appropriate to the circumstance in that instructions for the connecting of the back up air bank did not address the isolation of the defective bank of starting air.
(2.3) Operations Department Standing Order ODS0-17
" Auxiliary Operator Plant Tour and Operator Logs", Revision 6, was not prescribed by procedures appropriate to the circumstance in that the minimum acceptable voltage of 90 Volts for the Class 1E 125V battery could render the equipment inoperable.
(2.4) Annunciator Procedure ARP-09-8-4-11, Revision 2, EDG B engine trouble or shutdown and ARP-09-8-4-4 Revision 1, EDG B fuel tank level or transfer pump switch off normal procedures were not prescribed by appropriate procedures in that no specific directions were given to the operator to respond to the abnormality.
This is a Severity Level IV Violation.
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ENCLOSURE (2) TO JAFP-89-0689
. . Response to Notice of Violation 2.0 l (Continued) l RESPONSE TO NOTICE OF VIOLATION The Authority disagrees with the' conclusion that the procedures IMP-93.6, OP-22, ARP-09-8-4-11, and ARP-09-8-4-4 (identified in items 2.1, 2.2, and 2.4) are inadequate. The Authority believes that these procedures do meet the requirements of 10CFR 50 Appendix B, Criterion V, in that I
these procedures are appropriate to the circumstances for which they are required. A discussion of the specific concerns for each procedure is presented below:
(2.1) The calibration instructions in IMP-93.6 were based on the recommendations of the equipment manufacturer. These l recommendations did not include the use of a standard, so the procedure did not include this requirement.
(2.2) As noted in the inspection report, procedure OP-22 adequately addresses the normal transfer of the air reservoirs. It is not considered necessary for the procedure to address an abnormal plant condition, based on the known skills and training of the operating staff. In addition, each Emergency Diesel Generator (EDG) has redundant air reservoirs. The EDGs operate in two sets of two for redundancy. Moreover, each EDG can supply emergency power by itself.
(2.3) The use of 90 volts as the minimum acceptable voltage in ODSO-17 was an error apparently caused by association with the minimum 125V battery input to the plant UPS system, which is 90 volts.
(2.4) The current revisions of annunciator procedures ARP-09-8-4-11 and ARP-09-8-4-4 are considered to contain the appropriate level of direction. The Authority believes that the skills and training of the plant operating staff are sufficient to respond to abnormal conditions such as those l stated in the inspection report. l STEPS THAT HAVE BEEN TAKEN AND RESULTS ACHIEVED The following plant procedures have been revised to provide improved direction:
(2.1) Procedure IMP-93.6 was revised on June 21, 1989 to include the following:
- a. level instrument calibration to a standard.
- b. a three point calibration of the level indicating instruments l
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ENCLOSURE (2) TO JAFP-89-0689 Response to Notice of Violation 2.0 (Continued)
- c. a requirement to record the "as-left" conditions L d.' verification of level switch operability based on the-L correctly calibrated level instrument.
1 L This revision is considered to be an improvement of an acceptable condition.
(2.2) Procedure OP-22 was revised on May 26, 1989 to include specific operating instructions for shifting on-line reservoirs in both normal (with in-service reservoir pressure > 175 psig) and abnormal conditions (with in-service reservoir pressure (175 psig).
This revision is considered to be an improvement of an acceptable condition.
(2.3) Procedure ODS0-17 was revised on August 30, 1989 to reflect the correct minimum acceptable voltage of the 125VDC battery.
(2.4) Revisions to the subject annunciator procedures are presently-in development. The revised procedures, which will contain specific directions to respond to abnormal conditions, are scheduled for completion by December 31, 1989. These revisions are considered to be an improvement of an existing acceptable condition.
COMPLIANCE STATUS The Authority was in full compliance on August 30, 1989, upon approval of the revision of ODS0-17.
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ENCLOSURE (3) T0.JAFP-89-0689
. Response ~to Notice of Violation 3.0 NOTICE OF VIOLATION As a result.of.the inspection conducted on May 1 through May 26, 1989, and in accordance with the " General Statement of Policy and' Procedure for NRC Enforcement Action," 10 CFR Part 2, Appendix C, 53 Fed. Reg. 40019 (October 13,-1988),
the following violation was identified:
3.0 10CFR 50 Appendix B Criterion VI requires that measures be established to control the issuance of documents, such as drawings, which prescribe activities affecting quality to assure these documents are reviewed for adequacy, distributed and used at the location where'the activity is .
performed. NYPA QA Program Section 17.2.5 requires that activities affecting quality be prescribed by controlled drawings and accomplished with these drawings.
