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=Text=
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{{#Wiki_filter:_ . . .                                                                                   . . _
{{#Wiki_filter:_ . . .
                                                                                                          l
. . _
                  .
l
                                                                                                          I
.
                    O                                UNITED STATES
UNITED STATES
              [WCEG 'o,$                 NUCLEAR REGULATORY COMMISSION
[WCEG
            [                                           REGION 11
O
            $ .,         j                     101 MARIETTA STREET.N.W.
'o,$
            *           2                       ATLANTA. GEORGI A 30323
NUCLEAR REGULATORY COMMISSION
              \...../
[
          Report Nos.:     50-369/85-38 and 50-370/85-39
REGION 11
          Licensee:   Duke Power Company                         -
$ .,
                        422 South Church Street
j
                        Charlotte, NC 28242
101 MARIETTA STREET.N.W.
          Docket Nos.:     50-369 and 50-370                   License Nos.: NPF-9 and NPF-17
*
          Facility Name: McGuire 1 and 2
2
          Inspection Conducted:   ,
ATLANTA. GEORGI A 30323
                                      Oct er 15-17, 1985 and January 27-31, 1986
\\...../
            Inspectors:
Report Nos.:
                      B. T. ' Debs
50-369/85-38 and 50-370/85-39
                                        '
Licensee:
                                                                                    [!
Duke Power Company
                                                                                    Dite' Signed
-
                                                                                                  M
422 South Church Street
                      F. McCoy
Charlotte, NC 28242
                      S. D. Stadler
Docket Nos.:
                      W. Poertner
50-369 and 50-370
                      P. Moore
License Nos.:
          Accompanying Personnel: Gr y on L. Yoder, Ph.D. (ORNL)
NPF-9 and NPF-17
          Approved by           hw '. M             m                            b             f?6
Facility Name: McGuire 1 and 2
                                fson, Acting Section Chief                           Da'te Signed
Inspection Conducted: Oct er 15-17, 1985 and January 27-31, 1986
                          B.W)ionofReactorSafety
,
                          Divis
Inspectors:
                                                      SUMMARY
[! M
            Scope: This routine, unannounced inspection was in the area of Nuclear Service
'
          Water System Operability.
B. T. ' Debs
            Results:   Five violations were identified.
Dite' Signed
F. McCoy
S. D. Stadler
W. Poertner
P. Moore
Accompanying Personnel: Gr y on L. Yoder, Ph.D. (ORNL)
Approved by
hw '. M
b
f?6
m
B.W)ionofReactorSafety
fson, Acting Section Chief
Da'te Signed
Divis
SUMMARY
Scope: This routine, unannounced inspection was in the area of Nuclear Service
Water System Operability.
Results:
Five violations were identified.
i
i
                                                                                                          i
i
                                                                                                          ,
,
          8607100458 860727
8607100458 860727
          PDR     ADOCK 05000369
PDR
          G                     PDR
ADOCK 05000369
                                                                _
G
                                                                                      . _ ,           ___
PDR
._
-
_
. _ ,
___


    _.                           ._ .                                       _
_.
                                      REPORT DETAILS
._ .
1.     Persons Contacted                           -
_
        Licensee Employees
REPORT DETAILS
      +G. Vaughn, General Manager, Nuclear Stations
1.
  *+T. L. McConnell, McGuire Nuclear Station Manager
Persons Contacted
  *+R. L. Gill, McGuire Licensing
-
  *+E. O. McCraw, Compliance Engineer
Licensee Employees
  *+W.     J. Kronenwetter, Design Engineer
+G. Vaughn, General Manager, Nuclear Stations
  *+W.     M. Suslick, Associate Engineer
*+T. L. McConnell, McGuire Nuclear Station Manager
        Other licensee employees contacted included construction craftsmen,
*+R. L. Gill, McGuire Licensing
        engineers, technicians, operators, mechanics, security force members, and
*+E. O. McCraw, Compliance Engineer
        office personnel.
*+W. J. Kronenwetter, Design Engineer
        NRC Resident Inspectors
*+W. M. Suslick, Associate Engineer
  *+W. Orders, Senior Resident Inspector
Other licensee employees contacted included construction craftsmen,
        R. Pierson, Resident Inspector
engineers, technicians, operators, mechanics, security force members, and
      * Attended exit interview on 10/17/85
office personnel.
      + Attended exit interview on 01/31/86
NRC Resident Inspectors
2.     Exit Interview
*+W. Orders, Senior Resident Inspector
                                                                                    ,
R. Pierson, Resident Inspector
        The inspection scope and findings were summarized on October 17, 1985, and
* Attended exit interview on 10/17/85
        January 31, 1986, with those persons indicated in paragraph 1 above. The
+ Attended exit interview on 01/31/86
        inspector described the areas inspected and discussed in detail the inspec-
2.
        tion findings. No dissenting comments were received from the licensee.
Exit Interview
        The results of the inspection were discussed with utility management during
,
        a meeting in Atlanta on March 14, 1986. The details of this meeting are
The inspection scope and findings were summarized on October 17, 1985, and
        documented in Section 11 of this report.
January 31, 1986, with those persons indicated in paragraph 1 above.
        During the exit interview the enforcement findings were presented as
The
        preliminary and unresolved. Following NRC management review, the following
inspector described the areas inspected and discussed in detail the inspec-
        findings were determined:
tion findings.
        369/85-38-01, 370/85-39-01 Violation - Failure to adequately perform
No dissenting comments were received from the licensee.
        preoperational test on control room chiller - see paragraphs 7 and 8.
The results of the inspection were discussed with utility management during
        369/85-38-02, 370/85-39-02 Violation - Failure to implement and maintain
a meeting in Atlanta on March 14, 1986. The details of this meeting are
        procedures - see paragraphs 7 and 8.
documented in Section 11 of this report.
        369/85-38-03, 370/85-39-03 Violation - Failure to meet Technical Specifica-
During the exit interview the enforcement findings were presented as
        tion 3.7.4 for RN system operability - see paragraph 7.
preliminary and unresolved. Following NRC management review, the following
        369/85-38-04, 370/85-39-04 Violation - Failure to perform 10 CFR 50.59
findings were determined:
        evaluation on degraded equipment - see paragraph 8.
369/85-38-01, 370/85-39-01 Violation - Failure to adequately perform
preoperational test on control room chiller - see paragraphs 7 and 8.
369/85-38-02, 370/85-39-02 Violation - Failure to implement and maintain
procedures - see paragraphs 7 and 8.
369/85-38-03, 370/85-39-03 Violation - Failure to meet Technical Specifica-
tion 3.7.4 for RN system operability - see paragraph 7.
369/85-38-04, 370/85-39-04 Violation - Failure to perform 10 CFR 50.59
evaluation on degraded equipment - see paragraph 8.


r
r
                                          2
2
    369/85-38-05, 370/85-39-05 Violation - Failure to identi fy and correct
369/85-38-05, 370/85-39-05 Violation - Failure to identi fy and correct
    conditions adverse to quality as * required by 10 CFR 50, Appendix B,
conditions adverse to quality as * required by 10 CFR 50, Appendix B,
    Criterion XVI - see paragraph 12.
Criterion XVI - see paragraph 12.
    369/85-38-06, 370/85-39-06 Unresolved Item - NRC followup of licensee
369/85-38-06, 370/85-39-06 Unresolved Item - NRC followup of licensee
    response of April 25, 1986 - see paragraph 11.
response of April 25, 1986 - see paragraph 11.
  3. Licensee Action on Previous Enforcement Matters
3.
    This subject was not addressed in the inspection.
Licensee Action on Previous Enforcement Matters
  4. Unresolved Items
This subject was not addressed in the inspection.
    An Unresolved Item is a matter abcut which more information is required to
4.
    determine whether it is acceptable or may involve a violation or deviation.
Unresolved Items
    A new unresolved item identified during this inspection is discussed in
An Unresolved Item is a matter abcut which more information is required to
    Section 11.
determine whether it is acceptable or may involve a violation or deviation.
  5. Nuclear Service Water System Description
A new unresolved item identified during this inspection is discussed in
    The McGuire Final Safety Analysis Report (FSAR) states that the Nuclear
Section 11.
    Service Water (RN) System provides assured cooling water for various
5.
    Auxiliary Building and Reactor Building heat exchangers during all phases
Nuclear Service Water System Description
    of station operations.     Each unit has two redundant " essential headers"
The McGuire Final Safety Analysis Report (FSAR) states that the Nuclear
    serving two trains of equipment necessary for safe station shutdown, and a
Service Water (RN) System provides assured cooling water for various
    "non-essential header" serving equipment not required for safe shutdown. In
Auxiliary Building and Reactor Building heat exchangers during all phases
    conjunction with the Ultimate Heat Sink, comprised of Lake Norman and the
of station operations.
    Standby Nuclear Service Water Pond (SNSWP), the RN System is designed to
Each unit has two redundant " essential headers"
    meet design flow rates and pressure heads for normal station operation and
serving two trains of equipment necessary for safe station shutdown, and a
    also those flow rates and pressure heads required for safe station shutdown
"non-essential header" serving equipment not required for safe shutdown. In
    normally or as the result of a postulated Loss of Coolant Accident (LOCA).
conjunction with the Ultimate Heat Sink, comprised of Lake Norman and the
    The system is further designed to tolerate a single failure following a
Standby Nuclear Service Water Pond (SNSWP), the RN System is designed to
    LOCA, and/or seismic event causing loss of Lake Norman, and/or loss of
meet design flow rates and pressure heads for normal station operation and
      station power plus offsite power (station blackout). Sufficient margin is
also those flow rates and pressure heads required for safe station shutdown
    provided in the equipment design to accommodate anticipated corrosion and
normally or as the result of a postulated Loss of Coolant Accident (LOCA).
      fouling without degradation of system performance,                           j
The system is further designed to tolerate a single failure following a
  6. Summary of NRC Findings
LOCA, and/or seismic event causing loss of Lake Norman, and/or loss of
                                                                                  i
station power plus offsite power (station blackout). Sufficient margin is
    On October 4, 1985, the NRC Senior Resident Inspector reported to Region II   1
provided in the equipment design to accommodate anticipated corrosion and
    management that the 1A nuclear service water system, designated by the       j
fouling without degradation of system performance,
      licensee as the RN system, had failed to meet the acceptance criteria of
j
      its quarterly inservice test. Although the Technical Specificttion Action
6.
      Statement period of 72 hours expired on October 7,1985, both units con-
Summary of NRC Findings
      tinued operation at full power based on the licensee's contention that
i
      the 1A RN pump had been made operable by cross connecting it with the Unit 2
On October 4, 1985, the NRC Senior Resident Inspector reported to Region II
      2A RN pump. On October 10, 1985, NRC informed Duke Power Company (DPC) that
management that the 1A nuclear service water system, designated by the
      operation in the unit shared mode was an unacceptable unanalyzed condition.
j
      DPC restored unit separation and began justification for continuing opera-
licensee as the RN system, had failed to meet the acceptance criteria of
      tion with the apparently degraded pump.
its quarterly inservice test. Although the Technical Specificttion Action
                                                                                  1
Statement period of 72 hours expired on October 7,1985, both units con-
tinued operation at full power based on the licensee's contention that
the 1A RN pump had been made operable by cross connecting it with the Unit 2
2A RN pump. On October 10, 1985, NRC informed Duke Power Company (DPC) that
operation in the unit shared mode was an unacceptable unanalyzed condition.
DPC restored unit separation and began justification for continuing opera-
tion with the apparently degraded pump.


  7_
7_
      _ . _ _ _ _ _ _ _ _ _ _         _
_ . _ _ _ _ _ _ _ _ _ _
                                                    .
_
                                                          . _ .
.
                                                                    ..     . . . . ..           ..
. _
                                                                                                              .
.
                                                                                                                  .
..
                                                                  3
. . . .
                              Licensee representatives stated that they suspected that the pump was not
..
                              actually degraded, rather the pump discharge line flow orifice reading
..
                              was in error. One of the possible reasons stated was buildup of silt,
.
                              mud, or corrosion at the orifice.         Licensee representatives subsequently
.
                              stated several months later that the flow indication was erroneous and the
3
                              pump was not actually degraded.
Licensee representatives stated that they suspected that the pump was not
                              The NRC became concerned that if system fouling was that bad at the pump
actually degraded, rather the pump discharge line flow orifice reading
                              discharge, what was the status of the downstream components, especially heat
was in error. One of the possible reasons stated was buildup of silt,
                              exchangers. A reactive inspection was conducted October 15-17, 1985, to
mud, or corrosion at the orifice.
                              review these matters. Numerous phone conferences and letters were exchanged
Licensee representatives subsequently
                              in ensuing months, and a followup inspection was conducted January 27-31,
stated several months later that the flow indication was erroneous and the
                              1986.                                                                               l
pump was not actually degraded.
                              A summary of the major NRC findings presented in this report are as
The NRC became concerned that if system fouling was that bad at the pump
                              follows:
discharge, what was the status of the downstream components, especially heat
                              a.   Preop tests and subsequent surveillance tests performed in 1979 were
exchangers.
                                    not adequate to ascertain operability of RN components.
A reactive inspection was conducted October 15-17, 1985, to
                                        Several test procedures did not contain acceptance criteria. For
review these matters. Numerous phone conferences and letters were exchanged
                                        example, a quarterly test of RN heat exchanger 1-A on October 7,
in ensuing months, and a followup inspection was conducted January 27-31,
                                        1985, indicated potential fouling but the test procedure contained
1986.
                                        no acceptance criteria.     The potential fouling was apparently
l
                                        pursued only because of questions from the NRC and not addressed
A summary of the major NRC findings presented in this report are as
                                        by the licensee until October 14 when it was attributed to a
follows:
                                        faulty flow instrument. The heat exchanger was assumed to be
a.
;                                       operable during this period of evaluation.
Preop tests and subsequent surveillance tests performed in 1979 were
                                    *
not adequate to ascertain operability of RN components.
                                        Flow was not measured through control room air conditioner heat
Several test procedures did not contain acceptance criteria. For
                                        exchangers.
example, a quarterly test of RN heat exchanger 1-A on October 7,
                                        Test results were recorded in units of differential pressure when
1985, indicated potential fouling but the test procedure contained
                                        acceptance criteria were in units of flow rate.
no acceptance criteria.
                                    *
The potential fouling was apparently
                                        Heat transfer characteristics of heat exchangers were not normally
pursued only because of questions from the NRC and not addressed
                                        determined. Fouling factors or empirical tests could have been
by the licensee until October 14 when it was attributed to a
                                        used.
faulty flow instrument.
                                    *
The heat exchanger was assumed to be
                                        RN system was not originally preop tested in the most limiting
;
                                        post-LOCA configuration in that both trains were not aligned to
operable during this period of evaluation.
                                        simultaneously draw water from the Standby Nuclear Service Water
*
                                        Pond.
Flow was not measured through control room air conditioner heat
l                             b.   The positions of valves specified in preop test data were different
exchangers.
                                    from the positions in operating procedures.
Test results were recorded in units of differential pressure when
acceptance criteria were in units of flow rate.
*
Heat transfer characteristics of heat exchangers were not normally
determined.
Fouling factors or empirical tests could have been
used.
*
RN system was not originally preop tested in the most limiting
post-LOCA configuration in that both trains were not aligned to
simultaneously draw water from the Standby Nuclear Service Water
Pond.
l
b.
The positions of valves specified in preop test data were different
from the positions in operating procedures.
l
l
                              c.
'
'
c.
                                    The RN system had not been flow balanced since 1982 even though engi-
The RN system had not been flow balanced since 1982 even though engi-
                                    neering documents required it to be.
neering documents required it to be.
i
i
                                                                                                                a
a


