IR 05000387/1980032
| ML17139A080 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 01/21/1981 |
| From: | Gallo R, Mccabe E, Rhoads G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17139A076 | List: |
| References | |
| 50-387-80-32, NUDOCS 8104100609 | |
| Download: ML17139A080 (27) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I II P
"
~50-38 Docket No. 50-387 License No..CPPR-101 Priority Category Licensee:
Penns lvania Power 8 L't 2 North Ninth Street Allentown Penns 1 vania 18101 Facility Name:
Sus uehanna Steam Electric Station Inspection at:
Salem Township, Pennsylvania Inspection conducted:
November 10, 1980 - January 2,
1981 Inspectors:
a endor ess en nspector RA oa s, esl en nspec or tE 8~
d te signed 8-8j ate signed Approved by:
e
..-
c a e, le, ea or sectsSection II, ROSNS Branch date signed 1l >l I p(
date signed Ins ection Summar:
~li
'0
3 30, 1980 -
I 2, 1981 tR 0 t II
. 50-383/80-323
~A'
td.
R tl I
P tl tptp ld I
0 tll:I'.0 P
tl test review, preoperational test witnessing, operating staff training, and bulletin and circular followup.
The inspectors also performed plant tours and reviewed licensee actions on previously identified items.
The inspection involved 163 inspector-hours during both regular and backshift periods by the NRC resident in-spectors.
Results:
Of the five areas inspected, one item of noncompliance was identified,
~Fat1ure to hav'e" pr'oper1y approv.ed test procedure change's, Paragraph 1.e.)
Region I Form 12 (Rev.
Apr il 77)
8 1 0 4 100 (gag'
DETAILS 1.
Persons Contacted Penns lvania'Power and'Li ht Com an L.
R.
T.
D.
D.
F.
E.
E.
J.
J.
C.
H.
W.
L.
D.
R.
Adams, Plant Supervisor of Operations Bryam, Plant Supervisor of Maintenance Clymer, Site gAE Cassel, Senior Project Engineer Dunn, Resident Engineer Eisenhuth, Senior Compliance Engineer Figard, Assistant ISG Supervisor Gorski, Plant guality. Supervisor Graham, Plant Assistant Green, Operations guality Assurance Supervisor Jaffee, ISG gC Supervisor Keiser, Superintendent of Plant Lowthert, Plant Training Supervisor O'eill, Plant Technical Supervisor Thompson, Assistant Superintendent of Plant Webster, ISG Supervisor The inspectors also interviewed other PPSL employees, as well as employees of Bechtel, and General Electric Company.
2.
Licensee'Action on'Previousl Identified Items a.
(Closed)
Unresolved Item (387/79-31-04)
Connector External Dama e
e licensee respon e
o e
s concerns sn ua sty ssurance Action Request No. S-80-2 dated September 23, 1980.
The licensee stated that three specific connectors identified by the NRC had been inspected and repai,red'n accordance with General Electric Field Disposition Instruc-tion (FDI) - WJGO.
The licensee stated that the cause of the damage was determined to be improper or, excessive handling.
The licensee stated that FDI - WJGO was a comprehensive inspection program of Unit No.
1 cable con-nectors and that any.noted damage was repaired.
I The licensee further stated that it had recently implemented a Security Badge System to control access to the upper relay and cable spreading room, lower relay rooms, computer room, the lower relay room and the
- control room.
The inspector verified the implementation of the security badging system'or these areas.
No.additional cable damage has been identified by the NRC inspectors.
The inspector had no further questions on this matter at this tim b.
(Closed)
Unresol ved Item (387/79-40-05; 388/79-21-03)
Diesel Generator
g The licensee responded to FSAR guestion 040.65-x10 in FSAR Revision 15.
The licensee's response states that the modifications made on the diesel engines consisted of replacement of certain existing components with similar, improved components.
The licensee's response states that the replacing of these components was to eliminate long term wear problems with the rocker arm assembly and cracking problems of the air intake spring.
The licensee.
response states that major modifications were not made and therefore prototype qualification retesting is not required.
The response further states that the diesel engines will be subjected to site acceptance testing per IEEE 387-1972.
The response to the FSAR question is being reviewed by NRR as part of its routine safety evaluation review.
The inspector had no further questions on this matter at this time.
c.
(Closed)
Unresolved Item (387/80-01-03)
Design Interface.
The licensee responded to the NRC's specific concern, regarding RHR valve F020, in Bechtel guality Action Request (JAR) F-605 dated June 17, 1980.
The licensee's response states that meetings were held between General Electric and Bechtel on March 5, 1974.to review the RHR PAID M-151 as part of the NSSS interface review.
The inspector reviewed documentation ref-erenced in JAR F-605 and found that the valve in question had been dis-cussed at the meeting but no concern was raised regarding the valve size.
The licensee's response further states that Bechtel's flow calculation M151-26 determined that the head spray line containing valve F020 would meet the design flow of 1000 GPM.
The inspector reviewed FSAR drawings 5.4 - 14a and 14b to verify that maximum flow through F020 was 1000 GPM.
The inspector reviewed the following documentation Bechtel letter to G.E.
BLG-292 dated March 8, 1974.