Contrary to the above:
( 3. I') Drawing 11825-FE-1AJ, Rev. 6, indicated circuit breaker ratings for loads 23P-141, 27MOV-122 and 23MOV-25 as 20, 20 and 30 A, respectively. However, on May 22, 1989 field installations for these loads were 15, 30 and 40 A respectively.
(3.2) Drawing 11825-FE-1AJ, Rev. 6 indicated motor horsepower ratings for loads 23P-141 and 27MOV-122 as 1.0 and 0.66 hp, respectively. However, on May 22, 1989 field installations for'these loads were 1.3 and 2.89 hp, respectively.
l (3.3) Drawing 11825-FE-IS, Rev. 15 indicated motor horsepower rating for loads 10MOV-01,;B, 10MOV-089B and 10MOV-026B as 1.3, 4.0 and 3.9 hp, respectively. However, on May 22, 1989 field installations for these loads were 1.6, 1.0 and 2.0 hp, respectively.
This is a Severity Level IV Violation.
RESPONSE TO NOTICE OF VIOLATION )
The Authority agrees with this violation. The fundamental causes of this violation are as follows:
(3.1) The cause of the discrepancy in brecker rating for 23P-141 was an error in the field verification of the design drawing during original plant construction.
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L ENCLOSURE'(3') TO JAFP-89-0689 a
Response to Notice of Violation 3.0 (Continued)
- The cause of the discrepancies in breaker ratings for 27MOV-122 and 23MOV-25 was an inadvertent exchange of two MCC draw-out cubicles. This error occurred prior to 1978 and had gone' unnoticed during the installation of new valve and actuator 27MOV-122 by plant modification F1-80-028 in 1981.
(3.2) The cause of the discrepancy in the horsepower rating of 23P-141 was an error ~in the as-built verification of the design drawing during original plant construction.
The cause of the discrepancy in the horsepower rating for 27MOV-122_was an error in updating the drawing as a result of modification F1-80-028.
(3.3) The cause of the discrepancies in_the horsepower ratings of 10MOV-12B and 10MOV-89B was an error in the ,
as-built verification of the design drawing during original plant construction.
The apparent discrepancy in the horsepower rating of 10MOV-026B was due to a revision in progress as a result of a recently installed modification, F1-87-133. This modification replaced the existing 10MOV-26B valve and actuator. An as-built drawing had been issued to. revise the.
-drawing in accordance with the modification as indicated in
'the " Revision in Progress" listing for controlled drawings.
The process used to update the drawing is in accordance with procedure WACP 10.1.9, Control of Plant Drawings, Cable List,-Raceway List Piping Line Designation Tables.
CORRECTIVE STEPS TAKEN AND RESULTS ACHIEVED (3.1) Work Requests have been initiated to replace the three circuit breakers identified in item 3.1 with the size specified by the plant drawings. This work is scheduled for the 1989 fall maintenance outage, tentatively scheduled for completion in early October, 1989.
(3.2) A Drawing Change Request was issued on May 11, 1989 to correct drawing 11825-FE-1AJ.
(3.3)-A Drawing Change Request was issued on May 25, 1989 to correct drawing 11825-FE-1S.
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ENCLOSURE (3) TO,JAFP-89-0689
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Response to Notice of Violation 3.0 (Continued).
In addition, a-comprehensive' field verification walkdown of all AC and.DC Motor Control Centers has been initiated to verify the accuracy of the electrical one-line and MCC layout drawings. At present, this task is 20%' complete. A review of modification F1-80-028 has.been initiated to verify that all plant drawings affected by this modification have been properly updated. Both of these actions are scheduled for completion by December 31, 1989.
CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATION The Authority believes that the present administrative controls for revisions to drawings as a result of modifi-cations are appropriate. Because each of-the subject-
' drawing discrepancies occurred during construction, or as the result of an early plant modification, no further corrective action.is planned.
COMPLIANCE STATUS The Authority will be in full compliance by the end of the 1989 fall maintenance outage, tentatively scheduled for early October, 1989, upon installation.of the three circuit breakers and issuance of the revised drawings.
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' ENCLOSURE (4) TO JAFP-89-0689
-Response to Notice of Violation 4.0 NOTICE OF VIOLATION As a result of the inspection conducted on May 1 through May 26, 1989, and in accordance with the " General Statement of Policy and Procedure for NRC Enforcement Action", 10 CFR Part 2, Appendix C, 53 Fed. Reg. 40019 (October 13, 1988),
the following violation was-identified:
4.0
~
FitzPatrick Technical Specification Section 6.8A requires that written procedures be established, implemented and maintained that meet or exceed the requirements and rec-commendations of ANSI N18.7-1972 and Appendix A of NRC Regulatory Guide (RG) 1.33. ANSI N18.7-1972, Section 6.3.6 requires that procedures be provided for periodic cali-bration and testing of safety related alarm devices, sensors and protective circuits.