                                                                .--       -               - -       . . - .         .   . . _
.--
                                                                                              4
-
- -
. . - .
.
. . _
4
2
2
                                                    d.   The following heat exchanger fouling problems had occurred:
d.
                                                                  Containment spray heat exchanger 1A tested per IEB 81-03 showed
The following heat exchanger fouling problems had occurred:
;                                                                 increasing delta P from 20 psid- in 1983 to 29 psid in 1985.
Containment spray heat exchanger 1A tested per IEB 81-03 showed
                                                                  In October 1982, a containment ventilation heat exchanger would
;
                                                                  not function due to fouling.
increasing delta P from 20 psid- in 1983 to 29 psid in 1985.
                                                                  Periodic cleaning of control room air conditioning heat exchangers
In October 1982, a containment ventilation heat exchanger would
                                                                  had been necessary since 1982 due to fouling.
not function due to fouling.
                                                                  RCP motor coolers required cleaning three times during the period
Periodic cleaning of control room air conditioning heat exchangers
                                                                  1984 - 1985.
had been necessary since 1982 due to fouling.
                                                                  Unit 1 component cooling water heat exchanger observed to be
RCP motor coolers required cleaning three times during the period
                                                                  fouled in September 1984.
1984 - 1985.
;                                                   e.   Inservice testing of the 1A RN pump indicated degraded flow on
Unit 1 component cooling water heat exchanger observed to be
i                                                         October 4, 1985. Instead of entering a Technical Specification Action
fouled in September 1984.
                                                          Statement which would have required the operating unit to be brought       I
;
                                                          to the hot standby mode within six hours, the licensee inappropriately
e.
                                                          cross-connected RN train A and train B and continued to operate.
Inservice testing of the 1A RN pump indicated degraded flow on
                                                    f.   A flow balance test on RN train IA conducted on December 17, 1985,
i
                                                          revealed flow rates through several safety-related heat exchangers
October 4, 1985. Instead of entering a Technical Specification Action
                                                          to be below FSAR values. At the request of the NRC in January 1986,
Statement which would have required the operating unit to be brought
                                                          the licensee evaluated these test results pursuant to 10 CFR 50.59.
to the hot standby mode within six hours, the licensee inappropriately
                                                          This evaluation, which was based upon heat transfer tests by DPC and
cross-connected RN train A and train B and continued to operate.
                                                          calculations by Westinghouse, was completed and justified continued
f.
                                                          operation on January 14, 1986. .The licensee apparently assumed the
A flow balance test on RN train IA conducted on December 17, 1985,
                                                          system to be operable between December 17 and January 14.
revealed flow rates through several safety-related heat exchangers
to be below FSAR values. At the request of the NRC in January 1986,
the licensee evaluated these test results pursuant to 10 CFR 50.59.
This evaluation, which was based upon heat transfer tests by DPC and
calculations by Westinghouse, was completed and justified continued
operation on January 14, 1986. .The licensee apparently assumed the
system to be operable between December 17 and January 14.
I
Although it appears that RN heat exchangers were becoming progressively
'
'
I                                                  Although it appears that RN heat exchangers were becoming progressively
more fouled with time, the licensee did not recognize the symptoms or place
                                                    more fouled with time, the licensee did not recognize the symptoms or place
priority consideration on the overall system operability and associated
                                                    priority consideration on the overall system operability and associated
safety concerns.
                                                      safety concerns. Rather fouled components required for continued operation
Rather fouled components required for continued operation
                                                    were cleaned as needed but no regard shown for the status of dormant safety
were cleaned as needed but no regard shown for the status of dormant safety
                                                    equipment, such as the containment spray heat exchangers.
equipment, such as the containment spray heat exchangers.
                                                    When the concern was raised by the NRC, the licensee devoted significant
When the concern was raised by the NRC, the licensee devoted significant
                                                      resources toward correcting the problem. As a result, during the months
resources toward correcting the problem. As a result, during the months
                                                    of investigation, there were several instances when individual components
of investigation, there were several instances when individual components
!                                                   were found not to be capable of FSAR specified performance. On these
!
                                                    occasions, the licensee revised their accident analysis supporting calcula-
were found not to be capable of FSAR specified performance.
                                                      tions to justify continued power operation. This mode of operation complies
On these
                                                    with regulatory requirements but does not appear to represent to the NRC the
occasions, the licensee revised their accident analysis supporting calcula-
                                                    most conservative safety philosophy.
tions to justify continued power operation. This mode of operation complies
,
with regulatory requirements but does not appear to represent to the NRC the
most conservative safety philosophy.
,
4
4
  _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ . . _ _
.
.
.
. .


  - -         - _ . . .                 -           - -   - .     - - -       -
-
                                              5
-
      7. McGuire Nuclear Service Water System History
- _ .
          1979
. .
          Preoperational functional testing was completed by the licensee on July 25,
-
          1979, for the Unit 1 RN system and on November 12, 1982, for the Unit 2 RN
-
;         system. In January 1986, NRC Region II inspectors reviewed selected areas
-
- .
- - -
-
5
7.
McGuire Nuclear Service Water System History
1979
Preoperational functional testing was completed by the licensee on July 25,
1979, for the Unit 1 RN system and on November 12, 1982, for the Unit 2 RN
;
system.
In January 1986, NRC Region II inspectors reviewed selected areas
of preoperational test packages for both Units 1 and 2 RN systems.
,
,
          of preoperational test packages for both Units 1 and 2 RN systems.
:
:
l         It was noted that during the conduct of the Unit 1 preoperational tests
l
j         of nuclear service water, the safety evaluation section (8) of the major
It was noted that during the conduct of the Unit 1 preoperational tests
1        procedure form was marked as not applicable. Administrative Plant Manual,
j
;       Section 4.2.4.1(e) requires that prior to procedure use, a safety evaluation
of nuclear service water, the safety evaluation section (8) of the major
;         of major changes to a procedure shall be performed. Examples of the major
procedure form was marked as not applicable.
          changes made to the preoperational procedures included changes to the
Administrative Plant Manual,
          minimum acceptable RN flow criteria, initial RN system configuration at
1
;
Section 4.2.4.1(e) requires that prior to procedure use, a safety evaluation
;
of major changes to a procedure shall be performed.
Examples of the major
changes made to the preoperational procedures included changes to the
minimum acceptable RN flow criteria, initial RN system configuration at
test initiation, and the methods utilized to determine component flows.
>
>
          test initiation, and the methods utilized to determine component flows.
The use of "not applicable" for safety evaluations was allowed by a licensee
          The use of "not applicable" for safety evaluations was allowed by a licensee
internal memorandum dated September 14, 1979.
          internal memorandum dated September 14, 1979.     The memorandum deleted the
The memorandum deleted the
          procedural requirement for a safety evaluation prior to fuel load.
procedural requirement for a safety evaluation prior to fuel load.
The primary objective of the nuclear service water preoperational functional
,
,
          The primary objective of the nuclear service water preoperational functional
test was to verify that the system could supply designed cooling water flow
          test was to verify that the system could supply designed cooling water flow
j
j         to variou., components and to set each component throttle valve to provide
to variou., components and to set each component throttle valve to provide
          the proper flow rate. Adequate system and component flow was to be verified
the proper flow rate. Adequate system and component flow was to be verified
j         for all modes of operation.
j
for all modes of operation.
i
i
i
i        One of the safety related RN loads during post-LOCA conditions is the
One of the safety related RN loads during post-LOCA conditions is the
l         control room air conditioner which requires a minimum flow of 789 GPM as
l
j         stated in McGuire FSAR Table 9.2.2-1(8). During the RN preoperational
control room air conditioner which requires a minimum flow of 789 GPM as
,
j
          test for Unit 1, the flow to the control room air conditioning was unable
stated in McGuire FSAR Table 9.2.2-1(8).
During the RN preoperational
test for Unit 1, the flow to the control room air conditioning was unable
,
I
I
          to be determined due to problems encountered with the installed instrumen-
to be determined due to problems encountered with the installed instrumen-
l         tation. Subsequently, a major change to the preoperational test procedure,
l
!       TP/1/A/1400/01, was approved by the licensee to delete the requirement to
tation. Subsequently, a major change to the preoperational test procedure,
          verify the minimum RN flow of 789 GPM.     The change to the preoperational
!
          test was justified by the licensee on the basis that the flow control valve
TP/1/A/1400/01, was approved by the licensee to delete the requirement to
          is air operated and fails open during accident conditions. This justifi-
verify the minimum RN flow of 789 GPM.
          cation assumed that there were no internal obstructions and that the wide
The change to the preoperational
          open valve flow would meet or exceed the FSAR~ required flow. Due to this
test was justified by the licensee on the basis that the flow control valve
          procedure revision, the subsequent RN preoperational test for Unit 2
is air operated and fails open during accident conditions.
          also did not verify adequate flow to the control room air conditioning.
This justifi-
          As stated later in this report, subsequent functional flow test data
cation assumed that there were no internal obstructions and that the wide
;         obtained in late 1985 and early 1986 indicated that the required 789 GPM
open valve flow would meet or exceed the FSAR~ required flow. Due to this
,        was not being met. Failure to test the aforementioned component represents
procedure revision, the subsequent RN preoperational test for Unit 2
also did not verify adequate flow to the control room air conditioning.
As stated later in this report, subsequent functional flow test data
;
obtained in late 1985 and early 1986 indicated that the required 789 GPM
was not being met. Failure to test the aforementioned component represents
,
1
1
          a violation of 10 CFR 50, Appendix B, Criteria XI which requires a test
a violation of 10 CFR 50, Appendix B, Criteria XI which requires a test
          program to be established to assure that all testing required to demon-
program to be established to assure that all testing required to demon-
a        strate system components perform satisfactorily in service (369/85-38-01,
strate system components perform satisfactorily in service (369/85-38-01,
          370/85-39-01).
a
'_
370/85-39-01).
'
--
-
- - - -
. -
- - -
- -


  -. .                             .   _                   -                     _--
-. .
                                                                                                  ,
.
                                                6                                                 ,
_
          The inspector noted that in several instances during the conduct of the
-
          preoperational tests of the RN system, the measured flows were stated
_--
          as differential pressure (psid) rather than flow (GPM). The engineers who
,
          performed the tests and the preoperational logs indicated this was due to               [
6
          problems experienced with the instaihd flow instrumentation. To continue               !
,
          the tests with the inoperable flow instrumentation, the licensee utilized               i
The inspector noted that in several instances during the conduct of the
          temporary differential pressure instrumentation.           The conversion from         j
preoperational tests of the RN system, the measured flows were stated
          differential pressure to GPM was not made on test data enclosures. To
as differential pressure (psid) rather than flow (GPM). The engineers who
          verify that the minimum FSAR flow results were achieved for the RN compo-
performed the tests and the preoperational logs indicated this was due to
          nents preoperationally tested, the inspector, in early 1986, requested that             ;
[
          the licensee convert the differential pressures to flows. In each case it               I
problems experienced with the instaihd flow instrumentation. To continue
          was verified, based on the licensee's calculations, that the minimum accept-           #
!
          able flow rates had been achieved as stated in McGuire FSAR table 9.2.2-1.
the tests with the inoperable flow instrumentation, the licensee utilized
          The values from that table appear later in this report.                                 ,
i
          To assure minimum RN component flows, including adequate flow to the con-               I
temporary differential pressure instrumentation.
          tainment spray heat exchangers during design LOCA conditions, the normally             :
The conversion from
                                                                                                  '
j
          throttled valves associated with each RN component were required to be set
differential pressure to GPM was not made on test data enclosures.
          during p'reoperational testing of the RN system.     These throttled positions         ,
To
          established during preoperational testing were to be incorporated into                 l
verify that the minimum FSAR flow results were achieved for the RN compo-
          operating and surveillance procedures to protect these throttled settings               !
nents preoperationally tested, the inspector, in early 1986, requested that
                                                                                                  '
;
          during future operations. The inspectors noted that, in some cases and
the licensee convert the differential pressures to flows. In each case it
          particularly for Unit 1, the throttled valve positions listed in the                   ;
I
                                                                                                  '
#
          licensee's RN operating procedures and their locked valve verification
was verified, based on the licensee's calculations, that the minimum accept-
          procedures were not consistent with earlier preoperational "as left" data.
able flow rates had been achieved as stated in McGuire FSAR table 9.2.2-1.
          It was noted, however, for those throttled valves reviewed, the operational
The values from that table appear later in this report.
          positions were further open than the "as left" preoperational test posi-
,
          tions.     The licensee acknowledged these discrepancies and committed to
I
          revise the operational procedures to meet those valve settings established             ;
To assure minimum RN component flows, including adequate flow to the con-
          during recent 1985 and 1986 RN flow testing.
tainment spray heat exchangers during design LOCA conditions, the normally
          The inspection team noted that since 1976 the licensee has had a functional             r
:
          system description for the RN system. Section 5 of this system description             :
'
          (MCSD-0138.00) states that annually each essential RN train must be checked           !
throttled valves associated with each RN component were required to be set
          for proper throttling. Also, after any throttle valve is repositioned, the
during p'reoperational testing of the RN system.
          entire train must be checked for proper throttling. The system description.           A
These throttled positions
          then presents a detailed procedure to verify that the minimum flow condi-
,
          tions for operability of the safety related portion of the system are met.             ,
established during preoperational testing were to be incorporated into
l
operating and surveillance procedures to protect these throttled settings
!
'
during future operations. The inspectors noted that, in some cases and
particularly for Unit 1,
the throttled valve positions listed in the
;
licensee's RN operating procedures and their locked valve verification
'
procedures were not consistent with earlier preoperational "as left" data.
It was noted, however, for those throttled valves reviewed, the operational
positions were further open than the "as left" preoperational test posi-
tions.
The licensee acknowledged these discrepancies and committed to
revise the operational procedures to meet those valve settings established
;
during recent 1985 and 1986 RN flow testing.
The inspection team noted that since 1976 the licensee has had a functional
r
system description for the RN system. Section 5 of this system description
:
(MCSD-0138.00) states that annually each essential RN train must be checked
!
for proper throttling. Also, after any throttle valve is repositioned, the
entire train must be checked for proper throttling. The system description.
A
then presents a detailed procedure to verify that the minimum flow condi-
tions for operability of the safety related portion of the system are met.
The licensee had decided not to adopt the aforementioned recommendations.
,
l
4
4
          The licensee had decided not to adopt the aforementioned recommendations.              l
Consequently, no RN flow balances had been performed - beyond 1982 until
          Consequently, no RN flow balances had been performed - beyond 1982 until
'
'
          requested by the NRC in late 1985.       Functional system descriptions are
requested by the NRC in late 1985.
Functional system descriptions are
not used as procedures by licensees and, consequently, failure to follow
2
2
          not used as procedures by licensees and, consequently, failure to follow
'
'
MCSD-0138.00 is not considered to be a violation.
          MCSD-0138.00 is not considered to be a violation.           However, compliance
However, compliance
          with this document would have prevented the above violation.       However, The
with this document would have prevented the above violation.
          requirements to verify proper throttling position should have been in plant
However, The
          procedures.                                                                           ;
requirements to verify proper throttling position should have been in plant
          Failure to measure flow through components and failure to specify post-
procedures.
          tions of throttled valves in procedures represent examples of inadequate
;
        - .       _
Failure to measure flow through components and failure to specify post-
                                _
tions of throttled valves in procedures represent examples of inadequate
                                            .           _.       .-           --.     ,- . - ,
-
.
_
_
.
_.
.-
--.
,-
. - ,


