IR 05000338/1979028
| ML19249E624 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/07/1979 |
| From: | Kellogg P, Kidd M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19249E605 | List: |
| References | |
| 50-338-79-28, 50-339-79-35, NUDOCS 7910020059 | |
| Download: ML19249E624 (14) | |
Text
.
/g>RhtGygo UNITED STATES y' '
,.,.y,/g NUCLEAR REGULATORY COMMISSION
-
REGION 11
'"
r
101 M ARIETT A sT., N.W., SUITE 3100 o,
ATLANT A, GEORGI A 30303
%
j
- ...+
Report Nos. 50-338/79-28 and 50-339/79-35 Licensee: Virginia Electric and Power Company Richmond, Virginia 23261 Facility Name: North Anna, Units 1 and 2 Docket Nos. 50-338 and 50-339 License Nos. NPF-4 and CPPR-78 Inspection at North Anna S;te near Mineral, Virginia and Surry Power Station near Surry, Vit inia
.
.
inspector:
fT.
r
/'
V<. / 9
.
M. S. Kidd re'
Date Signed
//
f!7!77 tpproved by:
/
/
P. J.
' CliEl, RONS Branch Da te' Signed Inspection on May 14 through June 15, 1979 Unit 1 Areas Inspected This routine inspection by the resident inspector involved 18 man-hours on-site and at the Surry Power Station simulator of previously identified open and noncompliance items, responses to IE Bulletins, licensed operator simulator training, contingency planning, licensee events, and plant tours.
Unit 1 iindings No apparent items of noncompliance or deviations were identified.
Unit 2 Areas Inspected This routine inspection by the resident inspector involved 26 man-hours on-site of previously identified open items, witnessing of a preoperational test, responses to IE Bulletins, comparison of installed systems to FSAR, licensee reports per 10 CFR 50.55(e) and 10 CFR 21, and plant tours.
N os O o
191
,,
.,
h
'V
.
Unit 2 Findings Of the six areas inspected, no apparent items of noncompliance or devia-tions were identified in four areas.
One apparent item of noncompliance was identified in one area (Infraction-Fzilure to transmit information on a potentially reportable matter to persons responsible for reviewing the matter for reportability in a timely manner as required by procedure-Paragraph 14). One apparent deviation from FSAR commitments was identi-fied in one area (restricting orifices installed in the discharge lines for the outside recirculation spray pumps limit flow to 3,000 gpm each versus the FSAR assumed flow of 3,640 gpm each-Paragraph 9.a).
.
,
.
%
rf
'
DETAILS 1.
Persons Contacted Licensee Employees
- W. R. Cartwright, Station Manager
- S. L. Harvey, Operating Supervisor
- J. D. Kellams, Superintendent, Station Operations D. G. McLain, Engineering Supervisor J. W. Ogren, Administrative Supervisor
- E. R. Smith, Technical Services Superintendent D. E. Thomas, Supervisor, Electrical Maintenance C. F. Wirecoff, Security Supervisor Other licensee employees contacted during this inspection included six licensed operators, three office personnel, and two QA staff members.
- Attended one or more exit interviews.
2.
Exit Interview The inspection scope and findings were summarized en May 25, June 8 and June 15, 1979 for those persons indicated in Paragraph I above. The apparent Infraction defined in paragraph 14 was discussed June 8, 1979.
Station management acknowledged the inspector's comments but stated that the event in question had been reviewed on-site and determined to be non-reportable. The apparent Deviation defined in paragraph 9.a was discussed June 15, 1979.
3.
Licensee Action on Previous Inspection Findings (Closed) Infraction (338/79-01-08): Signin; of Design Change Package Prior to Completion. Corrective and preventive measures discus!2d in the licensee's response of April 23, 1979 were verified to be complete for the subject Design Change NAl-DC-78-40.
Administrative Procedure 104.0 was revised April 24, 1979, to provide for recording all revisions on the package cover sheet. This will assure review of all revisions to assure completion prior to signing the package to denote completion. QC audit N-79-31 verified completion of all revisions of the package. This item is closed.
lov'c7 C
,
,u
-2-
'
4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Preoperational Test Witnessing-Unit 2 On May 31, 1979, the inspector observed performance testing of casing cooling pump 2-RS-P-3B per 2-PO-38.3, " Casing Cooling System".
