ML20214Q561
| ML20214Q561 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 05/29/1987 |
| From: | Bundy H, Carpenter D, Chamberlain D, Constable G, Cummins J, Hildebrand E, Johnson W, William Jones, Luehman J, Madsen G, Greg Pick, Reis T, Tapia J, Renee Taylor NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20214Q539 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-***, TASK-TM 50-498-87-08, 50-498-87-8, 50-499-87-08, 50-499-87-8, GL-83-28, NUDOCS 8706050075 | |
| Download: ML20214Q561 (80) | |
See also: IR 05000498/1987008
Text
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APPENDIX B
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report: 50-498/87-08 Construction Permits: CPPR-128
50-499/87-08 CPPR-129
Dockets: 50-498
50-499
Licensee: Houston Lighting & Power Company (HL&P)
P. O. B)x 1700
Houston, Texas 77001
Facility Name: South Texas Project, Units 1 and 2 (STP)
Inspection At: STP, Matagorda County, Texas
Inspection Conducted: March 9 through April 10, 1987
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Inspectors:
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.Ma7penter, Senior Resident Inspector D(te /
/f'7
Project Section C, Reactor Projects Branch
w
- flD
,T. heis, Resident Inspector, Project
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Daf.e
Section C, Reactor Projects Branch
.fl2f/S7
H. F. Bundy, Project lYispector, Project Date
Section C, Reactor Projects Branch
WP
J. I. Tai >1a, ReactorInspector, Operations
f/17/N 7
Date
Section, Reactor Safety Branch
G706050075 870529
PDR ADOCK 05000498
E_ __0 PDR __ _ _ _ _ _ _ _ _ _ _
._ _ ___-________________ - - ________ ___ ___________
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2
h
J.~ G. Lu man, Seniof Resident Inspector
587/77
Date
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Project Section C, Reactor Projects Branch
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r J. E. Cummins, Senior Resident Inspector D(ate /2Fb 7
Project Section B, Reactor Projects Branch
l
fW
D. D. Chamtierlain, Serrior Resident Inspector
Tb W97
Date
/ Project Section A, Reactor Projects Branch
.fbfA'?
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W. D. Johns'on~~Senio Resident inspector Date
Project Section B, Reactor Projects Branch
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fAllfff7
G. L. Madsen, Reae dr Inspector, Operations (Tate ~
Section, Reactor Safety Branch
. B. Jon s, Reside ( Inspector, Project ate
Section A, Reactor Projects Branch
- . Y% Al,14 A
G. A. c , Reactor inspector, Operations Dat(
Secti n, Reactor Safety Branch
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/ $
C P. fit Tde' brand, React 6r Inspector Date
29 ~ ~7
Operations Section, Reactor Safety Branch
P
R. G.
s b, Project Inspector, Project
4/o
Dite/
Section , Reactor Projects Branch
Consultants: M. Bishop, F. Jagger, N. Jensen, R. Picker, J. Stachew,
J. Seherman, J. McGhee; EG&G Idaho Inc.
NRC Coop
Student: J. Lara
Approved:
i4.
_ -c<
onstable, Chief, Project Section C
("
Da(e
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Reactor Projects Branch
Inspection Summary
Inspection Conducted March 9 through April 10, 1987 (Report 50-498/87-08;
$0-499/87-08)
Areas Inspected: Routine, unannounced inspection of Technical Specifications (TS),
the structural integrity and integrated leak rate tests (SIT and ILRT),
preoperational test procedures, preoperational test results, the startup
testing program, the as-built plant to documentation reconciliation, the
residual heat removal (RHR)/ component cooling water (CCW) water hammer incident,
operational staffing, training and qualification programs, the reactor coolant
system loss of cleanliness recovery program, the Three Mile Island (TMI) and
GL 83-28 action items, licensee action on previous inspection findings, site
tours, review of the manual trip circuit, and procedures review.
Results: Within the areas inspected, two violations of NRC requirements were
identifled (failure to follow precedures for testing and inadequate cleanliness
controls over an open reactor coc!}qt .S/ stem, paragraphs 8 and 11, respectively).
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DETAILS
1. Persons Contacted
- R. W. Chewning, Special Assistant Nuclear Group l
- S. M. Head, Lead Engineer, Licensing
- D. L. Smith, Management Services Manager
- G. L. Jarvela, Manager, Health and Safety Services
- M. A. Ludwig, Maintenance Manager
- T. E. Underwood, Chemistry Manager
- G. L. Parkey, Technical Support Manager
- J. W. Loesch, Plant Superintendent
- M. T. Sweigart, General Supervisor, Operations Quality Control (QC)
- W. H. Kinsey, Flant Manager
- J. J. Eldridge, Operations Supervisor
- W. P. Evans, Project Compliance Engineer
J. T. Westermeier, Project Manager
F. A. White, Lead Licensing Engineer
R. J. Daly, Startup Manager
J. D. Green, Operations Quality Assurance (QA) Manager
V. E. Geiger, Nuclear Assurance Manager
M. Robinson, Director, Independent Safety Evaluation Group
D. L. Cody, Manager, Nuclear Training
M. E. Smith, Outage Manager
J. Hooper, Employment Counselor
D. Leazur, Reactor Performance Supervisor
T. Godsey, Technical Support Engineer
- Denotes those individuals attending the exit interview conducted on
April 10, 1987.
The NRC inspector also interviewed other personnel of HL&p, Bechtel Power
Corporation, and Ebasco Service, Inc.
2. TS Review
The NRC inspectors and EG&G Idaho consultants reviewed the Proof and
Review copy dated February 12, 1987, of the STP Unit 1 TS. In performing
this review, the following techniques were employed:
o Comparison with NUREG-0452, Revision 5, " Standard Technical
Specifications for Westinghouse Pressurized Water Reactors."
o Comparison with TSs for other recently licensed Westinghouse plants
(Wolf Creek Generating Station and Byron, Unit 1).
o Walkdown of selected systems and components to verify as-built
configurations were reflected in the TS.
o Verification that numerical values of setpoints, operating criteria,
and equipment operating parameters agreed with FSAR and/or
engineering specification values.
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Resulting comments were discussed with the licensee and forwarded to the
NRR project manager by memorandom on March 25, 1987. l
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No violations or deviations were identified,
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An inspection was conducted of the containment SIT procedures, test
performance and test results in order to determine consistency with
regulatory requirements and licensee commitments. The purpose of the SIT
is to demonstrate the ability of the containment structure to withstand
internal loads imposed by pressurizing to 1.15 times the design pressure
of 56.6 psig or 65.0 psig. Bechtel Specification Nos. 20019S50009,
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Revision 2, " Instrumentation For Structural Integrity Test of Containment
Structures," and 2C0195S1013, Revision 1, " Specification For The
Performance of Structural Integrity Test on Reactor Containment Building,
Unit 1," were reviewed by the NRC inspector.
Prior to pressurization of the containment, the NRC inspector toured the
containment and inspected the placement of instrumentation for the SIT.
The containment was subsequently pressurized in five equal pressure
increments. During the 1-hour hold periods between pressure levels,
strains, and deflections were recorded. Surface crack patterns of cracks
. larger than 0.01 in width were recorded at atmospheric pressure before the
test, at the maximum pressure level, and at atmospheric pressure after the
f test.
The NRC inspector monitored the acquisition of data during the maximum
pressure level holding period. Subsequent data analysis determined that
the deflection pattern and strain measurements of the containment were
within predetermined design acceptance criteria. No reportable cracks
were identified.
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At the completion of the SIT, the containment was depressurized to
31.8 psig to perform the ILRT.
The preoperational containment ILRT conducted using the Absolute Method
(as described in ANSI N45.4-1972, " Leakage Rate Testing of Containment
l Structures for Nuclear Power Plants," and ANSI /ANS-56.8 1981, " Containment
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System Leakage Testing Requirements,") was also addressed during this
inspection. The inspection involved procedure and records review, test
witnessing, and independent calculations by the NRC inspector. This ILRT
was conducted in accordance with approved procedures and satisfied the
specified acceptance criteria contained in 10 CFR Part 50, Appendix J,
" Primary Reactor Containment Leakage Testing for Water Cooled Power
Reactors," and in the Plant TS.
Preoperational Test Procedure No. 1-RC-P-03, Revision 0, " Containment
Integrated Leak Rate Test," incorporates the referenced requirements and
criteria. This procedure was reviewed by the NRC inspector and no
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discrepancies from the specified requirements and criteria were noted.
The review provided verification that the following test attributes were
correctly addressed:
o Containment interior and exterior requirements specified
o Instrument locations justified by area surveys
o Instrument calibration requirements specified
o Instrument loss / test abort criteria delineated
o Instrument error analysis performed
o Type B and C test results correction to Type A test results specified
o Venting of internal isolated volumes required
o Isolation valve closing mode specified to be the normal mode
o Proper postaccident system alignment to prevent creation of
artificial leakage barriers specified
o Quality control inspection specified
o Test log entries required for repairs needed to complete test
o Acceptance criteria specified
o Data acquisition requirements specified
o Data analysis technique specified
o Method of depressurization specified
The NRC. inspectors verified that the instrument calibration certifications
traceable to the U.S. National Bureau of Standards for the resistance
temperature detectors, humidity measuring devices, pressure gauges, and
the flowmeter used in the verification test had been reviewed. The
guidelines of ANSI /ANS-56.8-1981 were used to select the instruments for
the ILRT. The formula from the Instrumentation Selection Guide (ISG) was
used during the ILRT to ensure that the data acquisition system accuracy
was sufficient to provide reliable test results. This formula utilizes
the systematic error of each sensor to determine an overall value for the
data acquisition system. The instrumentation system for the ILRT was
based on a computer controlled data acquisition system capable of reading
all sensors rapidly, storing the information and then outputting to the
computer for conversions and calculation of the data. Bechtel Calculation
No. 2R569MC5887, Revision 0, "CLRT Volume Fractions," provided the
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calculational basis for the volumetric distribution of the resistance
temperature defectors located throughout the containment. This
calculation was reviewed by the NRC inspector. -
Af ter a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 31.8 psig to allow for degassing of
structures and components inside containment subsequent to the SIT,
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pressurization of the containment for the ILRT commenced. After the
- internal pressure reached 37.5 psig, the compressors were shut down and
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isolated and the stabilization period commenced. The atmosphere is
considered stabilized when the rate of change of containment temperature
,
averaged over the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> minus the rate of change in containment
temperature averaged over the last hour is less than 0.5 F/ hour. After
the stabilization criteria was satisfied, the ILRT test director declared
the start of the official 24-hour test. Continuation of the test
indicated convergence of the calculated leak rate and the upper confidence
limit below the allowable leakage. At completion of the 24-hour test, the-
4-hour superimposed leak verification test was performed. The NRC
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inspector also witnessed this portion of the ILRT and the result between
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the calculated and imposed leakages was found to be within the 25 percent
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allowable leakage (La) limit.
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Subsequent to the performance of the test, the NRC inspector obtained the
! raw data and computed the leakage rate in accordance with the Mass Point
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Data Analysis technique. The computations performed were compared with
! the licensee's results for the purpose of verifying the calculational [
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procedure and confirming the results. This analytical technique confirmed
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the acceptability of the results obtained by the licensee. The data
! providing the as-left values for the type B and C tests were also
, reviewed.
No violations or deviations were identified,
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4. Preoperational Test Procedures
, The NRC inspector reviewed preoperational test procedures which
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demonstrated the response of the plant's engineered safety features under
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both normal accident and accident coincident with the loss of offsite
l power conditions. The tests were scheduled to be performed in late
j April 1987. The tests were reviewed for compliance with FSAR commitments
! and adherence to Regulatory Guide 1.68 principles. Within the scope of
i the inspection, the procedures were found to comply with the stated
requirements. The specific tests reviewed were:
1-SF-P-01, Safeguards Systems Response - No Blackout
1-SF-P-02, Safeguards Systems Response - Plant Blackout
No violations or deviations were identified.
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5. Preoperational Test Results Review
The NRC inspectors began the review of results for selected completed
preoperational tests. The reviews were done to verify that the testing of
systems met specified acceptance criteria, problems encountered during
testing were properly resolved, appropriate reviews of tests results were
performed by the licensee, and approved administrative controls were
followed during the conduct of testing. The tests reviewed included:
1-MS-P-02-01, Main Steam Isolation Valve Logic
1-MS-P-03-01, Main Steam Power Operated Relief Valves and Main Steam
Dump
1-PK-P-03-01, IE AC Power Distribution Train C
1-RC-P-01-01, RCS Cold Hydrostatic Test
l- 1-RC-P-05-00, RCS Pressurizer Relief Tank
l 1-RC-P-13-00, RC Pump Check
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4 1-RH-P-01-01, RHR System Train A
1-RH-P-02-01, RHR System Train B
1-RH-P-03-01, RHR System Train C
1-RM-P-01-00, Reactor Makeup Water System
1-RS-P-02-00, Rod Control System
1-SI-P-01-01, SI Train A/B/C and Common Logic
1-SI-P-02-00, SI Accumulators
1-SI-P-04-00, SI Train (A,B,C) Performance
1-SP-P-01-00, Reactor Protection Logic
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, 1-SP-P-02-00, Reactor Protection Master Relay
1-VA-P-02-01, 120V AC Class IE Vital Power Channel II
Based on the reviews of the above tests, the NRC inspectors had the
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following comments.
a. The NRC inspectors noted a large number of instances in which the
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administrative requirements of Startup Administrative
Instruction (SAI) 18, "Preoperational Testing," were not followed.
The instances included failure to use the test change notices (TCN)
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to correct procedure errors, failure to make chronological log
entries for events specified in SAI 18, and a person other than the
person making a verification entry in a test procedure correcting
that entry. These administrative problems were discussed with the
licensee's startup manager and members of his staff. They stated
}thattherequirementsofSAI18havebeengraduallyputinplace
through the five revisions of the procedure, and that future
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preoperational test procedures should demonstrate stricter adherence
to the present administrative requirements.
b. In at least two procedures (1-SI-P-01 and 1-PK-P-01), TCNs were
generated to change test requirements, but the TCNs made no
references to the document or drawing that justified the change.
Justification for changing Steps 7.9.67 and -71 of 1-SI-P-01 from 15
to 19.4 and Step 7.5.95.2 of 1-PK-P-01 from black to red is
considered an open item pending HL&P providing the needed
justification and is identified as 498/8708-03.
c. The acceptance criteria for Steps 7.3.2, 7.3.41, 7.3.44, 7.4.20,
7.6.2, and 7.9.41 of 1-SI-P-04 were not met during testing, and the
NRC inspectors had two concerns with the dispositioning of these
nonconforming conditions. First, in each case, the test engineer
signed-off the step even though the acceptance criteria were not met.
The NRC inspectors were told by the licensee that the accepted
practice is to sign off the step as having been performed whether or
not acceptance criteria were met. The NRC inspectors could not find
any procedural guidance to support or prohibit this practice.
In the case of each step specified above, once unacceptable test
results were obtained, testing continued. SSP-8 requires that for
testing to continue after discovering a nonconforming condition, a
'\ conditional release must be obtained from the startup manager and the
quality assurance manager or a determination must be made that the
nonconformance will not affect continued testing. There was no
documentation of either of these conditions being met, and the NRC
inspectors could find no procedural guidance concerning who could
make the above determination.
The licensee explained that, although not documented, in each case
the test engineer made such a determination and, though not
proceduralized, a decision on the part of the test engineer was what
was intended by SSP-8. The NRC inspectors emphasized the need to
properly document the basis for continuing testing. This relates, in
part, to a and b above.
d. The NRC inspectors noted that in almost all procedures reviewed, the
" Witness" blocks, used for verifying the removal of temporary jumpers
used for testing, were signed days, weeks, and, in one case, months
after the test. This practice raised two questions. First, how
could someone witness jumper removal days after it was performed?
Second, how was the licensee realizing the full intent of the second
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verification of jumper removal if a missed jumper could remain in the
system for long periods, potentially invalidating further testing?
The licensee responded to the first question by saying that the
initial blocks for jumper removal should have been more properly
labelled " Verification" rather-than " Witness." In response to the
second question, the licensee will review the practice of delaying
verification of jumper removal.
Two specific cases of contradicting verification dates will require
formal resolution. In 1-RC-P-13, QC verification signatures were
made 1 day later than the test engineer's verification in
Sections 7.5.28-30 and -32, and 7.6.73, .74, .76, .77, .79, and
,
.80. In 1-MS-P-02, QC verification signatures were made 1 day
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earlier than the test engineer's verification in Sections 7.1,
7.16.32 .39, 7.17.1 .36, 7.19, and 7.20. These two cases of
contradicting entries are considered an open item (498/8708-04).
6. Startup Testing Program
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During this inspection period, the NRC inspectors began reviewing the
startup testing' program. The NRC inspectors crovided the licensee
a listing of 23 Regulatory Guide 1.68 line items which could not be
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closed. Major issues were as follows:
. a. Licensee's system for taking credit for preoperational testing and
i some surveillance-tests as coverage for RG 1.68 line items.
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b. The need for additional involvement of operations and Plant
- Operations Review Committee (PORC) in the evaluation of acceptability
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of preoperational tests for satisfaction of RG 1.68 requirements.
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The licensee committed to developing additional information which will be
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tracked as Open Item 498/8708-05.
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No violations or deviations were identified.
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7. As-Built Plant to Documentation Reconciliation
The NRC inspectors inspected selected systems to determine that they were
L installed in accordance with commitments contained in the FSAR and
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referenced drawings and specifications. In performing this inspection,
mechanical and fluid system walkdowns were performed. Also, draf t TS
surveillance test procedures were checked to be sure they could be
accomplished for the as-built system. Portions of the controls and
instrumentation were verified to conform to the descriptions contained in
the FSAR. The following systems were selected for walkdowns:
o Reactor Coolant System (RCS)
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o : Chemical Volume Control System
o Reactor Water Makeup System
o RHR System
o .High. Head Safety Injection System
o Low Head Safety Injection System
o Accumulator Injection System
o Containment Spray & Spray Additive Systems
o Control Room Emergency Air Cleanup System
o Auxiliary Feedwater System
o Emergency Diesel Generator Support Systems - Air, Lube Oil, and
Fuel Oil
o 480 VAC MCC EIA1
o Fuel Building Exhaust (HVAC) System
o -Containment Ventilation Subsystems
o Essential Cooling Water System
Selected systems generally conformed to the FSAR descriptions. For the
few discrepancies observed, the NRC inspectors verified that appropriate
design change packages existed and action had been taken to update the
FSAR drawings and descriptions.
- No violations or deviations were observed.
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'8. RHR/CCW Water Hammer Incident
a. Description of Incident
- During hot functional. testing (HFT) on Unit 1, the RCS was being
- maintained at 350 F and 350 psig with four reactor coolant pumps in
i operation. CCW was in service with all three pumps operating and all
! three RHR heat exchangers on line. Trains A and C of the essential
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cooling water (ECW) were in service supplying only Trains A and C
ECW/CCW heat exchangers. Minimal heat loads and mixing of CCW in the
i common portions of the system allowed operation of the B train CCW
i pump without its associated ECW train in operation. Preparations
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were underway to place Trains B and C RHR in service to support
Preoperational Test 1-RH-P-04, "RHR Thermal Performance Test."
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The incident occurred while an operator was placing two trains of RHR
in service to support the performance of the preoperational test.