Bechtel letter to G.E.
BLG-316 dated March 22, 1974.
G.E.
Drawings 761E29A Revision 7; 761E232A Revision The inspector also reviewed the licensee's response to IE Circular 79-11 "Design/Construction Interface Problem."
The results of that review are documented in paragraph 7.b.(1) of thjs report.
The inspector had no further questions on this matter at this time.
d.
(Open) Unresolved Item (387/80-14-05)
Review of Audit Reports.
The inspector reviewed the status of Site Audit No. 80.
The Resident NQA Engineer issued a
memo dated December 2,
1980 reassigning the remaining open items from the Superintendent of Plant to the Quality Control Super-viso The Quality Control Supervisor stated that Quality Control Pro-cedures (QCP's)
had been submitted to the Manager NQA and that the QCP's should resolve the reamining Audit No. 80 findings.
The inspector requested some verification that the QCP's did address the audit findings.
This item remains open pending review of that verification.
e.
(Open) Inspector Follow-up Item (387/80-14-09)
250 VDC Battery Pre-opera iona
.
es
.
evision The inspector had witnessed a portion of the battery service test and noted a Nonconformance Report (NCR-80-117)
had been issued against an un-calibrated thermometer, designated EM 134, being used to take temperature readings during the test.
The inspector stated he would review the dis-position of the NCR and the test results.
On November 20, 1980 the in-spector reviewed the disposition of NCR 80-117 and the Certification of Calibration Number,17532-0095 from American Electronic Laboratories, Inc.
dated August 21, 1980 stating the thermometer had passed the calibration check satisfactorily.
On December 23, 1980 the inspector reviewed the test results for P88. 1 Revision 1,
and noted the Superintendent of Plant had not yet approved the test results.
This item remains open pending NRC review of the approved P88.1 test results.
During this review the inspector determined that Technical Change Notices (TCN's) were written on two revisions of Form AD7.7-4 (Revision 0 and Revision 1).
Revision 0 of the Form required the Superintendent of Plant to sign block
signifying he approved the TCN, but Revision 1 had no approval signature requirement for the Superintendent.
The inspector then reviewed Startup Administrative Manual Procedure AD7.7 Revision
and 1 titled Preoperational/Acceptance Test Implementation.
Revision
was in effect from January 23, 1978 to May 1, 1980.
Form AD7.7-4 Revision 0 was an enclosure to this revision.
Revision 1 of AD7.7 went into effect on May 1, 1980 and has been in effect since that date.
Revision 1 to Form AD7.7-4 was an enclosure to this form.
The inspector noted that Section 5.3.2 to the procedure
"TCN Review and Approval" had been changed in Revision 1 to delete the requirement for the Superinten-.
dent to approve TCN's to Preoperational Tests stating that the Test Re-view Board (TRB) had the responsibility for reviewing and approving TCN's.
The inspector reviewed FSAR Subsection 14.2.4.3 titled "Procedure Modi-fication" which states, in part,
"Review"and approval requirements
.-
for TCN's are the same as for the original procedure as described in FSAR Subsection 14.2.3.2."
FSAR Subsection 14.2.3.2 states, in part, that the TRB is responsible for recommending approval of test procedures, and that upon completion of review and inclusion of required changes preoperational tests can be submitted for approval by the Superintendent of Plant.
On December 29, 1980 the inspector reviewed TCN's written for Preoperational Test P5. 1C Revision 1 "ESS 480 Volt Motor Control Centers and Auxiliaries."
TCN numbers 001 and 002 to this test were reviewed and approved by the TRB on September 4,
1980.
The results of this test were then approved by
. the Superintendent of Plant on September 4,
1980 without him approving the TCN's.
The inspector informed the Operations guality Assurance Engineer that this was contrary to the requirements of 10 CFR 50 Appendix-B, Criterion VI and was considered a violation.
This matter was discussed with the Superintendent of Plant at an exit interview on January 5,
1981.
(387/80-32-01)
(Open)
Item of Noncompliance (387/80-14-10) Identification and Disposition of Nonconforming Items.
By letter to the NRC dated September 10, 1980 the licensee responded to the Item of Noncompliance.
The inspector discussed this response with licensee representatives and an acknowledgement letter, dated November 10, 1980, was sent to the licensee.
The NRC position is that the licensee must document nonconforming conditions identified after turnover from Bechtel to the licensee.
The licensee revised Startup Administrative Procedure AD6.6 Revision 5 "Startup Work Request" to require the ISG Co-ordinator to forward a copy of Startup Work Request to PPSL guality Control for nonconformance evaluatio The licensee is in the process of revising guality Assurance Procedure SP-11 "Control of Nonconformances" to accomodate contractor documentation of nonconforming conditions in lieu of PPSL documentation of the non-conforming condition.
The inspector stated that the licensee must incor-porate some method of obtaining corrective action from the appropriate contractor and must use this information in trend analysis of the contractor.
This item remains open pending review of the licensee's approved pro-cedures.
g.
(Open)
Unresolved Item (387/80-20-04)
Alarm Testin
.
'
The inspector discussed alarm testing with licensee personnel to.determine-if present alarm check procedures verified all alarm conditions and set-points.
