ML20134M526: Difference between revisions
StriderTol (talk | contribs) (StriderTol Bot change) |
StriderTol (talk | contribs) (StriderTol Bot change) |
||
(10 intermediate revisions by the same user not shown) | |||
Line 1: | Line 1: | ||
{{Adams | |||
| number = ML20134M526 | |||
| issue date = 08/26/1985 | |||
| title = Insp Rept 50-382/85-20 on 850601-0731.Violation Noted: Failure to Meet Operational Mode Requirements Per Tech Spec 4.0.4 & Failure to Conduct 10CFR50.59 Review of Design Criteria for Control Room Habitability | |||
| author name = Constable G, Flippo T, Johnson A, Jones W | |||
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) | |||
| addressee name = | |||
| addressee affiliation = | |||
| docket = 05000382 | |||
| license number = | |||
| contact person = | |||
| document report number = 50-382-85-20, NUDOCS 8509040157 | |||
| package number = ML20134M511 | |||
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | |||
| page count = 14 | |||
}} | |||
See also: [[see also::IR 05000382/1985020]] | |||
=Text= | |||
{{#Wiki_filter:. .. . . . . . . . _ _ _ . . . __ | |||
- | |||
_ .- , | |||
e | |||
i | |||
APPENDIX B. | |||
'' | |||
U. S. NUCLEAR REGULATORY C012tISSION | |||
REGION IV | |||
i | |||
4 NRC Inspection Report: 50-382/85-20 License: NPF-38 ,, | |||
. | |||
. Docket: 50-382 | |||
! t | |||
,- Licensee: Louisiana Power & Light Company (LP&L) | |||
; 142 De.'.n.ronde Street , | |||
-New Orleans, Louisiana '70174 | |||
Facility-Name: Waterford Steam Electric Station, Unit 3 | |||
Inspection At: ,Taft, Louisiana | |||
Inspection' Conducted: June 1 through July 31, 1985 ! | |||
, | |||
Inspectors: I h (b. b | |||
T. A. Flippo,' Msident Inspector | |||
@-#,-95 | |||
Date I | |||
. | |||
' | |||
D. b - m2 S-N-Af | |||
W. B. Jongs, Reactor Inspector Date | |||
, f7 . JW tY ~ol & ~<?[ | |||
' | |||
A. R. Johnson, Reactor Inspector Date | |||
. i | |||
, , | |||
I- A'ssisting. | |||
Personnel:- -Howard Onorato, Energy, Incorporated | |||
>c Daniel Sanow,' Energy, Incorporated ' | |||
' | |||
K. D. Metcalf, EC&G Idaho, Inc. | |||
i | |||
Ipproved[ c, / 4 | |||
- | |||
/r-26-2[ | |||
d. L.' Constable Chief Date | |||
Resctor_ Project'Section C | |||
, | |||
j | |||
I | |||
8509040157 850828' 2 | |||
PDR | |||
G | |||
ADOCK 0% | |||
, | |||
, | |||
, _ _ _. _ - . , _ . _ . . _ . _ _ , , _ _ . _ , . . _ . . , . . | |||
. - _ _ . - | |||
. - . . .. - . . -.- . . . . . .- . | |||
I | |||
^ | |||
. . | |||
- | |||
! | |||
i | |||
- 2 | |||
: | |||
1 | |||
I Inspection Summary | |||
! Inspection Conducted June 1 through July 31, 1985 (Report 50-382/85-20) | |||
> Areas Inspected: Routine, announced inspection of: (1) Phase III Test | |||
Witnessing, (2) Test Results Evaluation, (3) Surveillance Testing and | |||
Calibration Control, (4) Station Batteries, (5) Control of Design Changes and | |||
, | |||
i | |||
; Modifications, (6) Audits, (7) Phase III Quality Activities, (8) Auditor and | |||
1- Inspector Training, (9) Control Room-Ventilation System Emergency Outside Air | |||
i | |||
Intake Valves, and (10) Operational Mode Changes. The inspection involved | |||
i 448 inspector-hours onsite by three NRC inspectors and three contract | |||
j consultants. | |||
i | |||
Results: Within the areas inspected, two violations were identified, | |||
i | |||
1 | |||
1 | |||
4 | |||
i | |||
i | |||
4 | |||
t | |||
I | |||
k- | |||
L | |||
f | |||
r | |||
4 | |||
i | |||
, - , . , - - , m .,- .r, ... ~s-,,,,mwn--se-wn-en ,,.--- ,,-, wen.,"--w e, ,-* - -- - . . e , | |||
- | |||
, | |||
. | |||
. | |||
3 | |||
Details | |||
1. Persons Contacted | |||
Principal Licensee Employees | |||
*R. S. Leddick, Senior Vice President, Nuclear Operations | |||
*R. P. Barkhurst, Plant Manager, Nuclear | |||
*T. F. Gerrets, Corporate QA Manager | |||
*S. A. Alleman, Assistant Plant Manager, Plant Technical Staff | |||
*J. R. McGaha, Assistant Plant Manager, Operations and Maintenance | |||
*L. M. Meyers, Operations Superintendent | |||
*J. N. Woods, QC Manager | |||
*A. S. Lockhart, Site Quality Manager | |||
, | |||
*R. F. Burski, Engineering and Nuclear Safety Manager | |||
K. L.. Brewster, Onsite Licensing Engineer | |||
G. E. Wuller, Onsite Licensing Coordinator | |||
*Present at exit interviews. | |||
In addition to the above personnel, the NRC inspectors held discussions ' | |||
with various operations, engineering, technical support, maintenance, and | |||
administrative members of the licensee's staff. | |||
2. Plant Status | |||
The Waterford 3 site is presently in the startup testing phase. The 100% | |||
testing plateau has been completed and the nuclear steam supply system | |||
(NSSS) warranty run needs to be completed. The plant is in an outage to | |||
- | |||
perform replacement of the main generator rotor retaining rings. | |||
3. Phase III Test Witnessing | |||
The NRC inspectors observed the performance of portions of the following | |||
. Phase III tests: | |||
SIT-TP-705 Nuclear and Thermal Power Calibration | |||
SIT-TP-716 Core Performance Record | |||
SIT-TP-718 Variable Tavg Test | |||
During the performance of the test, the NRC inspectors verified the | |||
following: | |||
a. The personnel conducting the test were cognizant of the test | |||
acceptance criteria, precautions, and prerequisites prior to | |||
beginning the test. | |||
. | |||
4 | |||
g | |||
. | |||
. | |||
, | |||
.(- | |||
4 | |||
b. The test was conducted.in accordance with an approved procedure and | |||
the test procedure was used and signed off by personnel conducting | |||
the test. | |||
c. Data was collected and recorded as requested by the test procedure | |||
instructions. | |||
No violations or deviations were identified. | |||
4. Test Results Evaluation | |||
The NRC inspectors reviewed Phase III test results to verify that: | |||
, | |||
a. All changes, including deletions to the test program, had been | |||
' | |||
reviewed for conformance to the requirements established in the FSAR | |||
and Regulatory Guide 1.68. | |||
Deficiencies had been adequately addressed and corrective action | |||
' | |||
-b. | |||
completed, | |||
c. The licensee had correctly analyzed the test data and verified that | |||
it met the established acceptance criteria. | |||
d. The startup organization as well as the plant operating review | |||