>- .The controlling HFT procedure directed that this operation be
performed in accordance with Operations Work Order
Request (0 WOR) 1-RO-RH-236. This 0 WOR is a modified version of the
normal Operating Procedure IP0P02-RH-0001, Residual Heat Removal
System Operation, for use during HFT only. While performing this i
procedure, the operator failed to close the RHR Heat Exchanger Outlet -
Valves (RH-HCV-865/866) as directed by the procedure. This resulted
in initiation of flow of hot RCS fluid (350 ) through the RHR Heat
Exchangers when the Cold Leg Injection valves were subsequently
opened. Since CCW to the heat exchangers had been terminated at the
verbal direction of the Startup Test Director to enhance heatup of
the RHR loops, no cooling medium was provided and rapid heatup and
void formation of the CCW in the heat exchangers occurred.
L When the RHR Heat Exchanger CCW Outlet Isolation valves were
reopened, CCW flow was established to the C Train RHR Heat Exchanger
only. This occurred because CCW Pump B had been secured to clear
flow alarms on CCW Trains B and C. This resulted in Train B RHR Heat
Exchanger being without cooling water for a longer period of time.
This prolonged no flow condition permitted a greater degree of
heating and void formation in the B RHR Heat Exchanger which
contributed to a' greater degree of water hammer on that train.
f
This incident was caused by the operator's failure to follow
procedures in that he skipped two steps in Procedure 1-RO-RH-236 and
the Startup Test Director's failure to follow procedure by verbally
directing a change.in the approved preoperational test procedures
instead of using the authorized TCN method. These actions constitute
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an apparent violation (498/8708-01). This incident resulted in minor
water hammer in Train C CCW and a greater water in Train B CCW. The
Train B water hammer resulted in damage to pipe hanger, supports, and
the CCW piping itself. The system was not breached nor was it
overstressed, based on engineering analysis. The licensee took
prompt corrective action on problem analysis and system repairs.
Further review of this incident will be documented by the licensee in
the final 10 CFR 50.55(e) Report and Station Problem Report.
,
b. Technical Review
l The NRC inspector examined two runs of CCW system piping supplying
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the jacket side of RHR Exchanger 1B that was involved in a water
hammer incident on March 11, 1987, during a preoperational test.
Based on the examination and interviews with licensee personnel, the
NRC inspection initially concluded that there was no apparent damage
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to either the piping or connected components. Several pipe supports
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were damaged to varying degrees which have been identified and
- addressed on nonconformance reports. The licensee reported the event
- pursuant to 10 CFR 50.55(e) on March 12, 1987. Pending conclusion of
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the licensee's evaluation of this matter, which is to be documented
in a report to the NRC, this matter will be considered an open item
(498/8708-06).
No additional violations or deviations were observed.
9. Operational Staffing
The NRC conductad a review of the organizational staffing and staff
>
qualifications of personnel assigned to the operational phase for Unit 1.
Discussions with appropriate personnel, reviews of organization charts,
reviews of 55 personnel resumes, and reviews of certification records
indicated that the operational staffing was in accordance with the STP
FSAR and the licensee's commitments to ANSI N18.1 with the following noted
exceptions:
a. The Manager of Reactor Operations position is vacant. This vacancy
and the overall level of nuclear plant operational experience
requires further evaluation.
b. The Independent Safety Engineering Group (ISEG) has not been fully
staffed. The Director of ISEG has been named; however, the
additional four engineers have not been selected. The licensee is
committed to the formation of the ISEG prior to fuel loading of
Unit 1.
c. The Operational QA organization described in Amendment 54 to the FSAR
differs from the in place organization. Discussion revealed that an
Amendment 58 to the FSAR is in preparation for submittal to NRR. A
review of the draft Amendment 58 revealed the proposed organizational
submittal to be in accordance with the existing organizational
structure.
Filling the Reactor Operations position, completing evaluation of
operational experience, and implementation of ISEG will be tracked
collectively as Open Item 498/8708-07.
No violations or deviations were identified.
10. Training and Qualification Programs
The NRC inspectors reviewed the licensee's training programs for licensed
operator and nonlicensed staff to verify that regulatory requirements and
license commitments are being met or that programs have been developed to
implement the training requirements and commitments. The administrative
programs to ensure that classroom and simulator training is based on
up-to-date training materials that reflect the as-built condition of the
plant and approved procedures were also reviewed.
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The licensee has clearly established responsibilities for administering
the training programs including evaluating, scheduling, assigning
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qualified instructors, examining, retraining, and record keeping. The
Manager, Nuclear Training oversees the implementation of the training
program and is assisted through direct line responsibility by the Training
Manager, Operation Training Division; Training Manager, Staff Training
Division; and the General Supervisor, Program / Administration Support
Section. The NRC inspector noted after discussions with training
management personnel that the licensee has established a strong commitment
to provide qualified personrel to operate and maintain STP. One example
of this commitment is that the licensee is presently greater than
70 percent complete with the program development for the 10 Institute for
Nuclear power Operations (INPO) training programs needed for
accreditation.
The NRC inspectors reviewed the licensed operator training program and
verified that the following program elements have been established and
implemented when required for:
o new reactor operators (RO);
o upgrading R0 licenses to senior reactor operator (SRO) licenses;
o qualifying instructors and shift technical advisors (STAS); and
o requalification of R0s and SR0s in accordance with the requirements
The licensee has demonstrated their commitment to provide an effective
' licensed operator training program as evidenced by the NRC first cold
license exams. The results of the exams were that 41 of 46 SRO candidates
passed the exam and the only R0 candidate passed the exam. The first
requalification cycle for these individuals began in April of this year.
The licensee has completed the first cycle training material for the
requalification program. The first group of plant equipment operators is
presently going through the RO training program. The progress of each
license candidate is being evaluated by weekly written and/or oral exams.
The results of these evaluations are used to determine what additional
training the individual requires or if any other actions should be taken.
The licensee also utilizes a fully operational reactor simulator located
at the STP site. The presence of the simulator should enhance the overall
effectiveness of both the initial and requalification training programs.
j
l The NRC inspectors reviewed the licensee's program for nonlicensed staff
i training and verified that personnel are being trained in the areas of
administrative controls, industrial safety, fire fighting, and QA. The
licensee demonstrated a strong commitment to the training of nonlicensed
- staff personnel; however, development of the on-the-job training (0JT)
l
program has not been completed. The complete development of this program
j will require a concerted effort between the applicable user groups and the
t training department. The NRC inspectors noted that the licensee has
dedicated several training personnel to complete the OJT program in the
j
past few months.
!
!
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15
Because of the manhours required to complete the 0JT program, the licensee
does not believe the task can be completed adequately prior to their
projected fuel load date. The NRC staff places great importance on
licensees' programs which provide training for personnel in the
performance of specific tasks such as complicated surveillance tests,
major equipment repair, major plant systems tests, and other special
procedures for which the DJT program is an integral part of that training
program. Based on the importance of the OJT program, the licensee must
provide supplemental training to plant personnel in the area that will be
covered by the DJT program until the OJT program has been developed and
implemented. The implementation of the supplemental training program
pending completion of the OJT program is an open item (498/8708-08). The
training of QC inspectors is presently being carried out through the QA
department. The training department has planned to take over the training
of QA/QC inspectors and to also provide training for engineering support
personnel such as ISEG. The development of these training programs will
be reviewed during subsequent NRC inspections.
The NRC inspectors also reviewed the licensees administrative programs for
ensuring that classroom and simulator training is based on up-to-date
training materials that reflect the as-built condition of the plant and
approved operating procedures as well as events or conditions identified
by the NRC, INPO, or other facilities. The licensee has established a
program for implementing changes to the facility and procedures into the
lesson plans. In addition, STP personnel can request training through
their supervisor through the use of the " Request For Training Assistance"
form, Modifications to the simulator were also reviewed and found to lag
only 5 months behind changes to the Unit 1 control panel which
demonstrated that plant modifications were being incorporated into the
simulator.
No violations or deviations were identified. ,
11. Reactor Coolant System - Loss of Cleanliness Recovery Program
On April 6,1987, the resident inspector accompanied by a regional reactor
inspector entered the reactor coolant system to independently verify
damage incurred by the resistance temperature detectors (RTDs) and to
inspect for screen remnants lost during HFT. During their entrance, the
NRC inspectors noted a general disregard of the principles of foreign
material exclusion and personnel and material accountability. At that
time various work, internal to the reactor coolant system, was being
performed via a startup work request (SWR). Rework via an SWR subsequent
to system turnover from the construction activity requires the
implementation of SSP-22. Section 2.2 of SSP-22 clearly defines the
applicability of the procedure.
Section 5.4.2 of SSP-22 applies to cleanliness controls of
,
systems / components after turnover from the construction activity and
requires the following:
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.. .
16
o 5.4.2.1 Performance of activities associated with a SWR shall
require additional attention to internal cleanliness as
outlined within this section.
o 5.4.2.2 Upon receipt.of a SWR, the appropriate discipline field
engineer (DFE) shall determine the following to preclude
entry of debris / contaminants into a system, if activity
exposes internal surfaces:
(a) Specific area controls required.
(b) Special methods to be utilized for entering the
system.
(c) Special methods required for maintaining system
cleanliness during performance of the activities.
The NRC inspectors discovered these requirements were clearly violated.
The reactor cavity and all hot and cold legs were open to a cluttered
construction type environment. Workmen entering the system did not have
their personal items and tools lanyarded to themselves and area controls
were insufficient. This is an apparent violation of Criterion V of
Appendix B of 10 CFR 50, which is implemented by the STP project GA plan
Section 5.2.1, in that the licensee failed to follow established
procedures for maintaining cleanliness (498/8708-02).
This situation was brought to the attention of project management the same
evening and management's response was that controls were deliberately
dropped because recleaning would have to be performed subsequent to
further work on the system. On April 7, 1987, Operations QC was notified
of the situation. They agreed that the ongoing work was not in
conformance with SSP 22 and issued Nonconformance Report (NCR) SN-03315 on
April 8, 1987. The NCR reiterated the deviations from SSP 22 and
ANSI N45.2.1-1973. On the evening of April 7, 1987, the NRC inspectors
further discussed the situation with the project manager, the deputy
project manager, and the startup manager. The NRC inspectors acknowledged
understanding of the logic behind the decision to abort cleanliness and
accountability control, but expressed concerns that cleanliness level B
was not formally relaxed and that present cleanliness controls were
inadequate and not in compliance with approved procedures.
No other violations or deviations were observed.
12. TMI and GL 83-28 Action Items
The NRC inspector and consultants examined the licensee's conformance with
the requirements set forth in NUREG-0737 and GL 83-28. The following
describes the NRC's position, observations, conclusions, and current
status of the TMI and GL 83-28 action items:
. .
17
(Closed) Generic Letter 83-28 Item 1.1 Post-Trip Review
Licensees and applicants shall describe their program for ensuring that
unscheduled shutdowns are analyzed and that a determination is made that
the plant-can be restarted safely.
A review was conducted of the applicant's response to item 1.1 of GL 83-28
(Salem ATWS) Post-Trip review process. It was determined by a point by
point review, that the STP Procedure OPGP03-ZO-0022, " Post-Trip Review,"
meets the requirements of 83-28 Item 1.1 in all areas.
Item 1.1 of GL 83-28 is considered closed.
(0 pen) TMI Item II.K.1.10 Operability Status of Safety-Related Systems
Review and modify as necessary your maintenance and test procedures to
ensure that they require:
o Verification, by test or inspection, of the operability of redundant
safety-related systems prior to the removal of any safety-related
system from service
o Verification of the operability of all safety-related systems when
they are returned to service following maintenance or testing
o Explicit notification of involved reactor operational personnel
whenever a safety-related system is removed from and returned to
service
The following procedures were examined to see if they fulfill the
requirements stated in the NRC position:
o OPGP03-Z0-0004, Part 4.10.6.1
o OPGP03-Z0-0001, Part 5.2.1.a
o OPGP03-ZO-0004, Part 4.10.7
o OPGP03-Z0-0003, Parts 4.10.3 and 4.10.23
.
l The examined procedures meet the intent of the NRC position, with one
'
deficiency. The wording in OPGP03-ZO-0004 and OPGP03-ZO-0001 uses
"should" where a stronger "shall" would be more appropriate since the
verification in question is mandatory, not optional. Until the wording
l choice of "should" rather than "shall" is resolveu, this is an open item
i (493/8708-09).
! (Closed) TMI Item I.C.6 Verification of Correct Performance of Operating
Activities
i
The applicant will have a procedure for verifying the correct performance
of operating activities. The Shift Supervisor, or, in the absence of the
Shift Supervisor, the Unit Supervisor (an SRO) should be responsible for
releasing equipment for testing, maintenance, or modifications. Following
, .
18
such activities, a qualified person from the shift crew (who does not have
to be a licensed operator) should be assigned to independently verify the
proper positioning of valves, circuit breakers, and control switches of
the systems that are important to safety,
o STP Procedure OPGP03-ZA-0010, Revision 2, " Plant Procedure
Compliance,' Implementation, and Review," sets forth the methods and
requirements for ensuring proper procedure compliance and independent
verifications.
o STP Procedure OPGP03-ZA-0039, Revision 3, " Plant Procedures Writers
Guide,"_ provides the guidelines for writing a procedure to include
the proper precautions (e.g. permissions required and independent
verifications).
o STP Procedure OPGP03-Z0-0005, Revision 0, " Reactor Operations
Division Conduct of Operations," was reviewed. The review indicated
the Shift Supervisor and Unit Supervisor have proper authority to
release equipment for testing, maintenance, or modifications.
Based on the above information it is concluded that the applicant has met
the requirements for Item I.C.6. Therefore, Item I.C.6 is considered
closed.
(Closed) TMI Item I.C.5. Feedback of Operating Experiences
Each applicant for an operating license shall prepare procedures to assure
that operating information pertinent to plant safety originating both
within and outside the utility organization is continually supplied to the
operators and other personnel and incorporated into training and
retraining programs.
o An interview was conducted with the Lead Engineer of the STP
Regulatory Compliance Group. During the interview, he displayed the
methods used to track, classify, and route inside and outside events
to the various facility departments according to STP
Interdepartmental Procedure IP-2.2Q, Revision 2, " Operating
Experience /In-House Experience Review." He also displayed, from the
License Commitment Tracking System (LCTS), how the operations
department responds to action items of the LCTS.
o The Training Department procedures for LCTS action items was
described by the Manager of Operator Training. The methods used by
the Training Department cover both initial training and retraining of
operators.
On the basis of the above information, it is concluded that the applicant
has coa: plied with NUREG-0737 Item I.C.5. Therefore, Item I.C.S. is
considered closed.
. .
19
(Closed) TMI Item I.C.8 Pilot Monitoring of Selected Emergency Procedures
for Near-Term Operating License Applicants
Applicants will be required to correct any deficiencies identified by an
NRC sample audit of selected emergency operating procedure (EOPs) before
full power operation.
o STP SER (NUREG-0781) Section 13.5.2.3 states, "This pilot monitoring
program was used on an interim basis for evaluation of applicant's
E0P's before staff approval of generic technical guidelines and staff
development of the long-term program for the upgrading of E0P's.
This is no longer necessary because the NRC has approved the
Westinghouse Emergency Response Guidelines (ERG's) and the applicant
has committed to develop E0P's based on the ERG's."
o The applicant has in place all applicable E0Ps, as recommended by the
Westinghouse ERGS. The effective dates for these procedures are
01-05-87 (42 procedures) and 02-23-87 (6 procedures).
o Four STP E0Ps were reviewed and found to be in compliance with the
format, intent, and content of the Westinghouse ERGS. These
procedures are:
o IPOP05-E0-E000, Revision 1, " Reactor Trip or Safety Injection"
o IPOP05-E0-E010, Revision 1, " Loss of Reactor or Secondary
Coolant"
o 1 POP 05-E0-ES02, Revision 1, " Natural Circulation Cooldown"
o 1 POP 05-E0-EC00, Revision 1, " Loss of All AC Power"
Item I.C.8 is considered closed.
[0 pen) TMI Item I.D.1 Control Room Design Review
All licensees and applicants are required to conduct a detailed control
room design review (DCRDR) to identify and correct design deficiencies.
o STP Safety Evaluation Report (SER), NUREG 0781, states in Section 18
that the applicant shows a commitment to comply with the requirements
of this item. However, the SER further states that to complete the
DCRDR activities, the following items must be resolved:
o Provide the results of the verification and validation program
for the final E0Ps to confirm that the instrumentation and
control needs have been adequately identified and satisfied.
o Provide the results of the investigation of the green
Roto-tellite indicating lights in the control room under actual
operating conditions.
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. - , _ - - - . . - - _ . . - . _ _ - . - - -
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20
o Provide the results of the surveys of the lighting, sound,
meter, and communication system, when planned work in the
control room is completed.
o A letter from M. R. Wisenburg, STP Deputy Project Manager to
. Vincent S. Noonan of the NRC Pressurized Water Reactor (PWR) Project
Directorate No. 5, dated December 26, 1986, (ST-HL-AE-1864) contains
Addendt.m 1 to the Human Engineering Deficiency (HED) Resolution
Report and Addendum 2 to the Executive Summary. These documents
include the results of the green Roto-tellite indicating light
investigations in the control room under actual operating conditions
(Pages A-7 and A-8 of the HED report and Page S-1 of the Executive
Summary) and conclude that, after modifications, the visibility of
these lights under actual operating conditions is acceptable.
o A letter from M. R. Wisenburg, STP Deputy Project Manager to
Vincent S. Noonan of the NRC PWR Project Directorate No. 5, dated
December 23,1986, (ST-HL-AE-1860) contains the initial submittal of
the E0P Validation Report. Control panel deficiencies and problems
identified during the E0P validation each have an associated
resolution identified with them. E0P problems encountered during the
validation process are documented, and their resolutions are
.
addressed in the STP Procedures Generation Package (PGP).
Item I.D.1 remains open (498/8708-10) pending submittal of the results of
the surveys of the lighting, sound, meter, and communication system
l (Closed) TMI Item I.A.1.1 Shif t Technical Advisor (STA)
Each licensee shall provide an on-shift technical advisor to the Shift
- Supervisor. The STA may serve more than one unit at a multiunit site if
'
qualified to perform the advisor function for the various units.
I o STP SER 13.2.2 provided the conclusion that the-STA Training program
', was acceptable.
l o A review of STP Station Procedure OPGP03-ZO-0008, Revision 0, " Shift
'
Technical Advisor," outlining the duties, responsibilities, training,
experience, and retraining, was conducted and found to satisfy the
- requirements of NUREG 0737.
! o An interview was performed with an STP Reactor Engineer qualified as
'
the staffing levels of the STAS (present and near future). The
interview provided information that the STA program is following the
training program and station procedures.
Item I.A.11 is considered closed.
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21
.( Closed) TMI Item I.A.2.1.4 Immediate Upgrade of R0 and SRO Training and
Qualifications - Modify Training
Licensees and applicants shall review their training programs and upgrade
them, as necessary, to include items relating to TMI-2 lessons learned.