The inspector reviewed the following procedures:
MT-RC-001, Revision 0 Electromechanical and Static Relays MT-RC-002, Revision 0 Electrical Transducers or Indicating Meters IC-DC-400, Revision 0 Switch/Bistable Calibration/Calibration Check Procedure IC-LC-001, Revision 1 Instrument Loop Calibration Check Procedure Startup Administrative Procedures AD6.5 - Revision 3 Control of Initial Instrument Calibration and Analog Loop Test AD7.5 Revision 8 Preoperational/Acceptance Test Procedure Format and Content Startup Technical Procedure TP1.9 Revision 3 Digital Control Scheme Testing J-701 Revision 13 Instrument Index Setpoint Report The inspector could not determine if all alarm conditions and setpoints were fully tested.
Loop and instrument testing and alarm annunciator testing are distinct test functions and preoperational test procedures do not verify all alarms.
The inspector discussed the above with licensee management.
The inspector determined that no Chapter 14.2. 12. 1 FSAR Change Request had been submitted and the Test Review Board had not reviewed TP1.9 as of January 2,
198 The inspector expressed his concern regarding the noted discrepancies between current site practices and FSAR commitments.
This item remains open pending further review by the NRC.
(Open) Inspector Follow-u Item 387 80-20-07 Control Rod Dr've cram ssc arge nstrument Vo ume.
During a review of the CRD Hydraulics System the inspector had noted some discrepancies between the FSAR Section 4.6. 1. 1.2.4.2.5 description of the scram discharge instrument volume and the as-built system.
The inspector reviewed this section of the FSAR during this inspection period in con-junction with the review of the CRD Hydraulic Preoperational Test.
The inspector noted that no changes had been made to this FSAR section.
This item remains open pending review of the change to the FSAR more accu-rately describing as-built system.
(Open) Inspector Follow-u Item 387 80-24-01 Reacto ys em reo erationa Test Review.
The inspector had reviewed the procedure and noted that Section 7.3.10(6)
listed three light emitting diodes (LEOs) which were supposed to energize within certain time constraints.
The time constraints were given in thou-sands of a second.
A note under this section stated that it was not necessary to verify the time shown down to the precision given, but did not specify any other precision necessary.
The inspector had questioned what the acceptable precision was.
On November 14, 1980 the inspector re-viewed a memorandum from the Startup Test Engineer dated October 22, 1980 which stated that a change to the note would be made to state a precision of tenths of a second for acceptance criteria.
The inspector also discussed the testing of an internal cooling fan located inside the rod status display.
The testing was performed in section 7.3. 1, but did not test setpoints for the automatic starting and stopping of the fan.
The inspector questioned whether this fan had design specifications, and whether it was installed to meet the temperature requirements stated in FSAR 3. 11 for Nuclear Steam Supply System Supplied Equipment.
The inspector stated if there was a temperature requirement then this setpoint should be tested.
This item will remain open pending review of the change to the timing pre-cision and a resolution of the fan setpoin j. (0 en)
Ins ector Follow-u Item 387/80-24-05)
Residual Heat Removal RHR Preo erationa Test P 9. 1 Revision Dra t The inspector had reviewed this test and reported a number of discrepancies between the test and the FSAR commitments.
On November 17, 1980 the in-spector discussed these comments with the General Electric Startup Control and Instrumentation Engineer.
During this discussion he stated that the as-built system and the FSAR system description were not in complete agreement.
The following discrepancies were noted:
(1)
FSAR Sections 7.3. l. la.1.6.3 and 7.3.1.1a.1.6.4 discusses the initiation circuits, logic and sequencing of low pressure coolant injection (LPCI) initiation and FSAR Figure 7.3-5 shows a block diagram of this circuitry.
These', sections conflict with each other on what is needed to initiate LPCI mode of RHR.
A review of 791E419WJ, Revision 2 -
RHR elementary diagrams shows the following exist:
(a) Initiation signal which repositions valves in the RHR train responds to a high drywell pressure signal and/or a low reactor water level signal.
(b) Start signal for RHR pumps consists of a low reactor water level signal and/or a high drywell pressure signal with a confirmatory low reactor pressure signal.
(c) LPCI injection valve opens on an initiation signal and a
confirmatory low reactor pressure signal.
The G.E. Startup Control and Instrumentation Engineer stated this is how the system is installed.
(2)
FSAR Section 7.3. 1. 1a. 16.5 states that a 2/3 core coverage reactor vessel water level signal is needed to manually initiate containment spray.
This is no longer an interlock.
(3)
FSAR Section 7.3.1. 1a.6. 11.2 states that LPCI initiation is auto-matic and no operator action is necessary for 10 minutes.
The, system will not operate automatically if the shutdown cooling mode with pressure less than 135 psig. Preoperational Test P49. 1-Revision 0 (Draft) Section 7.3.7(30)
demonstrates this.
(4)
FSAR Section 7.3. 1. 1a. 16.3 and FSAR Section 7.3.2a. 1.2. 1.9 state that manual initiation of system will start RHR pumps.