committee (PORC) had reviewed and accepted the test results. | |||
The following test packages were reviewed: | |||
; SIT-TP-705 Nuclear and Thermal Power Calibration | |||
SIT-TP-716 Core Performance Record | |||
SIT-TP-717 CPC/COLSS Verification | |||
, | |||
SIT-TP-726 Remote Reactor Trip with Subsequent Remote Plant ~ | |||
Cooldown | |||
SIT-TP-727 Total Loss of Flow Trip - Natural Circulation | |||
SIT-TP-740 100% Turbine Trip | |||
. | |||
The,NRC inspectors determined that each of the above test packages was | |||
properly reviewed by the licensee and met the applicable acceptance | |||
criteria. | |||
, | |||
No violations or deviations were identified. | |||
~ | |||
, | |||
) | |||
y z .. _ | |||
. . . . | |||
* | |||
.< | |||
a- | |||
5 | |||
~5.~ Surveillance Testing and Calibration Control | |||
The purpose of this portion of the inspection was to ascertain whether | |||
LP&L had developed and implemented programs for control and evaluation of | |||
surveillance testing, instrumentation calibration not covered by Technical | |||
Specification, and inservice inspection. | |||
The Preventive Maintenance Schedule System (PMSS) computer provides | |||
' for scheduling of most preventive maintenance. The schedules, as | |||
established in the data base, were not reviewed. The criteria for | |||
i establishing the schedules were reviewed during the Technical | |||
, Specification surveillance program review and the calibration control | |||
review. The PMSS computer provides an accurate means of tracking and | |||
; scheduling surveillances and preventive maintenance. | |||
a. Technical Specification Surveillance Program | |||
An inspection was conducted of the licensee's Technical Specification | |||
, | |||
Surveillance Program. Areas examined included the following: | |||
(1) Establishment of a master index and cross reference | |||
(2) Assignment of duties and responsibilities | |||
. | |||
(3) Proper documentation of data | |||
(4) Review of completed procedures for implementation | |||
The following procedures were reviewed by the NRC inspector: | |||
UNT-7-004 Technical Specification Surveillance Control, Rev. 2 | |||
OP-903-035 Containment Spray Pump Operability Check, Rev. 5 - | |||
Technical Specifications 4.6.2.lc and 4.0.5 | |||
OP-903-031 Containment Integrity Check, Rev. 2 - Technical | |||
Specification 4.6.1.la | |||
OP-903-021 RCS Water Inventory Balance, Rev. 2 - Technical | |||
. Specifications 4.4.5.2.la and 4.4.5.2.1d | |||
OP-903-032 Quarterly ISI Valve Test, Rev.1 - Technical | |||
l | |||
Specification 4.0.5 | |||
NE-5-103 COLSS Margin Alarm and Penalty Factor | |||
Verification, Rev. 1 - Technical Specifications | |||
4.2.1.3, 4.2.3.2c, and 4.2.4.3 | |||
: | |||
! | |||
T | |||
<,n | |||
y , m, r - - | |||
, - - - , e m es e - - | |||
- . . . - | |||
< . | |||
,. | |||
, | |||
. | |||
T | |||
6 | |||
2 | |||
MI-3-441 Turbine Generator Overspeed Protection System | |||
Calibration, Rev. 1 - Technical Specification | |||
4.3.4.2c | |||
MI-3-384 Condensate Vacuum Pump Discharge' Radiation | |||
Monitor, Rev. 3 - Technical Specification | |||
4.3.3.11, Table 4.3-9, Items 3a and 3d | |||
MI-3-101 Linear Power Channel Calibration,'Rev. 1 - | |||
Technical Specification 4.3.1.1, Table 4.3-1, Item 2 | |||
MI-3-201 Plant Protection System Calibration, Rev. 3 - | |||
Technical Specification 4.3.1.1, Table 4.3-1, | |||
Items 3, 4, 5, 6, 7, 8, 9, 10, 11, 14, 15, and 16 | |||
ME-3-210 Station Battery Bank'and Charger (Quarterly), | |||
Revision 2 - Technical Specifications 4.8.2.lbl, | |||
4.8.2.lb2, 4.8.2.lb3, and 4.8.2.2 | |||
ME-3-100 Fire Pump Diesel Starting Battery (Weekly). Rev. 2 | |||
Technical Specifications 4.7.20.1.3a1 and | |||
4.7.10.1.3a2 | |||
ME-3-010 Hydrogen Recombiner Temperature and Power | |||
Measurement, Rev. 2'- Technical Specification | |||
4.6.4.2a | |||
- | |||
- MM-3-015 Emergency Diesel Engine Inspection, Rev. 2 - | |||
Technical Specifications 4.8.1.1.2d1 and 4.8.1.2 | |||
: | |||
MM-3-033 Computer Room Halon 1301 Fire Suppression System | |||
Flow Test, Rev. 0 - Technical Specification | |||
4.7.10.3c2 | |||
OP-903-100 MOV Bypass Overload Test, Rev. 2 - Technical | |||
' Specification 4.8.4.2a2 | |||
No violations or deviations were identified. | |||
f.lthough no violations were noted, the following two concerns were | |||
identified during the inspection: | |||
(1) Technical Specification cross reference in UNT-7-004 identifies | |||
NE-2-102 as the procedure for completing surveillance 4.2.2.2a. | |||
. The cross reference also identifies SIT-TP-725 as the procedure | |||
for initial performance. Procedure NE-2-102 does not exist. It | |||
has not been written. Since SIT-TP-725 was used for initial | |||
performance, the surveillance is current. No means of tracking | |||
' | |||
.. | |||
4 | |||
- | |||
y y y - - - --r | |||
. . . . . - | |||
- | |||
. | |||
. | |||
7 | |||
has been established to ensure that NE-2-102 is issued prior to | |||
the next required performance date. | |||
(2) Technical Specification cross reference in UNT-7-004 identifies | |||
OP-903-100 as the procedure for completing surveillance | |||
4.8.4.2al. This procedure addresses the requirements of | |||
surveillance 4.8.4.2a2. The MOV overload bypass devices | |||
required to be tested are listed on Table 3.8-2. All of these | |||
devices are tested in Procedure OP-903-100 under the 4.8.4.2a2 | |||
criteria. None of the devices are governed by the 4.8.4.2a1 | |||
criteria. A Technical Specification change should be noted on | |||
' | |||
the cross raference and/or Procedure OP-903-100 that none of the | |||
MOV overload bypass devices are governed by Surveillance | |||
4.8.4.2al. | |||
: b. Instrumentation Calibration Not Covered by Technical Specifications | |||
An inspection was conducted of the licensee's instrumentation | |||
calibration program. Areas examined included the following items: | |||
(1) Establishment of a master calibration schedule | |||
(2) Assignment of duties and responsibilities | |||
.(3) Proper documentation data | |||
The following procedures were reviewed by the NRC inspector: | |||
MD-1-004 Preventive Maintenance Scheduling, Rev. 6 | |||
MD-1-015 Administrative Controls of Measuring and Test | |||
Equipment, Rev. O | |||