Any necessary training program modification must be in place prior to fuel
load.
o STP SER (NUREG 0781) concludes in Section 13.2.1.3 that the STP
training program meets the requirements of I.A.2.1.
o STP TS 6.4.1 requires that the training program shall include the
requirements of Sections A and C of Enclosure 1 of the March 28,
1980, NRC (H. R. Denton) letter to applicants and licensees
concerning qualification of reactor operators,
o STP Interdepartmental Procedure IP-8.8, Licensed Operator Training
Program, stipulates operator and instructor selection criteria and
licensed operator training requirements. These were reviewed and
found to be consistent with the requirements of STP TS 6.4.1.
o STP Interdepartmental Procedure IP-8.9, Licensed Operator
Requalification, provides instruction for the conduct of licensed SRO
and RO requalification. This procedure was reviewed and found to be
consistent with the requirements of TS 6.4.1.
Item I.A.2.1.4 is considered closed.
[Open) TMI Item I.C.4 Control Room Access
Licensees are to assure instructions are in place which cover the
authority and responsibilities of the person in charge of control room
access, and establish clear lines of authority and responsibility in the
control room during emergencies,
o STP' SER (NUREG-0781) Section 13.5.1.2 states that the applicant has
committed to limit access to the control room,
o STP Procedure OPGP03-ZO-0005, Revision 0, Reactor Operations Division
Conduct of Operations, Section 4.1, was reviewed with the following
findings.
o Section 4.1.1 states that if the Unit or Shift Supervisor deems
the number of people in the control room to be excessive, he has
authority to direct excess people to leave and may require
personnel who need further entry to obtain his prior approval,
o Section 4.1.2 states that the individual screening for control
room access shall be controlled by the security system.
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22
o Section 4.1.3 states that during a reactor trip or other plant
abnormality "it is recommended that all personnel not directly
involved with the recovery leave the control room. Assistance
from other persons shall be determined and requested by the
Shift Supervisor, Unit Supervisor, or the Reactor Operator."
The language used in Section 4.1.3 of OPGP03-Z0-0005 is not strong enough
to " establish clear lines of authority and responsibility in the control
room during emergencies." Conspicuously absent are directions which
require all nonessential personnel to leave the control room during
emergencies; instructions which stipulate who, specifically, is
responsible and authorized to control access during emergencies; and
instructions which specify who, specifically, may enter the control room
during emergencies (e.g. licensed operators, the STA, the NRC resident
inspector). Item I.C.4 remains open (498/8708-11) pending clarification
of control room access requirements during emergencies.
(Closed) TMI Item I.C.2 Shift and Relief Turnover Procedures
Licensees are to revise plant procedures for shift and relief turnover to
ensure that each oncoming shift is made aware of critical plant status
information and system availability,
o STP procedure OPGP03-ZA-0063, " Reactor Operations Shift Turnover,"
was reviewed and found to meet the intent of NUREG 0694, Item I.C.2,
as the applicant had committed to in the STP SER Item 13.5.1.4.
o Direct observations were made of the shift relief process, including
discussion with on-duty R0, as to how the turnover check lists were
used and process of turnover accomplished. In addition, the
on-coming crew briefing was observed. The observations indicated
that the procedure noted above is being used to accomplish shift
turnovers and personnel are being made aware of system availability
and critical plant status information prior to assuming control of
the plant.
1.C.2 is considered closed.
(Closed) TMI Item I.C.7 NSSS Vendor Review of Procedures
Applicants are required to obtain reactor vendor review of their low
power, power ascension, and emergency procedures as a further verification
4
of the adequacy of the procedures.
o STP SER (NUREG-0781) 13.5.2.3 was reviewed and states that because
the applicant has committed to implement procedures based on the NRC
approved ERGS, the staff does not consider an additional NSSS vendor
review of the E0Ps necessary. Furthermore, this section states that
STP committed to have the low power testing procedures and power
ascension testing procedures reviewed by Westinghouse.
_ _
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- -_ .-. . _ - _. _ - _ - _ _- -
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23
I
o Four STP E0Ps were reviewed and found to be in compliance with the
format, intent, and content of the Westinghouse ERGS. These
procedures are:
i o IP0P05-E0-E000, Revision 1, Reactor Trip or Safety Injection
o IPOP05-E0-E010, Revision 1, Loss of Reactor or Secondary Coolant
,
o 1 POP 05-E0-E502, Revision 1, Natural Circulation Cooldown
- o IPOP05-E0-EC00, Revision 1, Loss of All AC Power
,
o The Lead Engineer, STP Regulatory Compliance Group, indicated after
4
conversations with an STP Technical Support Supervisor that all
IPEPO4-series procedures (Initial Startup Test Procedures) were
reviewed by Westinghouse. This conforms to the requirement in STP
Procedure OPGP03-ZA-0002 Plant Procedures, Addendum 2, Step 3, which
states that Westinghouse shall be designated to review and comment on
the initial issuance of the initial startup test procedures.
o An STP Reactor Engineer indicated in conversation with the Lead
Engineer, STP Regulatory Compliance Group, that all, or parts of, the
following procedures were submitted to Westinghouse for review and
comment, based upon design features which are unique to STP:
o IPOP05-E0-FRC1 Response to Inadequate Core Cooling
,
o IPOP05-E0-FRC2 Response to Degraded Core Cooling
l o IP0P05-E0-F002 Core Cooling Safety Function Status Tree
a o IP0P05-E0-E000 Reactor Trip or Safety Injection
j o IPOP05-E0-ES13 Transfer to Cold Leg Recirculation
'
Additionally, he indicated that informal constructive interchange between
STP and Westinghouse was ongoing thrauchout the development of the E0Ps.
i TMI Item I.C 7 is considered closed.
,
(0 pen) TMI_ Item _II.E.1.2 Auxiliary Feedwater System Automatic Initiation
! and Flow Indications
i
, The applicant will provide an auxiliary feedwater system (AFWS),
, initiation and flow monitoring capabilities.
.
. o A review was conducted of the STP SER. The staff concludes that the
l AFWS meets the guidelines of NUREG-0737 concerning reliability and
that the AFWS is compatible with staff guidance for unavailability
per Standard Review Plan (SRP) Section 10.4.9.
.
,
o An interview was conducted with two STP startup engineers (SEs) to
i determine the functional capabilities demonstrated by the AFWS system
! during the recent HFT. Both SEs stated, when questioned, that the
AFWS logic, power supplies, flow indication, and valves functioned as
j designed. However, the integrated safeguards test (IST) which will
i
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complete the automatic circuitry testing in regard to loss of power,
auto sequenci19, and automatic safety injection actuation remains to
be completed,
o A review was conducted of the training department lesson plans and
system descriptions involving the AFWS to determine if NUREG-0737
items were addressed. The training department's lesson plans covered
all areas of the AFWS design, operation, and procedures. The depth
of training was describe!d as systems training, classroom work, and
simulatortrainingbythhManagerOperationsTrainingandSupervisor
Operations Training. ;
Item II.E.1.2 remains open (498/8708-12) pending the completion of the IST
and submittal of results and acceptance by NRC.
(Close_d) TMI Item II.E.3.1 Emergency Power for Pressurizer Heaters / Upgrade
Power Supply
Applicants shall provide the capability to supply, from either the offsite
power source or the emergency power source (when off site power is not
available), a predetermined number of pressurizer heaters and associated
controls necessary to establish and maintain natural circulation at hot
standby conditions,
o STP Procedure IPOP05-E0-E000, Revision 1, states the conditions when
the pressurizer heaters are to be loaded onto the emergency diesels.
o An interview was conducted with three STP SEs. They indicated that
the pressurizer heater modifica tion was installed and functioning
properly on the Engineered Safety Features (ESF) power supply,
o STP Plant Electrical Drawing 9-E-PLAA-01, Revision 6, and
9-E-PLAA-01, No. 1, Revision 9, " Single Line Diagram 480V Class IE
Load Center," both show heater groups power coming from ESF supplie:.,
Heater Group "A" from Bus EIA1, and Heater Group "B" from Bus E1C1.
o A review of the training departnent lesson plans indicated that the
training department has ensured operators have received proper
training on the pressurizer heaters.
On the basis of the above information Item II.E.3.1 is considered closed.
(Closed) TMI Item I.A.1.2 Shift Supervisor Responsibilities - Delegate
Non-Safety _ Duties
Administrative functions that detract from, or are subordinate to, the
management responsibility for assuring the safe operation of the plant are
to be delegated to other operations personnel not on duty in the control
room.
_
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25
o Section 13.1 of the STP FSAR, Organizational Structure of Applicant,
was reviewed. It was found that provision is made in the STP
organization for an on-shift administrative aide who reports
functionally to the shift supervisor of the assigned work shift.
o The job description applying to the administrative aide position was
reviewed and found to include administrative responsibilities which
should significantly reduce the burden on the shift supervisor caused
.
by ancillary responsibilities.
!
o Four individuals are currently filling Administrative Aide positions
on shift. The Manager of Administrative Services stated that these
( individuals are assigned to a three-shift rotation such that an
Administrative Aide is assigned on-shift at all times.
Item I.A.1.2 is considered closed.
(Closed) TMI Item II.K.1.17 Trip Per Pressurizer Low Level Bistable
Facilities which use pressurizer pressure for automatic initiation of
safety injection into the reactor coolant system must trip the low
pressurizer level setpoints such that, when pressurizer pressure reaches
the low setpoint, safety injection will be initiated regardless of the
pressurizer level.
o Drawing No. 5Z-10-9-Z-42112, Revision 7, "SSPS Logic Diagram," was
reviewed. It was ascertained from this review that STP does not
utilize pressurizer water level coincident with pressurizer pressure
for automatic initiation of safety injection. When any two of four
channels of pressurizer pressure decrease to a predetermined value
which corresponds to the low pressurizer pressure safety injection
setpoint, automatic safety injection will be initiated. Therefore,
TMI Item II.K.1.17 does not apply to STP.
TMI Item II.K.1.17 is considered closed.
(Closed) TMI Item I.C.3 Shift Supervisor Responsibilities - Corporate
Directive
Applicants and licensees shall revise their procedures to assure that
duties, responsibilities, and authority of the shift supervisor and
control room operators are properly defined,
o STP SER (NUREG-0781) concludes in Section 13.5.1.1 that the
applicant's plans for organization and conduct of the operating shift
crews meet the requirements of 10 CFR 50.54 and NUREG-0694
Item I.C.3. It is the position of the staff, however, that a letter,
signed by the Vice President - Nuclear Plant Operations, be reissued
on an annual basis to all station personnel directing that each shift
. .
26
supervisor has the responsibility of directing the licen ed
activities of licensed operators on the supervisor's shift, pursuant
to 10 CFR 50.54(1).
o STP Procedure OGP03-ZO-005, " Reactor Operations Division Conduct of
Operations," Revision 0, was reviewed and found to include the
duties, responsibilities, and authorities of all licensed on-shift
personnel,
o A letter from J. H. Goldberg to all Nuclear Group Personnel, dated
February 18,1987, " Command Authority for Direction of Licensed
Operations Activities South Texas Project Electric Generating
Station," was reviewed. This letter was found to contain explicit
and clearly defined direction concerning the command authority and
responsibility with which the cognizant shift supervisor is charged.
TMI Item I.C.3 is considered closed.
(Closed) TMI Action Item I.C.1 Guidance for the Evaluation and Development
of Procedure for Transients and Accidents
The licensee should reanalyze and propose guidelines and revise procedures
for small break (SB) loss of coolant accident (LOCA), inadequate core
cooling, and transients and accidents.
o STP FSAR 13.5.2.1.4 indicates the Westinghouse Owner's Group (W0G)
and ERG would be the basis for STP E0Ps.
o SER 13.5.2.3 and NUREG-0737, Supplement 1, identify the need for and
current staff review of PGP. The staff response will be in a
subsequent supplement to the SER (NUREG-0781).
o SSER 1 and 2 indicate that the staff response to I.C.1 is still
pending.
o I.C.1.1.1, Small-break LOCA. The STP Station Procedure IPOP05-E0-E000,
Revision 1, approved January 2, 1987, by the plant manager is " Reactor
Trip or Safety injection," and is said to include the SB LOCA
'
response. Procedure IPOP05-E0-E000, Revision 1, does indeed include
SB response such as verify AFW, leck PRZR PORV, check if steam
generator (S/G) tubes are intact, verify RCS subcooling margin, verify
secondary heat sink, and check for reactor coolant pump (RCP) trip.
- There are contingency actions identified of the check results in
negative findings, for example, can't confirm secondary heat sink.
There are cautions to alert the operator to the need for designating
the condition as a site emergency condition when certain criteria are
met; and there are actions and contingency actions to support
maintenance of critical safety functions.
i
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. .
27
o I.C.1.2.B. Inadequate Core Cooling. The STP Station Procedures for
inadequate core cooling are:
o IP0P05-E0-F002, Revision 1, " Core Cooling Safety Function Status
Free," approved February 14, 1987, by the plant manager,
o IPOP05-E0-FRC1, Revision ^1, " Response to Inadequate Core
Cooling," approved January 2, 1987.
l o IPOP05-E0-FRC3, Revision 1, " Response to Saturated Core
l Cooling," approved January 2, 1987.
o 1 POP 05-E0-FRC2, Revision 1, " Core Exit TC's Less Than
755 degrees F."
The Core Cooling Safety Function Status Tree indicates which
procedure to activate depending on core exit thermocouples (TCs) less
than 1200 F, RCS subcooling, RVWL upper plenum, and core exit TCs
less than 755 F. The " Response to Inadequate Core Cooling" procedure
meets the requirements of GL 82-33, Supplement 1 to NUREG-0737,
" Requirements for Emergency Response Capability.
o Per GL 82-33, STP has developed " Plant Procedures Writer's Guides,"
for E0Ps as follows:
o OPGP03-ZA-0027, Revision 1, " Emergency Operating Procedures
Preparation, Approval, and Implementation," approved May 15,
1986, by the plant manager.
o OPOP01-ZA-0006, Revision 2, " Emergency Procedures Writer Guide
and Verification," approved August 14, 1986, by the plant
manager.
o Reviewed ST-HL-AE-1848, " Response to NRC Comments on the Procedures
Generation Package," in which an attachment demonstrated how
OPGP03-ZA-0027, Revision 1, (see 6 above) was used to achieve a
completed procedure for " Steam Generator Tube Rupture,"
OPOP09-E0-E030.
o Per GL 82-33, NRC approval of the PGP is not necessary prior to
upgrading and implementing the E0Ps (7.2.b.).
o HL&P All Sets and Volume Procedure Listing indicates that 48
individual E0Ps have been developed and approved to comply with
I.C.1.3.B. " Transients and Accidents Procedure Revision."
Based on the results of the above review the Item I.C.1 is considered
closed.
l
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28
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(0 pen) TMI Item I.A.1.3 Shift Manning
Applicants shall include in their administrative procedures provisions
governing required shift staffing to assure that qualified personnel are
readily available to man the operational shifts in the event of an
abnormal or emergency situation. These procedures shall also set forth a
policy, the objective of which is to operate the plant with the required ,
staff and develop working schedules such that the use of overtime is j
avoided, to the extent practicable, for those persons who perform '
safety-related functions (e.g. , SR0s, R0s, health physicists (HPs),
auxiliary operators (A0s), and key maintenance personnel),
o STP Procedure OPGP03-ZO-000S, Revision 0, " Reactor Operations
Division Conduct of Operations," was reviewed and found to be in
compliance with the staffing and overtime requirements of NUREG-0737,
Item I.A.1.3. Minimum staffing requirements are delineated for all
modes of plant operation including refueling operations, and
direction is stipulated which effectively limits the amount of
overtime authorized to be worked by any individual. TMI
Item I.A.1.3.2, Implement Minimum Shift Crew Requirements, is
considered closed.
o Interviewed the Manager of Management Services Business Support Group
on March 17, 1987. He indicated that a system for tracking
individual overtime to ensure limits are not exceeded is not yet in
place. A procedure for which he is responsible, OPGP02-ZA-0060,
" Overtime Approval Program," is being typed and should be approved
for issuance in the near future. This procedure, when implemented,
will facilitate tracking of overtime for SR0s, R0s, HPs, A0s, and key
maintenance personnel. TMI Item I.A.1.3.1, Limit Overtime, will
remain open (498/8708-13) pending implementation of a verifiable
system for supervisors and other responsible personnel to track
individual overtime worked, and thereby limit overtime usage to
within specified maximum requirements.
(0 pen) TMI Item I.D.2 Plant Safety Parameter Display Console
Each applicant and licensee shall install a safety parameter display
system (SPDS) that will display to operating personnel a minimum set of
parameters which define the safety status of the plant. This can be
attained through continuous indication of direct and derived variables as
necessary to assess plant safety status,
o Reviewed Letter ST-HL-HL-36428, dated February 12, 1987, concerning
an audit by NRC of the STP SPDS. This audit lists five deficiencies
(significant findings) in the SPDS. They are:
o A validation of the capability of the SPOS to rapidly assess the
safety status of the plant should be performed, preferably
before fuel load.
O e
29
o Parameters to determine the status of the (NUREG-0737)
Radiological Control Critical Safety Function (CSF) should be
included in the SPDS.
o Administrative controls sitould be placed on modification to the
SPDS software to insure that the system's capability to provide
a rapid and reliable assessment of plant safety status is not
jeopardized.
o The SPDS should provide continuous display of the status of the
CSFs.
o No formal review of system requirements versus NUREG-0737
requirements was performed.
o Reviewed Letter ST-HL-AE-1934, which is HL&P's response to NRC
concerning noted deficiencies in the SPDS. All of the responses to
audit concerns will require further review by HL&P.
Based upon the two letters above, concerns raised from an audit of SPDS
have not yet been resolved. This item will remain open (498/8708-14)
until the SPDS audit concerns are acceptably resolved.
(Closed) TMI Item II.K.3.12 Confirm Existence of Anticipatory Reactor Trip
Upon Turbine Trip
Licensees and applicants must confirm that their plants have an
anticipatory reactor trip upon a turbine trip. The licensee for any plant
where this trip is not present should provide a conceptional design and
evaluation for the installation of this trip.
o STP SER (NUREG-0781) concludes in Section 7.2.2.4 that the STP
design, which includes an anticipatory reactor trip on a turbine trip
above 50 percent of rated thermal power (P-9 interlock) is in
compliance with the TMI Action Plan Guidelines.
o Drawing No. 5Z-10-9Z-4211 " Reactor Trip Signals Logic Diagram" was
reviewed and found to include logic which feeds the automatic reactor
trip circuitry whenever turbine EHC fluid pressure drops below a
predetermined setpoint (2/3 coincidence) or all four turbine throttle
stop valves are closed (2/4 coincidence), and reactor power is above
a predetermined setpoint (P-9) as sensed by Nuclear Instrument
Channels N41, N42, N43, and N44 (2/4 coincidence).
TMI Item II.K.3.12 is considered closed.
(0 pen) TMI Item I.B.1.2 Evaluation of Organization and Management
Applicants shall establish an on-site ISEG to perform independent reviews
of plant operations.
