IR 05000423/1988024
| ML20235G126 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/16/1989 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235G115 | List: |
| References | |
| 50-423-88-24, IEB-87-002, IEB-87-2, NUDOCS 8902230128 | |
| Download: ML20235G126 (18) | |
Text
-
_.
_.
--
...
.
.
..
.,
.
.c p
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
'50-423/88-24 Docket No.
50-423 License No.
' Licensee:
Northeast-Nuclear Energy, Company P.O.- Box 270 Hartford, CT_ 06101-0270 Facility Name: Millstone Nuclear Power Station, Unit 3 Inspection At: Waterford, Connecticut Inspection Conducted:
December 20, 1988 through January 23, 1989 Reporting Inspector:
G. S. Barber, Millstone 3 Resident Inspector Inspectors:
W. J. Raymond, Senior Resident Inspector G. S. Barber, Millstone 3 Resident Inspector P. J. Habighorst, Millstone 2 Resident Inspector S. T. Barr, Reactor Engineer Approved by:
SO.k k h all4l&9 E. C. McCabe, Chief, Reactor Projects Section IB Date Inspection Summary:
Inspection on 12/20/88 - 1/23/89 Areas Inspected: Routine onsite inspection (133 hours0.00154 days <br />0.0369 hours <br />2.199074e-4 weeks <br />5.06065e-5 months <br />, 8 backshift hours) of plant operations; status of previous inspection findings; IFI 80-BU-11, Masonry Wall Design (Closed); TI 2500/26, Fastener Testing to Determine Conformance with Ap-plicable Material Specifications (IEB 87-02) (Closed); IFI 79-80-13, Cracking in Feedwater System Piping (Closed); IFI 86-22-03, health physics phone line to NRC Headquarters (Closed); MOV (motor-operated valve) overload and containment pene-tration breaker FSAR change; plant incident reports (PIRs); reactor / turbine trip caused by loss of Emergency Bus 34B Tie Breaker; service water leaks; TI 2515/86, Natural Circulation Cooldown (GL 81-21) (Closed); licensee event reports (LERs);
and surveillance.
Results: No unsafe conditions were identified. Three long standing open items and two temporary instructions were closed out.
8902230128 890216 PDR ADOCK 05000423 Q
PDC i
_ _ _ _.
. __.__ _ ____-__-___.-___---- -._ _ _ _ -
_--
-
-
-
--
. _ - - - - _ _ - -. - _ _
!
.
a c
.
!
.-
!
J
,
a l
I'
l TABLE OF CONTENTS
.
PAGE 1.0 Persons Contacted....................................................
I 2.0 S umma ry o f Fa c i l i ty Ac t i vi t i e s.......................................
3.0 Status-of Previous Inspection Findings...............................
3.1 (Closed) I FI 80-B0-11, Masonry Wall Design (92702)..............
3.2 (Closed) TI 2500/26, Fastener Testing to Determine Conformance with Applicable Material Specifications'(IEB 87-02) (71707)...
i 3.3 (Closed) IFI 79-BU-13, Cracking in Feedwater System Piping (62703).......................................................
3.4 (Closed) IFI 86-22-03, Installation and Test of HPN or
.
Commercial Backup Line to NRC Headquarters (71707)............
4.0 Review of Facility Activities........................................
l 4.1 MOV Overload and Containment Penetration Breaker FSAR Change (37700).......................................................
4.2 Steam Generator Sample Containment Isolation Valve Leakage (71707)...........................
...........................
4.3 Polyethylene High Integrity Container Design Inadequacy (84722).
5.0' Plant Operational Status Reviews (71707)..............................
5.1 Review of Plant Incident Reports (PIRs) (90712).................
,
6.0 Reactor / Turbine Trip caused by loss of Emergency Bus 34D Tie Breaker (93702)............................................................
7.0 Se rv i ce Wa te r Lea ks ( 37700)..........................................
8.0 (Closed) TI 2515/86, Natural Circulation Cooldown (GL 81-21) (92703).
,
9.0 Licen see Event Reports ( LERs) (92700)................................
10.0 Surveillance Testing (61726).........................................
11.0 Management Meetings (30703)..........................................
l
l i
i
!
!
I
.s
.,
.
-
'
_y
,
,
,,
l l
DETAILS 1.0 Persons Contacted
'
Inspection findings were discussed periodically with the supervisory and man-agement personnel identified below:
S. Scace, Station Superintendent C. Clement, Unit Superintendent, Unit 3 M. Gentry, Operations Supervisor R. Rothgeb, Maintenance Supervisor K. Burton, Staff Assistant to Unit Superintendent J. Harris, Engineering Supervisor D. McDaniel, Reactor Engineer R. Satchatello, Health Physics Supervisor
.M. Pearson, Operations Assistant R. Stotts, Operations Training Supervisor W. Potter, Requalification Training Staff 2.0. Summary of Facility Activities The plant operated at full power until 4:00'p.m... December 28 when a power.
decrease was commenced to mitiga'te the effects of high winds, incoming tides and a problem with the "B" traveling water screen. ' Power remained at 75%
until a plant trip occurred the following day at 5:10 p.m. due to loss of an emergency bus and holding power to the control rods (see Detail 6.0).
Elec-trical power was restored, the cause of the trip was corrected,.and startup began on December ~30.
Full power was achieved at 3:30 am, January 2.
The plant continued to operate at full power.
3.0 Status of Previous Inspection Findings 3.1 (Closed) IFI 80-BU-11, Masonry Wall Design Bulletin 80-11, Masonry Wall Design, asked licensees to evaluate the structural integrity of Seismic Category 1 Masonry Walls. As'a result, Stone and hebster Engineering. Company (SWEC) imposed a hold on further construction and attachments to Category 1 Masonry Walls (Hold #89, 5/25/81).
