IR 05000272/1980032
| ML18085B055 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/20/1981 |
| From: | Hill W, Keimig R, Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18085B054 | List: |
| References | |
| 50-272-80-32, 50-311-80-22, NUDOCS 8103190098 | |
| Download: ML18085B055 (19) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION I
50-272/80-32 Report Nos. __
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50-272 Docket No ~~D~P~R-~7~0----------
L i cense No DPR-75
~~~-----------------
Licensee:
Public Service Electric and Gas Company 80 Park Plaza - 15A Newark, New Jersey 07101 *
Faci 1 i ty Name: __
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Inspection At: __
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1980 W. M. Hill, Jr., Resideit Rea_ctor Inspector 50272-800721 50272-800801 50272-800805 50272-800806 50272-800810 50272-800829 50272-800830 50272-800830 50272-800908 50272-800925 50272-800910 50311-800830 50311-800907
' date
- b,/s1 date Inspection Surrmary
Ins ections on November 16 - December 31, 1980 (Combined Re ort Numbers 50-272/80-32 an 0-311/80-22)
.
Unit 1 Areas Inspected: Routine inspections by the resident inspectors of plant operations including tours of the facility; conformance with technical specifica-tions and operating parameters; log and record reviews; review of licensee events; IE Bulletins and Circulars; followup on license events; completion of refueling activities and startup testing; and followup on previous inspection items. The inspection involved 141 inspector-hours by the resident inspector Results:
No items of noncompliance were identifie Unit 2 Areas Inspected: Routine inspections by the resident inspectors of plant startup testing including tours of the facility; license requirements and technical specifications; IE Bulletins and Circulars; followup on licensee events; and, follow-up on previous inspection items. The inspections involved 55 inspector-hours by the resident NRC inspector Results:
No items of noncompliance were identifie s1oa1eo o~S
DETAILS 1. Persons Contacted J. Driscoll, Chief Engineer L. Fry, Station Operating Engineer J. Gallagher, Assistant Maintenance Engineer H~ Midura, Manager - Salem Generating Station P. Moeller, Nuclear Licensing Engineer F. Schnarr, Station Operating Engineer R. Silverio, Assistant to the Manager J. Stillman, Station QA Engineer R. Swetnam, Radiation Protection Engineer The inspector also interviewed.and talked with other licensee personnel during the course of the inspections including management, clerical, maintenance, operations, performance and quality assurance personne '.
2. Status of Previous.Inspection Items (Closed)
(Closed)
(Closed)
(Closed)
Unresolved Item (272/79-08-01) Records of SORC procedure review The inspector verified that records of SORC review for the procedures in question are available. In some cases, notably AP-15, Tagging Rules, the original issue was not subject to review by the Committe However, the most recent revision has been reviewe The inspector had no further questions on this ite Unresolved Item (272/80-13-01) Pennanent heat.trace for vent sample pipin By inspection of the piping and electrical panel lGl, the inspector confirmed that pennanent heat tracing is installed and energized on the sample piping. The inspector had no further questions on this ite Unresolved Item (272/80-13-06) Relocation of Pressure transmitters PT-403 and PT-405. The inspector confirmed that the two pressure transmitters had been raised to an elevation above the postulated flood level in containment and recalibrated. The inspector had no further questions on this ite Unresolved Item (272/79-28-05) Time delay settings of safety related circuit breakers. The inspector reviewed records of breaker testing conducted by the station maintenance department for 460 V and 230 V breakers and relay testing conducted by the Relay Department for trip relays associated with 4 KV switchgear. This review indicated that a continuing program of breaker testing is in place and further re-vealed that time delay drifts such as originally identified in IE Bulletin 79-11 have not been experienced in the equipment inst.alled at the Salem Station. Continued application of the program currently established appears to address the concerns identified in the Bulleti The inspector had no further questions on this ite (Closed) Unresolved Item (272/80-20-03} Seismic design of masonry block walls *
Concerns identified in followup of IE Bulletin 80-11 have been satis-factorily resolved. This item is addressed in detail in paragraph *
(Closed)*. Deviation (272/79-33-01) Seismic design of check valve 1DR7 *. As described in the licensee's response dated April 7, 1980, the valve was prototype tested to confirm adequate seismic capabilitie Additionally, the inspector verified, through interviews with per-sonnel, that a complete system review had been conducted to confirm
- that "specification breaks" were consistent with design and that components met the design requirements of the system applicatio The inspector had no further questions on this ite (Closed) Follow Item (311/80-17-03) Fonnalized corporate review of Salem LER' The inspector reviewed procedures EPDN 16.8, Salem Projects Manual 6.3.1, and Salem Projects Manual (draft) 3.1.4. In aggregate, the above documents provide a mechanism for routing significant *
event reports to cognizant personnel. A tracking mechanism withi the Engineering Department to confinn completion of review h~s been established through computer-based project control procedures. The inspector had no further questions on this ite (Closed) Unresolved Item (272/80-07-01) LOCA procedures for inadequate core cooling and SI pump shutoff hea The inspector reviewed EII-4.4, Loss of Coolant, Rev 10, dated September 19, 1980 to confirm that operator considerations/alternatives for inadequate core cooling have been included. Relative to SI pump shutoff head, the licensee refers to FSAR Section 6.2.2.1, which states, "This (152o:psi)
.