' | |||
MI-1-005 Administrative Controls of Calibration and | |||
Maintenance, Rev. 2 | |||
< | |||
MI-1-006 Calibration and Loop Check Frequency for Process | |||
Instrumentation, Rev. 2 | |||
MI-5-160' Calibration of Plant Protection System Test and | |||
Calibration Card and DVM, Rev. 1 | |||
: MI-5-211 Calibration of Control Valves and Accessories, | |||
Rev. 2 | |||
MI-5-518 Control Element Drive Mechanisms Air Temperature | |||
Calibration CDC-IT-5201A/B, Rev. 1 | |||
1 | |||
- - - - - , , , | |||
. - . . | |||
3 , | |||
.. - _ | |||
- ' x | |||
~ | |||
. . , . . | |||
, | |||
, | |||
+ | |||
. | |||
, | |||
1-' **g | |||
. . | |||
.-4 | |||
' | |||
' | |||
. , 8 | |||
' - | |||
MI-5-561 Reactor Regulating System Inspection and Test, | |||
~Rev. 1 | |||
, | |||
, | |||
'MI-5-610 Equipment Drain Tank Level Loop Check and | |||
i 3 Calibration BM-IL-0616, Rev.1 | |||
i | |||
No violations or deviations were identified. | |||
c. Inservice Inspection | |||
, | |||
An inspection was conducted of the licensee's inservice inspection | |||
program. Areas examined included the following items: | |||
(1) Assignment of duties and responsibilities | |||
(2) Control of Inservice inspection procedures | |||
, (3) Scheduling of tests pump and valve | |||
2 | |||
-(4) Documentation of results | |||
The following procedures and documents were reviewed by the | |||
inspector: | |||
UNT-7-020 Pump and Valve Inservice Testing, Rev. 1 | |||
PE-1-003 Control of Inservice Inspection, Rev. 1 | |||
' | |||
PE-1-004 Section XI Pump and Valve Reference | |||
< Data / Acceptance Criteria, Rev. 2 | |||
Section XI Repairs and Replacement, Rev. 2 | |||
~ | |||
PE-1-001 | |||
; | |||
LP&L Pump and Valve Inservice Test Plan | |||
Section XI Pump and Valve Reference Data / Acceptance Criteria | |||
Notebook | |||
No violations or deviations were identified. | |||
* | |||
. - 6. Station Batteries | |||
~ | |||
An inspection was: conducted of the licensee's station batteries. Areas | |||
examined included the following items: | |||
, | |||
a.', Visual inspection for' deterioration | |||
- ' | |||
bf. Technical Specification surveillance requirements | |||
- | |||
. | |||
- - t | |||
. | |||
' | |||
t g | |||
4 t 4 | |||
5 | |||
> | |||
' | |||
m - | |||
n- po -V | |||
1 g . ; -y y _ ,2 | |||
- . . | |||
* | |||
. | |||
, | |||
9 | |||
c. Maintenance guidelines | |||
A visual inspection of the station batteries was conducted. Areas | |||
examined included cleanliness and condition of the' batteries and | |||
their rooms. | |||
The following procedures were reviewed by the inspector: | |||
-ME-3-200- Station Battery Bank and Charger (Weekly), Rev. 2 | |||
ME-3-210 Station Battery Bank and Charger (Quarterly), Rev. 1 | |||
; ME-3-220 Station Battery Bank and Charger (18-Month), Rev. 3 | |||
' | |||
ME-3-230 Battery Service Test, Rev. 3 | |||
ME-3-240 Battery Performance Test, Rev. 2 | |||
ME-3-250 Station Battery Performance Evaluation, Rev.1 | |||
ME-3-201 Station Batte y and Charger (Weekly), Rev. 5 | |||
ME-4-213 Battery Intercell Connections, Rev. O | |||
ME-4-231 Station Battery Charging, Rev. 4 | |||
No violations or deviations were identified. | |||
' | |||
. 7. Control of Design Changes and Modifications | |||
The NRC inspector reviewed the licensee's nuclear operations management | |||
manual and Procedures PE-2-006 and PMP-302. These documents outline the | |||
requirements and responsibilities'for.the preparation, control, and review | |||
of station modifications from request through implementation and final | |||
closecut. The station modification package is the vehicle by which design | |||
changes and modifications are made and the use of the forms and documents | |||
that become_ a part of the station modification package (SMP) provide the | |||
required control of design changes. The initiating document from the | |||
station modification request (SMR) provides for the identification, | |||
review, evaluation, and approval of design input. Upon receiving the SMR | |||
the action engineer includes a checkoff list of possible inclusions in the | |||
station modification (SM). | |||
The NRC inspector ascertained the requirements for the assurance | |||
'that the changes do not involve an unreviewed' safety question is catisfied | |||
by the required inclusion in SMP of a nuclear safety review checklist. A , | |||
positive response to any 'of the' questions on the checklist require that a 1 | |||
1 | |||
* | |||
1 | |||
i | |||
i | |||
-i . | |||
' | |||
. | |||
. | |||
10 | |||
nuclear evaluation form be completed and included in the SMP. This form | |||
requires the design engineer to review and evaluate all the nuclear safety | |||
questions outlined in 10 CFR 50.59. | |||
An interview with two action engineers, was conducted and the NRC | |||
inspector found them to be knowledgeable of the safety /nonsafety-related | |||
classification requirements. | |||
The fire protection guidelines of RG 1.120 are similarly handled by the | |||
inclusion in the SMP of a fire protection / safe shutdown checklist. A | |||
positive response to any of the questions on the checklist and other | |||
. | |||
criteria requires that a fire protection / safe shutdown review analysis | |||
(FP/SSA) be prepared. This document reviews components and fire | |||
protection features for changes to such, and reviews the location of | |||
additional fire loading relative to the modification for impact to | |||
Appendix R to 10 CFR 50 to ensure the level of fire protection does not | |||
decrease. The fire protection / safe shutdown checklist and review form are | |||
a part of a new procedure FP-1-022 that is not released but will be | |||
implemented soon. | |||
A document control system.has been administered that controls the release | |||
and distribution of design change documents, controls changes to released | |||
and approved documents, and provides for the control and recalling of | |||
obsolete design change documents. The procedures referenced, particularly | |||
QP-006-001, detail the controls of released documents. | |||
When the SM is completed the administrative controls and Procedure | |||
PE-2-006 require that a work completion notice (WCN) be sent out after the | |||
operational documents are updated and the control room drawings are red | |||