It was observed that a current test procedure was in use and that a current system diagram was available. Selected valves were observed to be in their proper test positions. Valve lineups had been revised prior to starting the test via procedure Deviations; these appeared to be in order.
Preliminary data reduction revealed that the pump met its acceptance criteria of 1,000 gpm at 100 feet of discharge head. Final results will be reviewed as part of the normal inspection program. Within the areas observed, no items of noncompliance or deviations were observed.
6.
Test Engineer Qualifications-Unit 2 Education and work experience requirements for Unit 2 preoperational test These engineers are given in Section 14.0 of the Units 1 and 2 FSAR.
requirements are implemented by station administrative procedure (ADM)
102.0, " Qualifications of Preoperational Testing Personnel", which also Records of defines on-site training requirements for the test engineers.
previous education and work experience, along with on-site training records, were reviewed June 4, 1979 for three engineers recently certified as test engineers. These records were compared to the requirements of ADM 102.0 and FSAR Section 14.0 with no discrepancies resulting.
During the above review, it was noted that certification for the engineers had not been placed in their training folders maintained by the station training department per ADM 12.0, " Station Training". They were available on that date from the Engineering Supervisor. Following discussions on this matter, the training folders were updated and later reviewed by the inspector on June 11, 1979. There were no further questions at that time.
7.
Followup on Previously Identified Open Items Limitorque Valve Actuators (339/79-01-04)
a.
As noted in IE Report 50-338/79-01 and 50-339/79-01, several valves of the type described in IE circular 78-16 are installed in Units I and During the inspection period, signs warning of results of manual 2.
operation were observed to bsve been placed on the actuators for most of the Unit 2 valves in question; however, small plastic signs installed on handwheel extensions for valves MOV-2890A-D on the first level of the safeguards building had been destroyed.
Station personnel stated that a more permanent type sign was being investigated.
This item remains open.
'
1P<?
Lvc i
.
-3-g
.
'
b.
SYNSOC By-Laws (338/79-01-01)
/
As discussed in IE Report 50-338/79-01, the System Nuclear Safety and Operating Committee (SYNSOC) By-Laws were found to be outdated regarding titles of certain members and other company officials and did not address all audit subjects Review of the required by Unit 1 Technical Specifications.
By-Laws, revised April 24, 1979, revealed that these discrepan-cies had been corrected. This item is closed.
Regenerative Heat Exchanger Supports (339/79-28-05)
c.
A report per 10 CFR 50.55(e) and 10 CFR 21 concerning potential overstressing of the supports for the regenerative heat exchanger was submitted April 20, 1979 (serial number 181). The supports were modified under the Operations Maintenance system using procedures and instructions supplied by Westinghouse and by Engineering and Design Coordination Reports (E&DCR) PS-5239-2 and PS-5239A-2. Replacement supports were supplied by Joseph Oat Corporation, the manufacturer of the heat exchanger.
Overall control of the work was via maintenance report (MR)
N2 79-04200915 and mechanical maintenance procedure MMP-C-P-1, Observa-
" Repair and/or Replacement of Piping and Components".
tion of the heat exchanger and supports on May 18, 1979 revealed that the new supports had been installed as required. A review of the completed MR, procedure and associated documenta-tion on June 8,1979 resulted in no further questions or comments. Item 339/79-28-05 is closed. The same problem was closed for Unit 1 in IE Report 50-338/79-20.
d.
Reactor Coolant Pump Flow Splitters (339/79-21-05)
The final report for this matter per 10 CFR 50.55(e), serial number 191B dated May 4, 1979, stated that the Unit 2 flow straightening devices would be mechanically removed from the pump suction lines. On May 21, 1979, the inspector visually examined the pump suction lines for all three Unit 2 reactor coolant pumps following splitter plate removal and cleanup.
The following observations were made:
(1) Cut edges had been smoothed down.
(2) Piping minimum wall thickness had not been violated.
