IR 05000348/1988019

Revised Pages 8-12 to Insp Repts 50-348/88-19 & 50-364/88-19 Correcting Grammatical Errors & Incorrect Paragraphs

ML20207H146
Person / Time
Site:	Farley
Issue date:	08/17/1988
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20207G834	List: IR 05000348/1988019 ML20207G830
References
50-348-88-19, 50-364-88-19, NUDOCS 8808240370
Download: ML20207H146 (6)

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8 IR50-348,364/88-19 Corrected 8/11/88 FNP-0-AP-1, Development, Review, and Approval of Plant Procedures, Section 5, Procedures Safety Evaluatio Section 5 of this administrative procedure related to safety evalua-tions required during development, review and approval of plant procedures and revisions to procedures. A Nuclear Safety Evaluation Check List is prepared during procedure pre)aration by the individual responsible for the preparation. This c1eck list is used to evaluated whether or not a 10 CFR 50.59 evaluatien' is require Also, the check list is used to determine whether or not PORC review and NRC approval is required prior to procedure implementatio Paragraph 5.2 states when a written description safety evaluation is neede Included within Background, References, Bases and Conclusion.the description However, there are the follow amplifying information in paragraph 5.2 other than the listing of these item As noted durin contain additional guidance,g the for especially exit subInterview, paragraphthe listing should o.2.3,

"Bases." Such guidance might include statements relating to why the change does not constitute an unreviewed safety question. The check list Part 8, "Safety Evaluation," paragraph 4.1 through 4.7, lists only the specific criteri Therefore, the support'ng safety evaluation should state clearly and concisely why the question 'yes or no" can be answere The licensee management stated that the comments would be considered, FNP-0-AP-8, Design Modification Control Section 5, Production Change Requests, Section 6.3, Safety Evaluation

Section 6.3 of this administrative procedure requires that all design changes shall have a safety evaluation check list completed to

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determine the 10 CFR applicability. The check list is very similar to the check list in FNP-0-AP-1 used for procedures. The check list is used for design change (PCNs) or minor departure Paragraph 6.3.2 requires safety evaluation check lists and safety evaluations to be provided by the design organization responsible for design development. However, no instruction are included on the detail necessary to show a clear basis of the determination that an unreviewed safety question exists. The procedure needs to ba brought us to the standards used in FNP-0-AP-1, and improved as described a)ove for FNP-0-AP- !

Licensee management stated that the comments would be considered, h PCN 84-2609, Upgrade of Primary Meteorological Tower Instrumentation A discussion of the PCN was conducted with the Plant Technical l Manager and the Supervisor of Environmental and Emergency Planning and an Evaluation Engineer of the Plant Modifications Grou The 8909240370 800017 PDR  :

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9 IR50-348,364/88-19 Corrected 8/11/88 modification is to improve the reliability of the existing meteorological data system as described in the FSAR Section 2. An older system of instrumentation is replaced with Met One instrument In addition, backup wind speed and direction instruments are added to the 150 foot level. Replacement instruments are to meet the requirements of Regulatory Guide 1.23 as well as the accuracy requirements of FSAR, Table 2.3-10. Revision 1 to PCN 84-2609, safety evaluation noted that addition of the backup instruments is to provide backup data for temperature differential between the 35 foot and 200 foot elevation. The safety evaluation check list notes that FSAR, Table 2.3-10 will require a chang PORC minutes for the 1522nd meeting held on May 13, 1986, were reviewe The PORC reviewed the design changes and safety evaluations associated with this change. The PORC determined that no unreviewed safety questions were involved and recommended approva The licensee noted that the FSAR will be updated in the annual updated scheduled for July 198 As noted in the Revision 0 to PCN 84-2609, required safety evaluation,3revent if the additional instrumentation is intended thea Technical to Specification change wou plant from entering an LCO upon failure of the one-channe' windspeed and direction at the 150 foot elevatio At the exit interview, the licensee was advised that Technical Specification Table 3.3-8 appears to need updating for consistency with the Air Temperature Difference Instrumen The licensee agreed to consider the need for these changes, d. PCN 86-32496, Erection of Solidification and Dewatering Facility An interview was conducted with the Plant Technical Manager, the Health Physics Manager, and Engineer of the Plant Modification Grou study The modifications performed are a major by Southern Company project Services, Inc. Thewithstudy, an engineering"Des! ,

Criteria for Solidification / Dewatering Building for Farley Nuclear Plant," dated October 31, 1985, was reviewed. In addition, PORC  !

meeting minutes of August 8,1986, and March 10, 1987, for design change Revisions 0 and 21 were reviewed. In each set of minutes, the

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PORC determined that no unreviewed safety question was involve !