._ _ _ _ _ _ . .- _ . _ _ .. _-
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o STP SER (NUREG-0781) concludes in Section 13.4.4 that the HL&P
organizations which perform review and audit functions for STP
(including ISEG) are in conformance with applicable guidelines and
standards (including NUREG-0737).
o STP TS (proof and review copy) Section 6.2.3 regarding ISEG was
reviewed and found to be in conformance with the requirements of
NUREG-0737, Item I.B.1.2.
l
o STP ISEG Operating Philosophy Document (OPD), dated February 9,1987, f
was reviewed. This document, prepared by the Director - ISEG, and
'
approved by the Chairman - Nuclear Safety Review Board (NSRB),
4
delineates how the ISEG plans to carry out its functions. It does
not, however, present an item-by-item discussion of how the group
satisfies every stated requirement. The following ISEG OPD
inclusions, relevant to NUREG-0737, Item I.B.I.2, were found to be
in conformance with the requirements of Item I.B.1.2:
o Staffing and Reporting Relationship - The ISEG will have a
multi-disciplimary, five-member technical staff located at the
STP site. The ISEG Director reports to the chairman of the
NSRB, located in Houston, Texas. This reporting relationship
- provides the ISEG with a high-level forum for the review of its i
'
determinations.
o ISEG Functions - The ISEG will review the operations at STP with
- particular emphasis on assessing the activities which impact
i nuclear safety. These assessments will carry recommendations
for NSRB consideration which will focus on the root causes of
'
events, problems, or undesirable trends. The ISEG will maintain
appropriate relationships with other similar operating plants
and will participate in industry-sponsored groups which bring
together utility personnel performing functions similar to the
STP ISEG.
o On March 16, 1987, the Director - ISEG, was interviewed. He provided
information indicating that the specific procedures relative to
actual performance of ISEG duties have yet to be written, and that
four full-time engineers have yet to be integrated into the ISEG to
satisfy the minimum staffing requirement of NUREG-0737 Item I.B.1.2.
At that time, the ISEG will be considered fully operational.
TMI Item I.B.1.2 remains open (498/8708-15) pending completion of the
following two items which will make the ISEG operational:
o Attaining a minimum staff of five dedicated, full-time engineers
(including the Director - ISEG), located onsite at STP.
l o The issuance of approved procedures which specifically address the
- responsibilities and duties of the ISEG.
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(Closed) TMI Item II.B.1 Reactor Coolant System Vents
Each applicant shall install RCS and Reactor Vessel Head High point vents
remotely operated from the control room,
o STP FSAR Appendix 7A addresses NUREG-0737, Item II.B.1, in regard to
safety analysis of design for seismic and LOCA events, with
10 CFR 50.44 and 10 CFR 50.46 as the acceptance criteria.
o STP SER Section 5.4.12 concluded that the STP design of the RCS High
Point Vents was satisfactory and meets the requirements of
o STP procedure IPOP05-E0-FRI3, Revision 1 " Response to Voids in
Reactor Vessel," was reviewed to determine the degree of guidance
provided to the operator. This procedure does provide necessary
information to the operator for initiating and terminating vent usage
in emergency conditions.
o STP Procedure 10P02-RC-0003, Revision 1. " Filling and Venting the
Reactor Coolant System," and 10P03-ZG-0001, Revision 2, " Plant
Heatup," provide procedural steps for vent valve lineup and
operations when cold,
o STP TS 3.4.11 provides the operability requirements of the Reactor
Vessel Head Vent System (RVHVS).
o STP Orawing SR149F05001, Revision 6, shows the piping and valve
arrangement of the RVHVS. This arrangement meets the design
requirements as stated in the STP FSAR and was accepted by the staff
o The following electrical one-line diagrams were reviewed;
9-E-DJAA-01, Revision 8; 9-E-DJAC-01, Revision 8; 9-E-DJAE-01,
Revision 10; and 9-E-RC19-01, Revision 4. From these drawings it was
determined that the RVHVS is supplied power from a Class IE power
supply.
o The HL&P Company " Pump and Valve Inservice Test Plan" lists the RVHVS
valves with information provided as to class of valves, category of
system, fati position, normal position, and test requirements.
o A tour of the Unit 1 control room was made to determine if the RVHVS
could be operated from the control panels. Indication and control is
available from the control panels.
Item II.B.1 is considered closed.
_
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32
(0penj TMI Item _II.B.4 Training for Mitigating Core Damage
_
_
Applicants are required to develop a training program to teach the use of
installed equipment and systems to control or mitigate accidents in which
the core is severely damaged. They must then implement the training
program.
o STP FSAR Appendix 7A states that training is provided as described in
FSAR Section 13.2.1.1.
o The staff concludes in the STP SER that the applicant has met
item II.B.4, which includes all subjects in Enclosure 3 of
H. R. Denton's letter of March 28, 1980. Also, the STAS and all
operating personnel including licensed operators, appropriate
managers, instrument and control (I&C) technicians, HP technicians,
and chemistry technicians shall receive training commensurate with
their responsibilities,
o A review of two lesson plans (CLT 006.03, Revision 0, Small Break
LOCA and CLT 006.02, Revision 0, Post Accident Cooling) used to teach
Mitigating Core Damage to licensed operators (plant manager through
the operations chain) was conducted. The lesson plans met the
requirements of NUREG-0737.
o A telephone interview was conducted with the Staff Training Manager
to determine the training provided to the I&C technicians on
Mitigating Core Damage. He stated that the I&C technicians receive
training from Lesson Plan ICT9.21. This lesson plan adequately
covers the requirements for the !&C technicians.
o An interview was conducted with the Radiation Protection Supervisor
concerning training of HP personnel. He provided a completed, but
unapproved, lesson plan, RPT001.61, " Radiological Aspects of a Core
Damage Accident." The lesson plan coverage is adequate, but the
actual implementation time was not available. In addition, he
provided lesson plan CATTP phase 3, " Radiological Aspects of a Core
Damage Accident," for the chemistry technician, which is implemented
for all chemistry technicians and is adequate in coverage.
Item !!.B.4 is to remain open (498/8708-16) pending verification that the
HP training has been implemented.
f(Closed)T$ystemTMI
soIFio Item !!.K.3.1.8 Testing /_ Installation __of Automatic PORV
The applicant must provide a system which will automatically cause the
power operated relief valve (PORV) block valve to close when the RCS
pressure decays after the PORV has opened,
o FSAR Appendix 7A,Section II.K.3.1, states that automatic PORV block
valve closure is not required in the STP design. The basis for this
.- .. __._ -_-__ - , _ - . __ .. . _ - _ _ _
. .
I 33
i
is found in a study performed by Westinghouse for the WOG in response
i to TMI Item II.K.3.2 - Report On Overall Safety Effect of
Power-Operated Relief Valve Isolation System. The results of this
study, WCAP-9804, concluded that with the incorporation of specific
post-TMI modifications, which have been or will be impleme.ited on
STP, the reduction in PORV of LOCA frequency is such that an
automatic PORV block valve closure system is not required.
o STP SER (NUREG-0781) concludes in Section 15.6.1 that WCAP-9804 is
acceptable on a generic basis, and that the STP design, hardware, and
j
PORV setpoint is acceptable.
'
TMI Item II.K.3.1.B is considered closed.
(Closed) TMI Item II.E.4.1 Dedicated Hydrogen Penetrations
Plants using external hydrogen recombiners or purge systems for
post-accident combustible gas control must have containment penetrations
dedicated to that service only,
o STP design includes two redundant, 100 percent capacity electric
hydrogen recombiners located inside the containment. Therefore, the
requirement for dedicated penetrations for this system does not apply
to STP.
- o Procedure IPOP-CG-001, Electric Hydrogen Recombiners is in place,
which addresses post-LOCA startup, verification of operation, placing i
in standby, and returning the recombiners to pre-LOCA condition. ,
,
o The following procedures, which contain steps directing operation of
I the hydrogen recombiners when containment hydrogen concentrations
j reach specified levels, are in place:
o 1 POP 05-E0-FRI3, Revision 1 - Response to Voids in Reactor Vessel
'
o 1 POP 05-E0-FRC1, Revision 1 - Response to Inadequate Core Cooling
o 1 POP 05-E0-FRZ1, Revision 1 - Response to High Containment
Pressure
o In Procedure 1 POP 05-EO-FRZ1 Steps 7.3 and 10.2 both erroneously
.
'
refer to Procedure 1 POP 02-CM-0001. This procedure does not exist;
the references should be to IPOP02-CG-0001.
1 o In Procedure 1 POP 05-E0-FRC1, Step 8.3 erroneously refers to
Procedure 1 POP 02-CM-0001. Again, this procedure does not exist; the
reference should be to Procedure 1 POP 02-CG-0001.
TMI Item II.E.4.1 is considered closed. Open item (498/8708-17) will
i track the procedural errors noted above, it should be noted that the
j
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34
.
described error exists in all three copies of each procedure; i.e.,
supervisor's copy, primary operator's copy, and secondary operator's copy.
[
(0 pen]TMIItemII.E.4.2ContainmentIsolationDependability
Applicants must comply with the Containment Isolation Dependability by
providing improved diverse isolation, minimum containment pressure
setpoints for isolation, containment purge valve changes, and closure of
purge valves on high radiation,
o STP SER was reviewed with the following acceptable conclusions of the
i
staff: "the applicant has complied with the provisions regarding
i diversity in parameters sensed for initiation of containment
i isolation, identification of essential and nonessential systems,
, automatic isolation of nonessential systems, and closure of
{ containment purge and vent isolation valves on a high radiation
! signal."
. o By a letter dated October 30, 1985, the applicant committed to equip
!' the outboard isolation valves for supplementary purge system supply
and exhaust lines with pneumatic operators. Configuration Control
l Package (CCP) No. IMST0115 was reviewed to determine the status of
- valve installation. The CCP shows a Field Notification of Change
Completion dated March 4, 1987.
o A visual check of the plant showed that the valves stated above are
installed in the system indicated.
,
o STP TS were reviewed to verify the required position of the 48-inch
i
'
containment purge valves during power operation, startup, hot
standby, and hot shutdown. TS 3.6.1.7 requires that these valves be
sealed closed and TS 4.6.1.7 provides the surveillance requirements
that the valve positions be verified periodically.
Procedure IPSP03-ZQ-0002, Revision 0, " Routine Passive Instrument
Surveillances for Modes 1, 2, 3, and 4," covers the requirement of
l TS 4.6.1.7. ,
t
o Containment pressure setpoint that initiates containment isolation is ;
i to be reduced to the minimum value compatible with normal operating l
l conditions. The setpoint value and justification is to be provided
- by the applicant to the NRC staff in conjunction with the development
of the plant TS per STP SER Section 6.2.4 page 6-13. The setpoint
'
value currently included in the proof and review TS is 5.8 psig. The
NRC staff is reviewing this value for acceptability. ,
l o Procedure IPOP02-HC-0001, " Containment HVAC " was reviewed and found
I to be lacking a General Precaution on maintaining containment
'
I
ressure between the TS limits. Procedure IPOP02-HC-0003,
p' Supplementary Containment Purge," was reviewed and found to have
incorrect values for containment pressure in the General Precaution
section.
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The following subitems of II.E.4.2 are closed:
II.E.4.2.1-4
11.3.4.2.6
II.E.4.2.7 -
II.E.4.2.8
The following are to remain open pending resolution as follows:
o II.E.4.2.5B: All drawings identified in CCP 1-M-ST-0115 are issued
as revised drawing which incorporate the modification changes.
o Procedures IPOP02-HC-0001 and IP0P02-HC-0003 are revised to
incorporate statement and correct values, respectively, as above.
These items will be tracked collectively as Open Item 498/8708-18.
,
(0 pen) TMI Item I.G.1.3 Training Requirements During Low Power Testing
Training will be provided during low power test programs to provide " hands
on" experience for plant evaluation and off-normal events for each
operating shift. It is not expected that all tests will be required to be
conducted by each operating shift. Observation by one shift of training
of another shift may be acceptable. The results of this training will be
documented.
This item requires training during low power testing, and the reporting of
l the results. This item will remain open (498/8708-19) since it cannot be
! completed until low power testing is performed.
(0 pen) SALEM ATWS 2.2 Equipment Classifications and Vendor Interfac_e
(Programs For AITSafe_ty-Related Components), Generic Letter (GL) 83-28
l Licensees and applicants shall submit a description of their programs for
, safety-related equipment classification and vendor interface including
I criteria for identifying components, a description of the information
handling system, station personnel use of the handling system, management
controls, and a demonstration of design verification and qualification for
procurement.
o ST-HL-AE-1274, June 28, 1985, " Response to NRC Generic Letter 83-28,
! Required Actions Based on Generic Implications of Salem ATWS Events,"
2.2, commits to the requirements of GL 83-28, 2.2.
o The Project Q-List identifies systems, structures, and components
that are safety-related. The Project Q-List, 5A479NQ1000, is
consistent with FSAR Table 3.2.A-1, Balance of Plant Quality
Classification of Structures, systems and Components.
o The Project Q-List, Table 1, Item 1.20.0, does not identify any
Q-List items for the Gaseous Waste Processing System (GWPS).
i
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However, FSAR Table 15.7-1, GWPS Failure Analysis (P15.7-6),
identifies that the whole body gamma dose would exceed 0.5 Rem at
both the Exclusion Zone Boundary and the Low Population Zone
Boundary, and therefore, at the site boundary. FSAR Section 3.2 as
,
' well as the Project Q-List identifies that failure of systems, i
'
<
structures, or components that would result in greater than 0.5 Rem
'
a whole body dose at the site boundary should be Safety Class 3 if not
already Safety Class 1 or 2. The Q-List designation of non-nuclear
!
'
safety (NNS) for the GWPS appears incorrectly designated since Safety
, Class 3 would make it a Q-List item. The HL&P Licensing Group stated
'
that the GWPS was designed NNS per RG 1.26, Revision 3, C.2.d, which
t exempts radioactive waste management systems from the 0.5 Rem
criteria. HL&P will pursue a change to the FSAR Section 3.2, and the
Q-List to correct the statement that all systems are inclusive in the
,
0.5 Rem whole body dose criteria.
, o The licensee identified in ST-HL-AE-1274 that it intends to implement
!j the recommendations of the INPO Nuclear Utility Task Action
r
Committee (NUTAC) program for the handling of vendor technical
information, Vendor Equipment Technical Information Program (VETIP).
The NRC inspector interviewed a consultant for HL&P, Operations
l Support Engineering, on the vendor interface. The " Design
Finalization Program Executive Summary" statuses the startup field
validation of safety-related vendor manuals and transmittal of vendor
bulletins and advisories. The consultant stated that replacement
,
vendors and equipment are sought when an existing vendor ceases
l business. Interdepartmental Procedure, Control of Vendor Documents,
STPEGS IP-1.80, Revision 2, approved August 8, 1986, establishes a
'.
single program for the receipt, review, statusing, and distribution
,
of vendor supplied technical information. Interdepartmental
'
Procedure, Nonconformance Control, IP-4.10, approved September 2,
1986, describes resolving nonconformances.
t
o STP Station Procedure OPGP03-ZN-0003, Revision 8, approved March 17,
,
1987, by the Plant Manager, establishes guidance for assigning a
safety classification to maintenance. This procedure is also the
mechanism for corrective action for safety-related equipment that has
defects, deficiencies, deviations, or malfunctions.
o The applicant did not provide any response for the following:
o Plant and corporate managements' oversight activities over
safety-related structures, systems, or components,
o Planned and periodic audits over activities impacting
<
safety-related equipment.
o Verification that vendor-recommended modifications were
implemented on the Reactor Trip System (RTS) breakers.
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o Implementation of a preventive maintenance program for
components where there is inadequate traceability of component
performance to the vendor (vendor can't be identified, went out
of business, or won't supply required information),
i
o Procedures to provide instructions for assigning a safety
l classification to operating and surveillance procedures.
o GL 83-28, Item 2.2, will remain open (498/8708-20) pending resolution
of designation of GWPS components in the Q-List and responses to the
items immediately above.
(0 pen) SALEM ATWS, 3.1, Post-Maintenance Testing (Reactor Trip System
Components)
Applicants shall submit their review of test and maintenance procedures
and TS to assure post maintenance operability testing and submit their
check of vendor and engineering recommendations to ensure that any
appropriate guidance is included in test and maintenance procedures.
Applicant shall identify any TS requirements for post-maintenance testing
that degrade rather than enhance safety.
o STP's response to this position is in ST-HL-AE-1274 and commits to
implementing the requirements.
o EGG-NTA-7159, February 1986, was reviewed and it was found that the
applicant met the requirements of 3.1.3 -- that no TS for
post-maintenance testing degrade safety for the RTS.
o STP Station Procedure OPGP03-ZE-0020, Revision 0, approved January 7,
1987, by the plant manager, describes the post-maintenance testing
program, including initiating requests, criteria and responsibilities
for review and approval, criteria and responsibility for designating
the activity as safety-related, and guidance for determining the
testing to be performed. Section 5.3.1 of OPGP03-ZE-0020,
Determination of Applicable Components, gives guidance on_ identifying
components that should have post-maintenance testing. It was noted
that the master document for identifying safety-related components,
the Project Q-List, is not referenced in Section 5.3.1 of
OPGP03-ZE-0020.
o STP's, " Post-Maintenance Testing Reference Manual," Revision 0,
approved February 28, 1987, describes 17 types of post-maintenance
tests and lists their requirements (e.g., Auto Start Test, Pump
Operability Test, Valve Leakage Test, etc.).
o STP Station Procedure OPGP03-ZM-0003, Revision 8, approved March 17,
1987, by the Plant Manager, establishes criteria and responsibilities
for review and approval of maintenance.Section V of the Maintenance
Work Request Form requires a yes or no indication for i
post-maintenance testing.
,
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GL 83-28, Item 3.1, will remain open (498/8708-21) pending completion of a
change to Procedure OPGP03-ZE-0020 to include reference to the Q-List as
discussed above.
(0 pen) SALEM ATWS GL 83-28/4.5 Reactor Trip System Reliability (Safety
Functional Testing)
On-line functional testing of the reactor trip system, including
independent testing of the diverse trip features, shall be performed on
all plants. (On-line is Modes 1 through 6).
o The NRC inspector reviewed Station Procedure " SSPS Logic Train R
Functional Test," 1 PSP 02-SP-0001R, Revision 0, which was approved
January 25, 1987, by the plant manager.
o The Train S Procedure, IPSP02-SP-00015, Revision 0, was on the plant
procedure computer printout list and was approved.
o FSAR Section 7.2.2.2.3.10.4, Testing of Reactor Trip Breakers, and
Figure 7.1-2 are not consistent with IPSP02-SP-0001R. The bypass
breaker designations in the FSAR, 52/BYR and 52/BYS, have been
interchanged in the procedure, 1 PSP 02-SP-0001R. That is, in the
FSAR, Reactor Trip Breaker 52/RTR is bypassed by Bypass
Breaker 52/BYR and Breaker 52/RTS is bypassed by 52/BYS. But in
IPSP02-SP-0001R Reactor Breaker 52/RTR is bypassed by 52/BYS and
Reactor Breaker 52/RTS is bypassed by 52/BYR. The licensee indicated
that 1PSPS02-SP-0001R is written to agree with wording in the
,
Westinghouse Maintenance Manual and not wording in the FSAR. The
'
FSAR wording, 1 PSP 02-SP-0001R, and Westinghouse Maintenance Manual
should be made consistent to avoid misinterpretation by the operating
staff. On page 39 of 48 of IPSP02-SP-0001R, Step 7.9.12.b should
read " Train R Reactor Trip Breaker-0 pen / Tripped position" not
" Train S Reactor Trip Breaker-0 pen / Tripped position." This requires
a change to Procedure 1 PSP 02-SP-0001R and appropriate review and
approval per TS 6.5.1.6a before this item can be closed out.
o Preoperational Test Procedure 1-RS-P-03 " Reactor Trip Switchgear" was
reviewed. Section 7.6 of this procedure, performed December 12,
1986, satisfactorily demonstrated that opening a Reactor Trip Breaker
with its respective Bypass Breaker closed does not cause a loss of
voltage to the Rod Control Power Cabinets.