(3) No cracks were evident.
(4) Edger and ends of the remaining bosses had been tapered.
l$
L
-
.
.
-4-
.
(5) Cleanliness control was in effect and a final inspection by VEPC0 QA was scheduled prior to closure of the steam generator manways, and (6) Photographs of the piping interiors following splitter plate removals were in QA files.
The inspector had no comments relative to the observations made. Documentation associated with this removal effort was reviewed by another IE inspector, whose findings are given in IE Report 50-339/79-33.
8.
Followup of IE Bulletins The following IE Bulletins were discussed to determine whether they had been received, reviewed for applicability to Units 1 and 2, and responded to as requested in the Bulletins.
IEB 79-03, Longitudinal Weld Defects in ASME SA-312 Type 304 a.
Stainless Steel Pipe Spools A response for Unit 1, serial no. 156A, dated April 16, 1979, stated that no spool pieces of the type described in the Bulletin were installed in Unit 1.
One length of pipe was found in storage and placed on quality control " hold".
A response for Unit 2, serial number 344 dated May 10, 1979, stated that an extension until June 11, 1979 would be required in order to complete all necessary record reviews. The Unit 2 responce will be evaluated after receipt. Open item 339/79-28-06 remains open.
b.
IEB 79-04, Incorrect Weights for Swing Check Valves Manufactured by Velan Engineering Corporation The unit I response, serial number 221A, dated May 3, 1979 identified certain three inch check valves found to have weight in excess of those shown on vendor drawings. A supple-mental response dated May 23, 1979 stated that reanalysis of the affected piping and supports had shown the existing designs to be acceptable. This subject will be reviewed in more detail during a future inspection (open item 338/79-15-03).
The response for Unit 2, serial number 430, dated May 30, 1979, also identified check valves with erroneous weights, but states that the system designs are acceptable. This will also be reviewed in more detail, this item 339/79-21-03 remains open.
,,9
-
UvE d
.
.
'
During the review requested by the IEB, it was discovered that the three 12-inch accumulator discharge valves for each Unit had also been provided with incorrect weights shown on their drawings. The weights were assumed to be 3,050 pounds and were later found to be 3,950 pounds. This was reported as LER 79-073/01T-0 for Unit 1 and per the special provisions of IEB 79-04 by letter 221A on May 3, 1979. A report for Unit 2 was submitted May 24, 1979 (serial number 409). Review of these valves will be conducted as part of the Bulletin followup.
IEB 79-05, Nuclear Incident at Three Mile Island c.
This Bulletin was issued for information only for non-Babcock and Wilcox plants. The inspector confirmed that it had been received and reviewed by station management. Response to its sequel for Westinghouse plants,79-06A, was addressed in IE Report 50-338/79-20.
d.
IEB 79-07, Seismic Stress Analysis of Safety-Related Piping Responses for Unit 1 (serial number 289 dated April 24, 1979)
and Unit 2 (serial number 299 dated April 25, 1979) describe the methods by which seismic analyses were conducted. These responses will be reviewed at a later date. Open item 338/79-28-01 is identified for Unit 1; Unit 2 item 339/79-28-07 remains open.
IEB 79-08, Events Relevant to Boiling Water Reactors Identified e.
During Three Mile Island Incident This Bulletin was issued for information only to owners of PWR's.
The inspector verified that it had been received and reviewed by station management.
f.
IEB 79-09, Failures of GE Type AK-2 Circuit Breaker in Safety-Related Systems The Units 1 and 2 response, serial number 294 dated May 23, 1979, states that no breakers of the type described in the Bulletin were in use or planned for use in either unit. There were no further questions on this and open item 339/79-28-08 is closed.
1P'7
'
IL L
v
-.
-6-Comparison of Installed Systems to FSAR Descriptions-Unit 2 9.
This inspection effort was a continuation of that documented in IE 50-339/78-35 and 50-339/79-22 regarding the quench and Report recirculation spray systems and also covered the casing cooling system for outside recirculation spray pumps.
Quench and Recirculation Spray Systems a.