limitation on discharge pressure does not significantly reduce the effectiveness of the safety injection pumps since any break of sufficient size to require safety injection will reduce the coolant pressure below 1500 psi Additionally, the licensee considers the fact that the mini-flow lines back to the RWST 'do not isolate on safety injection.* Accordingly, no additional guidance is necessary in the emergency procedures. The inspector had no further questions on this ite (Closed) Unresolved Item (272/80-13-05) Supplemental report for LER 80-09/03 On December 9, 1980, the licensee submitted LER 80-09/03X-1 which describes the repairs made to the spent fuel pool weld joints. A number of joints were affixed with half-round channels welded in place. Effectiveness of this repair was demonstrated by a leak rate reduced to 26 ml/hr (average) as measured at the channel tell-tale drain The inspector had no further questions on this ite (Closed)
Noncompliance (272/79-32-05) Unauthorized tag-out of three diesel generators. The inspector reviewed corrective actions detailed in licensee correspondence dated March 6, 1980. The following were confinne4; restoring final tagging authori~y to Shift Supervisors only, revision of AP-15, and review of OM-15 with all supervisor Continuing observations indicated that the above actions have been effective in strengthening controls over tagging activities. The inspector had no further question *
- ' I
(Closed) Unresolved Item (272/80-23-02) Reactor Control Rod high drag resistance readin The inspector reviewed the data recorded during the perfonnance of the Rod Drop Time Measurement test The actual performance of the test was monitored by the inspector for several rods. The "drop times" were within the 1 imits pre-SITE
. scribed by the Technical Specifications. The inspector paid particular attention to the drop times of the.control rods that had experienced high resistance 11drag 11 readings during earlier testing. These times were not only within the Technical Speci-fication, but were consistent with the average drop times of the other control rod The inspector had no further questions on this ite. Shift Logs and Operating Records.
- a. The inspector reviewed the following plant procedures to determine the licensee established requirements in this area in preparation for a review of selected logs and record AP-5, Operating Practices, Revision 10, May 21, 1980; AP-6, Operational Incidents; Revision 6, February 22, 1979; AP-13, Control of Lifted Leads and Jumpers, Revision 4, February 11, 1980; Operations Directive Man~al; and, AP-15, Safety Tagging Program, Revision 1, November 21, 198 The inspector had no questions in this are b. Shift logs and operating records were reviewed to verify that:
Control room log sheet entries are filled out and initialled; Auxiliary log sheets are filled out and initialled; Log entries involving abnonnal conditions provide sufficient detail to communicate equipment status, lockout status, correction and restoration; Log book reviews are being conducted by the staff; Operating orders do not conflict with Technical Specification requirements; Incident reports detail no violation of Technical Specification LCO or reporting requirement; and, Logs and records were maintained 1n accordance with Technical Specifications and the procedures in 3.a abov *
c. The review included the following plant shift logs and operating records as indicated and discussed with licensee pe~sonnel:
Log No. 1 - Control Room Daily Log, November 16 - December 31, 1980 Log No. 6 - Primary Plant Log, November 16 - December 31, i980 Log No. 7 - Secondary Plant Log, November 16 - December 31, 1980 Log No. 8 - Unavailable Equipment Status Log, November 16 - December 31, 1980 Night Orders, November 15 - December 26, 1980 Lifted Lead and Jumper Log - All.active Nonconformance Reports for November 1980 The inspector had no questions relative to logs reviewed during this inspection perio *
4. Plant Tour a. During the course of the inspections, the inspector made observations and conducted multiple tours of plant areas, including the following; (1) Control Room (daily)
(2) Relay Rooms (3) Auxiliary Building (4) Vital Switchgear Rooms (5) Turbine Building (6)
Yard Areas (7) Radwaste Buildin9 (8) Penetration Areas (9) Control Point (10) Site Perimeter (11) Fuel Handling Building (12) Containment
~---
6 The following detenninations were made:
Monitoring instrumentation: The inspector verified that selected instruments were functional and demonstrated parameters within Technical Specificatiofi~limit Valve positions. The inspector verified that selected valves were in the position or condition required by Technical Specifications for the applicable plant mode. This verification included control board indication and field observation of valve position (Charging/
Safety Injection, Auxiliary Feedwater, and Containment Spray Systems).:
Radiation Controls. The inspector verified by observation that