                                      7
7
procedural controls and are, therefore, a violation of McGuire Technical
procedural controls and are, therefore, a violation of McGuire Technical
Specification 6.8.1 and 10 CFR 50, Appendix B, Criteria V which requires
Specification 6.8.1 and 10 CFR 50, Appendix B, Criteria V which requires
Line 375: Line 526:
tive methods to control these valves. Currently, these valves are verified
tive methods to control these valves. Currently, these valves are verified
locked every six months under the Locked Valve Verification Procedure
locked every six months under the Locked Valve Verification Procedure
4700/23. In addition, independent verification is utilized to ensure that
4700/23.
In addition, independent verification is utilized to ensure that
the valves are returned to the proper position following valve repositioning
the valves are returned to the proper position following valve repositioning
for maintenance or other activities.       Despite these positive controls, the
for maintenance or other activities.
Despite these positive controls, the
inspectors noted the following recent deficiencies in the licensee's control
inspectors noted the following recent deficiencies in the licensee's control
over these throttle valves:
over these throttle valves:
-
-
      The Locked Valve Verification Procedure requires that the operator
The Locked Valve Verification Procedure requires that the operator
      verify the valve to be locked. No verification of the actual throttle
verify the valve to be locked. No verification of the actual throttle
      position is required.
position is required.
-
-
      The valve locks utilized for RN throttled valves are chain locks.
The valve locks utilized for RN throttled valves are chain locks.
      These chain locks work well for wide open valves, but the slack in the
These chain locks work well for wide open valves, but the slack in the
      chains cannot ensure that a valve remains open 1/4 turn. A valve that
chains cannot ensure that a valve remains open 1/4 turn. A valve that
      is required to be open 1/4 to 3/8 turn could be locked in the full-
is required to be open 1/4 to 3/8 turn could be locked in the full-
      closed position without detection.
closed position without detection.
One potential solution identified by the licensee for better control of
One potential solution identified by the licensee for better control of
these throttle valves include the use of locking collars which are used on
these throttle valves include the use of locking collars which are used on
throttle valves in other systems.     Since the locking collars can be sized
throttle valves in other systems.
to ensure the exact valve opening desired, their use would provide positive
Since the locking collars can be sized
indication of valve position.
to ensure the exact valve opening desired, their use would provide positive
indication of valve position.
The licensee initiated a 10 CFR 50.72 notification to the NRC stating that
The licensee initiated a 10 CFR 50.72 notification to the NRC stating that
prior to January 27, 1986, the RN systems for Unit 1 and 2 had never been
prior to January 27, 1986, the RN systems for Unit 1 and 2 had never been
tested under the requisite design basis accident configuration.       Specifi-
tested under the requisite design basis accident configuration.
cally, the system valves had never been positioned to supply the required
Specifi-
flow to essential headers for Units 1 and 2 with the system taking suction
cally, the system valves had never been positioned to supply the required
solely from the Nuclear Service Water Pond.       This issue is discussed in
flow to essential headers for Units 1 and 2 with the system taking suction
Section 6. of this report.
solely from the Nuclear Service Water Pond.
1981
This issue is discussed in
In response to IE Bulletin 81-03 which addressed the potential fouling of
Section 6. of this report.
safety related heat exchangers by clam and shell debris, the licensee com-
1981
mitted to the NRC to monitor two RN supplied heat exchangers on a quarterly
In response to IE Bulletin 81-03 which addressed the potential fouling of
basis. One of these heat exchangers is the 1A Containment Spray (NS) heat
safety related heat exchangers by clam and shell debris, the licensee com-
exchanger.   Additionally in the licensee's response, it was stated that
mitted to the NRC to monitor two RN supplied heat exchangers on a quarterly
"if significant fouling is detected on these heat exchangers, other heat
basis. One of these heat exchangers is the 1A Containment Spray (NS) heat
exchangers in the RN system will be inspected." The licensee performed
exchanger.
their monitoring under procedure PT/1/A/4403/04. This procedure for the 1A
Additionally in the licensee's response, it was stated that
NS heat exchanger requires that the test be performed for a FSAR accident
"if significant fouling is detected on these heat exchangers, other heat
exchangers in the RN system will be inspected." The licensee performed
their monitoring under procedure PT/1/A/4403/04. This procedure for the 1A
NS heat exchanger requires that the test be performed for a FSAR accident


                        . _ .     - _ -         =           _ -.-                                   -_
. _ .
!
- _ -
=
_ -.-
-_
!
!
!
;
;
                                                            8
8
4
4
i
i
    RN flow of 5000 GPM to the NS heat exchanger and that the heat exchanger
RN flow of 5000 GPM to the NS heat exchanger and that the heat exchanger
'
differential pressure (D/P) be recorded.
In October 1985, the inspectors
,
,
'
reviewed the past test data which indicated the following:
    differential pressure (D/P) be recorded.                            In October 1985, the inspectors
DATE OF TEST
    reviewed the past test data which indicated the following:
D/P (PSID)
              DATE OF TEST                           D/P (PSID)
l
l             6/20/83                                   20
6/20/83
!             9/22/83                               Not Available
20
              10/2/83                                   23.5
!
!             1/18/84                                   25
9/22/83
              4/11/84                                   23
Not Available
              7/18/84                                   25
10/2/83
                                                        29.5
23.5
!
1/18/84
25
4/11/84
23
7/18/84
25
11/9/84
29.5
'
'
              11/9/84
2/28/85
              2/28/85                                   23.5
23.5
              6/27/85                                   25
6/27/85
              *10/7/85                                   29
25
  *RN flow was 4600 GPM
*10/7/85
      The test procedure did not specify criteria for determining "significant
29
      fouling" and, tnus, other components were not inspected as a result of
*RN flow was 4600 GPM
The test procedure did not specify criteria for determining "significant
fouling" and, tnus, other components were not inspected as a result of
these tests.
Further discussion of these findings appears later in this
'
'
      these tests.            Further discussion of these findings appears later in this
report under the section titled 1985.
      report under the section titled 1985.
l
l
1982
        1982
On October 22, 1982, the licensee identified that fouling of the RN supplied
      On October 22, 1982, the licensee identified that fouling of the RN supplied                                   ,
,
        lower containment ventilation heat exchangers was a problem which was
lower containment ventilation heat exchangers was a problem which was
      causing unacceptable temperature increases in the lower containment areas.
causing unacceptable temperature increases in the lower containment areas.
      This subsequently forced the units to operate at reduced reactor power
This subsequently forced the units to operate at reduced reactor power
      during certain seasonal conditions.                     In April 1983, the licensee attempted
during certain seasonal conditions.
      to add a penetrant / dispersant to the RN system in an attempt to clean lower
In April 1983, the licensee attempted
:     containment cooling units.                   The attempt was ineffectual. Eventually the                     ,
to add a penetrant / dispersant to the RN system in an attempt to clean lower
        licensee modified the coolers with a self-cleaning mechanism which corrected
:
        the problem.
containment cooling units.
      As a result of a control room air conditioning trip due to fouling of the RN
The attempt was ineffectual.
        supplied, safety related air conditioning chillers, the licensee established
Eventually the
        a cleaning threshold based on increasing air conditioning condenser pres-                                   i
,
        sures. On the following dates, these chillers have been rodded out to
licensee modified the coolers with a self-cleaning mechanism which corrected
      maintain operability.
the problem.
              TRAIN A                                     TRAIN B
As a result of a control room air conditioning trip due to fouling of the RN
              11/19/82                                           3/83
supplied, safety related air conditioning chillers, the licensee established
!             10/03/83                                     01/07/85
a cleaning threshold based on increasing air conditioning condenser pres-
              12/19/83                                     10/21/85
i
              05/30/84                                     11/05/85                                               i
sures. On the following dates, these chillers have been rodded out to
                10/31/84                                                                                             l
maintain operability.
              09/25/85                                                                                             l
TRAIN A
                10/24/85                                                                                             i
TRAIN B
              10/31/85
11/19/82
                                                                                                                    ,
3/83
    , - . - -         -         -
!
                                          -- ----.       .,                   , -     - - . - - , ,,     , - . . - -
10/03/83
01/07/85
12/19/83
10/21/85
05/30/84
11/05/85
i
10/31/84
l
09/25/85
l
10/24/85
i
10/31/85
,
, - . - -
-
-
-- ----.
.,
, -
- - . - - ,
,,
, - . . - -


                                            .
.
                                      9
9
  1984
1984
  In March 1984, the licensee began development of a heat exchanger perform-
In March 1984, the licensee began development of a heat exchanger perform-
  ance monitoring program. At the time of this inspection, Duke Power Company
ance monitoring program. At the time of this inspection, Duke Power Company
  had not fully implemented this program at their nuclear plant.
had not fully implemented this program at their nuclear plant.
  The inspector reviewed the section of the program which requires monitoring
The inspector reviewed the section of the program which requires monitoring
  the performance of heat exchangers such as those in the RN system. The
the performance of heat exchangers such as those in the RN system.
  program appeared to be very comprehensive with provisions for monitoring
The
  both flows and heat transfer capabilities, for increasing the frequency of
program appeared to be very comprehensive with provisions for monitoring
  monitoring as warranted, and for initiating corrective actions as necessary.
both flows and heat transfer capabilities, for increasing the frequency of
  Once fully implemented, this Performance Monitoring Program will be a major
monitoring as warranted, and for initiating corrective actions as necessary.
  improvement in the licensee's ability to monitor plant equipment perfor-
Once fully implemented, this Performance Monitoring Program will be a major
  mance and to promptly identify degraded performance. A key to the relative
improvement in the licensee's ability to monitor plant equipment perfor-
+ success of the program, however, will be the effectiveness and timeliness
mance and to promptly identify degraded performance. A key to the relative
  of corrective actions taken in response to an identified deficiency.       The
success of the program, however, will be the effectiveness and timeliness
  inspector noted that this corporate monitoring program was scheduled to be
+
  implemented in stages at the various plants. The RN heat exchangers were
of corrective actions taken in response to an identified deficiency.
  scheduled for performance monitoring implementation during the second phase
The
  of the program which will be several months into 1986.       As a result of
inspector noted that this corporate monitoring program was scheduled to be
  the fouling and degraded performance being experienced with the RN heat
implemented in stages at the various plants. The RN heat exchangers were
  exchangers and concerns expressed by the NRC, the licensee indicated this
scheduled for performance monitoring implementation during the second phase
  phase of the program will be implemented on a priority basis.
of the program which will be several months into 1986.
  Also in 1984, the licensee began to experience RN fouling problems in
As a result of
  their reactor coolant pump motor coolers. The licensee has performed the
the fouling and degraded performance being experienced with the RN heat
  following cleanings of these coolers on the dates indicated:
exchangers and concerns expressed by the NRC, the licensee indicated this
phase of the program will be implemented on a priority basis.
Also in 1984, the licensee began to experience RN fouling problems in
their reactor coolant pump motor coolers.
The licensee has performed the
following cleanings of these coolers on the dates indicated:
'
'
        UNIT 1                                   UNIT 2
UNIT 1
        12/31/84                                 8/10/84
UNIT 2
                                                11/08/85
12/31/84
  In September 1984, the licensee evaluated the Unit 1 Component Cooling (KC)
8/10/84
  heat exchanger for fouling, although, according to the licensee, there was
11/08/85
  no indication of reduced heat transfer or high differential pressure. As
In September 1984, the licensee evaluated the Unit 1 Component Cooling (KC)
  part of the evaluation, DPC engineering calculated a fouling factor for the
heat exchanger for fouling, although, according to the licensee, there was
  KC heat exchangers. These calculations were based on informal test data
no indication of reduced heat transfer or high differential pressure. As
  which appeared to the cognizant engineer as nonrepresentative. In November
part of the evaluation, DPC engineering calculated a fouling factor for the
  1984 the Unit 1 KC heat exchanger was cleaned. In June and July 1985, the
KC heat exchangers. These calculations were based on informal test data
which appeared to the cognizant engineer as nonrepresentative.
In November
1984 the Unit 1 KC heat exchanger was cleaned. In June and July 1985, the
Unit 2 KC heat exchanger was cleaned. Although visual inspection of the
,
,
  Unit 2 KC heat exchanger was cleaned. Although visual inspection of the
heat exchanger by DPC engineering did not support the calculated fouling
  heat exchanger by DPC engineering did not support the calculated fouling
factor (the calculated fouling factor appeared to be less conservative),
  factor (the calculated fouling factor appeared to be less conservative),
the licensee did not perform further evaluation of past operability of these
  the licensee did not perform further evaluation of past operability of these
heat exchangers.
  heat exchangers.
t
t