Prior to releasing the' hold (11/19/82), SWEC replaced the as-yet unbuilt masonry walls with reinforced concrete walls. Notes were added to SWEC drawings 12179-EA-1E and EA-1F to prohibit attachments to existing block walls. To expedite construction, a formal request was initiated to allow light objects (50 lbs or-less) to be attached to the existing block walls (12/82).
SWEC Calculation 12179-SE0-SE-52.56 veri-fied that minor attachments did not affect masonry wall integrity.
NRC Inspection Report 50-423/85-13 Detail 5.1.2.1 documented that the 50 lb limit was well within the allowable loads of the SWEC calculation. No inadequacies.were noted.
-
- _ -- - __-
e
.
.
s+
.
.
.
.
.
.. ;
L
~The licensee initiated a procedure to require inspection of masonry walls L
on a rotating basis. All walls are.to be inspected within 10 years per L
the applicable ~ Engineering Procedure. No inadequacies were noted.' '
In addition to-Bulletin 80-11 reviews, the licensee initiated a specific-review of the structural integrity of the battery. room masonry walls in-response to an allegation. The alleger contended that these walls did not meet the original design requirements. The allegation was substan-tiated and the licensee was issued a violation and a deviation for their
failure to meet design requirements.
In their September 17, 1987,re-sponse, the licensee showed that, although the battery walls failed to.
meet the original design requirements, they were designed to meet the design requirements of the control room masonry walls.
Furthermore, refined calculations showed that the subject walls did, in fact, meet FSAR design criteria. A December 4, 1987 NRC letter documented NRC ac-ceptance of the refined battery wall-calculation (see Inspection Report 50-423/87-15 for details) and noted no inadequacies. This item is closed.
3.2 (Closed) TI 2500/26, Fastener Testing to Determine Conformance with Applicable Material Specifications (IEB 87-02)-
The NRC issued IE Bulletin 87-02, Fastener Testing to Determine Conform-ance with Applicable Material Specifications, dated November 6, 1587, to request licensees to review their receipt inspection requirements for fasteners, and to determine through independent testing whether in-stock fasteners meet required mechanical and chemical specifications.
The licensee submitted his response to IE Bulletin 87-02 by letter dated January 12, 1988.
Inspector review found that letter responsive to the Bulletin (see Inspection Report 50-423/88-02). Out of 160 fasteners sampled, seven discrepancies were identified. Of the seven, two were found to be nonconforming.
Nonconformance Reports (NCRs) were written.
These two fasteners were dispositioned to "use as is" since their out-of-tolerance condition (within 2% of ASTM measurement accuracy) did not significantly affect their strength, ductility or corrosion resistance.
The licensee concluded that no additional actions were warranted on the fasteners in stock.
Inspector review noted no inadequacies. However, the inspector committed to further review of TI items 5.01 and 5.02 re-garding receipt inspection requirement prior to TI closure.
TI Items 5.01 and 5.02 called for a comparison of the licensee's receipt inspection program / procedures, maintenance warehouse procedures, and the licensee's descriptions provided in the bulletin response.
The inspector performed this comparison and noted that the bulletin response accurately summarizes the requirements of the following procedures, as appropriate:
QSD 3.08, Performance of Receipt Inspection Activities; and QSD 3.07, Preparation, Performance and Reporting of Source Inspections. No inade-quacies were noted. This TI is closed.
!
!
_
__
- -_ - _ _ _
-
..
.
-
.
.,
,
t L
l 3.3 (Closed) IFI 79-BU-13, Cracking in Feedwater System Piping On October 17, 1979, the NRC issued Revision 2 to IE Bulletin No. 79-13.
The revision described an inspection program to be implemented by licen-sees and applicants for an operating license to identify cracking indi-cations in feedwater (FW) systems. On 0ctober 19, 1982, the licensee
~
-
applied for an operating license for Millstone Unit No. 3.
On completion of the hot functional testing program and prior to fuel loading, the licensee performed the inspections described in Item 1 of Bulletin 79-13.
During the first refueling outage (October 30, 1987 to February-10, 1988),
volumetric examination of the feedwater nozzle-to pipe welds, of the feedwater piping welds to the first support and of the feedwater-to-containment penetration welds was performed per Item 2.a of the Bulletin.
In addition, a visual inspection of all feedwater system piping supports (except four rigid supports) and snubbers in containment was' performed to verify operability and conformance to design. The four-rigid feed-water system piping supports were inspected during the April 13, 1988 outage.
Preliminary evaluation by the licensee identified no unacceptable indi-cations. These preliminary results (except for the FW supports) were documented in the. licensee's March 21, 1988 submittal. The FW supports were found to be satisfactory during the April 13, 1988 outage inspection.
The final evaluation summarized all inservice examinations completed during the Cycle 01 refueling outage.
The summary listed 25 radiographs (RTs) and 72 visual examinations (VTs) as satisfactory after disposition-ing Unresolved Indication Reports (UIRs). The inspector reviewed a 10%
sample (3 RTs and 7 VTs) and noted that Inservice Inspection VIRs were written to disposition initially unacceptable findings.
Those findings ranged from loose nuts (later tightened) for piping supports to a crack identified by magnetic particle testing (weld repaired and retested satisfactorily).
The inspector noted that Engineering adequately dis-positioned the nonconforming conditions: the nonconformances were re-paired / corrected and reinspection were satisfactory.
3.4 (Closed) IFI 86-22-03, Installation and Test of HPN or Commercial Backup Line to NRC Headquarters This item was opened due to the lack of a dedicated Health Physics Net-work (HPN) phone line at Millstone Unit 3.
Two dedicated HPN commercial telephone lines have since been installed in the Unit 3 Operations Sup-port Center.