_control point procedures and posting requirements were being followe Plant housekeeping conditions. Observations relative to plant house-keeping identified no unsatisfactory conditions except as noted in paragraph 1 Fluid leak No fluid leaks were observed which had not been identi-fied by station personnel and for which corrective action had not been
. initiated, as necessar Piping vibratio No excessive piping vibrations were observed and no adverse conditions were note Selected pipe hangers and seismic restraints were observed and no adverse conditions were note Equipment tagging. The inspector selected plant components for which valid tagging requests were in effect and verified that the tags were in place and the equipment in' the condition specifie Control room annunciators. Selected lit annunciators were discussed with control room operators to verify that the reasons for them were understood and corrective action, if required, was being take By frequent observation through the inspection, the inspector verified that control room manning requirements of 10 CFR 50.54 (k) and the Technical Specifications were being me In addition, the inspector observed shift turnovers to verify that continuity of system status was maintained. The inspector periodically questioned shift personnel relative to their awareness of plant conditions and knowledge of emergency procedure Release On a sampling basis, the inspector verified that appropriate documentation, sampling, authorization, and monitoring instrumentation, were provided for effluent releases *
Technical Specifications. Through log review and direct observations during tours, the inspector verified compliance with selected Technical Specification Limiting Conditions for Operatio The following para-meters were sampled frequently:
RWST level, BAST level and temperature, containment temperature, boration flow path, shutdown margin, offsite power. * In addition, the inspector conducted periodic visual checks * protective instrumentation and inspection of electrical switchboards to confirm availability of safeguards equiµnen *
Security. During the course of these inspections, observations relative to protected and vital area security were made,' including access controls, boundary integrity, search, escort, and badgin No notable conditions were identifie c. The following acceptance criteria were used for the above items:
Technical Specifications Operation Directives Manual Inspector Judgement The inspector had no questions relative to tours made during this inspectio. Surveillance Testing
a. During this refueling outage, the licensee conducted surveillance testing to comply with the requirements of Technical Specifications 4.6.2.1.c.2 and 4.6.2.2.d. These Specifications require, respectively, verifying a spray additive tank eductor flow rate of 35 + 3.5.. gpm to each containment spray system with the spray pump operating in the recirculation mode, and verifying a NaOH solution flow rate of 7.3 + 0.7 gpm from the spray-additive tank through sample valve 1 CS 61 with the spray additive tank at 2.5.:t. 0.5 psig *.
These tests are done at a five year frequency and were last performed as part of the preoperational test progra The educto~ flow rate test was originally performed using the spray additive tank* filled with pure water as the source of supply to the eductor. The periodic test must be performed using the RWST through a test line as the source of additive water to the educto When per-forming the test in this manner, the flow rate achieved was approximately 57 gp The increase is attributed to the difference in standing head between the two sources of additive water. This was further demonstrated by repeating the test, using the RWST as an additive water source, and throttling the manual valve in the supply line to achieve equivalent spray additive tank pressure at the eductor supply point. The result in this case was a flow rate of approx-imately 35 gp The licensee has modified the surveillance procedure to account for the head difference and*uses 57.:t, 5.7 gpm as the acceptance value. The test was completed on December 12, 198 ;
The drain valve flow rate test demonstrates proper flow from the additive tank to the educto When performed by the licensee, flow rates of approximately 12 gpm using the test conditions prescribed in Technical Specifications were observed *
The licensee sought, and was granted, a change to the Technical Specification on December 11, 198 The change establishes 12 + 3 gpm as the acceptable drain flow rate. The licensee recorded 11.32 gpm when the test was completed on December 12, 198 The inspector had no further questions on the above item *
8 Amendment No. 27 to Facility Operating License DPR-70 (Unit 1) was issued November 28, 1980 and first received on site by the inspector on or about December 17, 1980. The licensee had delayed distribution of the amendment to record copy holders due to an administrative error in the original. At the time of receipt, Unit l was in Mode 3 (Hot Standby).