                                        10
10
    1985
1985
    On October 4, 1985, following in-service testing, the 1A Nuclear Service
On October 4,
    Water pump performance was found to be degraded. The pump curve generated
1985, following in-service testing, the 1A Nuclear Service
    from the test data deviated from the previously established base-line
Water pump performance was found to be degraded. The pump curve generated
    curve. Delivered flow was estimated to be approximately 85 percent of that
from the test data deviated from the previously established base-line
    required. Technical Specification (TS) 3.7.4 requires two loops of RN to
curve. Delivered flow was estimated to be approximately 85 percent of that
    be operable. With only one loop operable, they must restore both loops to
required. Technical Specification (TS) 3.7.4 requires two loops of RN to
    operable within 72 hours or be in hot standby within the next 6 hours and
be operable. With only one loop operable, they must restore both loops to
    cold shutdown within the following 30 hours.
operable within 72 hours or be in hot standby within the next 6 hours and
    The licensee performed a 10 CFR 50.59 analysis to justify cross connecting
cold shutdown within the following 30 hours.
    the 1A and the 2A RN trains in an attempt to boost IA RN flow.         After
The licensee performed a 10 CFR 50.59 analysis to justify cross connecting
    reviewing the 50.59 analysis and extensive interaction with the licensee,
the 1A and the 2A RN trains in an attempt to boost IA RN flow.
    the NRC Region II, on October 10, 1985, informed the licensee that the NRC
After
    considered the licensee was not meeting the requirement of TS 3.7.4 which
reviewing the 50.59 analysis and extensive interaction with the licensee,
    requires two operable RN loops since the 1A train was inoperable due to
the NRC Region II, on October 10, 1985, informed the licensee that the NRC
    a degraded pump and that the cross connected configuration could not be
considered the licensee was not meeting the requirement of TS 3.7.4 which
    justified by a 50.59 analysis since it represented the possibility of an
requires two operable RN loops since the 1A train was inoperable due to
    unreviewed safety question and, in effect, changed the Technical Specifica-
a degraded pump and that the cross connected configuration could not be
    tion.
justified by a 50.59 analysis since it represented the possibility of an
    The licensee's action to cross connect the 1A and 2A RN trains and to con-
unreviewed safety question and, in effect, changed the Technical Specifica-
    tinue two unit operation for greater than 72 hours was contrary to TS 3.7.4
tion.
    and, therefore, represents a violation (369/85-38-03, 370/85-39-03).
The licensee's action to cross connect the 1A and 2A RN trains and to con-
    During this time period, the licensee discovered that one of the cross
tinue two unit operation for greater than 72 hours was contrary to TS 3.7.4
    connect valves had an erroneous position indication.         Thus, the valve
and, therefore, represents a violation (369/85-38-03, 370/85-39-03).
    was actually closed when thought to be open.       This matter was discussed
During this time period, the licensee discovered that one of the cross
    previously in Region II Inspection Report 50-369/85-35, 50-370/85-36.
connect valves had an erroneous position indication.
    As a result of the interactions with the NRC, the licensee split the RN
Thus, the valve
    trains and took compensatory measures to continue operation of the 1A
was actually closed when thought to be open.
    train under reduced flow conditions.         Further details of the apparent
This matter was discussed
    degradation of the 1A RN pump are contained in NRC Region II Inspection
previously in Region II Inspection Report 50-369/85-35, 50-370/85-36.
    Report 50-369/85-37. As a result of the aforementioned event, during the
As a result of the interactions with the NRC, the licensee split the RN
  '
trains and took compensatory measures to continue operation of the 1A
    period of October 15-17, 1985, Region II inspectors reviewed the overall
train under reduced flow conditions.
    RN system performance in light of the recent event.
Further details of the apparent
    The inspectors reviewed the licensee's quarterly performance, PT/1/A/
degradation of the 1A RN pump are contained in NRC Region II Inspection
    4403/04, data on the 1A NS heat exchanger which was tabulated earlier in
Report 50-369/85-37. As a result of the aforementioned event, during the
      this report under 1981.
period of October 15-17, 1985, Region II inspectors reviewed the overall
    The following observations were made by the inspectors regarding PT/1/A/
'
      4403/04:
RN system performance in light of the recent event.
            The performance test lacked qualitative and quantitative acceptance
The inspectors reviewed the licensee's quarterly performance, PT/1/A/
            criteria.
4403/04, data on the 1A NS heat exchanger which was tabulated earlier in
      *     The test results suggest an increasing D/P across the 1A NS heat     l
this report under 1981.
            exchanger.                                                           ;
The following observations were made by the inspectors regarding PT/1/A/
                                                                                  i
4403/04:
                                                            -_        _.   -_.
The performance test lacked qualitative and quantitative acceptance
__      _       . _ _                        _ -
criteria.
*
The test results suggest an increasing D/P across the 1A NS heat
exchanger.
;
i
_
.
-
-_
_.
-_.


                                    11
11
      The pressure drop could not be measured at the required design basis
The pressure drop could not be measured at the required design basis
      accident RN flow of 5000 GPM for the 1A NS heat exchanger because this
accident RN flow of 5000 GPM for the 1A NS heat exchanger because this
      flow could not be achieved for the test performed on October 7,1985.
flow could not be achieved for the test performed on October 7,1985.
      The measured flow was recorded as 4600 gpm.
The measured flow was recorded as 4600 gpm.
*     The 1A NS heat exchanger outlet throttle valve was closed to the
*
      extent that the as found flow through this heat exhanger was 800 gpm.
The 1A NS heat exchanger outlet throttle valve was closed to the
      It appears doubtful that the required accident flow of 5000 gpm could
extent that the as found flow through this heat exhanger was 800 gpm.
      have been achieved with this as found valve position.
It appears doubtful that the required accident flow of 5000 gpm could
have been achieved with this as found valve position.
The licensee indicated that, at that time, a qualitative or quantitative
The licensee indicated that, at that time, a qualitative or quantitative
acceptance criteria had not been determined but that work had begun to
acceptance criteria had not been determined but that work had begun to
provide such criteria.       10 CFR 50 Appendix B Criteria V states that
provide such criteria.
10 CFR 50 Appendix B Criteria V states that
procedures shall include appropriate quantitative or qualitative acceptance
procedures shall include appropriate quantitative or qualitative acceptance
criteria for determining that important activities have been satisfactorily
criteria for determining that important activities have been satisfactorily
accomplished. Contrary to this regulation, PT/1/A/4403/04 did not contain
accomplished.
an appropriate acceptance criteria.       This represents another example of
Contrary to this regulation, PT/1/A/4403/04 did not contain
an appropriate acceptance criteria.
This represents another example of
violation (319/85-38-02, 50-370/85-39-02).
violation (319/85-38-02, 50-370/85-39-02).
Regarding the aforementioned increasing D/P across the 1A NS heat exchanger,
Regarding the aforementioned increasing D/P across the 1A NS heat exchanger,
Line 589: Line 812:
Regarding the low 1A NS heat exchanger RN flow recorded on October 7, 1985,
Regarding the low 1A NS heat exchanger RN flow recorded on October 7, 1985,
the licensee indicated that the low reading could have been a result of a
the licensee indicated that the low reading could have been a result of a
calibration problem.     As a result of the inspector's questioning, the
calibration problem.
licensee issued Work Request Number 65574 to check the calibration of the
As a result of the inspector's questioning, the
licensee issued Work Request Number 65574 to check the calibration of the
flow instrument used to obtain the recorded 4600 GPM. On October 14, 1986,
flow instrument used to obtain the recorded 4600 GPM. On October 14, 1986,
the calibration results indicated that, at a flow of 5000 GPM, the instru-
the calibration results indicated that, at a flow of 5000 GPM, the instru-
ment indicated 4820 GPM. The licensee then took action to recalibrate the
ment indicated 4820 GPM. The licensee then took action to recalibrate the
instrument.
instrument.
Based on the data reviewed and discussions with licensee personnel, the
Based on the data reviewed and discussions with licensee personnel, the
inspector stated the following concerns:
inspector stated the following concerns:
      Since the licensee did not have an acceptance criteria for the in-
Since the licensee did not have an acceptance criteria for the in-
      creased D/P, could the apparently increasing D/P suggest heat exchanger
creased D/P, could the apparently increasing D/P suggest heat exchanger
      fouling which may have reduced heat exchange capacity to an unaccept-
fouling which may have reduced heat exchange capacity to an unaccept-
      able level? Could system flow reductions due to fouling affect other
able level? Could system flow reductions due to fouling affect other
      RN system component performance? These concerns were discussed with
RN system component performance? These concerns were discussed with
      plant management on October 17, 1985. The inspector requested manage-
plant management on October 17, 1985. The inspector requested manage-
      ment to consider the feasibility of performing a RN system integrated
ment to consider the feasibility of performing a RN system integrated
      flow test to provide confidence that all RN safety related loads could
flow test to provide confidence that all RN safety related loads could
      be provided the requisite design basis flows. Additionally, the
be provided the requisite design basis flows.
      inspector discussed the feasibility of measuring the heat transfer
Additionally, the
      capability of the 1A NS heat exchanger.
inspector discussed the feasibility of measuring the heat transfer
                                              _
capability of the 1A NS heat exchanger.
_


                                                                                    - _ _ _ .
- _ _ _ .
                                      12
12
  After growing concern by NRC Region II regarding the current ability of
After growing concern by NRC Region II regarding the current ability of
  the Unit 1 RN system to perform its safety function under accident condi-
the Unit 1 RN system to perform its safety function under accident condi-
  tions, the licensee was requested, on October 18, 1985, to provide the NRC
tions, the licensee was requested, on October 18, 1985, to provide the NRC
  Region II Office with a statement of operability for the RN system.         On
Region II Office with a statement of operability for the RN system.
  October 23, 1985, the operability statement was received from tne licensee.
On
  This :tatement concluded that the RN system is operable and capable of
October 23, 1985, the operability statement was received from tne licensee.
  performing its intended safety function.
This :tatement concluded that the RN system is operable and capable of
  The statement of operability included an engineering evaluation by Duke
performing its intended safety function.
  Power Company. The evaluation summarized the results of a Westinghouse
The statement of operability included an engineering evaluation by Duke
  computer calculation which utilizes the LOTIC code. This code predicts
Power Company.
  containment pressure response from inputs including the heat transfer
The evaluation summarized the results of a Westinghouse
  capability (UA) of the containment spray and component cooling water heat
computer calculation which utilizes the LOTIC code.
  exchangers. The Duke Power engineering calculations used to determine
This code predicts
  the UA for the 1A NS heat exchanger assumed the same fouling factor which
containment pressure response from inputs including the heat transfer
  was calculated for the Component Cooling Water (KC) heat exchanger in
capability (UA) of the containment spray and component cooling water heat
  early 1985.   The inspectors expressed reservation over this assumption;
exchangers.
  questioning the credibility of applying the existing fouling factor for a
The Duke Power engineering calculations used to determine
  single pass horizontal type heat exchanger (KC heat exchanger) to the NS
the UA for the 1A NS heat exchanger assumed the same fouling factor which
  heat exchanger which is a vertical U-tube heat exchanger. Additionally,
was calculated for the Component Cooling Water (KC) heat exchanger in
  RN flows through the tubes of the KC heat exchanger unlike the NS heat
early 1985.
  exchanger where RN flow is on the shell side.       It, however, was agreed by
The inspectors expressed reservation over this assumption;
  the NRC that for lack of any other available data this approach was accept-
questioning the credibility of applying the existing fouling factor for a
  able until specific empirical data could be obtained.
single pass horizontal type heat exchanger (KC heat exchanger) to the NS
  Based on the aforementioned assumptions and calculation as utilized in
heat exchanger which is a vertical U-tube heat exchanger. Additionally,
  the LOTIC program (WCAP-8282), a maximum containment pressure of 13.3 psig
RN flows through the tubes of the KC heat exchanger unlike the NS heat
  was predicted during a design basis accident.         The McGuire containment
exchanger where RN flow is on the shell side.
  design pressure is 15.0 psig.
It, however, was agreed by
  In response to NRC concerns over the potential fouling and degradation
the NRC that for lack of any other available data this approach was accept-
  of the 1A NS heat exchanger, the licensee developed a performance test
able until specific empirical data could be obtained.
  PT/0/A/4208/01, Containment Spray Heat Exchanger Performance Test. The
Based on the aforementioned assumptions and calculation as utilized in
  purpose of the test was to:
the LOTIC program (WCAP-8282), a maximum containment pressure of 13.3 psig
        Determine if a high flow flush reduces the heat exchanger differential
was predicted during a design basis accident.
        pressure.
The McGuire containment
        Assure the structural integrity of the heat exchanger tubes.
design pressure is 15.0 psig.
        Determine the overall heat transfer coefficient and fouling factor of
In response to NRC concerns over the potential fouling and degradation
        the NS heat exchangers.
of the 1A NS heat exchanger, the licensee developed a performance test
  The McGuire FSAR analysis utilized a containment spray heat exchanger UA
PT/0/A/4208/01, Containment Spray Heat Exchanger Performance Test.
  of 2.J4 x 10' BTU-Hr-Deg. F. Empirical data from the aforementioned test
The
  indicated that an actual UA of 7.35 x 105 BTU-Hr-Deg. F existed under
purpose of the test was to:
Determine if a high flow flush reduces the heat exchanger differential
pressure.
Assure the structural integrity of the heat exchanger tubes.
Determine the overall heat transfer coefficient and fouling factor of
the NS heat exchangers.
The McGuire FSAR analysis utilized a containment spray heat exchanger UA
of 2.J4 x 10' BTU-Hr-Deg. F.
Empirical data from the aforementioned test
indicated that an actual UA of 7.35 x 105 BTU-Hr-Deg. F existed under
current plant conditions.
This information was provided to Westinghouse
,
on November 27, 1985, to perform a LOTIC run utilizing this data. A con-
tainment response model which is less conservative than the one used in
the FSAR analysis was used by Westinghouse (WCAP-10325) for this run. Use
, _ .
-
,
,
  current plant conditions. This information was provided to Westinghouse
  on November 27, 1985, to perform a LOTIC run utilizing this data. A con-
  tainment response model which is less conservative than the one used in
  the FSAR analysis was used by Westinghouse (WCAP-10325) for this run. Use
                                , _ .                                      -    ,


                                            13
13
      of this model was accepted by the NRC since this WCAP had been reviewed
of this model was accepted by the NRC since this WCAP had been reviewed
      and found technically sound by the NRC staff, although the NRC's Safety
and found technically sound by the NRC staff, although the NRC's Safety
      Evaluation Report had not been issued at that time.       This LOTIC run of
Evaluation Report had not been issued at that time.
      November 27, 1985, indicated that, for the aforementioned UA, a peak
This LOTIC run of
      containment pressure of 14.42 psig would be realized under design basis
November 27, 1985, indicated that, for the aforementioned UA, a peak
      accident conditions.
containment pressure of 14.42 psig would be realized under design basis
      In addition to the heat transfer test, the licensee performed a heat
accident conditions.
      exchanger tube integrity test using the tritium activity of the Refueling
In addition to the heat transfer test, the licensee performed a heat
      Water Storage tank (RWST) as a tracer source.       The test results indicated
exchanger tube integrity test using the tritium activity of the Refueling
      insignificant leakage.
Water Storage tank (RWST) as a tracer source.
      Several cleaning attempts using various chemical and hydraulic techniques
The test results indicated
      were employed by the licensee to clean the 1A NS heat exchanger. The latest
insignificant leakage.
      performance test results (January 28, 1986) indicate that a UA of 2.03 x
Several cleaning attempts using various chemical and hydraulic techniques
      10' BTV/Hr-Deg. F had been achieved.
were employed by the licensee to clean the 1A NS heat exchanger. The latest
      The inspectors viewed video tapes of the licensee's fiber optic inspection
performance test results (January 28, 1986) indicate that a UA of 2.03 x
      of the 1A NS heat exchanger. Approximately the first seven feet of the
10' BTV/Hr-Deg. F had been achieved.
      upper portion of -the tube bundle could be viewed. The tape indicated that
The inspectors viewed video tapes of the licensee's fiber optic inspection
      a fairly uniform silica deposit completely covered the tubes, prior to
of the 1A NS heat exchanger. Approximately the first seven feet of the
      cleaning.
upper portion of -the tube bundle could be viewed. The tape indicated that
      Confirmatory UA calculations were performed by the inspection team. These
a fairly uniform silica deposit completely covered the tubes, prior to
      calculations appear as Attachment 4 to this report. Those calculations
cleaning.
      closely approximate those of the licensee.
Confirmatory UA calculations were performed by the inspection team. These
calculations appear as Attachment 4 to this report.
Those calculations
closely approximate those of the licensee.
'
'
    8. Review Of Flow Balance Testing
8.
      The inspectors conducted a review of the RN flow balance testing conducted
Review Of Flow Balance Testing
      on December 17, 1985, January 27, 1986, and January 28, 1986, for Train IA
The inspectors conducted a review of the RN flow balance testing conducted
      of the Nuclear Service Water System. Additionally, flow balance testing
on December 17, 1985, January 27, 1986, and January 28, 1986, for Train IA
      conducted on January 30, 1986, for Train 1B of the Nuclear Service Water
of the Nuclear Service Water System.
      System was reviewed.
Additionally, flow balance testing
      The Train 1A flow balance test conducted on December 17, 1985, was in
conducted on January 30, 1986, for Train 1B of the Nuclear Service Water
      accordance with procedure TT/1/A/9100/105, Change 0 through Change 1.     The
System was reviewed.
      test provided for:
The Train 1A flow balance test conducted on December 17, 1985, was in
      -
accordance with procedure TT/1/A/9100/105, Change 0 through Change 1.
              Isolation of Train IA and 1B essential header.
The
      -
test provided for:
              The low level intake providing Train 1A suction.
-
      -
Isolation of Train IA and 1B essential header.
              Isolation of the Unit I non-essential header from Train 1A.
-
      -
The low level intake providing Train 1A suction.
              Control Room and Equipment Room A Train Cooling Chillers being supplied
-
              by Nuclear Service Water Train 1A.
Isolation of the Unit I non-essential header from Train 1A.
-
Control Room and Equipment Room A Train Cooling Chillers being supplied
by Nuclear Service Water Train 1A.
1
1
                                                                                              ,
,
  '
'
                                    .                         _     _             _ - , . .
.
_
_
_ - , .
.