lined to reflect the changes. This WCN alerts all the reviewing | |||
organizations that the modification is complete. This WCN is the official | |||
notice for the updating of plant procedures, operator training, and the | |||
posting of plant drawings that indicate a change is in place that effects | |||
the drawing. Return of WCN to the station modification coordinator (SMC) | |||
indicates the required documentation update has been completed except for | |||
the affected as-built drawings, which is done prior to SM closeout. | |||
The NRC inspector verified that the responsibility and method of reporting | |||
to the NRC of design changes that are safety-related is established and it | |||
will be an annual report filed 1 year after initial criticality by the | |||
licensing group. | |||
The inspection identified what seems to be a problem in the implementation | |||
of the program although there were no deviations from the established | |||
administrative control and procedure outlines. There is, however, a very I | |||
large backlog of SMP in the WCN and drawing update stage. There were, in | |||
' | |||
fact, only 17 SMPs completely closed out and in project files with all | |||
document updates done. There were 125 awaiting drawing update and 206 | |||
.. | |||
- _ . . . . ._ | |||
l '.'' g | |||
' | |||
. | |||
t | |||
11 | |||
SMPs completed but awaiting some other form of review or document update. | |||
This backlog of SMPs causes some problems in the operational documents | |||
such as the red line drawings, where in at least one case, 5 SMs were | |||
. posted on the drawing as being completed but not marked on the drawing, as | |||
well as 3 additional SMs marked up on the drawing. This represents a | |||
total of 8 SMs affecting 1 drawing without any of them incorporated on the | |||
; | |||
. drawing. This is considered an open item (8520-01). | |||
The NRC inspector reviewed the temporary modification to lifted leads and | |||
jumpers requirements and found the procedure UNT-5-004 covers all the | |||
requirements when used with additional form referenced in that procedure. | |||
An examination of the temporary modification log indicated it was up to | |||
date and that log entries were complete. | |||
A field examination of the reactor protective system, the emergency safety | |||
features, and the emergency diesel generator control cabinets revealed no | |||
i lifted leads or jumpers in place. | |||
No violations or deviations were identified. | |||
> | |||
8. QA Program Audits | |||
. | |||
This NRC inspection included activities for preparation and issue of the | |||
! audit schedules, development of an audit plan and checklists, review of | |||
objective evidence reviewed during the audit, audit report control and | |||
distribution, and responses _to audits. In general, the audit program was | |||
found to be an effective status of implementation. The Technical | |||
Specification audit program development and vendor audit programs are | |||
progressing satisfactorily and the operations QA program audit are being | |||
i | |||
satisfactorily implemented. The following audits were reviewed by the NRC | |||
inspector: | |||
Audits: 85-45 85-03 85-07 85-15 85-08 | |||
, | |||
85-10 84-22 85-13 85-01 | |||
Vendor Audits: Combustion Engineering | |||
i | |||
Desselle-Maggard | |||
Cardinal Industries | |||
Rockbestoes | |||
. Southern Vital Records | |||
McGraw-Edison | |||
General Electric | |||
Capitol Controls | |||
Cajun Co. . | |||
Siemen-Allis Co. | |||
Yarway | |||
-_. _ _ _ _ _ . | |||
, | |||
. . | |||
. | |||
12 | |||
During this portion of tne audit two items were noted which are considered | |||
an open item and an unresolved item, respectively. | |||
a. QAP 302, paragraph 5.1.2.c requires onsite contr6ctors be audited | |||
triennially. Beyond this, the program does not adequately | |||
address the following areas: . | |||
Auditing onsite contractors in a timely manner once onsite | |||
Establishing an onsite contractor audit schedule | |||
Notification and placement of contractors on a schedule | |||
Identification of work scope to be audited | |||
Tracking and timely closure of audit findings | |||
Procedure for removal from site if necessary | |||
A current example of this concern is Unitec Company. They were | |||
qualified in Jely 1984 to work onsite but the first audit was not | |||
completed until July 25, 1985. (0 pen Item 8520-02) | |||
b. A review of vendor audit files found two active vendors had not | |||
received the required annual evaluation. Southern Vital Records is | |||
missing a 1984 evaluation and Siemens-Allis Company is missing the | |||
1983 and 1984 evaluations. (Unresolved Item 8520-03) | |||
No violations or deviations were identified. | |||
9. Phase III Quality Activities | |||
The inspection in this area concluded that activities committed to are | |||
being effective!y implemented. Commitments include audits, procedure | |||
reviews, and test data reviews. In the event any test deficiencies are | |||
noted, they are tracked via the CIWA system. Plant quality does | |||
get involved with all required holdpoints as deemed required in CIWA | |||
resolutions. Phase III operations QA audits included 84-45, 84-46, 85-01, | |||
and 85-05. An audit of M&TE activities is scheduled for August, 1985. | |||
No violations or deviations were identified. | |||
10. Auditor and Inspector Training | |||
The inspection included verifying certification documents were current, | |||
all necessary supporting documentation was available, and responsibilities | |||
coincided with training and qualifications. All aspects of the training | |||
._ _ _ _ _ | |||
.* | |||
, | |||
13 | |||
program reviewed during this audit were found to be effectively | |||
implemented. | |||
No violations-or deviations were identified. | |||
11. Control Room Ventilation System Emergency Outside Air Intake Valves | |||
The FSAR discusses the design criteria for control room Nabitability in | |||
Chapter 6.4. The criteria used for location and power supply for the | |||
valves is discussed. The ability to operate the system from the control | |||
room with the loss of a vital bus is one of the design criteria. | |||
OP-03-014, " Control Room Heating and Ventilation," provided the normal | |||