This is not shown on FSAR Figure 7.3-10 Sheet (5)
FSAR Figure 7.3-10 Sheet 2 shows LPCI loop selection circuitry as being installed.
This is not installed at this facility.
This item remains open pending an evaluation of the resolution to these FSAR discrepancies and the inspectors original comments.
Closed Ins ector Follow-u Item 387 80-24-06 i e Prep erational Test.
During a review of the Fire Protection Test P13.1 Revision 1, the inspector noted some discrepancies between the test and the National Fire Protection
.
Association Code requirements.
These comments were incorporated into Revision 2 of the test procedure.
The inspector reviewed P13.1 Revision 2 to assure that all comments had been properly'corrected in the test.
The inspector had no further questions on this ~atter.
(Closed Unresolved Item 387 80-24-10 Startu Work Authorization The inspector reviewed Startup Administrative Procedure AD6.4 Revision
and Bechtel Field Procedure FP-G-19 Revision 5 regarding the processing of Startup Work Authorization (SWA).
AD6.4 requires the use of an SWA for all hardware identified under the PPSL guality Assurance Program.
The Bechtel Field Procedures references AD6.4 for processing of SWA's by Bechtel personnel.
The inspector discussed the above with plant staff, startup group and quality assurance personnel to verify under standing of this requirement.
The inspector identified no deviations from this re-quirement.
The inspector had no further questions on this matter at this time.
0 en Ins ector Follow-u Item 387 80-28-01 C
Test P51. 1 Revision 0 Draft.
The inspector had noted that various FSAR sections and figures and General Electric drawings and system descriptions were conflicting as to the in-puts to the core spray initiation circuitry.
The inspector discussed these conflicts with the General Electric Startup Operations Supervisor on November 19, 1980.
He stated that a Field Deviaiton Disposition Re-quest (FDDR) was to be issued in early 1981 to put a reactor vessel low pressure permissive signal into the initiation logic.
- He acknowledged that the FSAR and supporting documentation needed correction to accurately describe the initiation logic as it is now and how it will be after the FDDR is incorporated.
He also acknowledged that the Core Spray Preopera-tional Test as it is written now does 'ot test the logic that will be in place after the FDDR is incorporated and this logic will have to be retested.
This item remains open pending review of the changes to the FSAR and supporting documentation and the incorporation of the FDDR into the pre-operational testing of the syste n. '(Open)'nspector Follow-up Item (387/80-28-02)
Primary Containment s rumen as reopera rona es The inspector had reviewed the preoperational test and noted that the test did not provide a leak rate test to assure that the system contained ade-quate gas in storage for a 30 day supply after a postulated design basis accident.
On December 22, 1980 the inspector reviewed Technical Change Notice 003 to P25.1 and verified it added this criteria to the test.
The inspector had also noted that relief valve setpoints were not checked as par t of the preoperational test.
The inspector was concerned that. this did not meet the requirement of Regulatory Guide 1.80 Paragraph C.3.
The 1-icensee stated that setpoints are set and checked at vendor shops and that a Technical Procedure (TP1.5)
was used by the licensee to verify that the seal had not been broken since the vendor established the setpoint.
On December 19, 1980 the inspector reviewed Manufacturer's Data Report for Safety and Safety Relief Valves (Form NV-1) from J.
F. Lonergan Company for the following relief valves in the containment instrument gas system:
(1)
PSV-12648 - set pressure 200 psig.
(2)
PSV-12643 - set pressure 200 psig.
The data reports were certified to be in compliance and in accordance with design on August 8, 1979.
The inspector next reviewed the Bill of Materials Sheets from King-Knight Company of Emeryville, California for the following valves:
(1)
PSV-12641 - setpoint 180 psig.
(2)
PSV-12638A - setpoint 180 psig.
(3)
PSV-12633A - setpoint 180 psig.
(4)
PSV-12619A - setpoint 180 psig.
(5)
PSV-12638B - setpoint 180 psig.
(6)
PSV-12633B - setpoint 180 psig.
(7)
PSV-126198 - setpoint 180 psig.
(8)
PSV-12615A - setpoint 90 psig.
(9)
PSV-12615B - setpoint 90 psi Finally the inspector reviewed Susquehanna Work Authorization (SWA)
Number U-01518 requesting the following safety relief valves in the con-tainment instrument gas system be bench tested due to broken seal setting mechanisms:
(1)
PS V-12616A (2)
PS V-12616B (3)
PSV-12644 (4)
PS V-12645 (5)
PS V-12646 This SWA had not been dispositioned as of the inspector review on December 19, 1980.
The inspector verified this included all'safety relief valves in the instru-ment gas system.
The inspector still had concerns whether this vendor testing and licensee seal inspection met the intent of Regulatory Guide, 1.80.
This item will be readdressed after NRC management review.
0.
(0 en) Ins ector Follow-u Item (387/80-28-13)
Reactor Core Isolation 00 ln ys em reo era iona es
. 'vlslon The inspector had reviewed the RCIC preoperational test and had noted dis-crepancies between the test and the FSAR commitments.
On November 19, 1980 the inspector discussed these discrepancies with the General Electric (G.E.) Sto Lead Engineer.