                                      14
14
  -
-
        Securing of Nuclear Service Water Train 2A due to a condition of
Securing of Nuclear Service Water Train 2A due to a condition of
        Nuclear Service Water operability resultant from prior degraded pump
Nuclear Service Water operability resultant from prior degraded pump
        performance in Train IA when supplying Control and Equipment room
performance in Train IA when supplying Control and Equipment room
        cooling.
cooling.
  -
-
        Alignment of service water valves in accordance with a lineup that
Alignment of service water valves in accordance with a lineup that
        was consistent with actual Safety Injection and Containment Spray
was consistent with actual Safety Injection and Containment Spray
        conditions.
conditions.
  The Train 1A flow balance test conducted on January 27, 1986, was per-
The Train 1A flow balance test conducted on January 27, 1986, was per-
  formed in this same manner with the exception that Change 2 of _ procedure
formed in this same manner with the exception that Change 2 of _ procedure
  TT/1/A/9100/105 was also in effect which changed the Train IA suction from
TT/1/A/9100/105 was also in effect which changed the Train IA suction from
  the low level intake to the service water pond in order to duplicate the
the low level intake to the service water pond in order to duplicate the
  most restrictive condition of operation for testing.
most restrictive condition of operation for testing.
  Flow rates through those essential heat exchangers required to mitigate
Flow rates through those essential heat exchangers required to mitigate
  accident consequences during Safety Injection and Containment Spray were
accident consequences during Safety Injection and Containment Spray were
  measured during these tests and compared to target values which were
measured during these tests and compared to target values which were
  specified in the FSAR. Measurement results and comparisons for Train 1A
specified in the FSAR. Measurement results and comparisons for Train 1A
  tests are delineated in Table 1.
tests are delineated in Table 1.
:  The data for the December 17, 1985 test reflects that FSAR specified flow
The data for the December 17, 1985 test reflects that FSAR specified flow
  rate values could not be attained for the containment spray heat exchanger
:
  (4% degraded), control room chiller heat exchanger (10*s degraded), the
rate values could not be attained for the containment spray heat exchanger
  charging pump oil cooler (46% degraded), spent fuel pool pump room ai*
(4% degraded), control room chiller heat exchanger (10*s degraded), the
  handling unit (27% degraded), and containment spray pump room air handlirg
charging pump oil cooler (46% degraded), spent fuel pool pump room ai*
  unit (56% degraded).
handling unit (27% degraded), and containment spray pump room air handlirg
  Although the data from the December 17th test indicated multicomponent
unit (56% degraded).
  degradation, the licensee performed an informal evaluation to support
Although the data from the December 17th test indicated multicomponent
  continued operation. The results of this evaluation were not documented.
degradation, the licensee performed an informal evaluation to support
  Not until requested by the NRC in January 1986, did the licensee perform
continued operation. The results of this evaluation were not documented.
    a detailed engineering evaluation as required by 10 CFR 50.59. Failure
Not until requested by the NRC in January 1986, did the licensee perform
    to perform this requisite evaluation is considered a violation of the
a detailed engineering evaluation as required by 10 CFR 50.59.
    aforementioned 10 CFR 50.59 (369/85-38-04, 370/85-39-04).
Failure
    In an operability statement dated January 14, 1986, the licensee performed
to perform this requisite evaluation is considered a violation of the
    an engineering evaluation to demonstrate the adequacy of the tested perform-
aforementioned 10 CFR 50.59 (369/85-38-04, 370/85-39-04).
    ance of the charging pump oil cooler, the containment spray pump room air
In an operability statement dated January 14, 1986, the licensee performed
    handling unit and the spent fuel pool cooling pump room air handling unit
an engineering evaluation to demonstrate the adequacy of the tested perform-
  with the observed reduced flow rates. In the operability statement the
ance of the charging pump oil cooler, the containment spray pump room air
    licensee stated that the degraded containment spray heat exchanger flow
handling unit and the spent fuel pool cooling pump room air handling unit
  was adequate, and justified continued operation of Unit 1.     This operability
with the observed reduced flow rates.
    statement was based on the actual tested values of the thermal efficiency
In the operability statement the
    for this particular heat exchanger and a containment pressure calculation
licensee stated that the degraded containment spray heat exchanger flow
    performed by Westinghouse and forwarded to Duke Power Company by letter
was adequate, and justified continued operation of Unit 1.
    DAP-86-513 dated January 16, 1986. The Westinghouse calculation was based
This operability
    on assumptions which included the following:
statement was based on the actual tested values of the thermal efficiency
    *
for this particular heat exchanger and a containment pressure calculation
        An active sump volume of 90,000 cubic feet.
performed by Westinghouse and forwarded to Duke Power Company by letter
DAP-86-513 dated January 16, 1986. The Westinghouse calculation was based
on assumptions which included the following:
*
An active sump volume of 90,000 cubic feet.


                                      15
15
      A thermal efficiency heat transfer coefficient of VA=7.35 x 105
A thermal efficiency heat transfer coefficient of VA=7.35 x 105
      BTU-Hr-Deg. F for the containment spray heat exchanger and UA=1.64 x
BTU-Hr-Deg. F for the containment spray heat exchanger and UA=1.64 x
        10' BTU-HT-Hr-Deg. F for the RHR heat exchanger. The licensee stated
10' BTU-HT-Hr-Deg. F for the RHR heat exchanger.
        that, for the containment spray heat exchanger, this represented a 75%
The licensee stated
        reduction in the UA coefficient. This value was a conservative selec-
that, for the containment spray heat exchanger, this represented a 75%
        tion by the licensee since the testing performed on December 17, 1985,
reduction in the UA coefficient. This value was a conservative selec-
      demonstrated the UA value to be nearly 58% degraded.
tion by the licensee since the testing performed on December 17, 1985,
  Under these assumptions the Westinghouse calculation demonstrated that
demonstrated the UA value to be nearly 58% degraded.
  during a LOCA, containment pressure would remain below the containment
Under these assumptions the Westinghouse calculation demonstrated that
  design pressure of 15 psig with service water flow through the containment
during a LOCA, containment pressure would remain below the containment
  spray heat exchanger reduced to 4800 gpm.       The licensee, therefore, con-
design pressure of 15 psig with service water flow through the containment
  sidered that the results of their evaluations and calculations justified
spray heat exchanger reduced to 4800 gpm.
  continued operation of Unit 1.     The basis for this conclusion was reviewed
The licensee, therefore, con-
  and accepted by the NRC
sidered that the results of their evaluations and calculations justified
  Between December 17, 1985 and January 27, 1986, three cleaning cycles were
continued operation of Unit 1.
  accomplished on the RN side of the 1A containment spray heat exchanger. The
The basis for this conclusion was reviewed
  licensee concluded that heat exchanger thermal ef ficiency increased from
and accepted by the NRC
  42.1% to 74.7% as a result of these cleaning cycles. The affects of these
Between December 17, 1985 and January 27, 1986, three cleaning cycles were
  cleaning cycles is also demonstrated in the reduced RN header pressure
accomplished on the RN side of the 1A containment spray heat exchanger. The
  delineated in Table 1, for the flow balance test of January 27, 1986.
licensee concluded that heat exchanger thermal ef ficiency increased from
  The data in Table 1 for the January 27, 1986 test reflects that, even af ter
42.1% to 74.7% as a result of these cleaning cycles. The affects of these
  the cleaning evolutions, FSAR specified flow rate values could again not be
cleaning cycles is also demonstrated in the reduced RN header pressure
  attainec for the containment spray heat exchanger (2% degraded), control
delineated in Table 1, for the flow balance test of January 27, 1986.
  room chiller heat exchanger (0.5% degraded), Spent Fuel Pool Pump Room Air
The data in Table 1 for the January 27, 1986 test reflects that, even af ter
  Handling Unit (30% degraded), containment spray pump room air handling
the cleaning evolutions, FSAR specified flow rate values could again not be
  unit (56% degraded), diesel generator cooling water heat exchanger (8%
attainec for the containment spray heat exchanger (2% degraded), control
  degraded), and safety injection pump motor air handling unit (15% degraded).
room chiller heat exchanger (0.5% degraded), Spent Fuel Pool Pump Room Air
  Degradation of the charging pump cooling flcws was attributed to faulty
Handling Unit (30% degraded), containment spray pump room air handling
  flow indication which required instrument replacement.
unit (56% degraded), diesel generator cooling water heat exchanger (8%
  The licensee stated that as a result of this test,' Train 1A of nuclear
degraded), and safety injection pump motor air handling unit (15% degraded).
  service water v's declared inoperable pending resolution of the degraded
Degradation of the charging pump cooling flcws was attributed to faulty
  flow conditions and correction of the faulty flow indicator associated with
flow indication which required instrument replacement.
  the charging pump oil cooler.
The licensee stated that as a result of this test,' Train 1A of nuclear
  The inspectors noted that these flow balance tests were accomplished with
service water v's declared inoperable pending resolution of the degraded
  Unit 2 Train A secured which was not conservative with respect to the
flow conditions and correction of the faulty flow indicator associated with
  design basis accident. Worst case conditions should assume Unit 2 Train A
the charging pump oil cooler.
  providing unit coaldown loads during the operation of Unit 1 Train A to
The inspectors noted that these flow balance tests were accomplished with
  mitigate accident conditions. This in effect would reduce the net positive
Unit 2 Train A secured which was not conservative with respect to the
  suction head for Unit 1 Train A. The inspectors considered that testing
design basis accident. Worst case conditions should assume Unit 2 Train A
  should reflect this condition.       The licensee stated that on January 28,
providing unit coaldown loads during the operation of Unit 1 Train A to
mitigate accident conditions. This in effect would reduce the net positive
suction head for Unit 1 Train A.
The inspectors considered that testing
should reflect this condition.
The licensee stated that on January 28,
1986, another flow balarce would be performed and that Train 2A would
,
,
  1986, another flow balarce would be performed and that Train 2A would
service necessary cooldown loads for Unit 2 during this test.
  service necessary cooldown loads for Unit 2 during this test.
!
!


                                                                                P
P
                                    16
16
In conjunction with resolution of the degraded flow conditions reflected in
In conjunction with resolution of the degraded flow conditions reflected in
the service water Train 1A flow balance testing, the licensee had requested
the service water Train 1A flow balance testing, the licensee had requested
Line 824: Line 1,080:
the flow instrument for the charging pump oil cooler had been replaced, and
the flow instrument for the charging pump oil cooler had been replaced, and
the RN system took suction only from the SNSWP. The results of this test
the RN system took suction only from the SNSWP. The results of this test
are delineated in Table 2. The result of this test demonstrated that flow
are delineated in Table 2.
The result of this test demonstrated that flow
values through all heat exchangers were within the new acceptable values
values through all heat exchangers were within the new acceptable values
established by the licensee within the operability statement of January 14,
established by the licensee within the operability statement of January 14,
1986. On March 11, 1986, the licensee made a 10 CFR 50.72 notification
1986. On March 11, 1986, the licensee made a 10 CFR 50.72 notification
to the NRC stating that, prior to January 27, 1986, the RN system for
to the NRC stating that, prior to January 27, 1986, the RN system for
both units had never been tested under the requisite accident conditions
both units had never been tested under the requisite accident conditions
with all RN being supplied by the SNSWP. Apparently after addressing
with all RN being supplied by the SNSWP.
Apparently after addressing
both NRC and DPC engineering concerns regarding the desired RN flow test
both NRC and DPC engineering concerns regarding the desired RN flow test
system configuration, the licensee later realized that the preoperational
system configuration, the licensee later realized that the preoperational
test configuration had not tested the system under the design basis
test configuration had not tested the system under the design basis
accident configuration.     The aforementioned event represents another
accident configuration.
example of a violation of 10 CFR 50, Appendix B, Criterion XI (369/85-38-01,
The aforementioned event represents another
370/85-39-01).
example of a violation of 10 CFR 50, Appendix B, Criterion XI (369/85-38-01,
370/85-39-01).
The NRC later learned from the licensee that during the establishment of the
The NRC later learned from the licensee that during the establishment of the
flow test system configuration on January 28, 1986, the RN system entered
flow test system configuration on January 28, 1986, the RN system entered
a pressure transient. While base loading the RN pumps (gradually placing
a pressure transient. While base loading the RN pumps (gradually placing
requisite heat exchangers on the line), a significant decrease in RN header
requisite heat exchangers on the line), a significant decrease in RN header
pressure was experienced. This event was not allowed to go full term and
pressure was experienced. This event was not allowed to go full term and
was terminated by throttling down on large component flows. The test was
was terminated by throttling down on large component flows. The test was
repeated with the throttled valve positions and acceptable results were
repeated with the throttled valve positions and acceptable results were
obtained. On March 12, 1986, NRC Region II learned of the January 28 flow
obtained. On March 12, 1986, NRC Region II learned of the January 28 flow
transient shortly after DPC management had been informed of it.     The NRC
transient shortly after DPC management had been informed of it.
expressed concern regarding the transient since it suggests that, under
The NRC
actual accident conditions, the RN system's pumps could have lost net
expressed concern regarding the transient since it suggests that, under
positive suction head resulting in a loss of the ultimate heat sink for
actual accident conditions, the RN system's pumps could have lost net
both units. This concern is further addressed in Section 10 of this report.
positive suction head resulting in a loss of the ultimate heat sink for
                                                          -               -
both units. This concern is further addressed in Section 10 of this report.
                                                                              ~
-
-
~


                      . . _
. . _
;
;
                                              17
17
          The Nuclear Service Water Train IB flow balance test was conducted on
The Nuclear Service Water Train IB flow balance test was conducted on
,        January 30, 1986, with the same test methodology utilized for the
January 30, 1986, with the same test methodology utilized for the
          January 28, 1986 flow balance test for Train 1A. The results of this test
,
          are delineated in Table 3. The results of this test demonstrated that
January 28, 1986 flow balance test for Train 1A. The results of this test
;       established operability values could not be obtained for the Spent Fuel Pool
are delineated in Table 3.
          Pump Room Air Handling Unit (5% degraded) and the Residual Heat Removal Pump
The results of this test demonstrated that
,        Room Air Handling Unit (1% degraded). The licensee was advised by the NRC
;
          that prior to establishing Train IB as being fully operable, these degraded
established operability values could not be obtained for the Spent Fuel Pool
Pump Room Air Handling Unit (5% degraded) and the Residual Heat Removal Pump
Room Air Handling Unit (1% degraded). The licensee was advised by the NRC
,
that prior to establishing Train IB as being fully operable, these degraded
conditions would require further evaluation and resolution.
4
4
          conditions would require further evaluation and resolution.
;
;
o
o
Line 874: Line 1,139:
i
i
i
i
                                                                                ,  -- . . . .
-
  - -_ -         .,                     --     _. , . ,   . . - . . _   ,
-_
-
.,
--
_.
, . ,
. . - . . _
,
,
--
. . . .