lineup for these valves. Contrary to the FSAR, all valves were normally | |||
aligned closed. The NRC inspector found no evidence that a proper | |||
10 CFR 50.59 review was conducted to calculate dose rates which an | |||
operator would experience if these valves had been manually opened from | |||
outside the control room. When the licensee was informed of this dis- | |||
crepancy, a change was made to the procedure (June 26, 1985). | |||
This is considered a violation. | |||
12. Operational Mode Changes | |||
On June 11, 1985, Waterford 3 Steam Electric Station was in Mode 5 (cold | |||
shutdown) when operations personnel were performirg Surveillance Procedure | |||
OP-903-069, " Integrated Emergency Diesel Generator / Engineered Safety | |||
Features Test." As part of the above procedure, operations personnel were | |||
- | |||
attempting to prove the operability of the Emergency Diesel Generator "B" | |||
automatic load sequence timer as required by Technical Specification | |||
4.8.1.1.2.d.12. While testing Load Block 7, Relay S7X actuated in 121.6 | |||
seconds, which was outside the plus/minus 10% tolerance of the sequenced | |||
load block time (168 plus/minus 16.8 seconds). However, operations | |||
personnel did not review the test data until 1545 hours on June 20, 1985. | |||
Waterford 3 entered Mode 4 (hot shutdown) at 1028 hours on June 20, 1985, | |||
with Emergency Diesel Generator B inoperable. | |||
Technical Specification 4.0.4 requires that " Entry into an OPERATIONAL- - | |||
MODE or other specified conditions shall not be made unless the | |||
surveillance requirement (s) associated with the limiting condition for | |||
-operation'have been performed within the stated surveillance interval or | |||
as otherwise specified." | |||
A | |||
LP&L Operating Procedure OP-10.001, Revision 4, " General Plant | |||
Operations," requires that when entering Mode 4 (hot shutdown) both | |||
emergency diesel generators be operable. | |||
. _ _ _ . . ._- _ _ | |||
_ ._ . _ . . _. _ . . - . -_ _ | |||
. | |||
' | |||
, | |||
t;*8 | |||
> | |||
14 | |||
This is considered a violation. | |||
, | |||
13. Open Items- | |||
The following new items were identified during this reporting period: | |||
8520-01 Open Item Large Backlog of Incomplete SMPs (paragraph 7) | |||
8520-02 Open Item Contractor Audit Program Inadequate | |||
(paragraph 8a) | |||
! | |||
8520-03 Unresolved Vendor Audit Files Incomplete (paragraph 8b) | |||
Item | |||
8520-04 Violation Failure to Conduct Proper 10 CFR 50.59 Review | |||
(paragraph 11) | |||
8520-05 Violation Failure to Meet Opt.'ational Mode Requirements | |||
(paragraph 12) | |||
! 14. Site Tour | |||
i | |||
At various times during the course of this inspection period, the NRC | |||
inspectors conducted general tours of the reactor building, reactor | |||
auxiliary building, and turbine building to observe ongoing maintenance r | |||
and testing. | |||
No violations or deviations were identified. | |||
; | |||
15. Exit Interviews | |||
, | |||
The NRC inspectors met with the licensee representatives at various times ~ | |||
during the course of the inspection. The scope and, findings of the | |||
. inspection were reviewed. | |||
. | |||
t | |||
4 | |||
4 | |||
f | |||
i | |||
. | |||
- _ . , , . . _ . . -.- . | |||
- . . . . . _ . - , . . . . , . - - , - . . . | |||
}} |
Latest revision as of 11:36, 14 December 2021
ML20134M526 | |
Person / Time | |
---|---|
Site: | Waterford |
Issue date: | 08/26/1985 |
From: | Constable G, Flippo T, Andrea Johnson, William Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
To: | |
Shared Package | |
ML20134M511 | List: |
References | |
50-382-85-20, NUDOCS 8509040157 | |
Download: ML20134M526 (14) | |
See also: IR 05000382/1985020
Text
. .. . . . . . . . _ _ _ . . . __
-
_ .- ,
e
i
APPENDIX B.
U. S. NUCLEAR REGULATORY C012tISSION
REGION IV
i
4 NRC Inspection Report: 50-382/85-20 License: NPF-38 ,,
.
. Docket: 50-382
! t
,- Licensee: Louisiana Power & Light Company (LP&L)
- 142 De.'.n.ronde Street ,
-New Orleans, Louisiana '70174
Facility-Name: Waterford Steam Electric Station, Unit 3
Inspection At: ,Taft, Louisiana
Inspection' Conducted: June 1 through July 31, 1985 !
,
Inspectors: I h (b. b
T. A. Flippo,' Msident Inspector
@-#,-95
Date I
.
'
D. b - m2 S-N-Af
W. B. Jongs, Reactor Inspector Date
, f7 . JW tY ~ol & ~<?[
'
A. R. Johnson, Reactor Inspector Date
. i
, ,
I- A'ssisting.
Personnel:- -Howard Onorato, Energy, Incorporated
>c Daniel Sanow,' Energy, Incorporated '
'
K. D. Metcalf, EC&G Idaho, Inc.
i
Ipproved[ c, / 4
-
/r-26-2[
d. L.' Constable Chief Date
Resctor_ Project'Section C
,
j
I
8509040157 850828' 2
G
ADOCK 0%
,
,
, _ _ _. _ - . , _ . _ . . _ . _ _ , , _ _ . _ , . . _ . . , . .
. - _ _ . -
. - . . .. - . . -.- . . . . . .- .
I
^
. .
-
!
i
- 2
1
I Inspection Summary
! Inspection Conducted June 1 through July 31, 1985 (Report 50-382/85-20)
> Areas Inspected: Routine, announced inspection of: (1) Phase III Test
Witnessing, (2) Test Results Evaluation, (3) Surveillance Testing and
Calibration Control, (4) Station Batteries, (5) Control of Design Changes and
,
i
- Modifications, (6) Audits, (7) Phase III Quality Activities, (8) Auditor and
1- Inspector Training, (9) Control Room-Ventilation System Emergency Outside Air
i
Intake Valves, and (10) Operational Mode Changes. The inspection involved
i 448 inspector-hours onsite by three NRC inspectors and three contract
j consultants.
i
Results: Within the areas inspected, two violations were identified,
i
1
1
4
i
i
4
t
I
k-
L
f
r
4
i
, - , . , - - , m .,- .r, ... ~s-,,,,mwn--se-wn-en ,,.--- ,,-, wen.,"--w e, ,-* - -- - . . e ,
-
,
.
.
3
Details
1. Persons Contacted
Principal Licensee Employees
- R. S. Leddick, Senior Vice President, Nuclear Operations
- R. P. Barkhurst, Plant Manager, Nuclear
- T. F. Gerrets, Corporate QA Manager
- S. A. Alleman, Assistant Plant Manager, Plant Technical Staff
- J. R. McGaha, Assistant Plant Manager, Operations and Maintenance
- L. M. Meyers, Operations Superintendent
- J. N. Woods, QC Manager
- A. S. Lockhart, Site Quality Manager
,
- R. F. Burski, Engineering and Nuclear Safety Manager
K. L.. Brewster, Onsite Licensing Engineer
G. E. Wuller, Onsite Licensing Coordinator
- Present at exit interviews.