The licensee performs a monthly surveillance on the lines, i
including a visual inspection of the line's physical integrity and a i
telephone call to NRC Headquarters as an operational check.
The inspec-tor verified the performance of this surveillance by reviewing the lic-ensee's surveillance records for the previous year and telephoned the
!
NRC Headquarters Operations Center on this line.
Communications on this line were clear and intelligible.
No inadequacies were noted.
I
_ _ _. - -.. -. _ _ _. _. _ _ _ _.
- _ _ _ _ _ _. - _ _. - -
_
_. - -... - -
-
_. _. _ - _. _ _.. _. -
. _ _. -. - -. _
_.---__-__--------_.._._.-.__-.__-_--.-._.-_._.._.___-D
-.
- _ _ _ _ - _ - _ _ _ _ ___
_
_ _.
. - _ - _ _ _ _ _ _.
-
- __ -
__ __
__-__.
_.__
_ _ _ _
-
!
.
A
-
- -
.
,
-
..
I 4.0 Review of Facility Activities
)
' j 4.1 MOV Overload and Containment Penetration' Breaker FSAR Change
'
The~ licensee submitted Technical Specification. (TS) change request 28 :
to remove unnecessary tables from TSs. -LCOs are to remain unchanged ex-cept;for the deletion of the tables. The tables are to be included in
.
the applicable FSAR section.
Licensee changes to the tables could then be implemented in accordance with 10 CFR 50.59.
During review of the-amendment request, the NRR licensing project manager (LPM) identified errors 'in. the identification of components contained within FSAR table 8.3-7, Motor-Operated Valve (MOV) Thermal Overloads (0Ls) Bypassed Only Under Accident Conditions. A supply valve was listed as a return valve and vice versa. -The amendment request was tabled.
Because of.the errors,-the licensee agreed to a 100% review of.the three FSAR tables (Table 8.3-7, 8.3-8, and 8.3-9).
A three-man task force completed the review and the results were documented in Licensee Memor-andum GEE-88-415.
The inspector noted that the task force identified.
errors in all three tables. A corrected version of the FSAR tables was
'
issued. - The inspector verified that the current control room copy of the FSAR corrected the noted errors.
No inadequacies were noted.
4.2 Steam Generator Sample Containment Isolation Valve Leakage Four remotely operated valves (3SSR*CTV19A, B, C, D) provide containment (CNTMT) isolation for steam. generator (SG) sample lines. These valves receive a close signal when an automatic (auto) start signal.is sent'to the auxiliary feedwater (AFW) pumps.
Their function'in this case is to prevent diversion of a small portion of AFW flow to the sample' sink.
The licensee has been experiencing problems in getting the' valves to seat-fully closed.
On December 22, the licensee entered Technical Specification (TS) 3.6.3 because of excessive leakage through 3SSR*CTV198. The licensee's main-
.
tenance organization mechanically agitated the valve and was able to i
reseat it prior to expiration of the action statement.
Engineering per-formed a review to determine the acceptability of various leakage limits without compromising the safety function of the subject valves.
I The subject valves are Class B Containment penetrations as defined in Section 6.2.4.2 of the FSAR.
Per section 6.2.4.1.2 and Table 6.2-65 of the FSAR, 10 CFR 50, Appendix A, General Design Criteria (GDC) 16, 54, and 57 apply to the design of these valves. GDC 16 states, in part, that
.
"... containment and associated systems shall be provided to establish
!
,
an essentially leak tight barrier against the uncontrolled release of
'
radioactivity to the environment...". GDC 57 states that Class B pene-trations have a valve that is either locked closed, automatic, or capable of remote manual operation.
The subject valves are both automatic (they I
!
!
l-
'
_
I
.
.
.
.,
.
-5 I
I close on an auto start signal to the Aux Feedwater Pumps), and capable i
of remote operation from the control room. Most containment isolation
'
valves are tested for leak tightness per 10 CFR 50 Appendix J.
However, the subject valves are not subjected to Appendix J testing.
Four sources of leakage criteria were reviewed by the licensee as follows.
1.
The subject valves were purchased to meet Specification Spec-654.
According to the valve data sheet in Spec-654, these valves are de-fined as leakage class 4.
The spec gives acceptable leakage limits for this class of valve as 0.0005 cc/ min per inch of port diameter per psi differential. Calculation of the associated leakage limit results in a limit of less than 1 cc/ min.,0perating history has shown that the valves will not meet this limit.
2.
ASME Code Section XI, Article IWV 3426 recommends a valve seat leakage limit of 30D cc/hr., where D is the nominal valve size in inches.
For the subject valves, this limit would also be signifi-cantly less than 1 cc/ min.
3.
TS 3.4.6.2 limits leakage from RCS pressure isolation valves to 0.5 gpm per nominal inch of valve size, up to a maximum of 5 gpm at normal operating pressure (NOP). Applying this guideline would result in an allowable leak rate corresponding to approximately 50%
of normal flow.
4.
Since an achievable leakage limit could not be derived from existing sources, the safety analysis and design basis were reviewed by the licensee to determine the effect of leakage through the valves.
Tbsse valves receive a signal to auto-close upon an auto-start sig-r,al to the AFW pumps. The purpose of this is to ensure that, during an accident, AFW will not leak from the Steam Generators and degrade Tieat sink capability. The sample lines are 3/4" lines from the Steam Generator Blowdown lines, to 3/8" tubing upstream of the con-tainment penetration, to 3/4" lines through the penetration, through the 3/4" Containment Isolation valves (the valves have a 3/8" port),
and through a reducer to 3/8" lines to the sample sink.
Normal flow through the valves, according to the valve data sheet, is 2000 cc/
min. Maximum flow is 3000 cc/ min.
Operating history has shown that these valves pass approximately 2500 cc/ min under normal conditions.