In addition to other changes, the amendment adds surveillance requirements for. the four safety injection flow path Maximum and minimum flow values are established for certain configurations of both high and low head safety injection pump Following completion of modifications for the ECCS sub-systems that alter the subsystem flow characteristics, these maximum and
- minimum values must be verifie The inspector reviewed known system modifications with licensee representa-tives and concluded that no modifications affecting system flow characteristic have been made since preoperation testing. However, the inspector took the position that, with the new surveillance requirement, operability of the ECCS
. system in accordance with Technical Specification 3.0.4 could only be demonstra-ted by a showing that the preoperational test data met the criteria stated in *
the Specificatio Review of preoperational test data revealed that the measured values for low head safety injection fell 10 gpm short of. the required three loop sum of 463 gp For the high head pumps, the test technique used did not provide direct data which could be used to demonstrate the system flow rat The licensee further noted that the test is normally done with the system in cold shutdown and.the vessel head remove Based on available data for the high head pump, the licensee requested relief from NRC for this surveillance. Jlmendments to the Technical Specifications were approved on December 18 and 19, 1980, and provided the relief requeste The minimum flow criterion was changed from 463 gpm to 453 gp The high head surveillance tests were deferred until the next refueling outag In discussions with licensee management, the inspector. expressed his concern that a change to Technical Specifications was not available to cognizant operations personnel (Shift Supervisors) three weeks after issue, resulting in a Mode change from Cold Shutdown to Hot Standby without full knowledge that required surveillance testing had been completed. The licensee stated that the method of distribution will be modified to ensure timely transmittal
.of an advance copy to operating personne The inspector had no further questions relative to the abov In response to NRC correspondence dated February 23, 1980, relating to LWR primary coolant system pressure isolation valves, the licensee committed to conducting a leak rate testing program of check valves which isolated the high pressure reactor coolant system from low pressure safety injection pipin The requested test program is responsive to concerns r.aised in WASH-1400 relative to potential failure modes due to overpressurized low pressure piping as a result of check valve failures. The licensee identified several configurations in which double series check valves are used to provide* isolatio At the conclusion of this outage, with the plant in Mode 3, the described leak tests were conducte Below are listed the leak rates observe The licensee used an acceptance criterion of 1.0 gpm since that value is given in the Unit 2 Technical Specification *
The observed leak rates were:
Valve(s) *
ll-14SJ17
- lSJlSO llSJS SJ55 13SJ55 14SJ55 11SJ56 12SJ56*
13SJ56 14SJ56 11 and 13SJ43 12 and 14SJ43 11-14SJ14 SJ156 12SJ156 13SJ156 14SJ156 11SJ139 (*)
12SJ139 13SJ139 14SJ139 Leak rate 0.25 0.18.1213 0.182 0.364.8.4 0.1213.0.1.2.243 0.22 gpm gpm gpm gpm gpm gpm gpm gpm gpm gpm gpm gpm gpm (each)
gpm gpm gpm gpm gpm gpm gpm gpm Test pressure in each case was nonttal system pressure (2235 psig, nominal)
or safety injection pump discharge pressure (1520 psig, nominal).
When reduced pressure was used, the leak *rate value was corrected to full system pressure. Valve 11SJ139, denoted by (*) above, leaked in excess of 5 gpm until the disc was freed and the valve exercised several times by alternately reversing fl o In the absence of absolute leak rate limits, the licensee elected not to cool down to repair any of the above valves at this time. Testing will be repeated-during the next outage to ascertain trend Since no system leakage had been identified previously, and since the observed
.leak rates are within the relief capacity of the low pressure systems, the inspector had no questions at this tim * IE Bulletin and Circular Followup a. The IE Bulle~ins discussed below were reviewed to verify that:
Licensee management forwarded copies of the response to the bulletin to appropriate onsite management representative Infonnation diScussed in the licensee's reply was supported by facility records or by visual examination of the facilit Corrective action taken was effected as described in the repl The licensee's reply was prompt and within the time period described in the bulleti The review inciuded discussions with licensee: personnel and observation and review of items discussed in the details 'belo By correspondence dated October 5, 1979, the licensee responded to IE Bulletin 79-21, Temperature Effects on Level Measurements, for both unit The response details setpoint changes required to assure timely start of auxiliary feedwater and procedural changes to alert operators to potential level indication errors. The inspector verified that the stated actions had been implemented at the statio In addition, the inspector reviewed the basis used for setpoint deter-mination on each unit. Salem Unit 2 uses Barton Model 764 Lot 1 Trans-mitters in this application with insulated reference legs. The setpoint is detennined as follows:
0%
minimum trip point 3%
trip point margin 2%
trip channel accuracy 2%
reference leg heatup error 10%
transmitter environmental error 17%
setpoint for low-low level Salem Unit 1 uses Barton Model 764 Lot 2 transmitters in this application without an insulated reference leg*.