GL 83-28. Item 4.5, will remain open (498/8708-22) pending resolution of
inconsistent wording between FSAR and IPSP02-SP-0001R and completion of
IPS02-SP-001R procedure change.
(Closed) SALEM ATWS Item GL 83-28/2.1 Equipment Classification and Vendor
Interface (Reactor Trip System Components)
,
Licensees and applicants shall confirm that all components whose
j functioning is required to trip the reactor are identified as
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39
safety-related on documents, procedures, and information handling systems
used in the plant to control safety-related activities, including
maintenance, work orders, and parts replacement. In addition, for these
components, licensees and applicants shall establish, implement, and
maintain a continuing program to ensure that vendor information is
complete, current, and controlled throughout the life of the plant, and
appropriately referenced or incorporated in plant instructions and
procedures. Vendors of these components should be contacted and an
interface established. Where vendors can not be identified, have gone out
of business, or will not supply the information, the licensee or applicant
shall assure that sufficient attention is paid to equipment maintenance,
replacement, and repair, to compensate for the lack of vendor backup and
to assure RTS reliability. The vendor interface program shall include
periodic communication with vendors to assure that all applicable
information has been received. The program should use a system of
positive feedback with vendors for mailings containing technical
information. This could be accomplished by licensee acknowledgement for
receipt of technical mailings. The program shall also define the
interface and division of responsibilities among the licensees and tFe
nuclear and non-nuclear divisions of their vendors that provide service on
RTS components to assure that requisite control of and applicable
instructions for maintenance work are provided.
o The NRC inspectors examined Letter ST-HL-AE-1274, which states that
HL&P has reviewed all components whose functioning is required to
trip the reactor and these components have been properly classified
in the design documents.
o A technical evaluation report (Supplemental SER (SSER) 1, Appendix N)
states that applicable RTS components were verified to be properly
classified. SSER 1 states that the program meets the requirements of
GL 83-28/2.1 (Part 1) and is acceptable.
o A technical evaluation report (SSER 2, Appendix T) states that STP is
a participant in the Westinghouse interface program for the RTS.
SSER 2 states that the program meets the requirements of GL 83-28/2.1
(Part 2) and is acceptable.
Based upon the above information, this item is considered closed.
(Closed) Salem ATWS Item GL 83-28/1.2 Post-Trip Review - Data and
Information Capability
Licensees and applicants, shall have or have planned a capability to
record, recall, and display data and information to permit diagnosing the
causes of unscheduled reactor shutdowns prior to restart and for
ascertaining the proper functioning of safety-related equipment.
_ _ _ _ _ _ _ _ _
. . ... -
.
. .
40
o The NRC inspectors reviewed Letter ST-HL-AE-1274 dated June 28, 1985.
This letter contains a description of the capabilities of the Proteus
Computer System and The Emergency Response Facilities Data
Acquisition and Display System (ERFDADS).
o. STP SSER 1 includes a conclusion that the applicant's post-trip _ 1
review data and information capabilities are acceptable.
,
o The CRTs and typers associated with Proteus Computer System and
ERFDADS were inspected in the Unit 1 control room. The status of the
systems was discussed with a control room operator. Although testing
is continuing on the systems, the CRTs and typers are operating and
it appears the systems will function as described in
Letter ST-HL-AE-1274.
Documentation and inspection indicate the facility has planned the
required capability for post-trip data and information. This item is
considered closed.
{ Closed) Salem ATWS GL-83-28 Item 4.3 Reactor Trip System Reliability
(Automatic Actuation of Shunt Trip Attachment for Westinghouse and B&W
Plants)
Westinghouse and B&W reactors shall be modified by providing automatic RTS
actuation of the breaker shunt trip attachments. The shunt trip
attachments shall be considered safety-related (Class 1E).
o Procedure IPSP02-SP-00012, Revision 0, was reviewed. STP SER Item 5
of Page 15-22, requires that quality assurance criteria set forth in
Appendix B to 10 CFR 50 be met. During a meeting with the QC General
Supervisor and the Audit / Surveillance Supervisor the QA program was
described concerning operational testing of equipment. The program
meets the requirements as indicated in GL 83-28.
,
'
o Procedure IPSP03-RS-0002, Revision 0, " Manual Reactor Trip TAD 0T,"
was completed. It was determined that this procedure meets the
operability test requirements to verify contacts and wiring of the
manual trip circuit before startup after each refueling outage.
Item 4.3 is considered closed.
'
(0 pen) Salem ATWS Item GL 83-28/4.1 Reactor Trip System Reliability
(Vendor-Related Modifications)
All vendor-recommended reactor trip breaker modifications shall be
reviewed to verify that either: (1) each modification has, in fact, been
implemented, or (2) a written evaluation of the technical reasons for not
,
implementing a modification exists.
<
!
o STP has Westinghouse type DS-416 reactor trip breakers installed.
STP SSER 1 (NUREG-0781 Supplement No. 1) states in Section 15.8.2
, .- __ ._. _ . . _ _ _ . __ __ _ _ _ - . _ _ _ , _ , _ , _. - _.
. . _
-. .
41
under Action Item 4.1 that the applicant has committed to implement
all vendor-related modifications before. fuel loading, and that the
applicant's position on Item 4.1 is acceptable.
o Letter ST-HL-AE-1274 was reviewed. It indicated on Page 14 of 18 in
i
Section 4.1 that "HL&P has been informed by Westinghouse that a
~
design discrepancy had been identified in the undervoltage attachment
-
and that Westinghouse intended to replace the undervoltage
attachments on DS-416 reactor trip switchgear. Field change
notices (FCNs) have been issued by Westinghouse for installation and
adjustment of the replacements."
o STP FCN TGXM-10563, Shop Order No. 386, indicates that the
undervoltage trip assemblies in the raattor trip switchgear have been
replaced. This FCN was closed on January 28, 1986.
o Westinghouse Nuclear Service Division Technical Bulletin 83-03 was
reviewed and found to provide the Westinghouse recommendation for a
single method of independently verifying the function of the shunt
^
trip and the undervoltage trip mechanisms of the Reactor Trip
Breakers. It was not intended that a utility would follow this
general procedure verbatim, but would first parform a thorough review
of the general procedure against the plant specific system.
. o STP Procedure ISP02-SP-0001R, Revision 0, " Solid State Protection
System (SSPS) Logic Train R Functional Test," date January 29, 1987,
was reviewed and found to include appropriate procedure steps fo"
testing the Undervoltage Trip Attachment and the Auto Shunt Trip
Function independently of one another.
o Westinghouse Nuclear Service Division Technical Bulletin 84-02 was
reviewed and found to advise all affected plants of a condition which
may exist on DS or DSL Class 1E circuit breakers used for Reactor
Trip or Safeguards Power Breakers. This condition, which should be
. investigated, involves rotential wire damage on the left side,
.
particularly ,in the vicinity of the wire retainer which forms the
extreme lef t coundary of the breaker, f acing the breaker front.
Instruction for dealing with this conditicn (replacement of damaged
wires and providing additional rigidity and mechanical support) are
i' also included in this fechnical Bulletin. STP Procedure OPMP05-NA-0008,
Revision 1, Westinghouse 480 Volt Breaker Test, dated February 24,
,- 1987, was reviewed and found to include a procedural step to insoect
breaker wiring, witn a precautionary note dealing with the left side,
facing the front of-the breaker. Westinghouse Nuclear Service
Division Technical Bulletin No. NSD-TB-84-02, Revision 1, "DS/DSL
Breakers - Potential Wire Damage For Reactor Trip or Safeguards Power
Breaker," is referenced in this procedure. An attempt was made to
ascertain that inspection of Reactor Trip Breakers had actually been
performed and the inspection results documented, but this information
was not made available.
._.
_, __, ,_. -
.__ ___ _ _ _ _ _ _ _ _ - ____ _ . _ . ~.. _. -
.- - - .-
-
.i
42
Salem ATWS Item GL 83-28/4.1 will remain open (498/8708-23) pending STP
submittal of verifiable evidence that inspection of the Reactor Trip
Breakers for potential wire damage has been completed and any deficiencies
noted have been corrected.
13. Licensee Action on Previous Inspection Findings
(Closed) 498/8620-01 - This item concerned the use of ambiguous terms such
as " extent necessary" for criteria in certain preoperational procedures
without offering guidance as to what constitutes " extent necessary." The
specific procedures have been changed to delete such phases. The
procedure changes were reviewed by the NRC inspectors and deemed
sati sfactory. This item is considered closed.
(Closed) 498/8620-02 - This item concerned using acceptance criteria in
, procedures without guidance as to what is acceptable. The specific case
involved fuel handling trolleys and hoists which were required to be
operated without excessive vibration. The criteria in this instance was
found to be superfluous as such abnormalities are routinely watched for
and covered by SAI 18, paragraph 5.3.4. The procedure was changed and
deemed satisfactory by the NRC inspectors.
14. Site Tours
The NRC inspectors conducted site tours independently. These tours were
made to assess the protection on inplace safety-related equipment, plant
status, and to observe testing. The areas toured included: Unit 1 -
Mechanical and Electrical Auxiliary Building (MEAB), Reactor Containment
Building (RCB), Fuel Handling Building (FHB), and the Emergency Diesel
Generator Building. With the exception of the violation noted in
paragraph 11, the NRC inspectors noted a marked improvement in the
maintenance of areas and equipment turned over to the Nuclear Plant
Operations Department (NPOD).
r
! No additional violations or deviations were identified.
15. Review of The Manual Trip Circuit
,
Because of a drawing error relative to the electrical location of the
manual trip circuit in relation to the output transistors Q3 and Q4 in a
'
l similar Westinghouse designed plant with a SSPS, the NRC inspector inspected
,
facility Drawings 14926-0387(2)00171-BWN and 14926-0387(2)00172-BWN.
'
These drawings correctly depict the manual trip circuit downstream of the
output transistors Q3 and Q4. .
16. Procedures Review
The NRC procedure review team by review of selected procedures, personnel
interviews, and system walkdowns, assessed the applicant's procedures for
adequate administrative controls, technical accuracy, and compliance with
, , ,--r--. , - . , , , , , , , -,- - - . , --,n- , - , - - - . ,,, , , , - , --n- - ,--,n.
. . . ___ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ __
4
o. .
-
43
regulatory requirements. A selected number of the procedures reviewed
were field verified by walking down the applicable system and verifying
that the procedure reflected as-built conditions.
Procedures in~ the following areas were selected for review:
i
'
a. Plant
b. Operating
c. Maintenance
d. Emergency Operating
e. Off-Normal Operating and Alarm Response
f. Surveillance
Following is a discussion of the NRC inspection effort in each of'these
- areas including a list of the specific procedures reviewed and NRC
i inspector observations and findings. (The above letters identify the
discussion for the corresponding procedure category below.):
'~
a. Plant Procedures
(1) Review Of Program For 10 CFR Part 50.59 Safety Evaluations
- The NRC inspector reviewed administrative procedures to
ascertain whether responsibilities have been assigned to
i appropriate personnel to ensure that plant procedures will be
reviewed, updated, and approved as required and to ensure that
- the revision process includes provisions for 10 CFR 50.59
considerations.
Procedures reviewed were as follows:
o OPGP03-ZA-0002, Revision 8, " Plant Procedures"
,
l o OPGP03-ZA-0003, Revision 4, " License Compliance Review"
I .o OPGP03-ZA-0010, Revision 2, " Plant Procedure Compliance,
Implementation and Review"
o OPGP03-ZA-0033, Revision 0, "10 CFR 50.59 Safety
- Evaluations"
'~
These procedures define the procedure revision process for all
plant procedures with the exception of plant organization, plant
- policies, and emergency operating procedures. They define the
'
responsibilities of personnel to ensure that each procedure is
reviewed within a 24-month interval and is revised, if required,
4 and that revisions are approved by the plant manager.
NRC inspector observations / concerns are discussed below:
l
l
I
-
_ . _ _ . . _ , _ _ .
.- .
.- ,
44
'
(a) OPGP02-ZA-0010, Revision 2, " Plant Procedure Compliance,
Implementation and Review." This procedure requires that
plant procedures be reviewed within 24-month intervals.
This review of plant procedures determines whether a
'
revision needs to be generated. However, once a procedure
revision has been determined to be necessary, there are no
guidelines to ensure that the procedure will be revised in
a timely manner. These guidelines should be written to
ensure that a procedure that is determined, by the review,
to need revising is not used while it is being revised.
This is an open item (498/8708-24).
F
(b) OPGP03-ZA-0033, Revision 0, "10 CFR 50.59 Safety
Evaluations." This procedure is included in the procedure
review process to ensure that procedure reviews include
10 CFR 50.59 considerations. The inspector noted that the
technical support superintendent is responsible for the
review of 10 CFR 50.59 Safety Evaluations. However, there
is no technical support superintendent within the
operations department. Pending correction, this is an onen
item (498/8708-25).
(c) Procedures OPMP08-ZI-0065, Revision 0, " Field Testing .of
Power Supplies and Over Voltage Protectors," and
OPMP07-SP-0001, " Revision 0, "SSPS Decoder Printed Circuit
Board Test and Rework." The NRC inspector determined that
these procedures had been reviewed on July 29, 1986, but
they had not been revised at the time of the NRC inspection
8 months later, even though the applicable procedure
biennial review form, OPGP03-ZA-0012-2, stated that these
procedures required revision to correct inadequacies. This
L is contrary to Step 3.3.2.3 of Procedure OPGP03-ZA-0002,
Revision 3, " Plant Procedure Compliance, Implementation,
and Review" which states, "The cognizant DM shall ensure
that identified procedural inadequacies are corrected in
accordance with OPGP03-ZA-0002 (Plant Procedures)." This
failure will be tracked as an open item pending resolution
by the licensee (498/8708-26).
(2) Review Of Standing Order and Short Term Instruction Programs
The NRC inspector verified that administrative controls were
established for STP standing orders and short term
instructions (STIs).
The administrative controls for standing orders were defined in
PRO-1, Revision 2, " Standing Orders," which provided a mechanism
for their issue and distribution. The standing orders were
required to be reviewed and updated quarterly; however, there
was no mechanism to document the review. Pending the applicants
establishing a method to document the quarterly review of
_ _ _ . _ _ _ _ . - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . . _. ,__.
_ _ _ . .- __
'
, .
45
standing orders, this is an open item (498/8708-27).
,
Responsibility was assigned to the Unit 1 operations supervisor
to issue, distribute, review, and update standing orders. PRO-1
established a limitation on the type of instructions that could
be issued as a standing order.
The administrative controls for STIs were defined in PRO-25,
Revision 0, "Short Term Instructions," dated March 31, 1987.
The standing order provided a mechanism for the issue,
distribution, review, and updating of STIs. The shift
,
supervisor was assigned responsibility for control of STIs
including reviewing and maintaining them. The standing order
provided limitations on the type of activities that could be
accomplished by STIs. Issuance and cancellation of STIs may be
accomplished by signature of the shift supervisor, operations
supervisor, or the reactor operations manager. The NRC
inspector reviewed selected standing orders and STIs to verify
they met the guidance specified and did not perform activities
that should be covered by procedures. The applicant's Standing
Order PRO-1 states that standing orders shall not be used in
,
lieu of procedures. The below listed standing orders were
reviewed:
Number Revision Subject
'
PRO-1 2 Standing Orders
PRO-2 1 Required Reading Program
PRO-18 0 Event Reports
PRO-23 0 Work Control Guidelines
PRO-25 0 Short Term Instructions
From discussions with a shift supervisor and after comparison of
the guidance contained in Procedures OPGP03-ZA-0002, Revision 8,
" Plant Procedures" and OPGP03-ZA-0003, Revision 4, " License
Compliance Review," the NRC inspector determined that the shift
supervisor was knowledgeable of the steps necessary to make
temporary changes to procedures.
!
(3) Control Room Logbooks
, The NRC inspector verified that OPGP03-ZQ-0001, Revision 0,
i " Maintenance of Reactor Operations Logbooks," provided guidance
for preparation and review of logbooks. The procedure described
the usage, control, and type of logbooks maintained in the
control room. The requirements for retention in the quality
assurance vault and maintenance and storage in the control room
were specified.
NRC inspector observations made during the review of
OPGP03-ZQ-0001 are discussed below:
l
l
l
'
_ _ _ _ _
_ _ . _ . . --
._ __ _ _ _ .. _ _ _ . . _ _ _ _ __ _ _
. ._ - . - - - . .
l
.. ,
46
o There was a typographical error (typo) in Step 6.5.6 in
that the statement "only to reference the procedure" was
not needed. The step should match Step 6.2.7.
o The steps in Section 6.5 were apparently misnumbered since
Step 6.5.5 was missing.
Plant Procedure OPGP03-ZA-0063, Revision 0, " Reactor Operations
Shift Turnover," described how the shift relief and turnover was
'
to be conducted and what documentation was to be generated and
-
retained. This procedure detailed the requirement for a Shift
Relief and Turnover Log. Observation on this procedure is:
Typo in Step 4.1.3.6.3, " Shift turnover of prior to end of
shift," should not include the first "of!"
b. Operating Procedures
This area of inspection was conducted in order to confirm that plant
operating procedures are prepared and approved to control
safety-related operational activities. An index of all current
procedures dated March 10, 1987, was reviewed for plant operating
procedures identified in Regulatory Guide 1.33, Revision 2, in the
following categories:
. o Administrative Procedures
o General Plant Operating Procedures
, o Procedures for Startup, Operation, and Shutdown of
Safety-Related Systems
Also, a sample: review of plant operating procedures in the above
categories was conducted to verify that the appropriate format was
used and that each procedure was technically adequate-to accomplish
the stated purpose.
The results of the reviews in each category are documented below:
(1) Administrative Procedures
- (a) The following procedures were not issued as of March'13,
1987:
o OPGP03-ZA-0064, "NPOD Preshift Briefing"
o OPGP03-ZQ-0001, " Maintenance of Reactor Operations
!
Logbooks"
o OPGP03-CN-0001, " Radio Communication"
l
!
!
l
!
. _ _ - . - . - . ,_- - , _ , .. _ ,, . _ , , _ . _ . - . _ . , _ - . _ _ _ _ , , . _ _ . . - , , ,-- _
. . _ ~
-_______________ _
. .
47
o 0PGP03-CN-0002, " Telephone Communication"
o OPGP03-CN-0003, " Plant Public Address and Alarm
System"
o OPGP03-CN-0004, " Pocket Pager System"
o OPGP03-CN-0005, " Communications Console System"
o OPGP03-CN-0006, " Communications System Test Program"
Pending issuance of the above procedures, this is an open
item (498/8708-28).
'
(b) Procedures Reviewed
Procedure No. Revision Title
OPGP03-ZA-0002 8 " Plant Procedures"
0PGP03-ZA-0010 2 " Plant Procedure Compliance.
,
Implementation and Review"
l
OPGP03-ZO-0001 3 " Equipment Clearance"
i
OPGP03-ZO-0003 3 " Temporary Modifications and
Alterations
OPGP03-ZA-0055 0 " Plant Surveillance
Scheduling"
0PGP03-ZE-0004 3 " Plant Surveillance Program"
OPGP03-Z0-0004 0 "Flant Conduct of
Operations"
0PGP03-ZA-0033 0 "10 CFR 50.59 Safety
Evaluations"
(c) NRC Inspector Observations / Concerns
1) OPGP03-ZA-0002, Revision 8, " Plant Procedures"
o There is no limit on the number of " field change
requests" (FCRs) that can be issued / approved for
a procedure before the procedure must be revised.
o There is no requirement to incorporate permanent
FCRs into procedure' revisions.