If all four valves were open, about 3 gallons per minute would be lost through this flow path.
This would not affect the AFW system's ability to perform its safety function.
In the event of a Steam Generator Tube Rupture, these lines become a
direct path from the RCS to the sample sink.
The licensee evaluated the leakage through the valve, even with the valve fully open, as insignifi-cant. Also, any leakage would be to the Auxiliary Building, and not directly to the environs.
In the event of some other DBA, these lines
!
I
q
-- __. - _ _
_.
. _ _ - _ _ _ _ - - _
- - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _
.
.
..
.;
.
G
?
-
could potentially become open to the Containment atmosphere, if sheared.
However, due to the steam / air environment in Containment under' those
-
conditions, and due to the lower pressure differential across the valve (38 psid vs 2200 psid at normal operating pressure), leakage would be much less. Again, any. leakage through the valve. would be to the the Auxiliary Building.
In light of the above analysis, Engineering recommended a change be written to SP 3611A.1 to allow up to 100 cc/ min leakage through the valve while still considering it to meet TS 3.6.3.
This limit was felt to provide ~the needed' operational flexibility while still providing assur-ance that the~ valve was essentially shut.
Later measurement of leakage from'these valves found that leakage exceeded 100 c:/ min. TS 3.6.3 was again entered for the leaky va'ves.
Further Engineering review concluded that, as long as the valves ir.$cated shut (by stem movement or remote /
local indication), the leakage limit could be expanded up to full flow through the valve.
The inspector questioned this conclusion.
The licensee stated that a major factor in their assessment was that the=
lines do not contain radioactive fluids nor are they open to containment atmosphere.
The diverse and redundant AFW pumps assure that the_ required flow is provided to the steam generators even with all the sample lines fully o' pen.
Post-accident addition to the source term with these valves open would be inconsequential. Therefore, as long as the valves are capable of being closed for their containment isolation function, the -
amount of leakage through the valves will not increase the consequences of postulated accidents.
This matter _is unresolved (88-24-01) pending further NRC review of'the following.
a.
The isolation valves being allowed to pass up to full flow when shut.
b.
The practicability of and need for installing isolation valves cap-able.of performing the isolation function and meeting the purchase specification.
4.3 Polyethylene High Integrity Container Design Inadequacy High Integrity Containers (HICs) are used to provide structural stability to low level radioactive waste. NRC regulations establish three cate-gories of low-level waste: Class A, B and C.
Class-A waste is the least, radioactive.
Class B or C wastes must be structurally stable: they must generally maintain their physical form and dimensions under expected disposal conditions.
Structural stability provides assurance that the waste and soil overburden will not settle over time and trap water in the settlements. Structural stability can be provided by the waste form
-
- _-
...
- _
__
_ _ _ _ _.
.
.
.
.
.
.
._
[
-
p
.
'
,
.
l
itself, by processing the.wa'ste to a stab 1'e form, or by placing the waste in a container or'a structure that will provide the stability.- Stability.
for 300 years is sought to prevent the need for active maintenance.
The NRC staff has concluded that certain HICs have not been shown to meet the 300 year structural stability criterion.
Since the concerns about-these containers are associated with very long-term stability, no imme-diate action is necessary on. containers used in the past. However, longer.-
term monitoring and maintenance of sites where these containers are buried may be necessary.
Based on NRC review of late-1983 and 1984 sub'ittals by three-companies m
on their high-density polyethylene high integrity container (HIC) designs, the NRC has determined that the designs have not been shown to meet Part 61 of the Commissions' regulations for the disposal of t, lass B and C low-level radioactive waste.
The three companies that submitted HIC designs to NRC for approval were Chem-Nuclear systems, Inc. of Columbia, SC; TFC Nuclear Associates, Inc., Moorestown, NJ; and Westinghouse-Hitt-man Nuclear Inc.,'Moorestown, NJ. The conclusion that these three de-signs have not been shown to meet the requirements for 300 year struc-tural stability and were therefore not approved for use for disposal of i
Class B and C waste is based on the following.
(1) Polyethylene is subject to plastic deformation (creep) that can lead to creep-rupture and contribute to the initiation of buckling.
(2) High-density polyethylene may be unacceptably degraded by embrittle-ment due to aging and/or irradiation.
The NRC's conclusions regarding the three designs submitted do not apply to HIC designs which use polyethylene for corrosion resistance only or which provide structural stability using another material or system.
Polyethylene HICs might, for example, be acceptable for low-level waste
,
disposal where an engineered structure provides the structural stability.
The NRC's findings were documented in a " Technical Evaluation Report Re-lated to the Topical Reports on High Integrity Containers Made with High-Density Polyethylene" dated December 1988.
Copies of the report were sent to the three companies that submitted their designs for approval and to the States of South Carolina, Washington and Nevada.
(These States have regulatory authority over the three operating low-level waste disposal facilities.) A copy of the transmittal letter was given to the licensee.
Polyethylene HICs have been used for disposal of low-level radioactive waste since 1980. After issuance of the Commission's low-level waste regulations in Part 61 in 1982, the containers continued to be accepted for disposal while awaiting NRC review.
The three States with regulatory I
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_
_
-_-
. _ - _ _ - _ _ _ _ _ _ _ _ _ _
_
. _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ - ___
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _
..
.
.
,
1.
.
.,
authority'over operating disposal sites are evaluating the impact of the NRC position regarding high-density polyethylene HICs on site operations.
The NRC staff is discussing implementation of the findings with the States.
The inspector. informed the licensee of the unacceptability of-the high density polyethylene HICs.
The licensee stated that they have only used one of.the suspect HICs.
It was loaded in late 1987 and is still await-
- ing classification based on sample activity. The manufacturer, Chem-Nuclear Systems, Inc. (CNSI) has provided no direction to the licensee on any additional administrative controls.