The setpoint for Unit 1 is determined as follows:
0%
minimum trip point _
3%
trip point margin 2%
trip channel accuracy 8.6%
reference leg heatup error with margin 3.4%
transmitter environmental error 17%
setpoint for low-low level For each unit, the setpoint has been established as nominally 18%, which is consistent with the respective Technical Specification In each case, a 3% margin is provide The inspector had no further questions on this ite **
c. * By correspondence dated June 13, July 18, and October 20, 1980, the licensee responded for both units to IE Bulletin 80-06, Engineered Safety Features (ESF) Reset Controls, which deals with the potential for re-* *
opening of safeguards-associated valves on reset of a safeguards signal.*
The licensee's responses indicate a potential for this problem on the following valves; MS 18's, FP 147, and SJ 53. Following appropriate modifications, retest is accomplished by performing the periodic surveillance procedure for isolation valve..
The inspector reviewed surveillance procedures, on Unit 1, for Phase A and B isolation. These procedures now include a verification step following reset of the isolation signal to confirm that none of the listed valves
. move from the post-accident posftion. All valves listed in the Bulletin response, as well as the remaining isolation valves, were included in this test *.
Similar review of Unit 2 test data will be conducted during a subsequent inspection prior to restart (311/80-22-01). By correspondence dated March 19, and April 11, 1980, the licensee responded for both units to IE Bulletin 80-03, Loss of Charcoal From Standard Type II, 2 Inch, Tray Adsorber Cells. The licensee describes the construction of adsorber cells used at Sal em which are different from. the 11probl em 11 cells, and* details the results of inspections conducted. The conclusion is drawn that potential for loss of charcoal incidental to storage, handling, and use is not presen As further verification, the inspector reviewed filter testing data generated by the vendor relative to installed HEPA and charcoal units in Unit 1. Data for adsorbers in the Auxiliary Building, Control Area, and Fuel Handling Building ventilation systems were reviewe In each case, filter efficiencies greater than 99.76% at flQw rates consistent with Technical Specifications were achieve The inspector had no further questions on this ite By correspondence dated July 2, November 4, December.4, and December 10, 1980, the licensee responded for both units, to IE Bulletin 80-11, Masonry Wall Desig The licensee identified eleven walls in the two units which could adversely affect safety-related systems or structures through failure during seismic conditions. The licensee's findings are further outlined in Licensee Event Report 80-56/0lP for Unit 1, dated December 1, 198 The licensee identifies in detail in the December 4 (Unit 1) and December 10, 1980 (Unit 2) submittals the walls in question, the proposed solutions in terms of reinforcement, and the analysis techniques used to demonstrate acceptable performance during seismic events. It is noted that most of the masonry walls in question were not intended to be permanent plant walls and were not originally designed with seismic considerations. Of particular concern is tne 11A 11 building wall and the west truck bay, which are permanent structures, not seismically designed, and which are adjacent to the Unit 1 Auxiliary Feedwater Storage Tank and Refueling Water Storage Tan ~-----
The wall construction, as-built conditions, proposed reinforcement, and the acceptance criteria used for reanalysis were discussed with the staff at meetings held November 20, and December 5, 198 At those meetings,*
the staff noted that acceptance criteria being used by the licensee were different from those proposed in draft by the staff. Further review of the analysis by the NRC staff is continuing. It was determined, however, that sufficient agreement existed to permit restart of Unit 1 once the modifications identified by the licensee**s analysis had been complete These understandings were documented in correspondence to the licensee from NRC Region I dated December 9, 1980. Since the licensee's final response to the Bulletfn was later than the 180 days authorized, the December 9,. 1980 letter also confinned that Unit 1 restart was contfngent on making the submittal and completion of identified modification The inspector verified, based on a sampling inspection using the licensee's
~ubmittal as reference, that all identified modifications to plant masonry walls had been completed prior to critical operation of Unit L No failure to complete the identified modifications was identifie No additional masonry walls beyond those identified by the licensee were foun The inspector had no further questions on this item at this time. Further review of licensee. analysis by NRC (NRR) staff may result in future
- modification As stated in the licensee's submittals, current modifica-tions do not preclude subsequent changes that may be require By correspondence dated September 23, and September 29, 1980, the licensee responded for both units to IE Bulletin 80-18, Maintenance of Adequate Minimum Flow Through Centrifugal Charging Pumps Following Secondary Side High Energy Line Ruptur The licensee acknowledges the potential problem with dead head operation of Charging Pumps in a steam break situatio To retain the guidelines for termination criteria already establish,ed, the following steps have been incorporated into Emergency Instruction I-4.6, Loss of Secondary Coolant; if pressure exceeds 2235 psig, open the PORV, if pressure remains above 2235, reset the safety injection and shutdown the charging pumps, if pressure subsequently falls to 2000 psig, start both charging pump These steps are short tenn considerations to address the item of concern Long term solutions are still being evaluated. This will remain an open item for inspection once a long term solution in terms of hardware or procedure is developed and implemented (272/80-32-01). The IE Circulars discussed below were reviewed to verify that they had been reviewed for applicability by cognizant management, and appropriate action initiate The inspection included discussions with licensee personnel, observation, and review of items discussed in the details belo Protection From Toxic Gas Hazards. Licensee review of this Ci-rcular is documented in a memorandum dated October 23, 198 Reference. is made to a 1 icensing submittal dated July 1, 1980 and the conclusion is made that conformance to SRP sections 2.2.1, 2.2.2, 2.2.3, and 6.4 as well as Regulatory Guides 1.78 and 1.95 exists with no f.urther modification necessar *
80-04 Securing of Threaded Locking Devices on Safety-Related Equipmen.