    . _ - _ .
. _ - _ .
        _         _      _ . _ _ . _ . . .   _ _ . _ . _               _ _ . _ _ _ _ _ _ . - .
_
                                                                                  _                        _m   = = . _ _ . _             _ . _ . . . _ _ , . _ . .       ._. _ .               . __ -. _ . _ _ , _ _ , _ .
_ . _ _ . _ . . .
!                                                                                                                                                                                                                             ,
_ _ . _ . _
                                                                                                                                                                                                                              t
_ _ . _ _ _ _ _ _ . - .
_m
= = . _ _ . _
_ . _ . . . _ _ , . _ . .
._. _ .
. __
-. _ . _ _ , _ _ , _ .
_
_
!
,
t
TABLE 1
;
Results of Heat Exchanger Flows and Comparison to FSAR Target Values During Nuclear Service Water
Train 1A Flow Balance Testing of December 17, 1985 and Janua ry 27, 1986.
December 17, 1985 Data
Janua ry 27, 1986
Ta rget Flow
flow Rate
Header
Flow Rate
Header Pressure
Heat Exchanger
Rate (CPM)
(CPM)
Pressure (psig)
(CPM)
(psig)
1.
Component Cooling Water
8000
0000
67.5
8000
56
'
2.
Conta inment Spray
5000
4800
67.5
4887
56
l
3.
Diesel Generator Cooling
900
900
67.5
830
56
;
;
                                                                                                        TABLE 1
Water
    Results of Heat Exchanger Flows and Comparison to FSAR Target Values During Nuclear Service Water
i
    Train 1A Flow Balance Testing of December 17, 1985 and Janua ry 27, 1986.
4.
                                                                                          December 17, 1985 Data                                                          Janua ry 27, 1986
Control Room Chiller
                                                            Ta rget Flow                flow Rate              Header                                      Flow Rate                Header Pressure
789
    Heat Exchanger                                          Rate (CPM)                        (CPM)            Pressure (psig)                                        (CPM)              (psig)
707
    1.        Component Cooling Water                        8000                              0000          67.5                                                  8000                56
67.5
                                                                                                                                                                                                                                '
785
    2.        Conta inment Spray                            5000                              4800          67.5                                                  4887                56
56
l  3.        Diesel Generator Cooling                        900                                900          67.5                                                    830                56
5.
;              Water                                                                                                                                                                                                          i
Cha rg i ng Pump Oi l Coo l e r
    4.         Control Room Chiller                           789                               707         67.5                                                   785               56
28
    5.         Cha rg i ng Pump Oi l Coo l e r                 28                                 15         67.5                                                     3               56
15
i   6.         Safety injection Pump                           20                                 21         67.5                                                     17               56
67.5
                Oil Cooler
3
56
i
6.
Safety injection Pump
20
21
67.5
17
56
Oil Cooler
c
c
7.
Spent fuel Pool Pump
20
14.7
67.5
14
56
'
'
    7.        Spent fuel Pool Pump                            20                              14.7          67.5                                                    14                56      *
*
                Air Handling Unit
Air Handling Unit
    8.         Conta inment Spray Pump                         45                                 20         67.5                                                     20               56
8.
                Air Handling Unit
Conta inment Spray Pump
    9.         Residual Heat Removal                           45                                 51         67.5                                                     52               56
45
                Pump Air Handling Unit                                                                                                                                                                                         ,
20
67.5
20
56
Air Handling Unit
9.
Residual Heat Removal
45
51
67.5
52
56
Pump Air Handling Unit
,
i
i
  1
1
:
:
$
$
  1
1
1
1
                                                                                                                              -_ - _ -
-
                                              -    ,,           - , - -.   ,                       ,   -                             . -           -                                     .,
,,
- , - -.
,
,
-
-_ - _ -
-
.,
. -


~ - ..       - -- - . - - - . .             . . ~ . . . - . . . . - . ~ . .       . ~ - . - . . _ . - .       . - . . _ . . - - - - - . - - _ _ . - - _ - - - -                           . - - - -
~ - ..
                                                                                                                                                                                                          *
- -- - . - - - . .
                                                                                                                                                                                                          l
. . ~ . . . - . . . . - . ~ . .
                                                                                      TABLE 2
. ~ - . - . . _ .
      Results of Heat Exchanger Flows and Comparison to FSAR Target Values and Licensee Established Operability
- .
      Values During Nuclear Service Water Train 1A Flow Balance Testing of January 28, 1986.
. - . . _ . . - - - - - . - - _ _ . - - _ - - - -
                                                                                                                                      January 28, 1986 DATA
. - - - -
                                                                              Licensee Established
*
                                            Target flow                       Ope ra b i l i ty Va l ue                     Flow Rate                               Header Pressure
l
      Heat Exchanger                       From FSAR (CPM)                   ' for Flow (GPM)                                   (GPM)                                       (psig)
TABLE 2
                                                                                                                                                                                                          ,
Results of Heat Exchanger Flows and Comparison to FSAR Target Values and Licensee Established Operability
      1 Component Cooling Water               8000                                   6000+                                 6000                                             62.5                       ;
Values During Nuclear Service Water Train 1A Flow Balance Testing of January 28, 1986.
      2. Containment Spray                   5000                                   3800+                                 3970                                             62.5
January 28, 1986 DATA
      3. Diesel Generator Cooling               900                                   900                                 950                                               62.5
E
            Water
Licensee Established
      4. Control Room Chiller                   789                                   789                                 946                                               62.5
Target flow
      5. Cha rg i ng Pump Oi l Coo le r .         28                                     15                               22                                               62.5
Ope ra b i l i ty Va l ue
      6. Safety injection Pump                   20                                     20                                 23                                               62.5
Flow Rate
            Oil Cooler                                                                                                                                                               ,
Header Pressure
      7. Spent fuel Pool Pump                     20                                 14.7                                   19                                               62.5
Heat Exchanger
            Air Handling Unit
From FSAR (CPM)
      8. Conta inment Spray Pump                   45                               20.0                                 23.5                                               62.5
' for Flow (GPM)
            Air Handling Unit
(GPM)
      9. Residual Heat Remova l                   45                               45.0                                 64.5                                               62.5
(psig)
            Pump Air Handling Unit
,
      + Based on assumption that Containment Spray Heat Exchanger Thermal Efficiency is greater than or equal to 70%.
1 Component Cooling Water
        The rma l performance data reflects it is currently 74.7% and past history indicates degradation will inc rea se
8000
        due to fouling.
6000+
                                                                                                          . - _ . - , - .-                         _ .. ._,____,_ _._ _._ _.         . ,,_         ,_
6000
62.5
;
2.
Containment Spray
5000
3800+
3970
62.5
3.
Diesel Generator Cooling
900
900
950
62.5
Water
4.
Control Room Chiller
789
789
946
62.5
5.
Cha rg i ng Pump Oi l Coo le r .
28
15
22
62.5
6.
Safety injection Pump
20
20
23
62.5
Oil Cooler
,
7.
Spent fuel Pool Pump
20
14.7
19
62.5
Air Handling Unit
8.
Conta inment Spray Pump
45
20.0
23.5
62.5
Air Handling Unit
9.
Residual Heat Remova l
45
45.0
64.5
62.5
Pump Air Handling Unit
+ Based on assumption that Containment Spray Heat Exchanger Thermal Efficiency is greater than or equal to 70%.
The rma l performance data reflects it is currently 74.7% and past history indicates degradation will inc rea se
due to fouling.
. - _ . - , - .-
_ .. ._,____,_ _._ _._ _.
.
,,_
,_


  - - - _ _ _ - - .           - _ _ . _ - . . - - . . _ - . _ _ - . ..           . - -         _ . _ - _ . - - . - - - ~ . - _
- - - _ _ _ - - .
                                                                                                                .                              _
- _ _ . _ -
                                                                                                                                                  . - . . _ . - _ . . . . . -- -..
. .
                                                                                                                                                                                    I
- - . . _ - . _ _ - . ..
                                                                                        TABLE 3
. - -
                    Results of Heat Exchangor Flows and Comparison to FSAR Target Values and Licensee Established Operability
_ . _ -
                    Values During Nuclear Service Water Train IB Flow Balance Testing of January 30, 1986.
_ . - - . - - - ~ . - _
                                                                                                                                      January 28, 1986 DATA
_
I                                                                                 Licensee Established
. - . . _ . - _ . . . .
j                         -
.
                                                                  Ta rge t flow   Ope ra b i l i ty Va l ue                     Flow Rate       Header Pressure
-- -..
                    Heat Exchanger                                From FSAR (GPM) for Flow (GPM)                                    (GPM)                  (psig)
.
I
TABLE 3
Results of Heat Exchangor Flows and Comparison to FSAR Target Values and Licensee Established Operability
Values During Nuclear Service Water Train IB Flow Balance Testing of January 30, 1986.
January 28, 1986 DATA
I
Licensee Established
j
-
Ta rge t flow
Ope ra b i l i ty Va l ue
Flow Rate
Header Pressure
j
j
Heat Exchanger
From FSAR (GPM)
for Flow (GPM)
(GPM)
(psig)
I
I
                    1 Component Cooling Water                           8000               6000+                                 6900                     52
1 Component Cooling Water
                    2. Conta inment Spray                             5000               3800+                                 5000                       52
8000
                    3. Diesel Generator Cooling                         900               900                                 920                       52
6000+
                        Water
6900
j                   4. Control Room Chiller           *
52
                                                                        789               789                                 912                       52
2.
i                   5. Cha rg ing Pump Oi l Coo le r                     28                 15                                   20                       52
Conta inment Spray
,                  6. Safety injection Pump                             20                 20                                 28.2                       52
5000
3800+
5000
52
3.
Diesel Generator Cooling
900
900
920
52
Water
j
4.
Control Room Chiller
789
789
912
52
*
i
5.
Cha rg ing Pump Oi l Coo le r
28
15
20
52
6.
Safety injection Pump
20
20
28.2
52
,
-
-
                        Oil Cooler
Oil Cooler
                    7. Spent ruel Pool Pump                             20               14.7                                   14                       52
7.
                        Air Handling Unit
Spent ruel Pool Pump
20
14.7
14
52
Air Handling Unit
f
f
,                  8. Containment Spray Pump                           45               20.0                                   46                       52
8.
Containment Spray Pump
45
20.0
46
52
,
'
'
                        Air Handling Unit
Air Handling Unit
                    9. Residual Heat Removal                             45               45.0                                 44.5                       52
9.
                        Pump Ai r Handling Unit
Residual Heat Removal
                    + Based on assumption that Containment Spray Heat Exchanger Thermal Erriciency is greater than or equal to 70%.
45
                    Thermal performance data reflects it is currently 74.7% and past history indicates dearadation will inc rea se
45.0
                    due to rouling.
44.5
52
Pump Ai r Handling Unit
+ Based on assumption that Containment Spray Heat Exchanger Thermal Erriciency is greater than or equal to 70%.
Thermal performance data reflects it is currently 74.7% and past history indicates dearadation will inc rea se
due to rouling.
I
I
l
l
                                                                                                        ,                                                 . - - .
,
                                                                                                                                                                      7
. - - .
7


                                    18
18
Following performance of the nuclear service water train IA flow balance
Following performance of the nuclear service water train IA flow balance
test of January 28, 1986, the inspector observed a train IA Diesel Generator
test of January 28, 1986, the inspector observed a train IA Diesel Generator
operability test.   During performance of this test, the inspectors noted
operability test.
During performance of this test, the inspectors noted
that the flow indicator for service water flow through the diesel generator
that the flow indicator for service water flow through the diesel generator
cooling water heat exchanger was off scale high (greater than 1000 gallons
cooling water heat exchanger was off scale high (greater than 1000 gallons
per minute) rather than indicating an expected value of 900 gallons per
per minute) rather than indicating an expected value of 900 gallons per
minute.   Interviews with licensee personnel who had performed the earlier
minute.
Interviews with licensee personnel who had performed the earlier
Train 1A flow balance test reflected that, during test restoration, valve
Train 1A flow balance test reflected that, during test restoration, valve
IRN73A was left in the test position rather than being returned to the
IRN73A was left in the test position rather than being returned to the
normal position.     The test position for this valve is " throttled to 900
normal position.
The test position for this valve is " throttled to 900
gallons per minute in the test lineup configuration". The normal position
gallons per minute in the test lineup configuration". The normal position
for this valve is " throttled to 900 gallons per minute in the normal lineup
for this valve is " throttled to 900 gallons per minute in the normal lineup
Line 1,007: Line 1,505:
The inspectors noted that restoration of the 1A train service water diesel
The inspectors noted that restoration of the 1A train service water diesel
generator heat exchanger outlet isolation valve (IRN73A) and the 1A train
generator heat exchanger outlet isolation valve (IRN73A) and the 1A train
service water containment spray heat exchanger outlet isolation valve
service water containment spray heat exchanger outlet isolation valve
(IRN137A) to their normal throttled positions could result in insufficient
(IRN137A) to their normal throttled positions could result in insufficient
nuclear service water flow being supplied to the diesel generator heat
nuclear service water flow being supplied to the diesel generator heat
exchanger and containment spray heat exchanger when the containment spray
exchanger and containment spray heat exchanger when the containment spray
heat exchanger is placed on line during transfer to cold leg recirculation
heat exchanger is placed on line during transfer to cold leg recirculation
unless specific operator actions were taken to ensure proper flow through
unless specific operator actions were taken to ensure proper flow through
these heat exchangers. A review of the emergency operating procedures for
these heat exchangers. A review of the emergency operating procedures for
safety injection (EP/1/A/5000/01, EP/2/A/5000/01) and for transfer to cold
safety injection (EP/1/A/5000/01, EP/2/A/5000/01) and for transfer to cold
leg recirculation (EP/1/A/5000/2.3, EP/2/A/5000/2.3) reflected that provi-
leg recirculation (EP/1/A/5000/2.3, EP/2/A/5000/2.3) reflected that provi-
sions were not established to assure proper service flow through the diesel
sions were not established to assure proper service flow through the diesel
generator cooling water heat exchanger and containment spray heat exchanger
generator cooling water heat exchanger and containment spray heat exchanger
when these component were required during accident conditions.           These
when these component were required during accident conditions.
inadequacies in the emergency operating procedures are considered a fourth
These
example of violation 369/85-38-02, 370/85-39-02, failure to properly
inadequacies in the emergency operating procedures are considered a fourth
establish and implement procedures.
example of violation 369/85-38-02, 370/85-39-02, failure to properly
During the course of this inspection, test procedure TT/1/A/9100/105, RN
establish and implement procedures.
Train 1A Flow Verification, was revised, and test procedure TT/1/A/9100/107,
During the course of this inspection, test procedure TT/1/A/9100/105, RN
RN Train IB Flow Verification, was written to leave the service water outlet
Train 1A Flow Verification, was revised, and test procedure TT/1/A/9100/107,
isolation valve to the containment spray heat exchangers in the tested
RN Train IB Flow Verification, was written to leave the service water outlet
throttle position.     Additionally, licensee actions were initiated to
isolation valve to the containment spray heat exchangers in the tested
revise emergency operating procedures EP/1/A/5000/01, EP/2/A/5000/01,
throttle position.
EP/1/A/5000/2.3, EP/2/A/5000/2.3 in order to establish adequate service
Additionally, licensee actions were initiated to
water flow through the diesel generator heat exchanger and containment
revise emergency operating procedures EP/1/A/5000/01,
spray heat exchanger, during safety injection and transfer to cold leg
EP/2/A/5000/01,
recirculation.
EP/1/A/5000/2.3, EP/2/A/5000/2.3 in order to establish adequate service
water flow through the diesel generator heat exchanger and containment
spray heat exchanger, during safety injection and transfer to cold leg
recirculation.