The licensee believes that all Class B and C waste in HICs is currently being' packed in concrete containers by CNSI upon receipt.
The inspector had no further questions on this issue.
5.0 Plant Operational Status Reviews
'
The inspector reviewed plant operations and the operational status of safety i
systems to verify safe operation in accordance with technical specifications I
and plant operating procedures. Actions to meet technical specification re-quirements when equipment was inoperable were reviewed to verify the limiting conditions for operations were met.
Plant logs and control room indicators were reviewed to identify changes in operational status since the last review and to verify that changes in the status of equipment was properly communi-cated in the logs and records.
Control room instruments'were observed for correlation between channels, proper functioning, and conformance with tech-nical speci.fications. Alarm conditions were reviewed with control room opera-tors to verify proper response to off-normal conditions and to verify opera-tors were knowledgeable of plant status. Operators were found to be cognizant of control room indications and plant status.
Control room manning and shift staffing were compared to technical specification requirements.
No inade-quacies were identified.
The following activities were also addressed.
5.1 Review of Plant Incident Reports The plant incident reports (PIRs) listed below were reviewed to (i) de-termine the significance of the events; (ii) review the licensee's evalu tion of the events; (iii) verify the licensee's response and cor-rective actions were proper; and, (iv) verify that the licensee reported the events in accordance with applicable requirements, if required. The PIRs reviewed were: 215-88 dated 12/12/88, 217-88 dated 12/15/88, 218-88 dated 12/19/88, 220-88 dated 12/19/88, 223-88 dated 12/28/88, 227-88 dated 12/30/88, 2-89 dated 1/6/89, 5-89 dated 1/10/89, 6-89 dated 1/11/89, i
8-89 dated 1/15/89, 10-89 dated 1/16/89, and 12-89 dated 1/22/89.
I I
The following items warranted inspector followup:
l l
l'
'1 l
_ _. _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ -. _
-___
.
.
'
.
.
.,.
.
L
'
PIRs 216-88 dated 12/14/88, 219-88 dated 12/19/88, 222-88. dated 12/20/88, 224-88 dated 12/29/88, and 3-89 dated 1/6/89 all document instances where fire barriers were~either breached or degraded without establishing the required fire watches. A violation was written in Inspection Report 50-423/88-23 in this same area. The listed PIR.s along with the PIRs men-tioned in the violation bring the sum total number of fire watch problems--
to fourteen._ The continuance of these problems along with the prior problem history emphasizes-the need for effective corrective action.
Licensee corrective action will be reviewed further.
PIRs 221-88 dated 12/18/88, 4-89 dated 1/9/89, 7-89 dated 1/13/89,.11-89-dated 1/17/89 document four instances where fire dampers failed to fully close and engage their locking devices.
The licensee is appropriately incorporating all of these failures into one master PIR that will docu-ment all of the failures and provide a programmatic review to ensure-effective corrective action.
This type of review is preferred over that which results from a one-at-a-time fix.
This programmatic review will be completed by 6/18/89 per PIR 4-89 guidance. This issue will be re-viewed in' future inspections.
,
PIR 225-88 dated 12/29/88 documented a Reactor Trip caused by a loss of Bus.34B.
See detail 6.0 for more information.
PIR 226-88_ dated 12/29/88 documented a potential relay failure that could result in a loss of both emergency busses.
See detail 6.0 for more in-formation.
PIR 228-88 dated 12/30/88 documented that nitrogen accumulator pressure switches (PSs) for the feedwater (FW) containment isolation valves were originally procured and installed as non-safety-related parts. As cor-rective action, the switches are left isolated at a class break boundary except for readings that are taken once daily, on the midnight' shift.
The appropriate LCO is entered and exited when the readings are taken.
The licensee plans to either upgrade the existing PSs or install quali-fied switches. No inadequacies were noted.
PIR 1-89 dated 1/5/89 documents secondary sample containment isolation valves that were left open for greater than four hours without entering.
'
the proper action statement.
This item was reportable in accordance with 10 CFR 50.73(a)(2)(1)(b). An LER is to be written to document the lic-ensee's investigation and corrective action. No inadequacies were noted.
!
See detail 4.2 for more information.
6.0 Reactor / Turbine Trip Caused by Loss of Emergency Bus 34B Tie Breaker A reactor / turbine trip occurred at 5:10 p.m.,12/29 due to tripping of the j
tie breaker between balance-of plant bus 34B and emergency bus 340. The plant
"
responded as expected except for a large post-trip cooldown. The Main Steam l
.
__
9*
_ ___-_ _ - ___
'
.
.
..
..,
.
,
Isolation Valves (MSIVs) were shut to stop the cooldown and the plant was'
stabilized in hot shutdown.
The licensee is conducting a programmatic review of excessive post-trip cooldowns.
The plant was at 75% power prior to the trip with.4160V emergency busses 34C l
and 34D being supplied by the reserve station service transformer (RSST).
!-
The 34C and' 340 supplies had been shifted to the RSST to prevent a previously unanalyzed accident from occurring: Licensee Event Report (LER) 88-26-00 documented a switchyard breaker opening scenario that could cause a generator load rejection with the main generator breaker remaining closed. The LER assumed one switchyard breaker was inoperable; tripping of the second switch-yard breaker would cause a load reject and generator coastdown. With the normal electrical lineup, 4160V busses are supplied from the normal station service transformer (NSST). During a load rejection, due to the opening of both switchyard breakers, the generator would coast down to 3220V/40Hz, and'
a fast transfer from the NSST to the RSST would shift the 4160V loads from 3220V/40Hz to 4160V/60Hz in 6 cycles (0.1 second).
This action could damage running safety-related equipment on both emergency busses.