The 1icensee 1 s evaluation, dated August 14, 1980, *reviews all
- *
items listed in the Circular. Only the Limitorque operators.
were evaluated as requiring verification. A program to confirm staking of operators is being implemented. A number of these have previously been verifie Problems With HPCI Turbine Oil System. This Circular applies to BWR HPCI systems, however, a review was conducted to evaluate the relevance to steam-driven Auxiliary Feedwater Pump In view of turbine stop valve independence of the oil system (air operated),
no concern was identifie Problems With Plant Internal Communication~ Systems. The licensee's evaluation, dated May 1, 1980, concludes that all considerations in this Circular are adequately addressed at the Salem Sit Emergency Diesel Generator Lube Oil Cooler Failures. The licensee's evaluation, dated August 20, 1980, concludes that inhibitor practices at Salem are consistent with manufacturer's recommendations. Additiona operational considerations are recommended to the station and are currently under revie Valve-Shaft-To-Actuator Key May Fall Out of Place When Mounted Below Horizontal Axi The licensee's evaluation, dated May 21, 1980, concludes that this concern is not applicable to Salem due to an opening and shoulder design which precludes fall out of the ke Radioactive Contamination of Plant Demineralized Water System and Resultant Internal Contamination of Personnel. The licensee's evaluation, SGS/M-SE-066, dated October 10, 1980, concludes, based on demineralized water system design and AP-24 controls, that this sequence of events is unlikel An additional recommendation is made to post signs at si:r1ks within the controlled access area cautioning against use for consumptio At the time of this in-spection, no such signs were identified. This is an open item..
(272/80-3~-02).
80-15 Loss of Reactor Coolant Pump Cooling and Natural Circulation Cool-dow The licensee's evaluation, SGS/M-SE-071, dated November 5, 1980, outlines the design differences between Byron-Jackson and Westinghouse pumps and concludes that this sequence of events is unlikely. A review of operating directives indicated that details of the event had been disseminated to all licensed personnel. A specific natural circulation cooldown procedure with an applicable operating envelope was not available for review prior to the con-clusion of this inspection. This an open item (272/80-32-03) *
- 80-17 Fuel Pin Damage Due to Water Jet from Baffle Plate Corne This Circular was evaluated on a generic basis by Westinghous A letter, dated November 3, 1980, details experience in pin damage and the apparent cause. The design configuration common to plants having experienced this fuel rod damage is not used at Salem and no concern is warrante CFR 50. 59 Safety Evaluat'ions for Changes to Radioactive Waste Treatment Systems. The licensee's evaluations of this Circular, dated September 10, 1980 and December 18, 1980, forward the in-formation to be included in station design change procedures, and state that the information will be included in a revision to Engineering and Construction Department directive EDD-1. A licensee representative :stated that EDD-1 will be revised by March 31, 1981. Confinriation that EDD-1 contains the guidance provided in this Circular is an* open item (272/80-32-04).
80-21 Regulation of Refueling Crews. The licensee's evaluation of this item, dated November 6, 198U, concludes that all considerations in theCircular have been covered in station procedures. The inspector has identified no failure to confonn to the guidance provided during observations of refueling operation Confirmation of Employee Qualifications. The licensee's evaluation dated December 24, 1980, concludes that, based on a review of PSE&G and contractor practices, the concerns raised by the Circular are adequately addresse The inspector had no further questions relative to Circulars reviewe.