.- ,
48
o The procedure does not describe a program for
controlling the expiration dates on FCRs (i.e.,
is the FCR deleted or must it be evaluated for
procedure incorporation).
o There are no administrative controls for the use
of FCRs (i.e., mark up controlled procedure to
reference FCRs, page for page replacement, sign
off on FCR or controlled procedure, etc.).
o The approved FCRs are not listed in the master
procedure listing /index.
o There is no required time limit for FCRs to
receive final approval. The administrative limit
of 14 days (should procedure) is not being
followed as evidenced by FCN 87-21 (initial
approval 1/21/87), FCN 87-055 (initial
approval 2/13/87), and FCN 87-064 (initial
approval 2/19/87) not receiving final approval as
of March 13, 1987.
Pending resolution of the above concerns related to
the control of temporary procedure changes, this is an
open item (498/8708-29).
2) OPGP03-Z0-0003, Revision 3, " Temporary Modifications
and Alterations"
,
o The procedure makes statements in Steps 1.2 and
1.3 which appear to allow other programs to be
used for temporary modifications. This appears
to allow circumventing of the required controls
o The procedure steps cannot be followed in
sequence as required by administrative procedure.
o There is no requirement for the shift supervisor
to acknowledge that the temporary modification
has been installed.
o The procedure allows the use of one tag on the
- outside of a panel door for several temporary
jumpers or lifted leads.
l-
- o The inclusion of the startup temporary alteration
! program in the plant operations program procedure
causes unnecessary complications of this
procedure. A simple reference to the startup
procedure and requirement for shift supervisor
l
!
t
< ..- , , , - - . ~ , - -
,.,-....---mm- . . , . - - - - . . - - - - - - , - . ~ . - . . , _
-
- _ _ . _. - _ - _ _ _ -_ _ _ _ _ .- _
o .
49
I
approval of temporary alterations and for all
temporary alterations to be converted to
temporary modifications would be sufficient. All
required temporary alterations should be
converted to temporary modifications prior to
operating license issuance. See
paragraph ~16.b.(2)(b)2) for further concerns
regarding modifications and alterations.
o The procedure allows the use of plastic screws
,
and washers for lifted leads with no inventory
requirements to assure that they are removed.
Pending resolution of the above concerns, this is an
open item (498/8708-30).
3) OPGP03-Z0-0004, Revision 0, " Plant Conduct of
Operations"
o Step 4.2.3 of this procedure indicated that only
4 an RO is required in the control room during
operating Modes 1 through 4 while TS
Section 6.2.2.b requires an RO in the control
room during all modes (with fuel in reactor) and
an R0 and SRO in the control room during Modes 1
through 4.
o Step 4.10.3 states that entry into limiting
conditions of operations (LCOs) as required by TS
be controlled in accordance with approved
procedures; however, there is no procedure for
tracking and controlling of LCOs by plant
operations.
'
Pending resolution of the above concerns, this is an
open item (498/8708-31).
4) A general concern identified during the review of
administrative procedures was the apparent overuse of
the word "should" which denotes a recommendation
rather than a strict requirement. The NRC inspector
is concerned that this will lead to procedure abuse
,
and a loss of control. This was discussed with
licensee management and discussed again during the NRC
exit meeting. The plant manager stated that
!
individual managers are expected to enforce the
"should" statements in procedures and that the
licensee QA organization would issue QA findings in
areas where lack of control is evidenced. The NRC
inspector cited the specific example of the final
, approval of FCNs to procedures exceeding the 14-day
, _. - _ . - - - .. - _ - _ . - . . -- .
. . .
50
i
time limit as a lack of control. Also, the following
examples of inappropriate "shoulds" in procedures were
noted by the NRC inspector:
t
o Step 4.10.3.1 of OPGP03-ZO-0004 states, "The
Shift Supervisor should be informed of TS
surveillance tests that have failed to meet their
, acceptance criteria . . . ." ,
1
o Step 4.10.1.2 states, "The operator who completed
the checklist should sign and date the checklist
and present the checklist to the supervisor who
directed the checklist be completed. For
components that are safety-related, supervisory
- personnel should direct that an independent
verification be performed. When the independent
verification is completed, the operator who
performed the independent verification should
sign and date the checklist."
,
o Step 4.10.1.3 states, "The completed checklist
should be reviewed by the shift supervisor, unit
supervisor, or chemical operations foreman, as
appropriate, to verify that the directed actions
were completed and to note any exceptions or
unusual conditions. The completed checklist
should be inserted in the control room system *
4 status file or the watchstation system status
'
file, as appropriate. The superseded checklist
>
should be forwarded for record retention or
i discarded as appropriate."
i
Pending resolution of this concern related to the use
~
of "shall" and "should", this is an open item
,
(498/8708-32).
I No additional concerns or comments were identified during
the review of administrative procedures.
(2) General Plant Operating Procedures
(a) Procedures Reviewed
i
Procedure No. Revision Title
,
'
OPGP03Z0-0022 0 " Post Trip Review"
1 POP 03-ZG-0001 2 " Plant Heatup"
1 POP 03-ZG-0004 1 " Reactor Startup"
,
4
5
F
, , . , - - - - ~ - ,- - - - r--r.--w, w # w , w way -, - - - -,_-..--.mi,, - , ,
-
..---,r--.r -- - . - , . --
__ _ _ _ . . _- _ _ . - _ _ _ _ _ . - - _ _ _ _. . ._. __ ___ _
m
i
i
! . .
51
4
j IP0P03-ZG-0005 0 " Power Operations"
IP0P03-ZG-0006 0 " Plant Shutdown To
,
Hot-Standby"
l (b) NRC Inspector Observations / Concerns
1) Generally, there were several inconsistencies noted
with signoffs of steps and sections (i.e., _some action
- steps not initialed, some prerequisites not signed
_
off, and the location of signoff blocks varied which
i
creates confusion as to what is being signed off).
- Pending resolution, this is an open item
(498/8708-33).
2) During the review of Procedure 1 POP 02-EW-0001,
Revision 2, " Essential Cooling Water Operation," it
was noted that FC 87-092 to the procedure had been
. issued because of " leads landed on plastic during
S/U." The NRC inspector asked for the temporary
'
.
modification documentation to support these lifted
leads. Temporary Modification Nos. T1-EW-8710, -8711,
and -8712 were provided to the NRC inspector. A
review of these modifications revealed that the
j emergency load sequencing capability of the essential
cooling water system components had been disabled by
the installed modifications. The " License Compliance
Review Forms" completed for the modifications
incorrectly stated these modifications did not involve
j
'
a change to the facility as described in_the FSAR and
do not require a change to the TS when, in fact, the
installed modifications do change the facility as
! described in the FSAR and would require a change to
the TS. These installed modifications would also
require a 10 CFR 50.59 Safety Evaluation to be
performed which was not accomplished. This failure to
- identify a facility modification which involved an
, unreviewed safety question was discussed with licensee
,
personnel and they stated that all temporary
i modifications existing at fuel load would have been
reevaluated per 10 CFR Part 50.59 for unreviewed
- safety question determinations. This will be
- evaluated by the NRC during closure of the Open
Item 498/8708-30 on conversion of temporary
,
alterations to temporary modifications. Pending the
4 licensee's response which should address the steps
i taken to assure that all previous License Compliance
j Reviews have been performed properly or that controls
- are in place to assure that needed
! reevaluations / reviews are performed, this is an open
item (498/8708-34).
$-
, - . . .- .. -. - - . - _ - . _ - - . . - . _ . - - . - - _ _ - - _ .
.m
-
. ..
52
3) IPOP03-ZG-0001, Revision 2, " Plant Heatup"
The prestartup checklist does not include related
activities such as emergency shutdown system readiness
or the performance of Procedure 1 POP 02-SI-0001,
" Safety Injection Accumulators," for filling and
venting the safety injection accumulators. This is an
open item (498/8708-35).
4) IPOP03-ZG-0004, Revision 1, " Reactor Startup"
There is no requirement to verify that the neutron
count rate on the source range instruments is above a
set minimum. This is an open item (498/8708-36).
5) 1 POP 03-ZG-0008, Revision 0, " Power Operations"
The purpose and scope section is incomplete. There.is
nothing listed after . . . "following:". This is an
open item (498/8708-37).
No other concerns were identified during the review of
general plant operating procedures.
(3) Procedures For Startup, Operation, and Shutdown Of Safety-Related
Systems
(a) The following procedures were not issued as of March 13,
1987:
o IP0P02-CG-0001, " Electric Hydrogen Recombiners"
o 1 POP 02-CZ-0001, " Electric Hydrogen Recombiners"
o IPOP02-DB-0005, " Technical Support Center Diesel
Generator"
o 1 POP 02-II-0001, "Moveaule Incore Detector System
Operation"
o 1 POP 02-SB-0001, " Steam Generator Blowdown System"
Pending issuance of these procedures, this is an open item
(498/8708-38).
(b) Procedures Reviewed
Procedure No. Revision Title
1 POP 02-SI-0002 2 " Safety Injection System
Normal Lineup"
.. .. _ _ _ .- .
_ _ _ . . _ - . . _
. .
',
53
'
.1 POP 02-AF-0001 1 " Auxiliary Feedwater"
1 POP 02-CS-0001 1 " Containment Spray
Standby Line-up"
- IPOP02-CV-0001 1 " Makeup To the Reactor
Coolant System"
1 POP 02-DG-0001 1 " Emergency Diesel -
' Generator No. 11"
1 POP 02-EW-0001 2 " Essential Cooling Water
j
Operation"
i
IPOP02-RH-0001 1 " Residual Heat Removal
System Operation"
1 POP 02-SI-0001 1 " Safety Injection
- -
. (c) Generic NRC Inspector Observations / Concerns
l o There are no procedural instructions provided for
j filling and venting certain systems (i.e., auxiliary
, feedwater and safety injection systems). This is an
l open item (498/8708-39).
l- o There are no procedural requirements provided for
disposition of exceptions identified during system
'
lineups. There is a page for listing exceptions but
no requirement to evaluate each identified exception.
.
This is an open item (498/8708-40).
!
~
! o There were inconsistencies between the procedures
l- valve lineups for identifying that pipe caps were
l installed on vents, drains, and test connections. At
! least one procedure identified pipe caps while others
did not. This is an open item (498/8708-41).
l 0 It is recommended that operating procedures for TS 1
i
'
systems reference the applicable TS sections. It was
noted that some procedures did reference TS sections
I,
'
and others did not. This is an open item
(498/8708-42).
! o There were inconsistencies with required signoffs on
! different procedures (i.e., IPOP02-CS-0001 did not
require signoffs for control board lineups while other
procedures did). This is an open item (498/8708-43).
l
i
i
, ,
,, e. , - ~ - ---- --w . , _ . .n, ,~,,,,. - _ . - , , - . . , . . . ,,,_._y. ,m .-,,.mn -, m ,w _ ww,,. ,-,._.,,-.,..,,---,,--,n, -,-
.. -- - _ - . .- -.
j. . .-
54
,
c. Maintenance Procedures
The NRC inspectors reviewed selected applicant maintenance and
administrative procedures to verify that the procedures were in the
appropriate format, that they were technically adequate, and that
maintenance activities would be controlled in accordance with
regulatory requirements. The procedures were divided into the areas
of maintenance and measuring and test equipment (M&TE) for this
review. A number of Unit 1/ common maintenance procedures (as
indicated in the table below) had not been issued at the time of the
i NRC procedure review. Prior to licensee issuance, the NRC will
review the number and types of procedures not issued to determine if
this would impact plant operations. The numbers in the table below
were taken from a computer maintenance procedure listing dated
April 7, 1987. Pending a subsequent NRC review, this is an open
item (498/8708-44).
Total No.
Total No. of In Review Total No.
Procedure Volume Procedures Process Apprcved
"
PMP01-Maintenance Adminis-
trative Procedures 5 1 3
! PMP02-Maintenance General
Procedures 8 0 7
PMP05-Electrical Maintenance
i Procedures 438 21 384
PMP06-Metrology Lab
Calibration Procedures 321 46 236
i
PMP07-I&C Maintenance Procedures 155 1 51
- PMP08-I&C Calibration Procedures 613 27 328
4
PSP 02-I&C Functional Test
Surveillance Procedures 129 17 119
PSPO4-Mechanical Surveillance
Procedures 3 1 1
'
PSP 05-I&C Calibration Surveil-
lance Procedures 160 8 142
! PSP 06-Electrical Surveillance
- Procedures 22 1 20
!
PSP 13-Response Time Surveil-
- lance Procedures 20 0 0
! .
!
!
.-- -, , - . - - - . - , - - , - - . . - . , , , - _ , . . . - .._ - - - . - . - . . - - , . . . - - - , - .-
. .
55
(1) Maintenance Procedures
(a) Procedures Reviewed
Procedure No. Revision . Title
OPGP03-Z0-0007 2 " Conduct of Maintenance"
OPMP01-ZA-0004 4 " Maintenance Procedures"
OPGP03-ZM-0003 7 " Maintenance Work Request
Program"
OPGP03-ZA-0010 2 " Plant Procedure Compliance,
Implementation, and Review"
1 PSP 06-DJ-0001 0 "125 Volt Class IE Battery
7-Day Surveillance Test"
l
1 PSP 06-DJ-0002 0 "125 Volt Class 1E Battery
Quarterly Surveillance-
Test"
OPGP03-ZO-0004 0 " Plant Conduct of Operations"
OPMP04-AF-0001 3 " Auxiliary Feedwater Pump
Maintenance"
OPMP04-CC-0001 3 " Component Cooling Water
Pump Maintenance"
0PMP04-DG-0004 0 " Standby Diesel Generator
Starting Air Compressor
Maintenance"
OPMP04-DG-0005 0 " Standby Diesel Generator
Maintenance"
OPMP04-FC-0001 1 " Spent Fuel Pit Cooling Pump
Maintenance"
OPMP04-FW-0003 1 "Atwood-Morrill Air Assist
20-Inch Feedwater Check
Valve Maintenance"
OPMP04-J F-0001 0 " Fuel Handling Machine
Inspection"
OPMP04-MS-0001 1 " Main Steam Safety Valve
Removal and Installation"
l
.
.. .
i
. .
56
l
OPMP04-MS-0002 2 " Main Steam Dump Valve
Actuator Maintenance"
0PMP04-MS-0003 1 " Main Steam Dump Valve
Actuator Removal and
Reinstallation"
1PMP05-DJ-006 0 " Battery Charger
Maintenance-Class 1E
125 VDC Distribution Panels"
1PMP05-VA-002 0 " Inverter /Reclifier Mainten-
ance Westinghouse 7.5 KVA"
1PMP05-PM-1101 0 "Switchgear Maintenance-
MCCEIAl"
1PMP05-PK-1014 0 "Switchgear Maintenance-
Bus E1A Cubicle 14"
1PMP05-PK-2014 0 "Switchgear Maintenance-
Bus EIB Cubicle 14"
1PMP05-PK-3014 0 "Switchgear Maintenance-
Bus E1C Cubicle 14"
OPMP04-SN-0002 1 " Hydraulic Snubber Removal
and Installation, Fluid
Addition and Sampling"
0PMP04-SN-0006 0 " Anchor Darling Model 151
Mechanical Snubber
Maintenance"
OPMP04-RC-0003 1 " Reactor Coolant Pump
Maintenance"
OPMP04-RC-0007 0 " Pressurizer Spray Valve
Maintenance"
OPMP04-RX-0001 1 " Reactor Vessel Head Removal
For Non-Rapid Refueling"
OPMP04-SI-0002 1 "High Head Safety Injection
Pump Maintenance"
OPMPO4-ZG-0006 3 "Limitorque Operator
Removal and Installation"
OPMPO4-ZG-0017 0 " Pacific Gate Valve Mainten-
ance (Bolted Bonnet)"
.
. .
57
(b) NRC Inspector Observations / Concerns
1) OPGP03-ZM-0003, Revision 7, " Maintenance Work Request
Program"
o Step 2.5 rather than stating "see addenda" could
more appropriately specify the applicable addenda
for this step.
o Step 3.1.1 is worded in such a way that a number
of people other than the shift supervisor
(control room) would have the authority to
release installed systems for maintenance. The
work start approval authority needs to be more
clearly defined so that the shift supervisor is
clearly the control point for releasing installed
components / equipment. Pending clarification of
the approval authority for release of systems
that could affect the plant, this is an open
item (498/8708-45).
o Step 4.1.7 should include words to the effect
that deletions should be lined once through,
initialed, and dated.
2) OPGP03-ZO-0004, Revision 0, " Plant Conduct of
Operations"
Step 4.10.6 states, "The shift supervisor or Chemical
Operations Foreman, as applicable shall authorize work
start approval for all maintenance, test, and other
activities that may affect the operation of the plant
or the status of plant structures, systems, and
components." The "as applicable" should be more
definitive to insure the shift supervisor controls
installed components / equipment. Pending clarification
of the "as applicable," this is an open
'
item (498/8708-46).
3) OPMP04-AF-0001, Revision 3, " Auxiliary Feedwater Pump
Maintenance"
" Documentation" section required by Addendum 3 of
Procedure OPMP01-ZA-0004 was missing.
4) OPMP04-CC-0001, Revision 3, " Component Cooling Water
Pump Maintenance"
o There was a possibility of placing wrong data on
data sheet due to dats Steps 5.12.5 and 5.12.4
being out of sequence.
. - --
. - ,
. . _ . . .. ~ _ . _ _ _ - .-. . _ _ - .. . _ ._
. .
58
'
!
,
o One fastener torque value was different from that
recommended in vendor manual. The procedure
required torquing of the motor mounting bolts to
324 ft-lbs while the vendor manual required
torquing to 70 ft-lbs.
.i Pending correction of the above items, this .is an open
- item (498/8708-47).
5) OPMP04-FC-0001, Revision 1, " Spent Fuel Pit Cooling
Pump Maintenance"
o " Documentation" section required by Addendum 3 of
Procedure OPMP01-ZA-0004 was missing.
4 o Part numbers in parenthesis after the part name
'
refer to items on the figure in Addendum 1;
I however, there was no reference to the addendum
at the start of the procedure (Step 5.7).
l o Steps 5.13.24 and 5.13.26 are incorrectly
i identified as 5.13.25 and 5.13.27 on the data
,
sheet.
,
o Step 5.12.9 should note the total indicated ,
runout to be less than 0.002-inch.
! Pending resolution of the above items, this is an open !
! item (498/8708-48).
! 6) OPMPO4-FW-0003, Revision 1, "Atwood-Morrill Air Assist
20-Inch Feedwater Check Valve Maintenance"
" Documentation" section as required by Addendum 3 of
Procedure OPMP01-ZA-0004 was missing.
7) OPMP04-JF-0001, Revision 0, " Fuel Handling Machine
Inspections"
o Items located on the data sheet did not have an
,.
asterisk and corresponding note nor any notation
'
within the main body to identify information
required to be recorded,
o Step 3.1 was apparently misnumbered, since on
the data sheet it was a chart for making not
n applicable undesired sections while in main body
of the report, it was a prerequisite for
obtaining a cleaning solvent.
.
i
w.-.m,.,----.-m--,-,%-+,.-,,.e,cy----,& --,,---.---,--------ewy -m y-----w-- , - - , -v-yy, --, , - - -r,,,,+w,-
.--n - ,, -e , - - - . tm-
,
. .