Interim corrective action was to transfer 4160V busses to the RSST if one of the switchyard breakers were to be opened.
Inspector review of the electrical drawings identified several relays which individually could fail and initiate the same coastdown scenario. These re-lays are associated with the pilot wire tripping scheme.
Current transformers outboard of the 13T and 14T switchyard breakers feed relay 87PWY which feeds relay 94PWY through an "0R" gate.
Relay 36 PWY also feeds the "0R" gate.
Relay 94PWY would initiate breaker trips for 13T and 14T without tripping the generator breaker.
Single failure of any of the above three relays could initiate the turbine coastdown scenario. The licensee is reviewing design change options.
The implemented design change will be reviewed in future inspections.
Response to the resident inspector's finding resulted in the licensee shifting busses 34C and 34D to the RSST at 4:17 p.m., 12/29. Then, the
'B' diesel was started for surveillance and closed in on the 340 bus. A current surge caused a directional overcurrent trip of the 34B/34D tie breaker (1120 amps out of the bus for 2 seconds) and deenergized the 348 bus. That caused Silicon Con-trolled Rectifiers (SCRs) providing holding current to the control rods to be deenergized. That caused the control rods to drop and cause a high nega-tive rate trip of the reactor and turbine.
Except for the post-trip cooldown, the plant responded as expected to the trip.
The inspector attended PORC meeting 3-88-182 to assess the licensee's evalu-
ation of the root cause of the reactor trip. The licensee postulated that,
'
when the 34B/34D bus tie breaker opened, the rod drive motor generator (MG)
set reversed current flow and tried to carry 480 volt bus 32N and 4160 volt bus 34B.
Identification of this scenario as the root cause was initially supported by other indications. The sequence-of-events (SOE) computer print-out showed the full length rods at the bottom of the core 112 milliseconds (msec) af ter the "A" reactor trip breaker (RTB) opened. A review of previous
- _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ - _ - _ _ - - _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ - - - -
-
.- - - - -
- -.---- -
--
.
.
- '
'
.
,
.
l.
SOE printouts shows that rods are at the bottom approximately 500 msec after opening the trip breakers. Therefore, the rods started falling prior to the reactor trip breakers opening.
Bus 34B was deenergized 650 msec prior to the
"A" RTB opening, supporting the conclusion that the 34B/34D bus tie breaker opened on a current surge and initiated the trip sequence.
Later Engineering review noted that bus 34B provided transformed power to 120 VAC panel 6N through circuits 5 and 19.
These circuits provide gating current to the SCRs that provide " hold" voltage to the control rods. A loss of these circuits by themselves should not have caused a trip because of a redundant power supply (PS). However, the redundant PS was malfunctioning when Bus 34B was lost and a plant trip occurred. The licensee repaired the defective PS and began a startup on December 30. The reactor was critical at 11:02 p.m.,
that same day and full power was reached at 3:30 a.m., January 2.
No unac-ceptable conditions were noted. Adequacy of maintenance of the PS which mal-functioned will be reviewed during a future inspection.
7.0 Service Water Leaks Over the past two years, approximately eight service water leaks have been identified, with four of the leaks occurring since July 1988. These leaks were documented in PIRs: 128-88 dated 7/3/88, 138-88 dated 7/24/88, 163-88 dated 9/22/88, and 176-88 dated 10/25/88.
In addition, the licensee has performed ultrasonic testing of a majority of 1.5 and 2.0 inch service water piping at suspect locations to determine the amount of wall loss experienced.
The licensee has established a trending and monitoring program to repair /re-place degraded piping.
Analysis of various leak locations has shown that the most probable piping location for wall reduction is on the inner diameter of the downstream side of a 45 or 90 degree elbow or tee connection. The licensee postulates that the change in direction caused by the tee or elbow sets up a strong, counter-rotating eddy current.
This eddy erodes the corrosion layer,. exposing fresh base metal (copper-nickel) to salt water.
That accelerates general and pit-ting corrosion.
This erosion-corrosion mechanism has caused the majority of service water leaks identified thus far.
The licensee monitoring program has identified piping wall losses at various locations.
During the initial inspection from June through September 1988, 80 to 90% of 1.5 and 2.0 inch piping was examined by Eddy Current Testing (ECT). The areas immediately surrounding 700 field welds (FWs) was examined and the degree of wall loss was documented. Ultrasonic testing (UT) was per-formed at the degraded locations because of the better accuracy of UT over ECT.
Degradations were mapped out around the affected FWs to characterize the flaws.
UT readings were taken within 3 inches (at 1 inch intervals) of each FW at readings of 0, 90, 180 and 270 degrees of pipe circumference. UT readings were compared against nominal wall to classify the degraded piping into categories A, B, and C, as follows.
._- _- _ -__ - ______ -
_ _ _ _
_ -__
__ _
-
k*
-
'
..
..
~
t
.
.
A.
Leaks with patches or locations with greater than 50% wall loss.
These are expected to leak before the second refueling outage (RF02).
B.
Locations with 25-50% wall loss. These may or may not leak before RF02.
More monthly. data is needed to make a more precise calculation.
C.
Locations with less than 25% wall loss. These are not expected to leak before RF02.
The 16 Category A and 10 Category B FWs found are scheduled to be repaired i
during the upcoming refueling outage. Disposition of the 6 Category C FWs will be delayed'pending determination of the wastage rate.
Present licensee practice is to temporarily repair leaks as they occur, with permanent repairs to be made during the refueling outage.
Licensee safety analysis, incorporat-ing seismic and fracture mechanism considerations, has concluded that the leakage potential does not represent a potential inability to meet the SW design bases.