                      -               .                                     _ _
-
.
_ _
!
!
.                                         19
19
  9. Changes to the McGuire Containment Pressure Response Model
.
    During the course of the licensee's engineering evaluations to justify
9.
    the apparent RN system degradation, many changes were made to the input
Changes to the McGuire Containment Pressure Response Model
    parameters used in the McGuire containment pressure response model.
During the course of the licensee's engineering evaluations to justify
    The following parameters have significant effect on peak containment
the apparent RN system degradation, many changes were made to the input
    pressure:
parameters used in the McGuire containment pressure response model.
            ice mass
The following parameters have significant effect on peak containment
    *
pressure:
            NS and KC heat exchanger UAs
ice mass
    *
*
            NS and KC heat exchanger tube and shell flows
NS and KC heat exchanger UAs
    *
NS and KC heat exchanger tube and shell flows
            mass and energy releases into containment
*
    *      auxiliary containment spray flow
*
    *
mass and energy releases into containment
            auxiliary containment spray actuation time
auxiliary containment spray flow
    *
*
            active containment sump volume
*
    Table 4 provides a chronology of these parameters and when each parameter
auxiliary containment spray actuation time
    was changed by Duke. Some values such as active containment sump are based
*
    on engineering judgement by Duke since calculations have not been completed
active containment sump volume
    to justify the value.
Table 4 provides a chronology of these parameters and when each parameter
                                        TABLE 4
was changed by Duke. Some values such as active containment sump are based
                  McGuire Containment Pressure Response Model Changes
on engineering judgement by Duke since calculations have not been completed
    Parameter                       10/31           11/28         1/17   1/28
to justify the value.
      Ice Mass                       2.220           2.220         2.220 2.220
TABLE 4
      (millions of LBM)
McGuire Containment Pressure Response Model Changes
    NS HX UA                         1.86           0.735         0.735 2.03
Parameter
      (millions of BTU /HR- F)
10/31
      KC HX UA                       5.00           5.00         5.00   2.98
11/28
      (millions of BTU /HR- F)
1/17
      NS/RN Flow (GPM)               5000           5000         4800   3800
1/28
      KC/RN Flow (GPM)               8000           8000         8000   6000
Ice Mass
      Mass and Energy Release         1974           1979         1979   1979
2.220
      Model (year)
2.220
      ND Containment Spray             1623           1623         1841   1841 '
2.220
                                                                                ,
2.220
      Flow (GPM)
(millions of LBM)
      NO Containment Spray           3000           3000         3000   3000
NS HX UA
      Actuation Time (SEC)
1.86
                                                                                  l
0.735
0.735
2.03
(millions of BTU /HR- F)
KC HX UA
5.00
5.00
5.00
2.98
(millions of BTU /HR- F)
NS/RN Flow (GPM)
5000
5000
4800
3800
KC/RN Flow (GPM)
8000
8000
8000
6000
Mass and Energy Release
1974
1979
1979
1979
Model (year)
ND Containment Spray
1623
1623
1841
1841
'
,
Flow (GPM)
NO Containment Spray
3000
3000
3000
3000
Actuation Time (SEC)


                                                                                  _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_
                                          20
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _
    Active Containment             46,500         46,500           90,000   90,000
20
    Sump Volume (FT3)
Active Containment
    Peak Pressure (Psig)           13.3           14.42           14.45     12.7
46,500
46,500
90,000
90,000
Sump Volume (FT3)
Peak Pressure (Psig)
13.3
14.42
14.45
12.7
10. RN System Walkdown
10. RN System Walkdown
    The inspectors conducted a detailed walkdown of portions of the Unit 1
The inspectors conducted a detailed walkdown of portions of the Unit 1
    Nuclear Service Water System. The inspectors reviewed the system operating
Nuclear Service Water System. The inspectors reviewed the system operating
    procedures, the valve checklist procedure and the system piping drawings.
procedures, the valve checklist procedure and the system piping drawings.
    The inspection was conducted to confirm that procedural valve lineups and
The inspection was conducted to confirm that procedural valve lineups and
    drawings matched the as-built configurations, to verify that equipment
drawings matched the as-built configurations, to verify that equipment
    conditions were satisfactory and items that might degrade performance were
conditions were satisfactory and items that might degrade performance were
    identified and evaluated, to verify that valves were in proper positions and
identified and evaluated, to verify that valves were in proper positions and
    locked if appropriate, and to verify that instrumentation was properly
locked if appropriate, and to verify that instrumentation was properly
    valved in.
valved in.
    The inspectors made the following observations. Valves 1RN 893 and 1RN 894,
The inspectors made the following observations. Valves 1RN 893 and 1RN 894,
    the inlets to the 1A1 and 1A2 Diesel generator Air Dryer and af ter dryer
the inlets to the 1A1 and 1A2 Diesel generator Air Dryer and af ter dryer
    respectively, were mislabeled.     Valve 1RN894 was labeled as IRN893. The
respectively, were mislabeled.
    Nuclear Service Water System valve checklist correctly described these
Valve 1RN894 was labeled as IRN893.
    valves and the licensee made arrangements to correct the label plates on the
The
    valves prior to the inspector leaving the site.
Nuclear Service Water System valve checklist correctly described these
    The inspector noted slight inaccuracies in the system piping diagrams, in
valves and the licensee made arrangements to correct the label plates on the
    that relief valve 1RN-295 is located upstream of flow element 5360 as
valves prior to the inspector leaving the site.
    opposed to downstream as indicated on DWG MC-1574-2.0 and vent valve IRN141
The inspector noted slight inaccuracies in the system piping diagrams, in
    is located upstream of flow element 5930 as opposed to downstream of the
that relief valve 1RN-295 is located upstream of flow element 5360 as
    flow element as indicated on DWG MC-1574-2.0.     The licensee made arrangements
opposed to downstream as indicated on DWG MC-1574-2.0 and vent valve IRN141
    to correct these inaccuracies prior to the inspectors leaving the site.
is located upstream of flow element 5930 as opposed to downstream of the
flow element as indicated on DWG MC-1574-2.0.
The licensee made arrangements
to correct these inaccuracies prior to the inspectors leaving the site.
11. Details of NRC/DPC Management Meeting Held on March 14, 1986
11. Details of NRC/DPC Management Meeting Held on March 14, 1986
    a.   Attendance at the Duke - NRC Management Conference on March 14, 1986,
a.
          held at DPC's request at the NRC's Region II Office included:
Attendance at the Duke - NRC Management Conference on March 14, 1986,
          Duke power Company
held at DPC's request at the NRC's Region II Office included:
          G. Vaughn, General Manager, Nuclear Stations
Duke power Company
          T. L. McConnell, McGuire Nuclear Station Manager                                                       ,
G. Vaughn, General Manager, Nuclear Stations
          W. A. Haller, Manager, Technical Services                                                               '
T. L. McConnell, McGuire Nuclear Station Manager
          R. L. Gill, McGuire Licensing                                                                           ,
,
          B. H. Hamilton, McGuire Superintendent of Technical Services                                             l
W. A. Haller, Manager, Technical Services
          J. E. Snyder, Supervising Engineer                                                                     !
'
          E. O. McCraw, Compliance Engineer
R. L. Gill, McGuire Licensing
          W. J. Kronenwetter, Design Engineer
,
          R. W. Revels, Design Engineer
B. H. Hamilton, McGuire Superintendent of Technical Services
          W. M. Suslick, Associate Engineer
J. E. Snyder, Supervising Engineer
E. O. McCraw, Compliance Engineer
W. J. Kronenwetter, Design Engineer
R. W. Revels, Design Engineer
W. M. Suslick, Associate Engineer
.
-


                                    21
21
,
Nuclear Regulatory Commission
      Nuclear Regulatory Commission
      R. D. Walker, Deputy Regional Administrator
      A. F. Gibson, Director, Division of Reactor Safety
      C. A. Julian, Chief, Operations Branch
      B. T. Debs, Acting, Chief, Operational Programs Section
i    M. V. Sinkule, Chief, Reactor Projects Section
      F. R. McCoy, Reactor Engineer
,    W. T. Orders, Senior Resident Inspector, McGuire                        '
:    C. W. Burger, Project Inspector
      C. L. Vanderniet, Reactor Engineer
i b. Members of the Duke Power Company staff met with members of the NRC
      Region II staff to discuss the status of the McGuire Units 1 and 2
      Nuclear Service Water System. A copy of the meeting agenda and DPC
      handouts appear as Attachments 1, 2, and 3 to this inspection report.
      DPC representatives stated that, from the information available to the
.    DPC staff, the Nuclear Service Water System had been and is currently
      operable. The NRC staff acknowledged that, once the NRC had surfaced
      concerns regarding the Nuclear Service Water System, the licensee has
      placed extensive resources on solving the problem.
      As a result of the aforementioned meeting, NRC representatives
      contacted DPC staff on March 24, 1986, to request additional informa-
      tion.    DPC staff agreed to formally submit a response by April 25,
,    1986, regarding the following seven requested items.
      -
            Provide the as-found and as-left RN flow balance test results for
;
            all RN trains.
      -
            Provide the as-found and as-left VA test results for all
            containment spray heat exchangers.
      -
            Provide an RN operability determination for early October 1985
            when RN flow was recorded as 800 GPM to the 1A containment spray
            heat exchangers.
      -
            Provide safety evaluation of the January 28, 1986 RN header
,
,
            pressure transient.
R. D. Walker, Deputy Regional Administrator
      -
A. F. Gibson, Director, Division of Reactor Safety
            Provide an RN operability determination based on the 1A contain-
C. A. Julian, Chief, Operations Branch
            ment spray heat exchanger throttle valve setting which existed    l
B. T. Debs, Acting, Chief, Operational Programs Section
                                                                              '
i
M. V. Sinkule, Chief, Reactor Projects Section
F. R. McCoy, Reactor Engineer
W. T. Orders, Senior Resident Inspector, McGuire
'
,
:
C. W. Burger, Project Inspector
C. L. Vanderniet, Reactor Engineer
i
b.
Members of the Duke Power Company staff met with members of the NRC
Region II staff to discuss the status of the McGuire Units 1 and 2
Nuclear Service Water System. A copy of the meeting agenda and DPC
handouts appear as Attachments 1, 2, and 3 to this inspection report.
DPC representatives stated that, from the information available to the
DPC staff, the Nuclear Service Water System had been and is currently
.
operable. The NRC staff acknowledged that, once the NRC had surfaced
concerns regarding the Nuclear Service Water System, the licensee has
placed extensive resources on solving the problem.
As a result of the aforementioned meeting, NRC representatives
contacted DPC staff on March 24, 1986, to request additional informa-
tion.
DPC staff agreed to formally submit a response by April 25,
1986, regarding the following seven requested items.
,
-
Provide the as-found and as-left RN flow balance test results for
;
all RN trains.
Provide the as-found and as-left VA test results for all
-
containment spray heat exchangers.
-
Provide an RN operability determination for early October 1985
when RN flow was recorded as 800 GPM to the 1A containment spray
heat exchangers.
Provide safety evaluation of the January 28, 1986 RN header
-
pressure transient.
,
,
            just prior to the first heat transfer test and based on expected
-
            flow under accident conditions prior to heat exchanger cleaning   )
Provide an RN operability determination based on the 1A contain-
            cycles.                                                           )
ment spray heat exchanger throttle valve setting which existed
                                                                                1
just prior to the first heat transfer test and based on expected
      -
'
            Provide the final parameters for use in the LOTIC program and
,
            their engineering basis.
flow under accident conditions prior to heat exchanger cleaning
)
cycles.
)
-
Provide the final parameters for use in the LOTIC program and
their engineering basis.
l
l
      -
Provide DPC plans to prevent a recurrence of these events.
            Provide DPC plans to prevent a recurrence of these events.
-
                        -.       -.             -.       . _ .
-.
-.
-- --
-.
. _ .


                _-                                         . -       -       -
_-
,
.
                                            22
-
              By memo of April 25, 1986, Duke Power Company responded to these
-
              requests. The responses contend that the RN system was continuously
-
              operable.   Inspectors will follow up on this information during a
,
                                                    -
22
              future inspection.
By memo of April 25, 1986, Duke Power Company responded to these
        The resolution of these matters represents unresolved item (369/85-38-06,
requests.
        370/85-39-06).
The responses contend that the RN system was continuously
    12. General Conclusions
operable.
1        During the operating history of the McGuire plant, the licensee has experi-
Inspectors will follow up on this information during a
        enced an increasing degradation of the RN system. It is apparent that
future inspection.
        the licensee has dealt with this situation on a case-by-case basis. Until
-
        prompted by the NRC, the licensee had not determined the full extent of
The resolution of these matters represents unresolved item (369/85-38-06,
        the RN system degradation or taken adequate corrective action to preclude
370/85-39-06).
i       repetition.   Although the licensee has recently dedicated significant
12. General Conclusions
        resources to addressing the problem, serious doubt exists regarding the
During the operating history of the McGuire plant, the licensee has experi-
        past operability of the RN system and those safety related systems, such as
1
        containment spray, for which RN is an ancillary system.       This doubt is
enced an increasing degradation of the RN system.
        fostered as a result of aggregate observations of significantly reduced
It is apparent that
;       heat transfer capability of various safety related heat exchangers, reduced
the licensee has dealt with this situation on a case-by-case basis. Until
  ,      RN flows, improper throttle valve settings, increased corrosion, and lack of
prompted by the NRC, the licensee had not determined the full extent of
  '
the RN system degradation or taken adequate corrective action to preclude
        adequate preoperational testing. This situation is contrary to 10 CFR 50,
i
        Anpendix B, Criterion XVI which requires that measures shall be established
repetition.
        tr assure that conditions adverse to quality, such as failures, malfunc-
Although the licensee has recently dedicated significant
.       tions, deficiencies, deviations, defective material and equipment, and
resources to addressing the problem, serious doubt exists regarding the
'
past operability of the RN system and those safety related systems, such as
'
containment spray, for which RN is an ancillary system.
        nonconformances are promptly identified and corrected.     In the case of
This doubt is
        significant conditions adverse to quality, the measures shall assure that
fostered as a result of aggregate observations of significantly reduced
'i
;
        the cause of the condition is determined and corrective action taken to
heat transfer capability of various safety related heat exchangers, reduced
        preclude repetition.     The identification of the significant condition
RN flows, improper throttle valve settings, increased corrosion, and lack of
        adverse to quality, the cause of the condition, and the corrective action
,
        taken shall be documented and reported to appropriate levels of management.
adequate preoperational testing.
        The licensee's failure to meet these requirements, in the case of the RN
This situation is contrary to 10 CFR 50,
'
Anpendix B, Criterion XVI which requires that measures shall be established
tr assure that conditions adverse to quality, such as failures, malfunc-
.
tions, deficiencies, deviations, defective material and equipment, and
'
nonconformances are promptly identified and corrected.
In the case of
'
significant conditions adverse to quality, the measures shall assure that
the cause of the condition is determined and corrective action taken to
'i
preclude repetition.
The identification of the significant condition
adverse to quality, the cause of the condition, and the corrective action
taken shall be documented and reported to appropriate levels of management.
The licensee's failure to meet these requirements, in the case of the RN
system, is a violation (369/85-38-05, 370/85-39-05).
,
,
        system, is a violation (369/85-38-05, 370/85-39-05).
)
)
,
,
1                                                                                     i
1
                                                                                      ,
i
                                                                                      '
,
'
;
;
l
l
:
:
i                                                                                   .
i
                                                                                      !
.
!