Licensee Event Reports (LER's) In Office Review of Licensee Event Reports The inspector reviewed LERs submitted to the NRC:RI office to verify that details of the event were clearly reported, including the accuracy of the description of cause and adequacy of corrective action. The inspector detennined whether further infonnation was required from the licensee, whether generic implications were involved, and whether the event warranted onsite followu The following LERs were reviewed:
Unit -1
80-42/03L Missed Surveillance - Fire Pump Diesel Starting Battery 80-43/03L Refueling Water Storage Tank Level Less Than 350,000 Gallons 80-44/03L Containment Air Lock Door Elevation 100' Inoperable 80-45/03L Air Particle Detector Inoperable due to Seized Pump 80-46/03L Failed Steam Flow/Feed Flow Mismatch Circuit - No. 12 Steam Generator
- --
80-47/03L Failure to post Fire Watch Within One Hour With Penetration Fire Barrier Inoperable
80-48/03L Calculated Quadrant Power Tilt Ratio Exceeded 1.02 80-49/03L No. 12 Charging Pump Inoperable Due to Lube Oil Cooler Leak 80-50/03L Inoperable Air Particulate Monitor Due to Failed Detector 80-51/03L Fuel Handling Area Ventilation System Inoperable 80-52/03L Missed Surveillance - In Service Inspection of Valves Unit 2
80-29/03L Pressurizer Overpressure Protection System Channel Failure 80-30/03L Both Safety Injection Pumps Operable With RCS Temperature Less Than 312°F
See paragraph (b).
b. Onsite Licensee Event Followup For those LERs selected for onsite followup (denoted by asterisks in detail Paragraph 7.a}, the inspector verified that the reporting requirements of Technical Specifications and Regulatory Guide 1.16 had been met, that appro-priate corrective action had been taken, that the event was reviewed by the licensee as required by AP-4, 6, and 7, and that continued operation of the facility was conducted in accordance with Technical Specification limit The following findings relate to the LERs reviewed on site:
Unit 1
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80-42/03L The inspector reviewed inspection order documentation for quarterly checks to confirm that a sign-off to document visual inspection has been include /03L This event is reviewed in detail in NRC Inspection Report 50-272/80-2 /03L The inspector confirmed that new signs are being posted on plant fire doors. Additionally, signs have been procured for placement at fire barrier penetration areas to caution personnel to contact shift supervision prior to breach of a barrier. These actions *should improve assurance that open fire barriers are covered by appropriate compensatory measure /03L The inspector confinned that ordering information for v-belts has been modified to require matched sets. This action should prevent recurrence of this even /03L The inspector confirmed that Operating Instruction OI-I-3.2, has been modified by permanent on-the-spot change P-11 to require verification that valve testing in accordance with Technical Specification 4.0.5 has been completed prior to mode chang *
Unit 2 80-29/03L The inspector verified that Design Change request 2SC0397 has been initiated to provide improved limit switches for POPS valve c. The following Licensee Event Reports detail events which required corrective action pursuant to the license or Technical Specifications:
50-272/80-42/03L 50-272/80-43/03L 50-272/80-47/03L 50-272/80-52/03L 50-311/80-30/03L The inspector had no further questions relative to the above LER.
Review of Periodic and Special Reports Updn receipt, periodic and special reports submitted by the licensee pursuant to Technical Specifications 6.9.l and 6.9.2 are reviewed by the inspecto This review includes the following considerations:
The report includes the infonnation required to be reported by NRC requirements; Test results and/or supporting information are consistent with design predictions and performance specifications; Planned corrective action is adequate for resolution of identified problems; and; Determination whether any infonnation in the report should be classified as an abnormal occurrenc *
Within the scope of the above, the following periodic reports were reviewed by the inspector:
Unit 1 Monthly Operating Report - August 1980 through November 1980 Unit 2*Monthly Operating Report - April 1980.through November 1980 The inspector noted that Unit 2 reports did not include refueling schedule in-formatio This has been corrected as of the November 1980 report. It was further noted that significant maintenance on Unit 2 was not being included in the monthly reports. The licensee stated that this was an oversight in accumu-lating the listing from the computer records of Work Orders. The licensee stated that all-Unit 2 maintenance not previously reported will be included in the next report. This will be confirmed by routine inspection *
The inspector had no further questions relative to the abov * Refueling
Unit 1 entered Mode 5 on November 25, 1980 with the bolting of the reactor pressure vessel hea Heat up to Mode 3 was completed on December 13, 1980 following resolution of the masonry wall problem addressed in paragraph 6 of
. this report. Criticality was achieved on December 21, 1980 and the generator first synchronized on December 2 On December 26-27, three reactor trips occurred as a result of low steam generator level. The first was due to loss of feedwater pump suction pressure as a result of higtf strainer differential pressure. The other two tri.ps were caused ~Y loss of level control while feeding steam generators under low flow conditions. The low level trips we~e increased to 18 % during this outag Coupled with operator inexperience in *this operating mode, these trips from low power do not raise a safety concern. At the end of the report period, power ascension with post-start-up testing was in progress. Reactor power was at 45 %.