Comparision of Revision 7 of station drawing FM-91A to FSAR Figure 6.2.2-1 dated July 7, 1977, resulted in observation of several discrepancies, identified as open item 339/78-35-02.
Revision 8 of FM-91A was recently compared to FSAR Figure 6.2-64, revised via Amendment 66 on March 23, 1979. This comparison, plus discussions with station personnel revealed that all significant differences had been resolved or changes to the drawing initiated via Preoperational Deficiency Reports.
Item 339/78-35-02 is considered closed.
Regarding the spray nozzles for inside and outside recircula-50-339/79-22, the tion spray systems discussion in Report inspector reviewed Engineering and Design Coordination Reports (E&DCR) P-2296A-2 (inside) and 25728A-2 (outside) and associated drawings and found the number of unplugged nozzles to be the same as given in FSAR Figure 6.2-64.
It was noted, however, that drawing FM-91A (valve operating number drawing) had not been revised to show the correct number of nozzles for each outside Later in the inspection period Revision 9 of spray header.
FM-91A, revised to show the proper number of nozzles, was received. There were no further questions in this area.
As part of the system comparison, the inspector observed the outside recirculation spray pump discharge lines restricting orifices RO-RS-200A and RO-RS-200B and questioned what size they were on June 7, 1979. Licensee personnel produced E&DCR P-688 and Field Quality Control Inspection Report 2-RS-20, signed off February 1, 1979, which showed that the orifice bores were sized to limit pump flows to 3,000 gpm each.
As 50-338/77-58, Details I, paragraph explained in IE Report 6.g.(3), the 3,000 gpm flow rate was an interim design later superceded by the installation of the casing cooling system for outside circulation spray pumps to overcome a net positive suction head (NPSH) problem. Upon correction of the NPSH problem, the flows were to be returned to their original design valves of about 3,700 gpm each (see FSAR page 6.2-70 and Table 6.2-40).
Following discussions on the subject, E&DCR P-688A-2 was initiated for Unit 2 on June 8, 1979 to replace the smaller bore orifices with full-sized bore flanges.
,-
t,qg Iuac s
.
.
.
%
-7-rf Theoutsiderecirculationspraysystemhpdbeenreleasedto VEPC0atthetimeofdisc6veryoftheigroperlysizedorifices and all preoperational testing on the system had been completed (no full flow tests are conducted on this system after orifice installation). Licensee personnel were informed that since the system was ready for operation and would not have had the flow capacity assumed in the FSAR, this situation appeared to deviate from an FSAR commitment (De ' tion 339/79-35-01).
VEPC0 and S&W were evaluating this
,er at the conclusion of
,
the inspection period to determine why an ESDCR written for Unit I to return flows to normal after casing cooling installa-tion for that Unit had not been applied to Unit 2.
This will be reviewed as part of the followup on the above Deviation.
b.
Casing Cooling System This system is shown on Figure 6.2-73 of the Units 1 and 2 FSAR and was installed to overcome NPSH problems on the outside recirculation spray pumps. This Figure was compared to Unit 2 station drawing FM-91B, Revision 3 and both drawings ompared to accessible portions of the as-built system. Severa. differ-ences were noted between the as-built system and the FSAR Figure. The system was four '- be in general agreement with FM-91B, Revision 3 and E&DCR's pending against it.
Discussions with station personnel revealed that the Unit I system, depicted in FSAR Figure 6.2-73 and the Unit I system are somewhat different, for example: reducers are used on the suction and discharge flanges on Unit 1 pumps, but not Unit 2; Unit I uses many gate valves in positions wherein Unit 2 has diaphragm valves. Following discussions on this matter, licensee person-nel were informed that the differe,.ces were not such that the FSAR would have to be updated, but that individual Unit prints should be accurate to this type of detail.
Installation of the casing cooling system and supply lines from the quench spray pump discharge lines to the suctions of inside recircula-tion spray pumps, as shown in FSAR Figure 6.2-64, completes modifications required to overcome NPSH problems for the fecirculation spray systems. This problem was initially reported per 10 CFR 21 per letter numbe-352 on August 15, 1977. Except for verification of quench spray to inside spray and the outside recirculation orifice sizes, the systems were found to be installed as described in the FSAR, thus open item 339/78-05-04 is closed. Correct orifice installation will be verified as part of followups of the apparent Deviation discussed in paragraph a above.