In addition to the above personnel, the NRC inspectors held discussions '
with various operations, engineering, technical support, maintenance, and
administrative members of the licensee's staff.
2. Plant Status
The Waterford 3 site is presently in the startup testing phase. The 100%
testing plateau has been completed and the nuclear steam supply system
(NSSS) warranty run needs to be completed. The plant is in an outage to
-
perform replacement of the main generator rotor retaining rings.
3. Phase III Test Witnessing
The NRC inspectors observed the performance of portions of the following
. Phase III tests:
SIT-TP-705 Nuclear and Thermal Power Calibration
SIT-TP-716 Core Performance Record
SIT-TP-718 Variable Tavg Test
During the performance of the test, the NRC inspectors verified the
following:
a. The personnel conducting the test were cognizant of the test
acceptance criteria, precautions, and prerequisites prior to
beginning the test.
.
4
g
.
.
,
.(-
4
b. The test was conducted.in accordance with an approved procedure and
the test procedure was used and signed off by personnel conducting
the test.
c. Data was collected and recorded as requested by the test procedure
instructions.
No violations or deviations were identified.
4. Test Results Evaluation
The NRC inspectors reviewed Phase III test results to verify that:
,
a. All changes, including deletions to the test program, had been
'
reviewed for conformance to the requirements established in the FSAR
Deficiencies had been adequately addressed and corrective action
'
-b.
completed,
c. The licensee had correctly analyzed the test data and verified that
it met the established acceptance criteria.
d. The startup organization as well as the plant operating review
committee (PORC) had reviewed and accepted the test results.
The following test packages were reviewed:
- SIT-TP-705 Nuclear and Thermal Power Calibration
SIT-TP-716 Core Performance Record
SIT-TP-717 CPC/COLSS Verification
,
SIT-TP-726 Remote Reactor Trip with Subsequent Remote Plant ~
Cooldown
SIT-TP-727 Total Loss of Flow Trip - Natural Circulation
SIT-TP-740 100% Turbine Trip
.
The,NRC inspectors determined that each of the above test packages was
properly reviewed by the licensee and met the applicable acceptance
criteria.
,
No violations or deviations were identified.
~
,
)
y z .. _
. . . .
.<
a-
5
~5.~ Surveillance Testing and Calibration Control
The purpose of this portion of the inspection was to ascertain whether
LP&L had developed and implemented programs for control and evaluation of
surveillance testing, instrumentation calibration not covered by Technical
Specification, and inservice inspection.
The Preventive Maintenance Schedule System (PMSS) computer provides
' for scheduling of most preventive maintenance. The schedules, as
established in the data base, were not reviewed. The criteria for
i establishing the schedules were reviewed during the Technical
, Specification surveillance program review and the calibration control
review. The PMSS computer provides an accurate means of tracking and
- scheduling surveillances and preventive maintenance.
a. Technical Specification Surveillance Program
An inspection was conducted of the licensee's Technical Specification
,
Surveillance Program. Areas examined included the following:
(1) Establishment of a master index and cross reference
(2) Assignment of duties and responsibilities
.
(3) Proper documentation of data
(4) Review of completed procedures for implementation
The following procedures were reviewed by the NRC inspector:
UNT-7-004 Technical Specification Surveillance Control, Rev. 2
OP-903-035 Containment Spray Pump Operability Check, Rev. 5 -
Technical Specifications 4.6.2.lc and 4.0.5
OP-903-031 Containment Integrity Check, Rev. 2 - Technical
Specification 4.6.1.la
OP-903-021 RCS Water Inventory Balance, Rev. 2 - Technical
. Specifications 4.4.5.2.la and 4.4.5.2.1d
OP-903-032 Quarterly ISI Valve Test, Rev.1 - Technical
l
Specification 4.0.5
NE-5-103 COLSS Margin Alarm and Penalty Factor
Verification, Rev. 1 - Technical Specifications 4.2.1.3, 4.2.3.2c, and 4.2.4.3
!
T
<,n
y , m, r - -
, - - - , e m es e - -
- . . . -
< .
,.
,
.
T
6
2
MI-3-441 Turbine Generator Overspeed Protection System
Calibration, Rev. 1 - Technical Specification 4.3.4.2c
MI-3-384 Condensate Vacuum Pump Discharge' Radiation
Monitor, Rev. 3 - Technical Specification 4.3.3.11, Table 4.3-9, Items 3a and 3d
MI-3-101 Linear Power Channel Calibration,'Rev. 1 -
Technical Specification 4.3.1.1, Table 4.3-1, Item 2
MI-3-201 Plant Protection System Calibration, Rev. 3 -
Technical Specification 4.3.1.1, Table 4.3-1,
Items 3, 4, 5, 6, 7, 8, 9, 10, 11, 14, 15, and 16
ME-3-210 Station Battery Bank'and Charger (Quarterly),
Revision 2 - Technical Specifications 4.8.2.lbl,
4.8.2.lb2, 4.8.2.lb3, and 4.8.2.2
ME-3-100 Fire Pump Diesel Starting Battery (Weekly). Rev. 2
Technical Specifications 4.7.20.1.3a1 and
4.7.10.1.3a2
ME-3-010 Hydrogen Recombiner Temperature and Power
Measurement, Rev. 2'- Technical Specification 4.6.4.2a
-
- MM-3-015 Emergency Diesel Engine Inspection, Rev. 2 -
Technical Specifications 4.8.1.1.2d1 and 4.8.1.2
MM-3-033 Computer Room Halon 1301 Fire Suppression System
Flow Test, Rev. 0 - Technical Specification 4.7.10.3c2
OP-903-100 MOV Bypass Overload Test, Rev. 2 - Technical
' Specification 4.8.4.2a2
No violations or deviations were identified.
f.lthough no violations were noted, the following two concerns were
identified during the inspection:
(1) Technical Specification cross reference in UNT-7-004 identifies
NE-2-102 as the procedure for completing surveillance 4.2.2.2a.
. The cross reference also identifies SIT-TP-725 as the procedure
for initial performance. Procedure NE-2-102 does not exist. It
has not been written. Since SIT-TP-725 was used for initial
performance, the surveillance is current. No means of tracking
'
..
4
-
y y y - - - --r
. . . . . -
-
.
.
7
has been established to ensure that NE-2-102 is issued prior to
the next required performance date.
(2) Technical Specification cross reference in UNT-7-004 identifies
OP-903-100 as the procedure for completing surveillance
4.8.4.2al. This procedure addresses the requirements of
surveillance 4.8.4.2a2. The MOV overload bypass devices
required to be tested are listed on Table 3.8-2. All of these
devices are tested in Procedure OP-903-100 under the 4.8.4.2a2
criteria. None of the devices are governed by the 4.8.4.2a1
criteria. A Technical Specification change should be noted on
'
the cross raference and/or Procedure OP-903-100 that none of the
MOV overload bypass devices are governed by Surveillance
4.8.4.2al.