The inspector reviewed the licensee's ECT and UT data and noted that the lic-ensee identified many degraded FWs. While reviewing the ECT~ data, the in-
-
'
spector noted that the uninspected. FWs (10% to 20%) were in hard to reach areas or were covered with lagging and the licensee elected not to inspect these areas. The inspector stated that, with the 5% degradation population that exists, there may be hidden flaws that will go through wall before the next outage. The licensee is evaluating this comment. This issue will be reviewed in future inspections.
8.0 (Closed) TI 2515/86, Natural Circulation Cooldown (GL 81-21)
This Temporary Instruction required verification of the licensee's implemen-tation of programs for the control of natural circulation (NC) cooldown in accordance with their commitment to Generic Letter (GL) 81-21. Based on an NC cooldown event at St. Lucie Unit 1 in 1980, multiplant action (MPA) B-66 was initiated for pressurized water reactor (PWR) licensees to implement pro-cedures and training programs to ensure the capability to deal with such events.
GL 81-21 specified that PWR licensees assess the procedures and training pro-grams, including a demonstration that a controlled NC cooldown would not re-suit in reactor vessel voiding, and describe plant training programs and emergency procedures that dealt with the prevention or mitigation of reactor vessel voiding. At the time GL 81-21 was issued, Millstone -Unit 3 had not
.
I yet' applied for an operating license, and the licensee was not required to provide this assessment.
Instead, MPA B-66 was incorporated into the normal
,
licensing review.
The purpose of TI 2515/86 was to verify that PWR licensees have implemented programs for the control of NC cooldown in accordance with GL 81-21.
The elements of this verification were:
,
1.
Verify plant-specific commitments and items requiring field verificatio _ _ _ _ _ _ _ _
-
>
.
. -
..
,
i j
-
]
1
.
.
.
l 2.
' Verify that the training program includes both classroom and simulator
.]
coverage 'of procedures on an NC cooldown by review of records, discus-sions with individuals, or observation of.similar activities for three licensed operators.
?
3.
. Verify that the licensee has emergency procedures regarding NC in ac-
cordance with GL 81-21.
- The licensee addressed natural circulation-cooldown as described in NUREG-1031, " Safety Evaluation. Report Related to the Operation of Millstone Nuclear
'
Power Station, Unit 3."
In Section 5.4.7.5 of NUREG-1031, the NRC indicated'
that confirmation of NC cooldown ability could be accomplished by referencing
,
results of tests from a plant similar in design to Millstone Unit 3.
In March.
l 1985, an NC cooldown test was performed at Diablo Canyon Unit 1, demonstrating
!
. that the plant could safely be taken to cold shutdown under NC conditions.
The licensee submitted a report to the NRC justifying the applicability of the results of that' test at Diablo Canyon to the Millstone Unit 3 design.
That report, " Millstone Nuclear Power Station, Unit 3, Natural Circulation System Comparison Report," dated November 6,1987, compared the plant systems and equipment that affect the natural circulation, cooldown and depressuriza-tion capabilities of Millstone Unit 3 with those of Diablo Canyon Unit 1.
At the NRC's request, this report was revised and re-submitted on August 3, 1988.
In a response dated October 18, 1988, the NRC staff concluded that their concerns regarding NC cooldown at Millstone Unit 3 had been adequately addressed, s
To verify the adequacy of the-licensee's NC cooldown training program, the inspector discussed natural circulation with three licensed operators.
The operators answered questions on natural circulation principles, the' applicable plant Emergency Operating Procedures, and on their confidence in the training they had received. The inspector found the operators to be knowledgeable in the subject matter. No inadequacies were noted.
On a separate occasion, the inspector discussed the training program with one of the licensee's operations training supervisors and with a member of'the
. licensee's requalification training staff. Also, the inspector reviewed in-di.vidual operators training records, licensee lesson plans and samples of test questions used for operator requalification. The inspector noted that the licensee provided a minimum of six weeks per year of requalification
<
training for each operator, with approximately half that time in the classroom and ths other half in simulator training.
Reviewing the previous year's les-son plans and attendance records, the inspector determined that 53 of the
!
plant's 57. licensed operators had received classroom requalification training in NC cooldown, and 55 of the licensed operators had experienced at least one simulator scenario iequiring the performance of an NC cooldown.
All dx of the operator absences were accounted for by the licensee, and the inspector
,~
verified records showed that all missed work had been made up through indi-vidual review assignments.
In addition, the inspector attended a requalifi-
cation training classroom session in which the NC emergency procedures were
'
reviewed.
This session covered the procedures step by step, with explanations
,
O ee
__m__._._.__u-_-_-___.--___m__.______m
_. _ _. _ _ _ _ _ _.
____m.
_ _. _ _ - _
-. _
+
.
'h,'
-
.
.
.
.
l
,
-
.
'
'
of what would.be occur. ring in the plant and what indications to be especially aware of in the control room.
No inadequacies in the licensee's training program were'noted.
'l
,
While assessing the licensee's ability to conduct an NC cooldown, the inspec-tor reviewed the Millstone Unit 3 Emergency Operating Procedures (E0Ps) used to control such an event. The-licensee has three E0Ps dedicated to an NC cooldown: E0P 35 ES-0.2, " Natural. Circulation Cooldown", E0P 35 ES-0.3, "Natu-ral Circulation Cooldown With Steam Void In Vessel (With RVLMS)", and E0P 35-ES-0.4, " Natural Circulation Cooldown With Steam Void In Vessel (Without RVLMS)." E0P 35 ES-0.2 compared favorably with the procedure used in the Diablo Canyon Unit 1 test documented in WCAP-11086, "Diablo Canyon Units 1 and 2 Natural Circulation / Boron Mixing /Cooldown Test Final Post Test Report, March 1986." Since the time GL 81-21 was issued, limited voiding in the reactor vessel upper head has been determined by the NRC staff to be a con-trollable evolution, provided it can be accomplished using all safety grade equipment with NRC-approved procedures and licensed operators trained in the use of those procedures.