59
o Step 6.1 is not numbered on the data sheet as it
should be to be consistent with other procedures.
Pending resolution of the above items, this is an open
item (498/8708-19).
8) IPMP05-DJ-006, ~ Revision 0, " Battery Charger
Maintenance-Class IE 125 V DC Distribution Panels"
Step 4.4 should specify minimum accuracy requirements
for M&TE.
9) IPMP05-JA-0002, Revision 0, " Inverter / Rectifier
Maintenance Westinghouse 7.5 KVA"
o Data Sheet -1, Step 6.9 should read "3CB Input DC
Breaker in lieu of 3CB AC Input DC Breaker."
o Step 4.3.1 M&TE requirements should include a
second torque wrench to be used when low torque
values are required.
o Step 6.17 should include instructions to include
data sheets from referenced procedures as part of
the required documentation.
Pending resolution of the above items, this is an open
item (498/8708-50).
10) IPMP05-PM-1101, Revision 0, "Switchgear Maintenance -
MCCEIAl"
o Step 4.3.2 should specify accuracy requirements
for M&TE.
o Torque wrench should be of 15-80 foot pound range
to correspond to Addendum 1.
o Quality Control should be notified prior to
performing any torquing on bolted connections.
- Pending resolution of the above items, this is an open
,
item (498/8708-51).
i 11) IPMP05-PK-2014, Revision 0, "Switchgear Maintenance -
Bus ElB Cubicle 14."
i- .
i Step 4.5.3 should include opening of breakers for
containment spray and component cooling water pumps.
i
!
i
i
- _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ ___ . _ _ _ .
a .
60
12) IPMP05-PK-3014, Revision 0, "Switchgear Maintenance -
Bus E1C Cubicle 14."
o Step 4.5.3 should place 4.16 KV Bus E1C Cubicle I
supply breaker in "Open" position and overcurrent
lockout relay in " Reset" position. Procedure
states them in reverse.
o Overcurrent lockout relay should be identified as
"86B" and not as " Breaker" under the " Comp. Name"
column in Step 4.5.3.
o "86A" relay should be identified as " Generator
Differential Lockout Relay" in Step 4.5.3. This
same comment applies to Procedures 1PMP05-PK-1014
and 2PMP05-PK-2014.
Pending resolution of the above items, this is an open
item (498/8708-52).
, 13) OPMP04-DG-0004, Revision 0, " Standby Diesel Generator
Starting Air Compressor Maintenance"
o " Documentation" section as required by Addendum 3
of Procedure OPMP01-ZA-0004 is missing.
o It was unclear as to what " GAP (New)" and
" Installed" meant in the table on the data sheet
for Steps 5.1.3.4 through 5.1.3.7 when the
procedure called for compression ring " gap" and
" location".
Pending resolution of the above items, this is an open
item (498/8708-53).
14) OPMP04-DG-0005, Revision 0, " Standby Diesel Generator
Engine Maintenance"
o " Documentation" section as required by Addendum 3
of Procedure OPMP01-ZA-0004 was missing.
o On Page 82, Step 5.7.5 currently reads 5.5.7.
o QIP designations were missing to left of QA/QC Rep
line in Steps 5.16.16.1, 5.24.12, 5.29.13,
5.34.7.1, 5.34.12, and 5.38 on data sheet.
o Step 5.16.5 (on page 26) would be clearer with
addition of "by tightening air starting valve
nut."
- - _ _ _ _ _ _ _ _ _ _ _ _ _ _
. o
61
o QIP designations were missing to left of
Step 5.33.5 in main body of procedure and on data
sheet.
o Step 5.29.8 regarding QIP should read 5.29.8 in
lieu of 5.8.29.
o Step 5.13.3 would be clarified it if referenced
figure 7.
o Section 5.16 should have QC verify cleanliness if
item would be enclosed after assembly,
o Step 5.16.5 was unclear as to the desired torque
value.
o Clarity would be increased if during reassembly
references to figures were included.
The applicant stated that this procedure was going to ,
be deleted and activities accomplished using other
procedures. Pending its deletion, this is an open
item (498/8708-54).
15) OPMP04-ZG-0017, Revision 0, " Pacific Gate Valve
Maintenance (Bolted Bonnet)"
o " Documentation" section as required by Addendum 3
of OPMP01-ZA-0004 was missing.
o It was unclear regarding which column was to be
used on the torque chart in Addendum 6.
Pending correction of the above items, this is an open
item (498/8708-55).
16) OPMP04-SN-0002, Revision 1, " Hydraulic Snubber Removal
and Installation, Fluid Addition and Sampling"
o Step 5.1.18 in the main body appeared as
Step 5.1.17 on the data sheet.
o " Documentation" section as required by Addendum 3
of OPMP01-ZA-0004 was missing.
17) OPMP04-SN-0006, Revision 0, " Anchor Darling Model 151
Mechanical Snubber Maintenance"
The note and corresponding asterisks to the left of
steps with data sheet entries was not used for all
but one other of the procedures in the OPMP04 group.
.
_
. .
62
Pending correction of the above item, this is an open
item (498/8708-56).
18) OPMP04-SI-0002, Revision 1, "High Head Safety
Injection Pump Maintenance"
o After reviewing the corresponding vendor manual,
there appeared to be a discrepancy in the
acceptance criteria for Step 5.12.3.50. The
criteria given was .008 to .012 inches and that
stated in the vendor manual for the " Head and
bowl bearing running clearance" was .013 to
.020 inches.
o Step 5.13.6 in the procedure should be asterisked
and the data sheet should contain entries
pertaining to torquing of the fourth stage
through seventeenth stage bowl fasteners.
o After comparing the vendor manual torque chart
to the procedure required torque values, the
inspector identified discrepancies in the
following steps:
-
5.13.5.7 and 5.13.6
-
5.13.8.2
-
5.13.9.2
-
5.13.10.10
o For the smaller torquing values called for in the
procedure there was not a prerequisite for a
torque wrench that has that value fall into the
upper half of the scale.
Pending resolution of the above items, this is an open
item (498/8707-57).
19) OPMP04-ZG-0006, Revision 3, "Limitorque Operator
Removal and Installation"
" Documentation" section required by 0PMP01-ZA-0004
was missing.
20) OPMP04-RC-0007, Revision 0, " Pressurizer Spray l
Valve Maintenance" l
" Documentation" section required by
OPMP01-ZA-0004 was missing.
_
. . . _ _ _ _ . _ _ - .__ _ - ___ _ _. _ . _ _ _. _ ____ _ __
i .- .
. 63
.
- 21) Generic Comments
i o Some of the procedures appeared to have too
many references. Elimination of references
that are not necessary for the performance
of the procedure would simplify the '
procedure and make it more_ user oriented.
'
Examples of this were Procedure OPMP03-ZO-0004,
i' Revision 0 (40 references), .
Procedure OPGP03-ZM-0003, Revision 7 (40
, references), and Procedure OPMP01-ZA-0004,
{ Revision 4 (11 references).
I
o The NRC inspector investigated the
j licensee's controls over torquing of
l- safety-related fasteners. Torquing of
4
fasteners at STP followed the information
,
located in each component's respective
! vendor manual. If there was no information
I given, but the determination that torquing
of fasteners would be required, then Plant
Procedure OPMP02-ZG-0004, " Fastener Torquing-
l and Detensioning," was required to be
- utilized.
I
l The NRC inspector compared 10 vendor manuals
- to their respective procedures out of a
l sample of 18 procedures. It was determined
I that the corrective maintenance procedures
I satisfactorily reflected the information
contained in the manuals.
- o The NRC inspector reviewed the licensee's
l
process for adding quality control hold
- points to corrective maintenance procedures.
- In the procedures reviewed, the licensee had
"
hold points to verify cleanliness of the
component and parts before enclosing the part
and/or system. QC was in the review cycle,
and had the option during maintenance work
i request review to add additional hold points
! as required.
5 o Referencing other procedures and documents
. by revision number could be confusing,
t Administrative controls could be used to
I require use of the latest revision, except
'
in the cases where a specific revision to a
document is applicable (i.e., a commitment).
i
l
.
,
'.----... *
_ - - _ _ . . . _ _ _.
.,.
i
i
. s.
,
&
64
(2) Procedures For Control of M&TE
The NRC' inspector reviewed plant maintenance procedures to
ascertain whether-the licensee's measuring and test control
program adequately provides. controls for the calibration,
testing, and checking of instrumentation and equipment.
Calibrating and testing procedures were reviewed to verify that
each was in the appropriate format as defined in administrative
control procedures and was technically adequate to accomplish the
stated purpose.
The NRC inspector reviewed Procedure OPGP03-ZM-0001, Revision 9,
a " Measuring and Test Equipment Control Program," which provides
, program guidelines for the control of M&TE. For the purposes of
' J' this procedure, M&TE does not include permanently installed
<
4
plant instrumentation. Calibration and testing of plant
r instruments is addressed by " Plant Instrumentation Sealing
Program," Revision 2. This procedure is applicable to devices
which require calibration or testing by the instrumentation and
i controls section of the maintenance division. This procedure
delineates controls for calibration and-status
verification / notification of installed permanent plant
instruments. Thirteen calibration and testing procedures were
i reviewed to verify conformance to administrative guidelines and
<
technical adequacy. The procedures reviewed were found to be
'
technically adequate. They defined step-by-step instructions to
accomplish their stated purpose. The NRC inspector did note
some procedures needed to be reviewed and revised to conform to
administrative guidelines. The NRC inspector held discussions
with various personnel during the review of the procedures.
4c These discussions indicated that they were familiar with program
guidelines and requirements. NRC inspector observations were
discussed with appropriate licensee personnel.
(a) Procedures Reviewed
,
Procedure No. Revision Title
IPMP05-YA-0001 1 " Vital Distribution Panel
Tests"
i
OPMP05-ZE-0104 0 " Frequency Transducer
Calibration"
OPMP05-ZE-0034 1 " Calibration of ITE-27
Relays"
{-
" Insulation Resistance
'
OPMP05-ZE-0203 2
!
Testing-4.16K and 13.8K
'
Volt Motors"
. ._ - - .._-_ ___ .. - _ _ . _ ~ _ - - - - _ _ _ _ _ - _ . - . . . - - ~-- _ _ . . _ -
_ _ _ _ - _ _ _ __-___-____ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
9
.. .
65
OPMP05-ZE-0206 0 " Potential Transformer Tests"
OPMP06-ZT-0186' 2 " Calibration of the
Westinghouse Relay Test Set"
0PMP06-ZT-0279 0 " Calibration of General
Resistance RTD-100 RTD
Simulator"
.0PMP06-ZT-0293 0 " Calibration of Fluke
80I-600 Clamp-On Current
Transformer"
'
OPMP07-SP-0001 0 "SSPS Decoder Printed Circuit
Board Test and Rework"
OPMP08-ZI-0011 2 " Generic Temperature Switch
l Calibration (Filled
-
Element)"
OPMP08-ZI-0065 0 " Field Testing Of Power
Supplies and Over Voltage
Protectors"
0PMP08ZI-0203' 2 " Pressure or Differential
Pressure Indicator
'
Calibration"
, s -n IPMP08-SI-0861 0 "RHR/LHSI Pump.1.A Discharge
,
c, -
Pressure Calibration
lo.lJ-
'
A?
(P-0861)"
', N f;
OPGP03-ZM-0011 2 " Plant Instrumentation
Scaling"
0PGP03-ZM-0016 0 " Installed Plant
Instrumentation Calibration
'
L
g. ]gf '
Verification Program"
l ,
(b) NRC Inspector Observations / Concern _s
"
1) OPMP05-ZE-0104, Revisfeb u, " Frequency Transducer
Calibration"
o Step 6.6.5 data sheet should read " Transducer
removed from service."
o Precautions should include instructions to verify
equipment clearance, if applicable.
.
_
. .
66
2) OPMP05-ZE-0034, Revision 1, " Calibration of ITE-27
Relays"
Precautions should include instructions to verify
equipment clearance,'if applicable.
3) OPMP05-ZE-0203, Revision 2, " Insulation Resistance
Testing-4.16 K and 13.8 K Volt Motors"
o Steps 6.12 through 6.15 should be incorporated
into Restoration / Documentation sections.
o Procedure should specify actions to take if
acceptance criteria is not met.
4) OPMP07-SP-0001, Revision 0, "SSPS Decoder Printed
Circuit Board Test and Rework"
o Procedure should include instructions on how to
complete and process test documentation.
o Precautions should include instructions to verify
equipment clearance, if applicable,
o Though this procedure is technically adequate to
accomplish its stated purpose, it should not be
used until it is revised due to inadequate
procedure guidelines and instructions. Procedure
was in the review and revision process at.the
time of the NRC inspection.
Pending completion of this procedure revision, this is
an open item (498/8708-58).
5) OPMP08-ZI-0011, Revision 2, " Generic Temperature
Switch Calibration (Filled Element)"
o Precautions should include instructions to verify
equipment clearance, if applicable,
o Step 7.4.3 states " proceed to Step 7.3," which is
incorrect.
6) IPMP08-SI-0861, Revision 0, "RHR/LHSI Pump 1A
Discharge Pressure Calibration (P-0861)."
Addendum 1 does not identify V2 per Step 7.3.1.2.
7) OPMP08-ZI-0065, Revision 0, " Field Testing of Power
Supplies and,0vervoltage Protectors"
_ - . . ._ _
'
. .
4~
67- l
1
l
This procedure should be clarified in the following
-areas:
o Procedure should define responsibilities of
personnel.
o Procedure should define actions to be taken if
-acceptance criteria is not met.
o Precautions should include
restoration / documentation instructions.
o Procedure should specify M&TE accuracy
requirements.
'
o Procedure should identify references where
acceptance criteria can-be found.
'
.'
o Procedure should not be used until a revision is
issued do to inadequate procedure guidelines and
instructions. Revision 0 is in the review and
revision process.
Pending resolution of the above items, this is an open
item (498/8708-59).
d. Emergency Operating Procedures
- (1)' Purpose
! The purpose of this inspection was to determine whether E0Ps had
been prepared in accordance with the PGP and whether they were
technically adequate to control safety-related functions in the
- event of system or component malfunction. At the time of this
inspection, the PGP had been submitted to the NRC Office of NRR,
but the NRC staff review of PGP was not yet complete.
i
(2) Procedures Reviewed
i The following procedures were reviewed during this inspection:
o OPGP03-ZA-0027, Revision 1, " Emergency Operating Procedures
Preparation, Approval, and Implementation"
o OPOP01-ZA-0006, Revision 2, " Emergency Procedures Writers
, Guide and Verification"
i
o 1 POP 05-E0-E000, Revision 1, " Reactor Trip or Safety
Injection"
,
'.
!
, - , . . - , . . . _ . . . . _, , _ . _ _ ~ _ - . _ , - . . - - _ _ . _ . . _ _ - - . . . . - . _ , . , _ . - . . . .
. _
._- _ . . . _
,
c.^ ,
-
S 68
s
o IPOP05-E0-E010, Revision 1, " Loss of Reactor or Secondary
Coolant"
o 1 POP 05-E0-EC00, Revision 1, " Loss of All AC Power"
o IPOP05-E0-FRH1, Revision 1, " Response to Loss of Secondary
Heat Sink"
(3) Status Of Completi2n
The NRC inspector compared the index of the applicant's E0Ps to
the index of the 40G ERGS. This index indicated that E0Ps had
been prepared ano approved for all WOG ERGS. One additional E0P
was in preparati>n in response to the fire hazards analysis.
This E0P was scheduled to be approved prior to loading fuel.
i ,
(4) Technical Adequicy
The NRC inspector assessed E0P technical adequacy by comparing
the E0Ps to the WOG ERGS and plant piping and instrumentation
drawings. While'no major technical inadequacies were
<
identified, several errors and discrepancies indicating the ,
-applicant's failure to develop E0Ps appropriate to the
. circumstances as required by 10 CFR 50 Appendix B, Criterion V,
I were identified. These are listed below:
(a) Statior, Procedure 1 POP 05-E0-E000, " Reactor Trip or Safety
Injection"
o Step 10 did not state how many essential cooling water
pumps should be running.
o Step 11.1 required ' transfer' of reactor containment
fan cooling to component cooling water. This step
should have verified that automatic transfer had
,
occurred.
l
j o Step 22.2 listed incorrect equipment designations for
the pressurizer spray valves,
o Step 31.2 incorrectly directed the secondary operator
to sample the steam generators. This function should
be performed by plant chemists, as directed by the
Unit Supervisor.
(b) Station Procedure IPOP05-E0-E010, " Loss of Reactor or
.
Secondary Coolant," had an incorrect reactor coolant system
pressure referenced in Entry Condition 3.
.
e e v -1* y,.-.~.-, - - .--- ,. .-,-y ,~. _.. - .,____, e--,-,..,,,,m. ,-,..r,,,,_,2 . - _ . _ _.---._,.-,.r._..-2.-_,,, c.,.v. - - r m.m.-e_.2
p
r
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69
(c) Station Procedure IPOP05-E0-EC00, " Loss of All AC Power,"
failed to include the desired entry step of the referenced
procedure in the contingency action statement of Step 27.1.
(d) Station Procedure IP0F05-E0-FRH1, "Resperse to Loss of
Secondary Heat Sink."
o The step 9 caution statement stated, ". . . establish
RCS heat by RCS bleed and feed." The word ' Removal'
should be inserted after ' heat'.
o The Step 19 caution statement contained an incorrect
procedure number reference.
o The note before Step 22 was repeated before Step 23,
deleting the proper note which should have been placed
before Step 23, "After closing a PORV, it may be
.necessary to wait for RCS pressure to increase to
permit stopping high-head SI pumps in Step 22."
Pending licensee response, the above deficiencies will be
tracked as an open item (498/8708-60).
(5) Deviations from WOG ERGS
Section II of the STP PGP dated May 16, 1985, stated, "As the
procedure writer prepared the E0P, deviation between the WOG
ERGS and the plant specific procedure caused by plant design or
preferred due to control board layout were documented on form
! (4) of the writer's guide. Reasons for deviations were also
documented on the same form." Section 2.3.1 of Station
Procedure OPOP01-ZA-0006, " Emergency Procedure Writers Guide and
Verification," states, "When preparing the E0Ps, situations may
arise where the intent specified by the guidelines may have to
be altered . . . due to the STP design. When this happens, the
, writer shall complete the documentation for changes to the WOG
'
Guidelines or 'EOP Step Justification / Verification Form' (-4)."
- The NRC inspector reviewed the E0P verification packages for
Revisions 0 and 1 of the E0Ps listed above. The Step
l'
Justification / Verification (SJ/V) forms completed during
procedure preparation and revision were included in these
packages. For a small number of deviations from the WOG ERGS,
the NRC inspector found that the SJ/V forms provided a very good
. justification for the deviation. However, many deviations from
! the WOG ERGS were not documented and justified on SJ/V forms and
'
there were inadequate basis or justification for scme of the
deviations which were documented on SJ/V forms. Pending
- licensee resolution, the lack of documentation o' deviations
from the WOG ERGS will be tracked as an open item (498/8708-61).
.
a
. ..
70
The NRC inspector found that SJ/V forms were generally not used
to document and provide a basis for plant specific information
~
inserted into the E0Ps where the WOG ERGS used a notation such
as, " Establish main feedwater flow [-Enter plant specific
means]."
It should be noted that the lack of SJ/V forms was frequently
, documented by. reviewers in the procedure verification process
using E0P Discrepancy / Comment forms from Station
Procedure OPOP01-ZA-0006. In these cases the resolution of the
comment provided some justification of the deviation from the
WOG ERGS.
(6)' Plant Specific Values
One plant specific value from each of the four E0Ps was selected
for verification. The reference for plant specific values.was
the HL&P Emergency Operation Procedure Setpoint Document,
Revision 1, dated November 10, 1986. No problems were
identified in this verification. Applicant representatives
informed the NRC inspector that Revision 2 of the Setpoint
Document has been issued but not yet incorporated into the E0Ps.
They plan to incorporate the latest setpoints into-the E0Ps
prior to loading fuel.
(7) Compliance With Writers Guide
The NRC inspector reviewed the four E0Ps listed above to
determine whether they had been written in accordance with the
guidance provided in Station Procedure OPOP01-ZA-0006,
" Emergency Procedure Writers Guide and Verification." General
conformance was noted with the exceptions listed below:
(a) Section 3.1.2.1 of the writers guide required that each
operator copy of an E0P present the user information and
steps on the left page and the non-user information on the
right page when opened. The NRC inspector noted that
action steps identified as being performed by the Unit
Supervisor were not included with the non-user information
on the right page of the operator's copies of the EOPs.
(b) Section 5.4 of the writers guide required maintenance of a
direct horizontal relationship between the related action
steps in the left column and the contingency action steps
in the right column. The NRC inspector found that the
contingency action step associated with action Step 5.3 of
Station Procedure IPOP05-EO-E010 was aligned horizontally
with action Step 5.2.
-(c) Section 17.4 of the writers guide required that missing
information shall be listed on a separate punchlist at the
. ,
71
end of the written procedure body. The NRC inspector found
that no punchlist was attached to Station
Procedure 1 POP 05-E0-EC00 although Steps 15 and 16 of this
procedure were missing information which should have been
identified on a punchlist. The missing information related
to contingency actions for filling the auxiliary feedwater
storage tank using the fire water system and local
operation of steam dump valves.
(d) Section 18.2 of the writers guide required that the
appropriate emergency action level be entered into the E0P
at the earliest possible point. The NRC inspector found
that Station Procedure IPOP05-EC-FHR1 contained no
reference to emergency action levels which had been
reached. Station Procedure OEPP01-ZA-0001, " Emergency
Classification," Addendum 3, indicated that reaching
Step 9.0 of 1 POP 05-E0-FRH1 was the emergency action level
for declaration of a General Emergency.
Pending licensee response, the above exceptions to the writers
guide will be tracked as an open item (498/8708-62).
(8) Verification and Validation
The NRC inspector reviewed the verification packages for
Revisions 0 and 1 of the E0Ps listed above, the Emergency
Operating Procedure Validation Report dated December 22, 1986,
and the checklists and deficiency sheets associated with the
validation program. An applicant representative stated that
the final validation report was in preparation at the time of
this inspection. It appeared that the verification and
validation program was conducted in accordance with the PGP and
associated plant procedures. However, some weakness in this
program was indicated by the dis ~crepancies discussed above. .
(9) Other Comments
(a) During the review of E0P verification packages, the NRC
inspector noted that one individual signed a License
Compliance Review Form (OPGP03-ZA-0003-1) as both preparer
and reviewer. While this action was not prohibited by
Plant Procedure OPGP03-ZA-0003, the NRC inspector stated
that this was not a generally accepted practice. Applicant
representatives stated that they had recognized this as a
problem and that corrective action was underway.
(b) Step 6.3.1.6 of Station Procedure OPGP03-ZA-0027 appeared
to be incomplete.
(c) The E0Ps included no statement of purpose or scope.
( NUREG-0899, Section 5.4.3 states, "Each E0P should contain
l
l
. .. , _
_ _ _ --
. . . - . . -- - . ~ . . .- ~ . -
l'
. .
72
.
a brief statement that describes what it is intended to
accomplish. In many cases it may be possible to include
the scope in the title of the E0P without making the title
too long." While the PGP indicated that the writers guide
- was based on NUREG-0899 (and other references), the writers
guide did not require E0Ps to have a statement of purpose
- or scope. The NRC inspector noted that the WOG ERGS each
i begin with a statement of purpose, some of which provide
considerably more information about the purpose of the.
procedure than does the procedure title. This comment is
,
expected to be resolved during the process of NRC review
and approval of the PGP.
i e. Off-Normal Operating and Alarm Response' Procedures
The NRC inspector reviewed selected applicant off-normal. operating
j- procedures and annunciator response procedures to verify they were in
'
the required format and that they were technically adequate to
perform the designated function. The NRC inspector walked down
selected procedures to verify that the 'as-built' conditions were
compatible with the plant procedures.
.
(1) Procedures Reviewed
(a) Off-Normal Procedures
,
t ~
Procedure No. Revision Title
1 POP 04-RC-0001 2 "High Reactor Coolant
,
System Activity"
i
1 POP 04-FW-0001 1 " Loss of Feedwater Flow
or Control"
1 POP 04-CR-0001 1 " Main Condenser-Loss of
Vacuum Off-Normal'
Procedures"
I IPOP04-RC-0002 1 " Loss of Reactor Coolant
Pump"
!
(b) Operating Procedures
a
4
Procedure No. Revision Title
4 1 POP 02-CV-0004 1 " Chemical & Volume
i Control System"
1
\
f
4
y -w r-'vm,w.--.",. ,,,-.,,,,---r-v -,,, -----,~-r,,,
- , rm, y--.,y--r--,,,---,,,.w,,,,,-e-v- r=w e ,- ,, m mw e me e w- ----m , sr e w -- --v- . ~w- m
_ _ _
. ~ _ - - . . . . . . . . - . --. .-- -. .- . .- -.
1
. .
73
- (c) Annunciator Response Instructions
IPOP09-AN-04M8-D-4 0 "LETDN HX OUTLET FLOW
HI/LO"
1 POP 09-AN-04M8-D-3 0 "LETDN HX TEMP HI DEMIN
DVRT"
1 POP 09-AN-04M8-D-1 0 " SEAL WTR INJ FILTER
DR HI"
1 POP 09-AN-04M8-C-4 0 "LETDN HX OUTLET PRESS
HI"
1 POP 09-AN-04M8-C-3 0 "LETDN HX OUTLET TEMP ,
HI"
1
1 POP 09-AN-04M8-C-2 0 "BTS DEMIN DP HI"
1 POP 09-AN-05M2-E-1 0 "RCP CCW FLOW LO"
1 POP 09-AN-05M2-E-2 0 "RCP TRIP"
1 POP 09-AN-05M2-E-3 0 "Rx VSL FLNGE LEAK TEMP
HI"
1 POP 09-AN-05M2-C-1 thru 0 "RCP UPPR OIL RSVR LVL
IPOP09-AN-05M2-C-4 HI/LO"
1 POP 09-AN-05M2-D-1 thru 0 "RCP LOWR OIL RSVR LVL-
1P0P09-AN-05M2-D-4 HI/LO"
1P0P09-AN-07M3-E-7 0 " MAIN COND VACUUM LO"
1 POP 09-AN-07M3-F-6 0 "F.W. S/V PMP L. O. AUX
PMP TRBL"
,
IPOP09-AM-07M3-F-8 0 " SEAL LEAK OFF TNK LVL
l HI"
l
l
These annunciator response instructions were walked down in
j- the plant.
' (2) NRC Inspector Observations / Concerns
i
. (a) IPOP02-CV-0004, Revision 1, " Chemical and Volume Control l
- System"
!
i
!
!
i
..,,,--,,-,1--- . , _ , , , , . . _., .,_.m --- ,,,.---- - ... ,_-_--,_.-_-....----m-_.,-..,,_,,,_-,.., , , , . . . , - . . _ _ , . , , . . , - _ , , ~ , - - . . . , , - -
-. ..
74
o In Step 11.3 there is an incorrect valve designation
in the valve lineup in that in Cation Bed Demin IA
valve line up should read "1*CV-129A" in lieu of
"1*CV-129B".
o In Step 11.5 'MAB' should be 'MEAB'.
(b) IPOP04-FW-0001, Revision 1, " Loss of Feedwater Flow or
Control"
According to the instructions in Steps 4.3 and 4.4 of
' Procedure OPOP01-ZA-0007, Revision 1, "Off-Normal
Procedures Writer's Guide," there is a conflicting use of
" Note" versus " Caution" statements. An example of this is
the following " Note" located after Step 4.3 which should be
a " Caution" statement:
" Reducing turbine load too rapidly may cause an unnecessary
reactor trip due to the effects of SG shrink."
Pending resolution, this is an open item (498/8708-63).
(c) IPOP04-RC-0002, Revision 1, " Loss of Reactor Coolant Pump"
The following " Note" which occurs at Step 4.0 in the
procedures is an instruction statement (i.e., action step).
This is contrary to the instructions in Step 4.4 of
OPOP01-ZA-0007, Revision 1, "Off-Normal Procedures Writer's
Guide," which states that notes should not be used as
instruction statements.
" Note: If Rx pwr is above P-8 and conditions exist calling
for an immediate RCP trip then trip the reactor first then
the RCP to ensure heat removal. I_f Rx power is less than
P-8, trip the RCP only."
Pending resolution, this is an open item (498/8708-64).
(d) OPOP09-AN-05M2-E-1, Revision 0, "RCP CCW Flow LO"
The immediate actions section (below) of the procedure
should also verify the remainder of the CCW system is in
service (i.e., per a POP or by looking at a control panel).
" Verify 1*CC-MOV-318, 1*CC-MOV-029, inlet isol. are open
and 1*CC-MOV-403, 1*CC-FV-4493, 1*CC-MOV-404, 1*CC-MOV-542,
outlet isolation valves are open."
(e) 1 POP 39-AN-05M2-E-3, Revision 0, "Rx VSL FLNGE LEAK TEMP HI"
_ - ____ - _ _____ -_____________-___ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
o o
75
3rocedure states " Place HS-3400 to 'Close' position as
firected by Unit Supervisor," however, the switch is not
libeled as HS-3400 on the control board. Applicant has
aided this item to the labels correction log.
(f) IPOP09-AM-07M3-F-8, Revision 0, " SEAL LEAKOFF TANK LVL HI"
Supplementary actions states to " Manually control level
using Bypass Valve 1-FW-0483;" however, Valve 1-FW-0483 is
installed under floor grating that makes it not readily
accessible.
Pending resolution, this is an open item (498/8708-65).
(g) In general, action statements are inconsistent in the
amount of detail provided for performing the action. Some
of the action statements only give general guidance as in
Step 5.1 of IPOPO4-FW-0001, Revision 1, " Loss of Feedwater
Flow or Control" which states:
" Ensure steam dumps are operating properly and verify that
T
AVG
is being matched with T Ref' "
This action statement should be more specific by
referencing instruments to be monitored. Examples of
action statements which provide specific instructions are
Steps 4.1 and 5.1 of Procedure IPOP04-CR-0001, Revision 1,
" Main Condenser-Loss of Vacuum Off-Normal Procedures,"
which state respectively:
o "4.1, Verify all vacuum pumps operating (hogging)
(CD-009)."
o "5.1, Verify steam supply to turbine seals (CP-008)."
Pending resolution, this is an open item (498/8708-66).
f. Surveillance Procedures
The NRC inspectors performed the following activities to verify that
the applicant had established adequate procedures to perform required
TS surveillances:
o Compared the proof-and-review version of STP Unit 1 TSs to the
STP index of surveillance procedures to verify that the
applicant had established or was establishing a procedure to
accomplish each surveillance required by the TSs.
o Performed in-depth review of selected effective / approved
surveillance procedures to verify that TS surveillance
requirements were satisfied.
- _ - _ _ _ _ _ _
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76
o Performed in plant walkdowns of selected surveillance procedures
to verify that as-built equipment and indicators reflected the
TS requirements.
(1) Procedures Reviewed
(a) Administrative
Procedure No. Revision Title
OPGP03-ZE-0005 3 " Plant Surveillance Procedure
Preparation"
(b) Instrumentation and Control (I&C) Functional
Procedure No. Revision Title
- 1 PSP 02-EH-6328, 0 " Turbine Thrott!e Valve
TA00T"
1 PSP 02-FW-0574 0 "SG-ID Narrow Range Level
Set 1 ACOT"
1 PSP 02-HC-0935 0 " Containment Pressure Set 3
ACOT"
1 PSP 02-MS-0506 1 " Turbine Impulse Chamber
Pressure Set 2 ACOT"
- 1 PSP 02-NI-0042 0 " Power Range Neutron Flux
Channel II ACOT"
1 PSP 02-RC-0427 0 "RCS Flow Loop 2 Set 1 ACOT"
1 PSP 02-SI-0952 0 " Accumulator 18 Level
Group IV ACOT"
1 PSP 02-SP-0001R 0 "SSPS Logic Train R
Functional Test"
1 PSP 02-RC-0452 0 "RCS Temperature Loop II
Set 1 ACOT"
1 PSP 02-RC-0466 0 " Pressurizer Level Set II
ACOT"
(c) System and Component
IPSP03-AF-0004 1 "AFW Pump II Reference Value
Measure"
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77
1 PSP 03-SI-0017 0 " Containment Spray Valve
Checklist"
1 PSP 03-CV-0007 0 " Boric Acid Transfer Pump 1A
Reference Valves Measures"
1 PSP 03-CV-0010 0 "Boration Flow Verification"
1 PSP 03-0G-0011 1 " Standby Diesel IZ Auto Start
on ESF Actuation Test
Signal"
1 PSP 03-EA-0002 0 "ESF Power Availability"
1 PSP 03-FP-0001 0 " Fire Protection System Valve
Operability Test"
1 PSP 03-RC-0006 0 " Reactor Coolant Inventory"
- 1 PSP 03-RH-0005 0 " Residual Heat Removal
Pump 1B Reference Valve I
Measurement"
- 1 PSP 03-RS-0002 0 " Manual Reactor Trip TADOT"
1 PSP 03-SI-0013 0 " Accumulator Isolation Valve
Verification"
- 1 PSP 03-SP-0001 0 " Remote Shutdown Monitoring
Instrumentation Channel
Check"
1 PSP 03-0G-0003 1 " Standby Diesel B Operability
Test"
1 PSP 03-RM-0001 1 "Raactor Makeup Water System
Valve Operability Test"
(d) I&C Calibration
Calibration"
1 PSP 05-FW-0503 0 "SG-IC Wide Range Level
Channel B Calibration"
1 PSP 05-AC-0936 0 " Feed Flow Loop I Set 3
Calibration"
1 PSP 05-NI-0031 0 " Source Range Channel I
Calibration"
. _ _ _ _ - - __
.- . - . _
o .
78
1 PSP 05-NI-0044 0 " Power Range Channel IV
Calibration"
1 PSP 05-RC-0417 0 "RCS Flow Loop I Set I
Calibration"
1 PSP 05-RC-0451 0 "RCS Temperature Loop I Set I
Calibration"
1 PSP 05-RC-0458 0 " Pressurizer Pressure Set IV
Calibration"
1 PSP 05-RC-0466 0 " Pressurizer Level Set II
Calibration"
1 PSP 05-SI-0954 1 " Accumulator 1C Level
Group III Calibration"
1 PSP 05-WL-0478 0 " Plant Liquid Waste Discharge
Flow Calibration"
1 PSP 05-CC-4503 b 0 "CCS Surge Tank Compartment A
Level Switch Calibration"
(e) Electrical
IPSP06-DG-0001 0 "Undervoltage Loss of Relay
Voltage Channel Calibration"
1 PSP 06-DJ-0004 0 "125 V Class 1E Battery
Service Surveillance Test"
- Procedures selected for in plant walkdown.
(2) NRC Inspector Observations / Concerns
(a) 1 PSP 03-SP-0001, Revision 1, " Remote Shutdown Monitoring
Instrumentation Channel Check"
The acceptance criteria in Step 7-1 on channel checks needs
to be more specific. Tolerances should be included so the
test parformer knows when an indicator is not functioning
properly. Presently the only criteria is that an
indication exists. The NRC inspector determined from
discussions with applicant personnel that the applicant
feels that since a quantitative assessment of the channel
behavior is not required by the TS that a tolerance is not
desired. The procedure does include a step that states
that if in the operators judgement the indications are in
error he is to report that information to the shift
supervisor.
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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79
(b) IPSP03-RH-0005, Revision 0, " Residual Heat Removal Pump 18
Reference Values Measurement"
This procedure lines up the system to run the RHR pump with
the suction valve closed. This did not appear to be a good
practice since the small volume of water within that closed
system could heat up rapidly and could cause cavitation and
possible water hammer problems. The NRC inspector
determined from discussion with applicant personnel that
the system design allows for cooling by the CCW system of
the water being pumped. The utility feels that this will
be sufficient to prevent cavitation problems due to heatup.
(c) OPGP03-ZE-0005, Revision 3, " Plant Surveillance Procedure
Preparation"
Step 3.2.4.b specifies that any LC0 which may be entered
during the performance of a surveillance test be inserted
in the pretest verification section of the procedure. This
was omitted from many of the procedures that were reviewed.
Applicant personnel stated that during the next revision to
PGP03-ZE-0005, " Plant Surveillance Procedure Preparation,"
the instructions for the contents of the pre-test
verification section will be modified.
Pending this revision, this is an open item (498/8708-67).
(d) 1 PSP 03-RC-0006, Revision 0, " Reactor Coolant Inventory,"
Procedure does not specifically address " leakage to RCP
seals." This leakage to the seals is stated in TS
Surveillance 4.4.6.2.1.c. Applicant states that TS
Surveillance 4.4.6.2.1.c is being deleted.
Pending this deletion and correcting the procedure, if
required, this is an open item (498/8708-68).
(e) IPSP03-SI-0013, Revision 0, " Accumulation Isolation Valve
Verification"
Step 5.3 states that an indicating light on the main
control board is used to verify power removed to the valve
operator. This verification should be done by breaker
position. The NRC inspector determined from discussion
with applicant personnel that the design of the controls on
the main control board provides for removal of control
power to the isolation valve and indication of such removal
and valve position. This is sufficient to ensure that
power has been removed to the isolation valve.
(f) IPSP03-SP-0001, Revision 0, " Remote Shutdown Monitoring
Instrumentation Channel Check"
_ _ - _ _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _
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80
Performance frequency for this procedure was listed as
quarterly in the surveillance procedure index computer
printout; however, TS 4.3.3.5, Table 4.3-6, requires the
surveillance to be performed monthly. Applicant revised
the surveillance testing master index so that the frequency
for this procedure was changed to monthly to comply with TS
requirements.
(g) The NRC inspector observed that QC should be involved in
performance of surveillance procedures. It was noted that
some procedures have QC involvement but most do not. The
NRC inspector determined from discussions with applicant
personnel that it is the applicants philosophy to have QC
be on the worker level. The applicant feels that the
training given the worker is sufficient to ensure quality
control and it is not necessary to include QC on all
procedures. The quality assurance department spot checks
the performance of procedures.
No violations or deviations were identified.
17. Exit Interview
The NRC resident inspectors met with -licensee representatives (denoted in
paragraph 1) on April 10, 1987, and summarized the scope and findings of
the inspection. Other meetings between NRC inspectors and licensee
management were held periodically during the inspection to discuss
identified concerns.
.. .. .