- b. Instrumentation Calibration Not Covered by Technical Specifications
An inspection was conducted of the licensee's instrumentation
calibration program. Areas examined included the following items:
(1) Establishment of a master calibration schedule
(2) Assignment of duties and responsibilities
.(3) Proper documentation data
The following procedures were reviewed by the NRC inspector:
MD-1-004 Preventive Maintenance Scheduling, Rev. 6
MD-1-015 Administrative Controls of Measuring and Test
Equipment, Rev. O
'
MI-1-005 Administrative Controls of Calibration and
Maintenance, Rev. 2
<
MI-1-006 Calibration and Loop Check Frequency for Process
Instrumentation, Rev. 2
MI-5-160' Calibration of Plant Protection System Test and
Calibration Card and DVM, Rev. 1
- MI-5-211 Calibration of Control Valves and Accessories,
Rev. 2
MI-5-518 Control Element Drive Mechanisms Air Temperature
Calibration CDC-IT-5201A/B, Rev. 1
1
- - - - - , , ,
. - . .
3 ,
.. - _
- ' x
~
. . , . .
,
,
+
.
,
1-' **g
. .
.-4
'
'
. , 8
' -
MI-5-561 Reactor Regulating System Inspection and Test,
~Rev. 1
,
,
'MI-5-610 Equipment Drain Tank Level Loop Check and
i 3 Calibration BM-IL-0616, Rev.1
i
No violations or deviations were identified.
c. Inservice Inspection
,
An inspection was conducted of the licensee's inservice inspection
program. Areas examined included the following items:
(1) Assignment of duties and responsibilities
(2) Control of Inservice inspection procedures
, (3) Scheduling of tests pump and valve
2
-(4) Documentation of results
The following procedures and documents were reviewed by the
inspector:
UNT-7-020 Pump and Valve Inservice Testing, Rev. 1
PE-1-003 Control of Inservice Inspection, Rev. 1
'
PE-1-004 Section XI Pump and Valve Reference
< Data / Acceptance Criteria, Rev. 2
Section XI Repairs and Replacement, Rev. 2
~
PE-1-001
LP&L Pump and Valve Inservice Test Plan
Section XI Pump and Valve Reference Data / Acceptance Criteria
Notebook
No violations or deviations were identified.
. - 6. Station Batteries
~
An inspection was: conducted of the licensee's station batteries. Areas
examined included the following items:
,
a.', Visual inspection for' deterioration
- '
bf. Technical Specification surveillance requirements
-
.
- - t
.
'
t g
4 t 4
5
>
'
m -
n- po -V
1 g . ; -y y _ ,2
- . .
.
,
9
c. Maintenance guidelines
A visual inspection of the station batteries was conducted. Areas
examined included cleanliness and condition of the' batteries and
their rooms.
The following procedures were reviewed by the inspector:
-ME-3-200- Station Battery Bank and Charger (Weekly), Rev. 2
ME-3-210 Station Battery Bank and Charger (Quarterly), Rev. 1
- ME-3-220 Station Battery Bank and Charger (18-Month), Rev. 3
'
ME-3-230 Battery Service Test, Rev. 3
ME-3-240 Battery Performance Test, Rev. 2
ME-3-250 Station Battery Performance Evaluation, Rev.1
ME-3-201 Station Batte y and Charger (Weekly), Rev. 5
ME-4-213 Battery Intercell Connections, Rev. O
ME-4-231 Station Battery Charging, Rev. 4
No violations or deviations were identified.
'
. 7. Control of Design Changes and Modifications
The NRC inspector reviewed the licensee's nuclear operations management
manual and Procedures PE-2-006 and PMP-302. These documents outline the
requirements and responsibilities'for.the preparation, control, and review
of station modifications from request through implementation and final
closecut. The station modification package is the vehicle by which design
changes and modifications are made and the use of the forms and documents
that become_ a part of the station modification package (SMP) provide the
required control of design changes. The initiating document from the
station modification request (SMR) provides for the identification,
review, evaluation, and approval of design input. Upon receiving the SMR
the action engineer includes a checkoff list of possible inclusions in the
station modification (SM).
The NRC inspector ascertained the requirements for the assurance
'that the changes do not involve an unreviewed' safety question is catisfied
by the required inclusion in SMP of a nuclear safety review checklist. A ,
positive response to any 'of the' questions on the checklist require that a 1
1
1
i
i
-i .
'
.
.
10
nuclear evaluation form be completed and included in the SMP. This form
requires the design engineer to review and evaluate all the nuclear safety
questions outlined in 10 CFR 50.59.
An interview with two action engineers, was conducted and the NRC
inspector found them to be knowledgeable of the safety /nonsafety-related
classification requirements.
The fire protection guidelines of RG 1.120 are similarly handled by the
inclusion in the SMP of a fire protection / safe shutdown checklist. A
positive response to any of the questions on the checklist and other
.
criteria requires that a fire protection / safe shutdown review analysis
(FP/SSA) be prepared. This document reviews components and fire
protection features for changes to such, and reviews the location of
additional fire loading relative to the modification for impact to
Appendix R to 10 CFR 50 to ensure the level of fire protection does not
decrease. The fire protection / safe shutdown checklist and review form are
a part of a new procedure FP-1-022 that is not released but will be
implemented soon.
A document control system.has been administered that controls the release
and distribution of design change documents, controls changes to released
and approved documents, and provides for the control and recalling of
obsolete design change documents. The procedures referenced, particularly
QP-006-001, detail the controls of released documents.
When the SM is completed the administrative controls and Procedure
PE-2-006 require that a work completion notice (WCN) be sent out after the
operational documents are updated and the control room drawings are red
lined to reflect the changes. This WCN alerts all the reviewing
organizations that the modification is complete. This WCN is the official
notice for the updating of plant procedures, operator training, and the
posting of plant drawings that indicate a change is in place that effects
the drawing. Return of WCN to the station modification coordinator (SMC)
indicates the required documentation update has been completed except for
the affected as-built drawings, which is done prior to SM closeout.
The NRC inspector verified that the responsibility and method of reporting
to the NRC of design changes that are safety-related is established and it
will be an annual report filed 1 year after initial criticality by the
licensing group.
The inspection identified what seems to be a problem in the implementation
of the program although there were no deviations from the established
administrative control and procedure outlines. There is, however, a very I
large backlog of SMP in the WCN and drawing update stage. There were, in
'
fact, only 17 SMPs completely closed out and in project files with all
document updates done. There were 125 awaiting drawing update and 206
..