1n(9
-
Iuua a
.
.
.
-8-10. Followup Operator Training on Three Mile Island Incident Findings regarding Units 1 and 2 operator training relative to Three Mile Island (TMI) following the issuance of IE Bulletin 79-06A were discussed in IE Report 50-338/79-20, Details, paragraph 7.
Subsequent to those initial training efforts, all licensed reactor and senior reactor operators attended the Surry Power Station simulator for additional training. General topics covered in the lesson plan for the simulator training included:
Discussion of TMI including design differences.
a.
b.
Lack of understanding of pressure / temperature relationships.
Demonstration of approximately what occurred at TMI.
c.
d.
Demonstration of effect on North Anna and Surry with various water volumes in the stecm generators.
Discussion of loss of coolant accidents, and e.
f.
Discussion of similar problems on Westinghouse plants.
On June 6, 1979, the inspector attended the last session for North Anna operators as an observer. Six licensed persons were in attendance.
The inspector concluded that the session demonstrated the difference in Westinghouse and Babcock and Wilcox plant behaviors to the TMI-type event quite well, along with accomplishing the other objectives listed above.
11.
Plans For Coping with Strikes Negotiations between VEPCO and the International Brotherhood of Electrical Workers (IBEW) concerning a new contract were in progress during the inspection period with results of balloting by union members expected at the end of May. During the latter part of May, the inspector discussed contingency plans which had been formulated by station management to be used in the event of a strike by IBEW.
IBEW members include operators, electrical and mechanical maintenance craftsmen, storeroom employees, and laborers. These discussions were directed toward ascertaining whether the following considera-tions had been covered:
Plant staffing to etat Technical Specification requirements, a.
including fire team members, and emergency plan team members.
fbb2 3 '[
__
.
.
-
.
.
_o_
rf l
b.
Arrangements had been made with local su ort agencies to ensure delivery of food and necessary s plies to the site as needed.
Provisions had been made with local law enforcement officials c.
to deal with non-docile strikers.
Regarding station security responsibilities during strikes by workers which could have impact on the operational unit, written procedures exist which provide appropriate instructionc for security personnel. These include security plan implementing procedures SPIP 14, " Support From Local Law Enforcement Agencies", and SPIP 22,
" Security During Station Emergencies", and General Order No. 27,
" Civil Disturbances". Plans for staffing of the station were delineated in an internal memorandum dated March 26, 1979.
It was noted that other plans, such as for delivery of food and supplies, had been completed, but not documented. The inspector observed that a checklist should be developed to ensure all aspects of necessary planning are covered for future contingencies and noted that this approach would be more efficient in that plans are redrawn on a recurring basis.
Station management acknowledged this comment.
There were no other comments in this area.
At the conclusion of the inspection period, union members were scheduled to vote on a revised proposal in late June 1979 12. Demonstration at Plant Site The inspector discussed licensee plans for dealing with a planned, nonviolent civil disobedience action to occur June 3, 1979 at the plant site on various dates during the inspection period. Plans for the demonstration, coordinated by the Virginia Sunshine Alliance, were well publicized in the vicinity of the plant. Station procedures for handling of such actions are defined in the security plan implementing procedures referenced in paragraph 11.
The discussions were held to verify that contacts had been made with local and state law enforcement officials and necessary assistance requested to assure that safety of the plant would not be eudangered.
All necessary contacts appeared to have been made. The inspector had no questions in this area.
13. Service Water Expansion Joints A problem involving rubber expansion joints in the Units 1 and 2 service water lines to and from component cooling heat exchangers
.
"*'
"
.
.
.
.
-10-was discussed in IE Report 50-338/79-20, Details, paragraph 6.
These joints were found to have an insufficient number of tie rods to prevent excessive lateral pipe movement. Or May 31, 1979, the inspector observed that the service water lines to and from the recirculation spray heat exchangers for each unit employed similar expansion joints which had no tie rods. A review of station drawings MSK-105-H7 and H8 for Unit I and MSK-105A1 and B8 for Unit 2 confirmed that no tie rods were required. This is due to the fact that the pipes are restrained on either side of the expansion joints. There were no further questions on this matter.