During this discussion the following FSAR problem was noted:'SAR section 5.4.6.2. 1.3 describes a
15 second time delay associated with valve HV1F045 and the low suction pressure trip signal.
According to G.E. elementary drawing. 791E421AF, Revision 7 this time'",del'ay.-
does not exist.
The inspector also discussed with the G.E. Sto Lead Engineer the need to test valves at their maximum differential pressure to assure proper opera-tion as described in the FSAR.
He stated this was a problem fo} many Nuclear Steam Supply Systems during the preoperational test phase since the reactor would not be at operating pressure.
He also stated that a'etter would be sent to the PP8L Startup Test Engineer stating what re-quirements/testing could not be performed during the preoperational phase testing and requesting such testing be incorporated into the startup testing phas This item remains open pending a review of. the incorporation of the in-spector's comments into the preoperational test, the change to the FSAR, and the letter from the G.E. Sto Lead Engineer to the PP8L Startup Test Engineer.
3.
Plant Tour The inspector conducted periodic tours of accessible areas in the plant during normal and backshift hours.
During these tours, the following items were evaluated:
Hot Work: Adequacy of fire prevention/protection measures used.
Fire Equipment: Operability and evidence of periodic inspection of fire suppression equipment.
Housekeeping:
Minimal accumulations of debris and maintenance of required cleanliness levels of systems under or following testing.
Equipment Preservation:
Maintenance or. special precautionary measures for installed equipment, as applicable.
Component Tagging:
Implementation and observance of equipment tagging for safety, equipment protections and jurisdiction.
Instrumentation:
Adequate protection for installed instrumentation.
Logs:
Completeness of logs maintained.
Security:
Adequate site construction security.
I Cable Installation:
Adequate precautions taken to prevent damage to in-stalled cables.
Communications:
Adequate public address system.
Equipment Maintenance and Controls: Corrective maintenance is performed in accordance with approved procedures, no unauthorized work activities on systems of equipment, no uncontrolled openings in previously cleaned or flushed systems or components.
No unacceptable items were identifie.
Prep erational'Test Procedure Review a.
Core S ray Pattern Test References (a)
(b (c d
(e)
Preoperational Test P51.1A, Revision 0 - Draft.
FSAR Section 6.3.2 FSAR Section 7.3. 1 FSAR Section 14.2.12.1 Regulatory Guide 1.68, Revision
(2)
The inspector reviewed reference (a) to ascertain whether design and operational commitments made'n references (b),'through (e) were ade-quately tested.
The inspector noted the following:
(a)
The test descr iptions of FSAR Section 14.2.12. 1 did not include a description of preoperational test P51.1A.
On November 24, 1980, the inspector was told by the cognizant Startup Engineer this was originally part of Preoperational Test P51. 1 and was therefore incorporated into the discussion of P51. 1.
(b)
FSAR section 14.2. 12.1 discusses a flow pattern test as part of P51. 1. It states that the test will demonstrate that the..
core spray system can deliver flow at adequate pressure to the reactor pressure vessel in an acceptable spray pattern.
Reference (a) performs this test, but the test does not provide a pressure limit as one of the acceptance criteria.
These comments were given to the Plant Assistant on November 25, 1980.
The resolution to these comments will be reviewed by an NRC inspector during a subsequent inspection.
(387/80-32-02)
b.
Standb Gas Treatment S stem (SGTS)
References'a)
(b)
(c)
(d)
(e)
(f)
(g)
Preoperational Test P70. 1, Revision 1, approved October 30, 1980 FSAR Section 6.5.1 FSAR Section 6.3.1 FSAR Section 11.5.2 FSAR Section 14.2.12.1 Regulatory Guide 1. 52, Revi si on
Regulatory Guide 1.68, Revision
(2)
The inspector reviewed reference (a) to ascertain whether design and
'perational commitments made in references (b) through (g) were ade-quately tested.
The following discrepancies were noted:
(a)
Reference (b) stated that one of the non-safety objectives of the SGTS was to filter and exhaust air'from the primary contain-ment pressure relief line.
This system does not exist at Susque-hanna Steam Electric Station.
(b)
FSAR Figure 9.4-9 shows the Main Steam Isolation Valve (MSIV) leak-age control system exhausting into the SGTS.
This is not correct since the system actually exhausts into the Zone 1 Equipment Compartment Exhaust System.
(c)
Reference (d) is not referenced in the preoperational test and describes automatic stops and starts of the SGTS due to high radia-tion in the SGTS vent exhaust.
This feature is not tested in the preoperational test.
(d)
Two dampers - HD-07543A and HD-07543B which are needed to function during a loss of coolent accident condition are not tested.
(e)
Reference (c) is not referenced in the preoperational test and describes numerous initiation/stop signals for the SGTS which are not tested.
(f)
Interlocks needed to start the SGTS manually are not tested.
(g)
No Regulatory Guides are used as references for the test.
(h)
Four times throughout the test - Test Section 7.3.5(7), 7.3.8(7),
7.3. 13(20)
and 7.3. 14(20)- the 'fan. motor.is.=stopped by opening dis-connect links under, load.
This appears to be a potential safety hazard.
These comments were given to the ISG Coordinator on November 25, 1980, and were discussed with the Assistant Superintendent of Plant on November 26, 1980.