- _ . . . . ._
l '. g
'
.
t
11
SMPs completed but awaiting some other form of review or document update.
This backlog of SMPs causes some problems in the operational documents
such as the red line drawings, where in at least one case, 5 SMs were
. posted on the drawing as being completed but not marked on the drawing, as
well as 3 additional SMs marked up on the drawing. This represents a
total of 8 SMs affecting 1 drawing without any of them incorporated on the
. drawing. This is considered an open item (8520-01).
The NRC inspector reviewed the temporary modification to lifted leads and
jumpers requirements and found the procedure UNT-5-004 covers all the
requirements when used with additional form referenced in that procedure.
An examination of the temporary modification log indicated it was up to
date and that log entries were complete.
A field examination of the reactor protective system, the emergency safety
features, and the emergency diesel generator control cabinets revealed no
i lifted leads or jumpers in place.
No violations or deviations were identified.
>
8. QA Program Audits
.
This NRC inspection included activities for preparation and issue of the
! audit schedules, development of an audit plan and checklists, review of
objective evidence reviewed during the audit, audit report control and
distribution, and responses _to audits. In general, the audit program was
found to be an effective status of implementation. The Technical
Specification audit program development and vendor audit programs are
progressing satisfactorily and the operations QA program audit are being
i
satisfactorily implemented. The following audits were reviewed by the NRC
inspector:
Audits: 85-45 85-03 85-07 85-15 85-08
,
85-10 84-22 85-13 85-01
Vendor Audits: Combustion Engineering
i
Desselle-Maggard
Cardinal Industries
Rockbestoes
. Southern Vital Records
McGraw-Edison
General Electric
Capitol Controls
Cajun Co. .
Siemen-Allis Co.
Yarway
-_. _ _ _ _ _ .
,
. .
.
12
During this portion of tne audit two items were noted which are considered
an open item and an unresolved item, respectively.
a. QAP 302, paragraph 5.1.2.c requires onsite contr6ctors be audited
triennially. Beyond this, the program does not adequately
address the following areas: .
Auditing onsite contractors in a timely manner once onsite
Establishing an onsite contractor audit schedule
Notification and placement of contractors on a schedule
Identification of work scope to be audited
Tracking and timely closure of audit findings
Procedure for removal from site if necessary
A current example of this concern is Unitec Company. They were
qualified in Jely 1984 to work onsite but the first audit was not
completed until July 25, 1985. (0 pen Item 8520-02)
b. A review of vendor audit files found two active vendors had not
received the required annual evaluation. Southern Vital Records is
missing a 1984 evaluation and Siemens-Allis Company is missing the
1983 and 1984 evaluations. (Unresolved Item 8520-03)
No violations or deviations were identified.
9. Phase III Quality Activities
The inspection in this area concluded that activities committed to are
being effective!y implemented. Commitments include audits, procedure
reviews, and test data reviews. In the event any test deficiencies are
noted, they are tracked via the CIWA system. Plant quality does
get involved with all required holdpoints as deemed required in CIWA
resolutions. Phase III operations QA audits included 84-45, 84-46, 85-01,
and 85-05. An audit of M&TE activities is scheduled for August, 1985.
No violations or deviations were identified.
10. Auditor and Inspector Training
The inspection included verifying certification documents were current,
all necessary supporting documentation was available, and responsibilities
coincided with training and qualifications. All aspects of the training
._ _ _ _ _
.*
,
13
program reviewed during this audit were found to be effectively
implemented.
No violations-or deviations were identified.
11. Control Room Ventilation System Emergency Outside Air Intake Valves
The FSAR discusses the design criteria for control room Nabitability in
Chapter 6.4. The criteria used for location and power supply for the
valves is discussed. The ability to operate the system from the control
room with the loss of a vital bus is one of the design criteria.
OP-03-014, " Control Room Heating and Ventilation," provided the normal
lineup for these valves. Contrary to the FSAR, all valves were normally
aligned closed. The NRC inspector found no evidence that a proper
10 CFR 50.59 review was conducted to calculate dose rates which an
operator would experience if these valves had been manually opened from
outside the control room. When the licensee was informed of this dis-
crepancy, a change was made to the procedure (June 26, 1985).
This is considered a violation.
12. Operational Mode Changes
On June 11, 1985, Waterford 3 Steam Electric Station was in Mode 5 (cold
shutdown) when operations personnel were performirg Surveillance Procedure
OP-903-069, " Integrated Emergency Diesel Generator / Engineered Safety
Features Test." As part of the above procedure, operations personnel were
-
attempting to prove the operability of the Emergency Diesel Generator "B"
automatic load sequence timer as required by Technical Specification 4.8.1.1.2.d.12. While testing Load Block 7, Relay S7X actuated in 121.6
seconds, which was outside the plus/minus 10% tolerance of the sequenced
load block time (168 plus/minus 16.8 seconds). However, operations
personnel did not review the test data until 1545 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.878725e-4 months <br /> on June 20, 1985.
Waterford 3 entered Mode 4 (hot shutdown) at 1028 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91154e-4 months <br /> on June 20, 1985,
with Emergency Diesel Generator B inoperable.
Technical Specification 4.0.4 requires that " Entry into an OPERATIONAL- -
MODE or other specified conditions shall not be made unless the
surveillance requirement (s) associated with the limiting condition for
-operation'have been performed within the stated surveillance interval or
as otherwise specified."
A
LP&L Operating Procedure OP-10.001, Revision 4, " General Plant
Operations," requires that when entering Mode 4 (hot shutdown) both
emergency diesel generators be operable.
. _ _ _ . . ._- _ _
_ ._ . _ . . _. _ . . - . -_ _
.
'
,
t;*8
>
14
This is considered a violation.
,
13. Open Items-
The following new items were identified during this reporting period:
8520-01 Open Item Large Backlog of Incomplete SMPs (paragraph 7)
8520-02 Open Item Contractor Audit Program Inadequate
(paragraph 8a)
!
8520-03 Unresolved Vendor Audit Files Incomplete (paragraph 8b)
Item
8520-04 Violation Failure to Conduct Proper 10 CFR 50.59 Review
(paragraph 11)
8520-05 Violation Failure to Meet Opt.'ational Mode Requirements
(paragraph 12)
! 14. Site Tour
i
At various times during the course of this inspection period, the NRC
inspectors conducted general tours of the reactor building, reactor
auxiliary building, and turbine building to observe ongoing maintenance r
and testing.
No violations or deviations were identified.
15. Exit Interviews
,
The NRC inspectors met with the licensee representatives at various times ~
during the course of the inspection. The scope and, findings of the
. inspection were reviewed.
.
t
4
4
f
i
.
- _ . , , . . _ . . -.- .
- . . . . . _ . - , . . . . , . - - , - . . .