Contrary to the above:
(4.1) On May 22, 1989, Class 1E 125 Vdc circuit breakers and battery charger voltage sensing relays had not been subject-ed to periodic testing and calibration as required.
(4.2) On May 22, 1989, no periodic testing was done on the air system check valve for the reactor building closed cooling water isolation valve.
This is.a Severity Level IV Violation.
RESPONSE TO NOTICE OF VIOLATION The Authority agrees with this violation. The fundamental causes of this violation are as follows:
(4.1) The Authority established a formal periodic maintenance and testing program in 1984. The Authority has been systematically adding components to this program since that time. The 125 Vdc circuit breakers had been identified for inclusion in the program prior to the inspection, however, actual testing had not begun.
The battery charger voltage sensing relay was inadvertently omitted from the list of components which require periodic calibration. This list was established using the plant 1
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ENCLOSURE (4) TO'JAFP-89-0689
' Response to Notice of Violation 4.0 (Continued)
Technical Specifications and the Master Equipment' List (MEL). The subject relay, classified as a subcomponent of the battery charger (and thus not having a unique component identifier), was not listed separately in the MEL.
Therefore, it was not included in the calibration list.
(4.2) As noted in the inspection report (paragraph 4.7.5),
the Authority has been- performing surveillance testing of the backup air supplies for the subject isolation valves.
The isolation valves were installed under Plant Modification No. F1-81-026 in 1983. The current surveillance test was based on the preoperational test from this modification.
The modification did not adequately address the safety significance or function of the air inlet check valve.
Consequently, neither the preoperational test nor the subsequent surveillance test fully demonstrated the valves' ability to function.
CORRECTIVE STEPS THAT HAVE BEEN TAKEN AND RESULTS ACHIEVED (4.1) The Authority has implemented a program for periodic testing of the Class IE 125 Vdc circuit breakers. The
. testing will begin in the 1989 fall maintenance outage and will be performed in accordance with existing plant procedure MP-200.16, Maintenance and Subcomponent Replace-ment of GE 7700 Series DC MCCs (BMCC). One DC motor control center has been selected for complete testing during the outage. Routine testing of other BMCCs will be conducted .,
during future outages.
To address the calibration of the battery charger voltage ,
sensing relays, the Authority has prepared procedure IMP 71.23, Station Battery Charger 71BC-1A(IB) Undervoltage and Ground Alarms, which was approved on 6/21/89. This procedure requires calibration of the relay every two years.
The calibration is scheduled for the 1989 fall maintenance outage.
(4.2) The Authority has developed procedure ST-1T, RBC Containment Isolation A0V Accumulator Check Valve Leak Test, to address the testing of the air system check valves for the Reactor Building closed cooling water isolation valves.
The procedure was approved on 8/30/89 and the testing will be performed during the 1989 fall maintenance outage.
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L . ENCLOSURE (4) TO JAFP-89-0689 l
L Response to Notice of Violation 4.0 l- (Continued) 1 QR_RECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIONS j (4.1) The Authority implemented a long term program for .
improving maintenance of all plant equipment in 1984. The i responsibilities of the Planned Maintenance Task Force I include validation of vendor technical manuals, review and evaluation of vendor maintenance recommendations, and ,
development of additional plant maintenance procedures as {
needed. The schedule and status of this program are provided to the Resident NRC Inspector on a quarterly basis.
This effort will identify any other components which require periodic maintenance and testing.
I The Authority will perform a review of all Control Room annunciators for safety-related equipment to verify that all sensing devices which require periodic calibration are included in the calibration program. This review will be completed by December 31, 1989.
(4.2) The deficiency identified in Modification-F1-81-026 is considered to be an isolated case, and no further corrective actions are planned.
COMPLIANCE STATUS (4.1) The Authority will be in full compliance by the end of the 1989 fall maintenance outage, upon implementation of periodic testing of the Class 1E 125Vdc circuit breakers, and calibration of the battery charger voltage sensing relay.
(4.2) The Authority vill be in full compliance by the end of the fall 1989 maintenance outage, upon completion of the testing of the air system check valves.