The resolution of these comments will be reviewed by an NRC 'inspector during a subsequent NRC Inspection.
(387/80-32-03)
c.
Automatic Depressurization System '(ADS)/Steam Relief Valves (SRV)
(1)
References (a)
Preoperational Test P83.1B Revision 1 approved November 27, 1980.
(b)
FSAR Section 1.2 (c)
FSAR Section 5.2.2 (d)
FSAR Section 5.4.13 (e)
FSAR Section 7.3.1 (f)
FSAR Section 14.2.12.1 (g)
Regulatory Guide 1.68 Revision
(2)
The inspector reviewed reference (a) to ascertain whether design and operational commitments made in references (b) through (g) was ade-quately tested.
The inspector noted the following discrepancies:
(a)
References (c) and (e) both discuss sizing requirements for the ADS accumulators.
These sizing requirements are not tested in the preoperational test.
(b)
Reference (c) states there is a maximum of a one-tenth (. 1)
second time delay from the initiation of the activation signal until the ADS valves start to open.
This is not proven as part of preoperational test.
(c)
The test does not show that those SRV's which can be controlled from the remote shutdown panel will not operate from the remote shutdown panel if the transfer control switches are in normal.
(d)'est section 7.3. 1 steps (79) through (93) are testing test jacks J1A and J2A..
The procedure interchanges, the use:of these jacks providing erroneous results.
These comments were given to the ISG Coordinator on December 5,
1980.
The resolution of these comments will be reviewed by an NRC in-spector during a subsequent inspection.
(397/80-32-04)
(3)
During the review of reference (a) it was noted that a flow in-dication system for the SRV's was not being tested.
This system is required by NUREG-0737 issued November, 1980 to be installed prior to issuance of a fuel-loading license.
The inspector noted that no system has yet been installed and when installation is com-pleted a preoperational test of this system will have to be completed.
This test will be reviewed during a subsequent NRC inspection.
(387/80-32-05)
d.
Control Rod Drive (CRD)
H draulics (1)
References (a)
Preoperational Test P55. 1 Revision 1 approved December 5,
1980.
(b)
FSAR Section 4.6.
(c)
FSAR Section 14.2.12.
(d)
Regulatory Guide 1.68 Revision
(e)
IE Bulletin 80-17 Supplement
(g)
IE Bulletin 80-17 Supplement
(h)
IE Bulletin 80-17 Supplement
(i)
NUREG 0619 (j)
FSAR guestion 211.43 (2)
The inspector reviewed reference (a) to ascertain whether design and operational commitments/requirements made in references (b) through (j) were adequately tested.
The following discrepancies were noted:
(a)
Reference (b) states that the CRD drive water pressure will be approximately 260 pounds psi square inch (psi) above reactor vessel pressure, and the CRD cooling water pressure will be approximately 20 psi above reactor vessel pressure.
The pre-operational test sets these and tests regulators at 250 + 5 psid and 30 + 5 psid respectfull (b)
(c)
(d)
(e)
(g)
(h)
The following alarms were not tested using the Technical Pro-cedure T.P. 1.9 for digital control scheme testing:
(i)
CRD high temperature alarm (ii) Accumulator high level alarm (iii) Accumulator low pressure alarm (iv) Rod overtravel alarm The following alarms were not tested in the, preoperational test:
(i)
CRD high temperature alarm (ii) CRD pump trip alarm (iii) CRD motor overload alarm (iv) Charging water pressure high alarm (v) Scram valve pilot air header high/low alarm Test Section 7.3.5(14)
couples rods, but no check is performed to assure coupling is successful.
Section 7.3.6 sets total cooling water flow at 63 + 2 gallons per minute (gpm).
'Reference (b} states cooling water flow to each of CRD's should be from.20 to.34 gpm.
This means for 185 CRD's the maximum flow should be 62.9 gpm.
The band established in the preoperational test could have cooling flow set above this maximum value.
Reference (i) stated the applicants for operating licenses must demonstrate that both CRD pumps can be run simultaneously with no NPSH limits being reached.
This was not tested.
References (i) and (j) indicated CRD flow to the vessel should be measured and monitored during the preoperational test.
This
'as not done on P55.1 Revision l.
Pressure equalizing valves 1F150A, and 1F150B were not tested in preoperational test in accordance with licensee answer to reference (j).
These comments were given to the ISG guality Control Inspector on December 19, 1980.
The resolution of these comments will be reviewed in a subsequent NRC inspection.
(387/80-32-06)
e.
Reactor Recirculation S stem (1)
References (a)
(b)
(c)
(d)
(e)
(g)
(h)
Preoperational Test P64.1 Revision 2 approved November 22, 1980..
FSAR Section 5.1 FSAR Section 5.4 FSAR Section 7.6.1a FSAR Section 7.7.1 FSAR Section 14.2.12.1 Regulatory Guide 1.68 Revision
General Electric Elementary Diagrams - 731E287MJ sheets 1 through 21, Revision
(2)
The inspector reviewed references (a)
and (h) to ascertain whether design and operational commitments/requirements made in references (b) through (g) were adequately tested.