Prior to this modification, the solidification and dewatering  !

process was performed in an open area between the refueling water storage tanks. Temporary connections were made each time the processes were performed. The new permanent structure (with all connections to reactor plant systems) allows the processes to be  ;

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performed more safely, more efficiently, and with improved radio- '

active protection measure :

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10 IR 50-348,164/88-19 Corrected 8/11/88 The building (49' wide, 82' long, 40' high) also provides storage during refueling (including studbolt storage), waste and decontamination equipment storage, solidification equipment storage, contaminated oil, paint, and scaffold storage, four spent resin liner pits with remote operating controls and view ports, and supercompaction equipment for dry active waste are housed in the new facilit Airborne radioactivity is processed and monitored in the Unit 1 Auxiliary Building HVAC System. The facility was designed against collapse due to wind, tornado and seismic loads combinations per Regulatory Guide 1.29, Revision 1, and the FSAR. Piping and valves were designed, fabricated, inspected, snd tested per ANSI 831.1 as required by Regulatory Guide 1.14', , hvision The "Safety Evaluation for Solidification /Deweering Facility for Farley Nuclear Plant (PCN 86-0-3496) Revision 3 dated August 1986, paragraph titled "Supplementary NRC Guidance Rev}ew", stated that the facility was reviewed against NRC guidance of IE Circular 80-18 and Generic Letter No. 81-38. Subparagraph C, "10 CFR Part 50" notes that the design and operation of the facility is in compliance with 10 CFR 50.59 and no unreviewed safety questions have been identifie Evaluation of potential exposures from direct radiation sources and potential radioactivity releases was given in subsequent paragraph The evaluation included discussions of potential liquid and gaseous events and concluded that:

(1) Consequences of a liquid spill are expected to be significantly loss than the open air process previously use '

(2) Consequence of dropping the resin filled liner with a subsequent airborne release will be less that the open-air configuration previously use (3) There is no significant potential for gaseous release due to fire or following a liquid spil l (4) Potential for a gaseous release will be less while filling the liner and will be minimized by use of the pressure blower taking suction from the closed pit and exhausting to the Unit 1 Auxiliary Building radwaste vent syste The evaluation goes on to state that in view of this evaluation, the facility is not a potential release pathway per the GDC-64 criteri The audit indicates that the licensee personnel who were interview 6d, as well as the procedures and safety evaluations performed, followed licensee guidance and were in conformance with the intent of 10 CFR 50.5 No items of concern were note '

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11 IR 50448,164/88-19 Corrected 8/11/88

PCN 887-0-4384, Replacement of Existing Commercial Grade Agastat Relay with Seismically Qualified Relay An interview was conducted with the Plant Technical Manager and Electrical Maintenance Department Group Supervisor. The General Office production change request PCR 87-0-4384, the Nuclear Safety Evaluation Checklists (Revisions 1 to 4) and the minutes for PORC Meeting No. 1711 aated August 4, 1987 were audite The initial PCN related to replacement of Agastat Model 7012 PA (commercial grade) in Unit 2 600V load center 2E, a safety related load center. Also, Agastat Model 7022 (commercial grade) relays throughout the plant were evaluated for replacement. The Nuclear Safety Evaluatioa Checklist, Revision 1, indicates that a change to the plant as described in the FSAR is "yes " For this reason, a FSAR Figure 8.3-10 and 8.3-13 safety (Unit 1 ) evaluation was performe required changin A figure will be added to the FSAR for ;

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Unit 2 specifying the correct relay mode The safety evaluation concludes that the new relay is qualified to a higher acceleration and when installed will not affect the load

center's original seismic qualification. One of the actual relays replaced was examine Licensee personnel advised that the relays

(commercial or seismically qualified) are identical to all intents i and purposo. The Nuclear Safety Evaluation Checklist for Revision 4

, to the PCN was reviewed for the replacement of other relays at 600V Emergency Load Center 1A,10, 2A, and 20. Similar safety conclusions ,

were mad No items of concern were noted requiring license) actio ; Procedure FNP-0-AP-76, Revision 4, Authorize Use of Morpholine / Boric