Inspection activities during this period included sampling verification that
. pre-requisites for Mode changes were met, that system operability was verified through appropriate and timely surveillance testing, that system lineups were correct to support system operability and that station procedures were followed in conducting the startu In addition, testing and documentation of testing were observed to confirm that the startup test program was being conducted in accordance with procedures and commitment Appropriate test records of conduct of testing were inspected for the following:
'i Initial criticality of Cycle III core ARO Flux mapping Rod Drop Time Measurement,- Hot Full Flow
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An early flux map, made at approximately 4 % power, indicated a core flux tilt of approximately 6-7 %.
Subsequent maps at 25 % and 45 % power continued to show a 3-4 % tilt in core flux distribution. At the conclusion of this inspection, power increases and flux mapping at 20 % thermal power intervals were being con-ducted. Continuing evaluation by the licensee and Westinghouse indicated that a design tilt of 2-2.5 % was expected. Analysis will continue during power escalation to confirm that core thennal limits are not approached. This item is inspected as part of the routine progra. Management Controls a~
By correspondence dated October 29, 1979, theflicensee responded to an NRC letter~ dated September 21, 1979 on the subject; Multiple*Equipment Failures 11\\
and Surveillance Testing Error Items of concern included apparent complacency of operators relative to surveillance testing, multiple equipment failures, and failures of common components (e.g. solenoid valves) *
The licensee relies on station procedures and training of operators and technicians. *Additional guidance was provided to all licensed operators and department heads in a memorandum dated June 24, 1980. This memorandum details previous events and further emphasises the operator's responsi-bilities. Department.heads are further encouraged to pursue system problems which could potentially cause safety system challenge The inspector also noted that the control room status panel design aids operators in detennining whether proposed testing would actuate any safety systems. It was further observed that the frequency of safety system challenges has tlecreased markedly since initial Unit 1 operatio The inspector.had no further questions relative to the licensee's response or the actions taken to address this concer * The licensee requested and received NRC approval to effect an organization*
change at the co.rperate *leve The change separates fossil and nuclear production at the level of general manage All quality assurance responsi-
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bility has been separated from production line responsibility. The General 1~ *
Manager - Corporate Quality Assurance reports directly to the Senior Vice-President - Energy Supply and Engineering. The change has been reflected in an admendment to the Technical Specifications. The inspector had no question relative to the reorganization or the qualification of the incumbent. Fire Protection The inspector observed the post installation pre-operational check of a newly installed HALON fire suppression system for Unit 1. The system discharges a fire suppression medium (HALON) into the electrical relay room upon manual or automatic initiation.* The* HALON primary and secondary systems are identical. The system was tested by remote manual actuation (electrical). During the testing observed, the primary system tripped, but HALON was intentionally prevented from dischargin The secondary system failed to trip. Investigation revealed a burned resistor in the internal circuitry. The secondary system was tripped locally, and the system discharged as intended. Subsequently, the defective resistor was replace Full operational status is awaiting modifications to improve room air circulation following actuation. This will reduce HALON stratification in the roo Discussion with licensee personnel indicates surveillance/preventative maintenance procedures will be performed to insure continued system operability. The inspec-tor had no further question regarding the fire protection syste. Design Changes and Modifications The licensee perfonned modifications to systems inside containment during this outage. The inspector reviewed some unit one design change packages and, on a sampling basis, determined the field installations to be consistent with the design chang The packages reviewed consisted of the following:
- Reactor Vessel Head Remote Operating Vent (EC 0545)
RHR Sump Modification (EC 0753)
-- * Hot Shutdown Panel/Pressurizer and Steam Generator Level Indicators (EC 0429)
Post LOCA Containment Sampling Capability (EC 0540A)
Each package reviewed was not necessarily the entire modification but, in some cases, was issued to accomplish work inside containment during this outag The inspector had no questions relative to these design change packages or field installations/modification ~ Maintenance (housekeeping)
The inspector continually monitored housekeeping activities during periods of high maintenance activity during this refueling outag In view of the heavy maintenance activity, the inspector found housekeeping generally acceptabl As a notable exception, a number of tne electrical and instrument panels, particularly in the relay rooms, had an accumulation of dust, dirt, and debris deposited on the bottom. The custodians who normally perfonned housekeeping activities were not allowed inside these panels unless specifically directe The licensee stated that the panels would be inspected and cleaned as necessar This is an unresolved item (272/80-32-05).
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Unresolved Items Areas for which more infonnation is required to determine acceptability are considered unresolved. Unresolved items are contained in Paragraph 6 and 13 of this repor. Exit Interview At periodic intervals during the course of this inspection, meetings were held with senior facility management to discuss inspection scope and findings.