The inspector noted the following discrepancies:
(a)
(b)
(c)
Reference (f) states that the containment atmosphere circulation'ystem and the instrument air system are needed for the test, but the test does not list them as 'a needed support system in section 7.2.
The motor operated valves in the recirculation system are designed to be reversible.
This means the operator can reverse the'irection of the valve't any time.
This design is not tested in the preoperational test.
FSAR Section 7.7. 1.3.3.4.7 discusses the speed limiter.
This discussion is limited to the 20/ limit if the recirculation discharge valve is not fully open, or if feedwater flow is less than 205 of rated.
This limiter also is in effect if reactor vessel level goes below Level 3 (+12.5").
This is.not discussed in the FSAR section.
Also a second limiter limits recircula-tion pump 'speed to 555 of rated speed.
This limiter and the signals that produce the limitation are, not discussed in the FSA (d)
The preoperational test does not transfer control of recircula-tion pump speed between the manual/automatic transfer stations, and the master controller as described in reference (e).
(e)
Reference (h) sheet 7 shows that shutting the recirculation pump suction valve will trip the recirculation motor generator drive motor breaker.
This is not tested during the preoperational test.
(f).Reference (h) sheet 13 and 21 appear to have the K10A and KlOB relay contacts labeled in error.
(g)
Test sections 7.3.18 and 7.3. 19 do not test all combinations of the one out of two taken twice logic for the RPT breaker trip circuit on turbine stop valve closure, or the two out of two logic for the turbine control valve closure in accordance with reference '(d)..
These comments were given to the ISG guality Control Inspector on December.18, 1980 for resolution.
The resolution to these comments will be reviewed during a subsequent NRC inspection.
(387/80-32-07)
f.
Plant Leak Detection S stem (1)
References (a)
Preoperational Test P76. 1 Revision 1 approved November 17, 1980.
(b)
FSAR Section 9.3.3 (c)
FSAR Section 14.2.12.1 (d)
FSAR guestion 211.101 (e)
Regulatory Guide 1.68 Revision
(2)
The inspector reviewed reference (a) to ascertain whether design and operational commitments/requirements made in references (b) through (e) were adequately tested.
The inspector noted the following discrepancies:
(a)
The procedure referenced no FSAR sections.
(b)
Reference (c) stated that system redundancy and electrical independence would be proven as part of preoperational test.
This is not done, nor is the system designed to be redundan,
(c)
Appendix D of.reference (a) lists no.circuit breakers being needed to be in any position for this test.
Since the test is checking electrical circuits the inspector was unable to clarify how the proper electrical power supplies would be available for the test.
These comments were given to the ISG Coordinator on November 20, 1980.
The resolution to these comments will be reviewed during a subsequent NRC inspection.
(387/80-32-08)
5.
FSAR Discre ancies On November 26, 1980 the inspector met with the Assistant Superintendent of Plant and discussed problems he had noted between the FSAR and the preopera-tional tests being reviewed.
The inspector stated that two categories of problems had been noted during his review.
First the FSAR has description, or figure discrepancies with no changes initiated.to update these discrep-ancies.
The inspector gave as examples the RHR/LPGI-system-.initiation.descrip-tion, the core spray initiation description, the plant leak detection test description, and the control rod hydraulics scram discharge instrument volume description.
Second the preoperational test does not demonstrate the full capabilities of the system as described in the FSAR..
The inspector gave as examples the stand-by gas treatment system test, and'he core spray pattern system test.
The Assistant Superintendent of Plant acknowledged these findings and stated he knew that probl,ems existed with both the FSAR needing to be updated, and a better review for completeness for the preoperational test.
He stated the problems would be resolved..
The inspector acknowledged this.
6.
Boilin Water Reactors
- Licensin Review Grou Meetin On December 2, 1980, the inspector participated in a meeting between the NRC and the Boiling Water Reactors - Licensing Review Group (BWR-LRG).
A summary of this meeting was issued in an NRC letter dated December 5,
1980 addressed to the six applicants which form this BWR-LRG.
7.
IE Bulletin and Circular Followu a.
Discussion IE Bulletins and Circulars issued to PP8L were reviewed to verify the following:
(I)
Bulletins and circulars received by PPSL corporate management were forwarded to appropriate individuals within the organization,
including station management, for information, review and/or corrective actions as required.
(2)
PP8L bulletin responses were submitted to the NRC within the specified time period.
(3)
Licensee reviews and evaluations of bulletins and circulars are complete and accurate, as supported by other facility records and by inspector observations of installed plant equipment.
(4)
Corrective actions specified in licensee bulletin responses or internal circular evaluation memoranda have been completed and/or responsibilities have been assigned for completion, b.,~Findin s
(1)
The inspector reviewed licensee actions relative to IE Circular
,
79-11 "Design/Construction Interface Problem."
The inspector re-viewed the following relative to this circular:
PP8L Internal letter PLI-11045 dated December ll, 1980 PPSL letters to Bechtel PLB-10201 dated August 2, 1979 PLB-10208 dated August 6, 1979 PLB-11864 dated August 29, 1980 Bechtel letters to PPSL BLP-11365 dated August 27, 1979 BLP-11398 dated September 7,
1979 Bechtel Engineering Procedures Manual (EPM) Appendix G - In-terim change 4 to Revision 9 dated May 20, 1980
-
- ..Startup Administrative Manual Procedure AD6.11 Revision
dated April 9, 1980 PP8L Engineering Work Instruction (WI) No.