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r ENCLOSURE (5) TO JAFP-89-0689 Response to Notice of Violation ~5.0 NOTICE OF VIOLATION
'As a result of the inspection conducted on May 1 through May 26, 1989,'and in accordance with the " General Statement'of Policy and Procedure for NRC Enforcement Action," 10 CFR Part 2, Appendix C, 53 Fed. Reg. 40019 (October 13,,.1988),
the following violation was identified:
L 5.0 10CFR 50, Appendix B, Criterion III requires that measures shall be established to assure that applicable regulatory requirements and design bases are correctly translated into specifications and instructions.
Contrary to the above:
(5.1) On May 22, 1989, measures were not established to ensure that applicable design bases are correctly translated into the instruction for plugging the floor drains in the emergency diesel generator rooms in that plugging all the-floor drains left the potential for spreading an oil fire to adjacent diesel rooms.
(5.2) On May 22, 1989, measures were not established to ensure that applicable design bases are correctly translated into the instruction for the Heating Ventilation and Air Conditioning (HVAC) design for the switchgear enclosures in that the HVAC system would not have functioned during a high energy line break accident.
This is a Severity Level IV Violation.
RESPONSE TO NOTICE OF VIOLATION The Authority agrees with this violation. The fundamental causes of this violation are as follows:
(5.1) As noted in the inspection report, (paragraph 4.4.2.3) the plugging of the Emergency Diesel Generator (EDG) Room floor drains occurred in or about 1981. This action was taken in response to a concern raised by the New York State Department of Environmental Conservation with regard to the possible release of chromates in the EDG jacket cooling water to the environment. At that time, the Authority failed to recognize this activity as a modification and consequently did not perform a safety evaluation as required by 10CFR 50.59. Because a modification package was not initiated, specifications and instructions relative to this activity were not developed.
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g !i; ENCLOSURE (5) TO JAFP-89-0689 Response to Notice of Violation 5.0 (Continued)
(5.2) In designing the cooling system for the environmental
~
enclosures, the ability of the_ cooling units to function in the HELB environment was properly considered. The specification for the cooling system did addrese its required capacity and the environmental conditions under which it must operate. The equipment manufacturer proposed demonstrating the systems' ability,by testing at normal conditions and providing analysis for the HELB conditions.
The Authority concurred with this approach, however, it did not review or approve the vendor's analysis. As such, the Authority did not exercise adequate control in the review of the vendor's documentation.
CORRECTIVE STEPS THAT HAVE BEEN TAKEN AND RESULTS ACHIEVED (5.1)'In late 1988, the Authority initiated an evaluation of the plugged floor drains relative to equipment flooding, in response to an INPO Significant Operating Experience Report. At the same time, efforts were increased to obtain a permit from the New York State Department of Environmental Conservation allowing the opening of the floor drains. In response to the concerns of the SSFI inspection team, a calculation was performed to demonstrate that a fire in one room would not spread into an adjacent room. The floor drain plugs were removed in May 15, 1989.
(5.2) In response to the inquiries of the inspection team regarding the design of the cooling system for the ,
environmental enclosures, the Authority requested, reviewed !
and approved the manufacturer's analysis for the HELB environment. As noted in the inspection report, contact.
with the manufacturer by the Authority led to discovery of a design error which would have prevented automatic reset after a trip of the equipment. Immediate modifications were made to the equipment control logic to correct this deficiency. 1 Page _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ -
ENCLOSURE (5) TO JAFP-89-0689 Response to Notice of Violation'5.0 (Continued)
CORRECTIVE' STEPS'THAT WILL BE TAKEN TO PREVENT FURTHER VIOLATIONS (5.1) It is the--Authority's position that the'present administrative controls for work activities and L modifications at the James A. FitzPatrick Nuclear Power-F Plant are appropriate. In particular, Work Activity Control Procedure WACP 10.1.3, Control of Jumpers, Lifted Leads, and. Temporary Modifications, requires that temporary.
L modifications be reviewed and evaluated by a licensed h operator,.the. Technical Services Department and Plant Operations. Review Committee and that when required a. Nuclear
, Safety Evaluation be prepared in accordance with 10CFR50.59.
Further, the procedure specifically identifies the plugging of_ floor drains as a temporary modification. No further corrective action is planned.
(5.2) The Authority has recently developed an improved design control program, which is documented in the Design Control Manual. Procedure No. DCM-11, Control, Review,.
Comment and Acceptance of Vendor Documents, states the requirements for review of vendor documentation. The program is scheduled to be implemented at the James A.
FitzPatrick Nuclear Power Plant by December. 31,'1989.
COMPLIANCE STATUS (5.1) The Authority was in full compliance with this item on May 15, 1989, upon removal of the drain plugs.
(5.2) The Authority was in full compliance with this item on May 31, 1989, upon acceptance of the vendor's operability analysis and modification of the equipment control logic.
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