Steam voids in the reactor vessel are' addressed in E0P 35 ES-0.3 and E0P 35 ES-0.4.
The operator is cautioned to be. aware of the indications of void formation in E0P:35 ES-0.2.
If void formation is detected, the operator is directed to either E0P 35 ES-0.3 or E0P 35 ES-0.4 No procedural deficiencies were noted.
This inspection reviewed Millstone Unit 3 plant facilities, operating proce-dures and operator training programs with respect to the licensee's ability
.
!
'
'
to' conduct this type of plant cooldown.
No inadequacies were found. This
. item'is closed.
9.0 Licensee Event Reports (LERs)
Licensee Event Reports (LERs)' submitted during the report period were reviewed to assess LER accuracy, the adequacy of corrective actions, compliance with 10 CFR 50.73 reporting requirements, and to determine if there were generic implications or if further information was required.
Selected corrective actions were reviewed for implementation and thoroughness.
The LERs reviewed were:
LER 88-24-00, Manual Reactor Trip due to Imminent Turbine Trip on Low
--
Condenser Vacuum, documented a trip due to storm fouling of the intake screens.
That caused a loss of two out of six circulating water (CW)
pumps.
Loss of the CW pumps led to unrecoverable loss of condenser vacuum, resulting in an operator-initiated manual reactor trip. The root cause of the trip was debris loading beyond the capabilities of the screen wash system.
Blockage and backup of debris in the sluicing system prevented continued effective backwash of the screens.
The licensee has installed an open clean-out port in the sluice water entry to the sluic-ing system to allow easier cleaning during stormy conditions.
No inade-quacies were note _
- - -
_
_-
- _ _ _ _ _ _ _ _ _ - _
- - _ - _ - _ _ _ _ _ -.. _ _ _
_ _ - _ _ _ - _ - - _ _ - _ _
.
.
a
...
i
,
,
'b
'
%
LER 87-42-01, Missed Intermediate Range / Power Range Surve111'ance d'ue to
--
~
Procedural' Inadequacy, documented a failure to perform a detector plateau.
'
curve during the surveillance per TS 4.3.1.1.
The root cause was the omission of the plateau curve requirement from the surveillance procedure.
Corrective action involved' reviewing all outstanding surveillance.
.against the Master Test ' Control List (MTCL):to verify that all TS re-quirements were addressed.
LER 86-58-02 ' reviewed 96 of the 1028.sur-veillances listed on the MTCL. Of the remaining 932, 404 required no-
,
action, 525 had minor' problems and 3 were evaluated as reportable and
.o were included'in this LER.
Surveillance Procedure adequacy was an.iden-tified weak area in the last SALP.
This issue will be reviewed:further during a management meeting in.early 1989.
LER 88-25-00, Failure to Post a Fire Watch with Degraded Fire Protection
--
Due to Procedural. Deficiency, documented the failure to. compensate for the "A" Train 4160V Switchgear fire protection system when its zone module switch was mispositioned. The root cause-'of this event was per-
,
,f sonnel error: the operator bumped the wrong switch during the test.
Licensee corrective. action involves listing the local' panel alarms ex-
>
pected+during the test. Applicable' procedures are to be corrected by March 1, 1989.
The inspector agreed that the' addition of expected alarms to the procedure should minimize the chance.for error. However, the use of the words " procedural deficiency" in the title is a misrepresentation
~
of what actually happened and should not be listed as part of the title.
>
The LER should descr.ibe what actually happened in each specific case,
,
This comment was noted by the licensee.
LER 88-26-00, Potential Damage to Safety-Related Equipment due to Design
--
Inadequacy, describes a sequence that could result.in a loss of both i
'
emergency busses. The inspector noted that a thorough licensee review of a startup test disclosed the potential for this undesirable coastdown scenario. However, the licensee's review was incomplete since it did
,
not consider the single failure of any plant component that could gene-i rate the same undesirable coastdown.
This item is described in detail in Section 7.0.
LER 88-27-00, Mislocated Firewatch due to Person.nel Error, describes a
--
failure to station a fire watch at the proper location.
Fire detection equipment for the HVAC unit in the south end of the ESF building failed its surveillance. The compensating fire watch checked the ESF building's south end but did not check the affected compartment.- The root cause i
of this event was personnel error: the Shift Supervisor (SS) misinter-preted the location of the HVAC unit.
The corrective action for this
.
LER involved counseling the SS on the importance of proper communications,
'
using the fire detection procedure to identify affected components, and changing the procedure to better identify affected components. No un-acceptable conditions were noted.
- _ - _ _ _ _ _ - - _ _ _
. - _
_--
_..
.
_- _ _ --
- _
_ -.
_
. - _ _ _
_ _ - _
_.
=
.
,
'
..
'
.
10.0 Surveillance Testing The inspector observed portions of surveillance tests to assess performance j
in accordance with approved procedures and Limiting Conditions of Operation,
"
removal and restoration of equipment, and deficiency review and resolution.
The following tests were reviewed:
SP 3440A01, Plant Startup Surveillance (NI Cal Check) dated 1/17/89.
--
,
'
--
SP 3601F.3, RCS Leakage, dated 12/29/88.
SP 3613F.1, Containment Purge and Exhaust Valve Checks, dated 1/24/88.
--
No inadequacies were noted.
l 11.0 Management Meetings Periodic meetings were held with station management to discuss inspection findings during the inspection period. A summary of findings was also dis-cussed at the conclusion of the inspection. A copy of the Table of Contents was given to the licensee to better understand the inspection areas covered.
No proprietary information was covered within the scope of the inspection.
Unless specifically listed in this report, no written material was given to the licensee during the inspection period.
I
!
!
. _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _