IR 05000312/1978005
| ML19309A384 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/03/1978 |
| From: | Crews J, Johnson P, Morrill P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML19309A369 | List: |
| References | |
| 50-312-78-05, 50-312-78-5, NUDOCS 8003270835 | |
| Download: ML19309A384 (6) | |
Text
..
.
.,
j
.
.
..
.
U. S. MJCLEAR REGULATDRY COMMISSION
.
0FFICE OF INSPECTION AND ENFORCEMEN'
- REGION V Report No.
50-312/78-05 Docket No.
50-312 License No.
DPR-54 Safeguards Group Licensee:
Sacramento Municipal Utility District P. O. Box 15830
'
Sacramento, California 95813
~
Facility Name:
Rancho Seco Inspection at:
Rancho Seco Inspection Conducted:
May 15-18. 1978 Inspectors:
(A M
'72 O
f opnson, Reactor Inspector 4 /> /W Date Signed M H.
- b.b P.
r1 1 Reactor Inspector Date Signed
.
Approved By:
.
/#lM Rh8
-
J.
Crewk,6hief,ReactorOperatior; sand
/ Date Signed Nuclear Support Branch
,
Sunnary:
-
' Inspection on May 15-18, 1978 (Report No. 50-312/78-05)
Areas Inspected: Routine, unannounced inspection of plant operations;
'
~
review and audit; procedures; safety limits, limiting conditions for operation, and limiting safety system settings; and licensee event reports.
The inspection involved 62 inspector-hours on-site by two NRC inspectors.
Results: One item of noncompliance (deficiency - no documentation of surycil-
,
lance test) was identified in one of the fiv,e areas inspected.
.
.
.
.
O 8003270 h [
'
l IE:V Form 219 (2)
..
.
.
..
.
..
-
. - _ - -
.-
-
,c
_
,
N
.
,
.
N
-
,
.
~
DETAILS 1.
Persons Contacted
.
- R. Rodriguez, Manager of Nuclear Operations
- L. Schwieger, Quality Assurance Director
- R. Colombo Technical Assistant N. Brock, Instrument Supervisor
- W. Ford, Operations Supervisor M. Carter, Shift Supervisor
- G. Coward, Senior Power Plant Engineer The inspectors also talked with and interviewed several other
,
.
licensee employees, including members of the technical and x
engineering staff, shift supervisors, and reactor operators.
,
<
'..
- Denotes those present at the exit interview.
'l
'
l 2.
Review cf Plant Operations
'
A review of plant operations was conducted, including examination of O
selected logs and records, operating orders, and " reportable occurrence" reports. During a. tour of the plant, observations were made regarding the status of alams, monitoring instruments, seismic restraints, valve positions, piping system conditions, housekeeping, control room manning, and other plant conditions.
The status of the plant was dis-cussed with licensed operatcrs on duty.
Examination of records related to removing equipment frbilfervice indicated that the "A" LP Injection valve, the "B" Nuclear Service Cooling Water Pump, and one of the two HP Injection recirculation Valves were removed from service for breaker inspection on April 6, 1978. Control Room and/or Shift Supervisor Logs documented appro-priate valve lineups and surveillance of redundant components with the exception of the redundant HP Injection recirculation valve.
Operating personnel who had removed the recirculation valve from service expressed a belief that the required surveillance had been performed, but no documentation of this testing was available. The licensee was advised during the exit interview that the lack of documentation was in noncompliance with Technical, Specifications requirenents (Paragraph 6.10.1.d).
., -
%
.
.
-_
_
__
_
.
.
.
.
.
-2-
,
.
3.
Review of Audit The inspector examined selected licensee records described below and discussed the content.of these records with various licensee personnel. Based on these examinations and discussions, the inspector verified that the activities of the piant Review Committee (PRC) and the Management Safety Review Committee (MSRC) were con-ducted in conformance with regulatory requirements.
PRC Meeting Minutes Numbers 388 through 488.
MSRC Meeting Minutes Numbers 59 through 63.
'
Suninary of MSRC Audit Activity, September 1977 through March 1978.
Audit Reports Nunbers 163 through 184.
No noncompliance items or deviations were identified.
4.
Procedures The inspector examined the following operating, emergency, and O
administrative procedures to verify that they.had been reviewed and approved in accordance with the technical specifications, that changes were made in conformance with the regulatory requirements, and that technical adequacy appeared consistent with the intended mode of operation. The inspector also verified that procedures were implemented in a timely manner and that changes were promul-gated to all holders of controlled copies of the Plant Operating Manual.
AP-22 Reporting of Reportable Occurrences AP-30 Entrance into Carbon Dioxide Protected Areas AP-32 Fire Drills and Fire Fighters' Training Program AP-700 Rancho Seco Personnel Training Program
.
M.6 Pressurizer Relief Valve Removal and Replacement M.20 Make-up and High Pressure Injection Pumps M.105 Charcoal Filter Tray Change Out MT.013 Control of Mechanical Measuring Devices EM-144 Biannual Testing of Protective and Control Relays I-105 Energizing and De-energizing Reactor Protection System D-6 Moderator Dilution
,,a
,. _
D.17 Loss of Control Room O
V
-.
. _.
- .
..
.
..
.
Annunciator Procedures:
I H2X
- CCW Surge TK LVL Hi-Lo H2FSA - Spray P. rop. Channel A Activation
,
Analog Charnel A Test or Module Withdrawn
'
-
P2PSA - R. C. Loop B Press Lo-Lo H2PSB - St. Gen. A Level Hi-Lo H2ES - 125V DC Bus B Trouble H3FPA - Fire Main Isols HRR Closed FPW 011-012
'
H3FPB - DSL DRVN Fire PMP 8HR RES.
SP 201.03B Monthly Surveillance of Plant Fire Pumps and i
Power Supplies SP 203.02A Quarterly High Pressure Injection Loop A, Inspection and Surveillance SP 206.03A Diesel Generator "A" Synchronization Surveillance
,
Test
-
SP 210.02 Main iteam Safety Valves j
No noncompliance items or deviations were identified.
,
5.
Review of Safety Limits, limiting Safety System Settings, and Limiting Conditions for Operation This inspection concludeo a review (commenced during inspection 50-312/78-03) of p.lant operation for compliance with safety limits, limiting safety system settings, and limiting conditions for opera-tion defined in tite Ranche Seco Technical Specifications.
The review included discussions with plant personnel; direct observa -
'
tions of operations and plant conditions; and examination of operating logs, recorder charts, and other pertinent documents.
The following technical specifications requirements were included in the review:
,
Reactivity Control and Power Distribution:
Paragraphs '.1.3.1 and 3.5.2.5.B,C c
Instrumentation: Paragraphs 2.1.2 and 2.3.1 (power-flow-imbalance)
Reactor Coolant System: Paragraphs 3.1.1.1.A,
,
3.1.2.3, and 3.1.9.2 e '
O
!
l l
--
-
. -.
. _.
-.
..
-
. -.
.
-
- _.
_.
. -.
.
.
-4-
.
'
Emergency Core Cooling System:
Paragraphs 3.2.1,
3.2.2, and 3.3.2
)
Plant and Electrical, Power Systems: Paragraph 3.4.2 l
During the review, a question was raised by the inspector concerning the power-flew-imbalance trip point. Calibration records for one se-
,
lected reactor protection channel confirmed that the trip setpoints
'
for 100% flow (4-pump operation) were consistent with Figure 2.3-2 of the Technical Specifications.
The records also confirmed a power trip setpoint of 1.05 times percent flow as measured flow is decreased,
,
consistent with Table 2.3-1.
Figure 2.3-2 shows a " maximum allowable setpoint" of 78% power for the 3-pump operating configuration, based upon a " typical" flow rate of 74.4% (bases, Specification 2.3.1). Reactor coolant flow with 3 pumps was measured by the licensee at 76.8% with 3 operating reactor coolant pumps at the
,
time of the inspection, which resulted in a power trip point (1.05 times flow) of 80.5%. The inspector observed that this was in excess of the setpoint allowed by Figure 2.3-2, although the in-creased flow upon which the setpoint was based appeared to alleviate
,
technical concern about the setpoint.. Licensee representatives l
maintained that Figure 2.3-2 was intended to depict " typical" values (Bases, Specification 2.3-1), and that Table 2.3-1 speci-fied the required trip point.
Paragraph 2.3.1 of the Technical l
Specifications states that. :The reactor protection system trip setting limits...shall be as stated'in Table 2.3-1 and Figure
!
2.3-2."
In view of the apparent disagreement between Table 2.3-1 and Figure 2.3-2 for reduced flow configurations, the inspector stated that the matter would be considered an unresolved item (see Paragraph 8 of this report). This matter was discussed dur-ing the exit interview.
No deviations or noncompliance items (other than the item listed aboveasunresolved)wereidentifici.
.
6.
Licensee Event Followug
,
The circumstances and corrective actions' described in Licensee l
Event Reports submitted since the previous inspection were veri-
!
fied. The details of each event had been' reported to facility
<
'
management and reviewed by the Plant Review Committee. Licensee Event Reports reviewed were as follows:
78-2 Low fluid level in hydraulic snubbers.
z 78-3, Low fluid level in hydraulic snubber.
78-5, Main steam safety valve.setpoint error.
.
-
'w,-,.--
e
+
=
,,..
-
-,m,,,-
..e..-,-.,,,-,-s---
--s
i.v-ga
.
.
..
.
.
..
-5-No noncompliance items or deviations were identified.
7.
Potential Protective System Loss of Ground
'
The inspector reviewed the facility's potential for loss of ground to the reactor protection system (RPS). This potential problem was described in a 10 CFR Part 21 report submitted by B&W on March 7, 1978. The review, which included discussion with facility
representatives and examination of RPS cabinets, showed that con-
'
siderable redundancy in ground connections has been provided.
Facility representatives stated that the Plant Review Comittee had reviewed an investigation of the question, including ground measure-ments taken for the various system components, and determined that adequate assurance against loss of ground existed.
8.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of
-
!
noncompliance, or deviations. An unresolved item identified during the inspection is discussed in Paragraph 5.
9.
Exit Interview
.
The inspectors met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on May 18, 1978.
The inspectors summarized the scope and findings of the inspection.
,
The licensee representative acknowledged the item of noncompliance identified during 'the inspection (Paragraph 2). During discussion of the power-flow-imbalance reactor trip setpoint, the licensee representative stated that the trip point would not be readjusted to 78% for the three-pump configuration until clarification of this I
item has been provided by the NRC. The inspector stated that this matter was considered to be in unresolved item (Paragraphs 5 and 8).
.
O
- -.
-
- -
..
_ _ - -
.
. -.
-.
.
-
-