21 Revision
Design Change Packages WJFK, WJIG, WJFB, WJAD, WJDV.
Based on the review of the above this circular is closed.
(2)
The inspector reviewed licensee actions relative to IE Circular 80-18 "10 CFR 50.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems."
The inspector reviewed the following relative to this circular:
PPSL Internal letter PLI-10481 dated October 31, 1980 Plant Staff Administrative Procedure AD-00-003, Revision 1,
"PORC Procedure" PPENPP No. 6, Revision
PPENPP No. 25, Revision
10 CFR 50.59 The inspector determined.that the licensee had not adequately addressed the following issue:
II The referenced licensee documents do not define how it is determined which proposed changes to the facility requi.re a
safety evaluation.
The inspector was informed that Plant Staff Administrative Procedure AD-00-037 "Plant Modification Pro-gram" would address this concern.
The inspector requested a
copy of AD-00-037 as soon as it was avilable for review.
This circular remains open pending that review.
(3)
The inspector reviewed licensee actions relative to IE Circular 80-21 "Regulation of Refueling Crews."
The inspector reviewed the following relative to this circular:
PPSL Internal letter PLI-9991 dated September 29, 1980 Plant Staff Reactor Engineering Procedure RE-81-032, Revision 0 Draft, "Refueling Operations" The inspector determined that the licensee had not adequately addressed the following issues:
The licensee's refueling procedures doe" not include a pro-vision that the Refueling Floor Shift Supervisor directly supervise, from the refueling deck,'ore alterations.
The inspector discussed this concern with the licensee's
.Reactor Engineer, who agreed to revise RE-81-032 to include such a
provision.
The licensee's Startup Procedure (ST-3)
Fuel Loading was not available for review.
This circular remains open pending further review of RE-81-032 and ST-3.
(4)
No Bulletins were presented to the Resident Inspectors for review during this inspectio.
0 eratin Staff Traininq 9.
a.
The inspector reviewed the status of Plant Staff Procedures available at the Susquehanna Training Center Simulator for training of licensed operators.
The inspector reviewed the procedures with respect to commitments in Plant Staff Administrative Procedure AD-00-001, Revision 2 "Procedure Program."
On November 12, 1980, the inspector found that certain procedures were in use with incomplete approvals.
Ten of twenty-seven Off-Normal Operating (ON) Procedures, seven of seven General Operating (GO) Procedures, and one of seventeen Operating Procedures (OP)
were identified as not meeting the approval requirements of AD-00-001.
The licensee subsequently revised AD-00-001 by issuing Procedure Change 80-33 to permit the use of procedures for training or simulator tryout with approval by the responsible section head.
On November 25, 1980, the inspector reviewed the aforementioned procedures and selectively re-viewed additional Operating Procedures to verify implementation of Pro-cedure Change 80-33.
No additional discrepancies were identified.
b.
The inspector reviewed Susquehanna Training Center Directives and PPSL Nuclear Department Instruction (NDI) 4.2. 1 Revision 0 "Licensed Operator Training and gualification Program."
The inspector reviewed these pro-cedures to verify the existence of a training program.
A review for adequacy of the training program will be done during a future. inspection.
No unacceptable conditions were identified.
NRC TMI Action Plan NUREG 0737 - Clarification of TMI Action Plan Requirements
- was issued by the NRC in November, 1980, and describes post-TMI requirements for operating reactors and for applicants for an operating license.
The inspector dis-cussed this NUREG with the Superintendent of Plant, and members of his staff on December 10, 1980.
The inspector pointed out to the Superintendent of Plant that many of these requirements stipulated that action must be taken by the licensee.
All action items were given completion date requirements by the NRC and the inspector noted that some of these completion dates were stated as prior to Safety Evaluation Report (SER) issuance.
The Superin-tendent acknowledged these remarks and stated the licensee was reviewing this document to incorporate all required actions into their schedule.
The inspectors noted that FSAR guestions 211.210, 211.230, 331. 11, 331.14, and 331.15 had been answered by stating that the answer would be provided in PP8L's response to NUREG-0578, TMI-2 Lessons Learned.
As of January 2,
1981, the licensee's response to HUREG-0578 had not been submitted.
To date, no NUREG-0578 issues have been inspecte.
Unresolved Items
Unresolved items are matters about which more information is required to as-certain whether they are acceptable items, items of noncompliance, or devia-tion.
Unresolved items disclosed during the inspection are discussed in para-graph 2.e.
11.
Exit Interviews At periodic intervals during the course of this inspection, meetings were held with facility management to discuss inspection and findings.
At the final exit interview on January 5,
1981 the inspector expressed his concern to the Superintendent of Plant regarding the lack of adequate correc-tive actions to resolve previousl'y identified NRC findings.
The i,nspector stated that it appeared that increased emphasis needed to be placed on resolu-tion of these findings.