! Acid in Secondary Water Chemistry Control System I

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An interview was conducted with the Plant Technical Manager and the '

l Chemistry and Environmental Group Supervisor relating to secondar :

water chemistry control changes of Revision 4 "Conduct of Opera y i i

tions - Chemistry and Environmental Group." The asplicable safety i evaluation check list indicated "yes" to the question; "A change to the plant as described in the FSAR?" PORC meeting minutes No 1760 i dated November 3,1987, and the associated safety evaluation were :

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Licensed Condition 2.C(3)(g) requires the licensee to implement a l secondary water chemistry monitoring program. The program had been j

' inspected most recently on August 17-21, 1987 as described in NRC Inspection Report No. 87-21, dated September 16 1987. The inspector concluded at the time, that the licensee was, aware of concerns

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relating to maintaining the integrity of the primary coolant pressure boundary as well as the remainder of the secondary cooling syste l y l l

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12 IR50-948,$64/88-19 Corrected 8/11/88 i FSAR Section 10.3.5, Water Chemistry, contains a brief description of the secondary system water chemistry controls previously used to minimize corrosion of the steam generator (SG) internals. This FSAR '

section will require revision to describe the use of morphpoline/

boric acid chemistry control. The previous all volatile treatment

(AVT) used ammonium hydroxide for pH control and hydrazine for ox gen scavenging as recommended by the SG Owner's Group. More recent ,

the NSSS vendor had recommended the use of boric acid injection or control of SG tube dentin Boric acid was used on Unit I since.

l 1983, and on Unit 2 since 198 FNP-0-AP-76, Revision 4, reflects changes in details of the proce- '

! dure AVT, AVT/ morpholine, and AVT/ morpholine / boric chemistry specifications were added. It provided for a formal documentation i

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memorandum of the selected secondary chemistry program. The 10 CFR 50.59 safety evaluation Bases section discusses the safety risks. in

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handling morpholine, the environmental impact (Alabama Department of :

Environmental Management approved), corrosive effects of morpholine on slant components, contribution to the total organic carbon, effect on )1owdown demineralizers, effect on laboratory analysis for silica, and effect on in-line instruments. The conclusion is that implemen-tation of morpholine / boric acid treatment dose not constitute an unreviewed safety question as defined by 10 CFR 50.5 '

However, the safety evaluation dose not specifically address the three criteria for a determination whether of not NRC approval must '

be obtained before implementing the change. The safety evaluation of the licensee clearly determined that the use of morpholine was saf A Westinghouse Nuclear Safety Evaluation Check List SECL-87-501, in the audit material clearly concludes that "the previously analyzed consequences of excessive corrosion (e.g., tube rupture, feedline break, and turbine missiles) have not been increased nor has the probability of such postulated events been previously analyzed. The safety factors used in design evaluations of the components including the pressure boundary stress analysis done in accordance with the l

ASME Boiler and Pressure Vessel Code remain valid. Therefore, the '

margin of safety has not been reduced."

i During subsequent conversations with the licensee, use of evaluations ;

for safety and separate evaluations for 10 CFR 50.59 determinations were discussed. A review against the criteria in 10 CFR 50.59 does not determine that a change 's. safe but that the change does or does not require NRC approval prior to Im

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Both reviews and evaluations important and necessary.plementatio Training in this area for

personnel examining the 10 CFR 50.59 determinations does not exist today at Farley sit Consid: ration for such training was !

recommende ;

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12a IR 50448,164/88-19 -,

Corrected 8/11/88 {

1 Unit 1 S'tartup from Refueling (71711)

The inspectors verified that adequate administrative procedures were-

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available to assure that systems disturbed or tested during the refueling  ;

outage were returned to operable-status before plant startup. Accessible portions of the auxiliary feedwater system and chemical and volume control -

system were inspected to verify: valves were in correct Lalignment; hangers and supports were made up properly; . major components were-properly-labeled, lubricated, cooled and no visible leakage exits; breakers were properly aligned; instrumentation calibration _ dates were current; support '

systems essential to system performance were operational; and,. housekeeping and cleanliness were adequately maintained. Portions of 4 these systems had been disturbed during the outage, but based on this inspection these systems appeared to have been returned to service in- '

accordance with the applicable procedures. Refer to E information on auxiliary feedwater pump lubrication. paragraph 4.'b for No discrepancies were identifie '

Portions of the unit startup operations were witnessed by the resident-  ;

inspectors and a regional based inspector. The ins)ectors verified that  :

the required core-physics tests were performed anc that the startup '

activities were conducted in accordance with the TS requirements. Refer  !

to NRC report 348,364/88-20 for additional comments on this are !

No violations or deviatio,is were identifie j

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