14.
Unit 2 Motor Control Center Failure On June 1, 1979, the inspector was verbally informed of the failure of Class 1 motor control center 2-EP-MC-11 under the reporting requirements of 10 CFR 50.55(e) and 10 CFR 21.
A written report was submitted June 6, 1979 (sett41 number 455) and supplemented June 14, 1979 by letter number 455A. Review of the event revealed that it had first been documented via Preoperational Deficiency Report Number 2-107 on May 16, 1979 by station nersonnel, but that information concerning it was not received by Power Station Engineering in the corporate office until May 29, 1979. The Project Engineer, or designee, is charged with initiating a review of such anomalies for reportability under 10 CFR 21 or 10 CFR 50.55(e) per paragraph 6.1 of Section 5.13 of VEPCO's Quality Assurance Manual-Engineering an Construction. Paragraph 6.1 states, in part, that "Each suspected failure to comply with the Atomic Energy Act or with NRC rules, orders, or licenses, and each deviation murt be reported on Form #883.8C.
The VEPCO employee, VEPC0 architect-enginett, constructor, NSS vendor or consultant who discovers the suspt-ted failure or deviation shall immediately forward the information to L.m Project Engineer, or his designee, for completion of Part 1 of form #883.8C".
Station personnel were informed during the management interview on June 8, 1979 that it appeared that the referenced reqairement was not followed in that information was not expiditiously transmitted to the Project Engineer and that this appeared f.o be in noncompliance with Criterion XVI of Appendix b to 10 CFR 5) (Infraction-339/79-35-02).
It was noted that this appea.-ed to be similar to the Infraction denoted in IE Report 50-339/'9-28. Station management acknowledged the inspector's comments but s cated that the event had been reviewed at the station and felt to le non-reportable.
Technical aspects of this event will be reviewed at a later date.
15. Plant Tours Tours of selected plant areas were conducted on May 24, 1979 and other dates during the inspection period in conjunction with other oea
%
/
.
.
.
-11-
.
inspection activities. During the tours, the following items, as available, were observed:
a.
Fire Equipment Operability and evidence of periodic inspection of fire suppres-sica equipment.
b.
Housekeeping Minimal accumulations of debris and maintenance of required cleanliness levels in systems under or following testing.
Equipment Preservation c.
Maintenance of special preservative measures for installed equipment as applicable.
d.
Component Tagging Implementation and observance of equipment tagging for safety or equipment protection.
e.
Communication Effectiveness of public address system in all areas toured.
f.
Equipment Controls Effectiveness of jurisdictional controls in precluding unauthor-ized work on systems turned over for initial operations or preoperational testing.
g.
Foreign Material Exclusion Maintenance of controls to assure systems which have been cleaned and flushed are not reopened to admit foreign material, b.
Security Implementation of security provisions for both Units.
Within the above areas, no items of noncompliance or deviations were observed when compared to the applicable station pregrams and procedures.
.
1?
^ -
Iuuc o s
.
.
.
-12-
.
If. Rosemount Pressure Transmitters On May 25, 1979, the inspector was requested by Region II IE to inquire as to the use of Rosemount type 1153 Series I pressure transmitters at North Anna 1 and 2.
These transmitters had been identified by Rosemount, Inc. as being potentially defective in that under certain conditions they might not actuate when low voltage DC is applied. Under worst case conditions, they might require as much as 30 volts DC to actuate. Discussions with station personnel revealed that VEPC0 had been notified of this problem by Rosemount by TWX to Surry on May 18, 1979 with a copy of the TWX being received at North Anna on May 22, 1979. Review by station personnel revealed that none of the transmitters were installed in Unit I and two were recently installed in Unit 2 as part of a fire protection program revision to enable monitoring of pressurizer level and pressure from a redundant plant area.
Station personnel determined this to be non-reportable in that a 40 volt DC system is used. The !.nspector had no further questions at that time.
-
.
D f Ital
-
.