              .       .   _ - _ . .
.
  '.'.
.
                                        ATTACHMENT 1
_
    !
-
                                DUKE POWER /NRC REGION 11
_ . .
' . ' .
ATTACHMENT 1
!
DUKE POWER /NRC REGION 11
i
MEETING TO DISCUSS McGUIRE NUCLEAR STATION
NUCLEAR SERVICE WATER SYSTEM PERFORMANCE
MARCH 14, 1986
AGENDA
*
OPENING REMARKS
GERALD VAUGHN
*
OVERVIEW OF NUCLEAR SERVICE WATER SYSTEM
NEAL McCRAW
i
i
                      MEETING TO DISCUSS McGUIRE NUCLEAR STATION
'
                        NUCLEAR SERVICE WATER SYSTEM PERFORMANCE
*
                                          MARCH 14, 1986
NUCLEAR SERVICE WATER SYSTEM EXPERIENCE
                                              AGENDA
TONY McCONNELL
            * OPENING REMARKS                                GERALD VAUGHN
*
            * OVERVIEW OF NUCLEAR SERVICE WATER SYSTEM       NEAL McCRAW
RECENT OPERATIONAL EXPERIENCE
i
BILL SUSLICK
'
(10/04/85 TO PRESENT)
            * NUCLEAR SERVICE WATER SYSTEM EXPERIENCE        TONY McCONNELL
*
            * RECENT OPERATIONAL EXPERIENCE                 BILL SUSLICK
DESIGN CONSERVATISMS
              (10/04/85 TO PRESENT)
BILL KRONENWETTER
            * DESIGN CONSERVATISMS                           BILL KRONENWETTER
!
!
            * CLOSING REMARKS                               GERALD VAUGHN
*
CLOSING REMARKS
GERALD VAUGHN
;
;
i
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                                                                                                              LE
Fm
                                                                                                                                                          2
a
                                                                                                                                                            -
e
                                                                                                                                                                2
LE
                                                                                                                                                                2
2
                                                                                                                                                                      B
R
                                                                                                                                                                      R
@) Nl
                                                                                                                                                                              .c
CT
                                                                                                                                                                              i.
-
                                                                                                              CT                                                      A
A
                                                                        Fm                                    UA
i
                                                                                                                                                                n
A
                                                                                                                                                                  -
UA
                                                                                                                                                                              i
n
                                                                                                                                                                              A
.
                                  N                                                                          NW                                           -
N
                                                                                                                                                                              .
-
                                                      @) Nl                                                                                              s            :
NW
                                                                                                                                                          v   s     N     1
s
_                                                                                                                                                       S     i     W
:
                                                                            .
v
                                                                                                                                                            -         A
s
                                                                                                                                                                .-
N
                                                                                                                                                                c      t
1
                                                                                                                                                          c           I
W
                                                                                                                                                                      D
_
                                                                                                                                                                .
.
_                                                                                                                                           i
S
                                                                                                                          ;.       81
i
                                                                ER
A
                                                                                                            i
-
                                                                                                                                                a    l
.-
                                                                                                                                                        p
t
                                                                OE
c
                                                                TL                                                       X
I
                                      X                 d
c
                                                                0O
D
                                                                8
.
                                                                8    O
i
                                                                      C c                                                 :,       1 i'
_
                                                                                                                                              i
;.
                                                                            ns
81
                                                                            ik
a
                                                                            cE
i
                                                                            nL                                                                   nF
p
-
ER
                                        -
l
                                        F
OE
                                                a
TL
                                                E
X
                                                  8
X
                                                      M                X e.
0O
                                                                            A0
i
                                                                            u0
88 O
                                                                          CC
d
                                                                                  er
C c
                                                                                                                          X
:,
                                                                                                                          ; , le L
1 i'
                                                                                                                                    l
ns
                                                                                                                                            )i4
ik
                                                                                                                                                                                  .
n
                                                                                                                                                                                  0
M
                                                                                                                                                                                  . E
CC
                                                                                                                                                                                      s
cE
                                                                                                                                                                                      a
nL
                                            DC                            wtB
A0
                                            E
F
                                            EN
.
                                            PE
-
                                                                          sF
X
                                                                          C                                i
l
                                                                                                                                                ,#                                .L00
0
                                                                                                                                                                                      C
u0
                                            SR
F
                                                                                        S
-
                                                                                          ,
8
                                                                                                                                                rt   I
X e. r
                                                                                                                                                                                  C,
s
                                                                                                                          X
)i4
                                                                                                                                                                                  ,
a
                                    I
a
                                    ?
E
                                                    T
. E
                                                      )$                              h                                  :lls[
eB
                                                                                                                                            yi6
; , le L
                                                                                                                                            .
DC
                                                                                                                                                      I
wt
  -                                                              SX                                                                                  i!
.L
                                                                                                                                                m9
,#
                                  W
0
                                                                  MM                                        i
EN
                                *                                                                                        X
E
                                                                                                                                                      1
sF
  .
0
  _
C
                  .
PE
                        -
i
                                                                      .
C
                                                                    S. NUO
SR
                                                                    EArP
,
                                                                        UEO
rt
                                                                            tLI
C,
                                                                                      R
S
                                                                                                            "                              R        ".
I
                                                                                                                                                          S
)$
                                  H
h
                                                                                                          NN                       D g E.       -
yi6
                                                                                                                                        v
X
                                                      (            t
,
                                                                    cErC
I
                                                                                                                                    I
?
                                                                                                                                    Ul
T
                                                                                                                                    La X   M  c= L"
W
                                                                                                                                              ''
.
                                                                                                                                              =C
:lls[
  _
Ii!
                                                                    eLr/D                               S                          FDQ                                          2
SX
                                      X                             fBuAI .O S
-
                                                                    f
m
                                                                                      I
MM
                                                                                                                                                                                =
*
                                                                    M Aq
i
                              y8                           ,
9
                                                                    2    CElpC                                                                = 0' ,                            >
1
                                                                                                                                                                                >
X
                                                                                                                                                                                >
R
                                      o                                                                 S                                                                   ,
"
                                                                                    "                                                                                      S
.
_
.
                                    $o
.
                                              g
tLI
                                                                  CI
S
                                                                  2
_
                                                                    KM
S. U E O
                                                                          nE
R
                                                                                  R
".
                                                                                        S
NUO
                                                                                          e
EArP
                                                                                                C
NN
                                                                                                D
D g E.
                                                                                                E,
-
                                                                                                R.E
H
                                                                                                U.   ,
c L"
                                                                                                          M'                                                                        .
-
                                                                                                                                                                                        -
v
                                                                          oL
=
                                        3, s     , g6            ACRO                           i
IUl M
                                                                                                                    X
(
                                      -
tcErC
                                                        #J
La X
                                                          -
_
                                                                  I    KoO
=C
                                                                          nC                                                       cL                                                   _
''
  _
2
                                                                                                                                2  SIM
eLr/D
                                                                                                                                  BO
X
                                                                                                                                                                                        _
fBuAI .
                                                                                                                                                                                        _
S
                                                                                                                                                                                        _
FDQ
                              -                        }lL                                                         a
= 0' ,
                                                                                                                    E
I
                                                                                                                                1
M'
                                      X                                   mE
M Aq
                                                                                  R             4 -                T.
O S
                                                                                                                    U o.
f
                                                                                                                                i
=
                                                                                                                                                    o
CElpC
                                        j ;1
>
                                        -
2
                                                  I
y8
                                                      4l  -
,
                                                                  ACRO
"
                                                                  l    3oO
>>
                                                                            uL
,
                                                                          nC
o
                                                                                                n,
S
                                                                                                p.
S
                                                                                                    E              a.mw
$
                                                                                                                    r0
CI
                                                                                                                              c                  $
KM
                                                                                                                                                    o
C
                                                        6 J                                      a.
o
                                                                                                S.
DE,
                                                                                                                    c
R.
                                                                                                                                                  $g
U.
                                                        1LLXmR
.
                                                        1
E
                                                                  EMiE
g
                                                                                                A
R
                                                                                                        S
_
                                      X                           S
nE
                                                                  EN0AH
e
                                                                          LTI
X
                                                                                                        Y
-
                                                                                                        S            2                              X
, g6
                                                                        E0W                                           r
2
                                  H            3      )IDCC                                             V
oL
                                                                                                        R
S
                                                                                                                      t
,
                                                                                                                      s
_
                                              @ ((
3, s
                                                                                                                      '
ACRO
-
i
                                  D
-
                                                J
KoO
                                                                                                                      v
I
                          _                                               a                             &                                                                              _
SIM
                      g                             ,6                      ,M
nC
                                                                                        e
cL
                                                                                                                                                    X
#J
                                                                                                                                                                                        _
_
                                                                                                                                                                                        _
_
                                                                                                                                                                                        _
_
                                                                                                                                                                                        _
BO
                                                                                                                                                                                        _
2
                                  ]g           I
_
                                                                          y1                                                                                                            .
-
                                                                                                                                                                                        _
}lL
                                                                                                                                                                                        _
a
                                  D
1
                                                                                                                                                                                        .
X
.                                                                            3                                                                                                          .
R
                                                              J                                 * ,                                                 o
4
                                  X                   GL
E
                                                                                                          I
i
                    ,                         h
a.mw
                          ,
T.
                          -                                                                                           -
-
                                                                                                          [>
U o.
                                                      I
o
-
mE
                                                                                                    .
4l
                        yo LYr                                                                t S
uL
                                                                                              eNA
n,
                                                                                                                      c=
j ;1
                                                                                                                      S '
I
                                                  W                                           t
r0
                        S    r                    D          a                              Tm
$
                                                                                              i
ACRO
                                                                                                                        .
p.
                        S
-
                        *
c
                            u                    F J" t
-
_
3oO
                            S
l
                                                  XJu
Ea.
                                                  U
o
                                                  A           C
nC
                                                                                              S .A
6
                                                                                                                                                                                        _
c
                                                                                                                                                                                        _
J
                                                                                                                                                                                        _
$g
  - -~
S.
                                                                                                                                                                                        _
1LLXmR
                                                                                                                                                                                        -
A
              ){\     !                                                                                             '
S
                                                        i                     il           !                                   ,
EMiE
                                                                                                                                                  i
Y
1
X
X
2
S
LTI
S
EN0AH
E0W
V
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R
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H
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sR *
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t'
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S
                                                                                                        S
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                                                                                                        S
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sD
                                                                    -
aE
                                                                    -
S
                                                                    ~                                         .
E
                            -
EH
                                                              h-                                   g                             3
S"
                                                                                                                                8
S
                            - A
.
                                  0
EH
                                              3 3
E
                                                    1
-
                                                                A-
-
                                                                3                   0         ' ,                             1
.
                                                                                                                                    -
~
                                                                                                                                                  A
-
                                  4               4           4                   4             .                             -              W    Y
h-
                                                                                                                                2                 D L
g
                                                                                                                                  :l                   N
3
                                                                                                                                E                 :
- A
                                                                                                                                                      O
3 3
                                                                                                                                T       4       F
A -
                                                                                                                                A       7       F E
8
                                                                                                                                D       5       A S
-
                                                                                                                              l
1
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A
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0
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1
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3
                                                                  ,,       ?                                                     -
0
                                                                                                                                s        5       A   I
' ,
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-
                                                                                                                                a        1       N   l
W Y
                                                                                                                                R         .     D A
4
                                                                                                                                  -     C             R
4
                                                                                  (                     {                       S       1
4
                                                                                                                                        3        :   T
4
                                                                                  jd                                           T                 I
.
                                                                                                                                                  I
2
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D L
                                      g                     g
:l
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N
                                                                                                                                C
E
                                                                                                                                  -      y
:
                                                                                                                                        E
O
                                                                                                                                                  A
T
                                                                                                                                        t
4
                                                                                                                                                R
F
                                                                                                                                M        s      D
A
                                                                  Q
7
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F
                                6
E
                                1                                             g
D
                                                                              ~
5
                                                                            4
A S
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2
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U
                                                                                                H
l
                                            2h A
l
                                              4
.
                                                                                                )*5                                                     .
G
                                  3                     3                                       5
4
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4
                                        8
N
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1
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7
                                                    1
I
                                                                    9
,,
                                                                                                                                                    E
?
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-
                                                                                                                                                    U
5
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A
                                                                                                                                                    S
lI
                                                                                                                                                    E
sa
                  :     A                                                                                                                          R
1
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N
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l
                                ^                                 A 2   4                                                                         T
R
                                                                                                                                                            _
.
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D A
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-
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C
                                                              ,     3                                                                                I
R
                  3
(
                                                              ,
{
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S
                                                                                                                                                    I
:
                                                                                                                                                    I
T
                                                                        F                                                            .               A
13
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jd
                                                                        S                                                       U               TI M
T
                        L                                             N                                                       D             U
II
                      CF                                         I
S
                                                                        S                                                       E               DO
W
                      RF                                                                                                        C               E
g
                        U                      ,,                                                                             A               C1
g
                        S                                                                                                      I
-
                                                                                                                                B              lAH
y
                                                                                                                                                aN
A
                                                                                                                                                    I
g
                                                                                                                                t
g
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g
                                                                                                                                O              R2
C
                              g                                                                                                         , O
E
                                                                                                            I                        , .             T
R
                                                                                                                                S S             .I
M
                                                                                                                                              S N
t
                                                                                                                                2 2                 U
D
                                                                                                                                              2
Q
                                                                                                                                R R                 R
s
                                    a                                                                   2                       OO R O
a
                                  g                                                                                           I I
g
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A
                                                                                                                                              I
6
                                                                                                                                T T                 T
g
                                    4                                                                     s                     I I T II
1
                                    0                                                                   1                     N I I NI
~
                                    5                                                                   5                     UU             I     U
4
                                    1                                                                   1                     1     0
2
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1
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2h
                                                                                                                                              1
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                                                                                                _.
_.
  '.: .
'.:
  ^*
.
^*
4
4
                                NUCLEAR SERVICE WATER SYSTEM EXPERIENCE
NUCLEAR SERVICE WATER SYSTEM EXPERIENCE
                    l. ESTABLISHMENT OF BASIS FOR RN SYSTEM OPERABILITY
l.
                        *   RN SYSTEM PRE-OPERATION FUNCTIONAL TEST COMPLETION DATES
ESTABLISHMENT OF BASIS FOR RN SYSTEM OPERABILITY
                                    7/25/79                       -
*
                                                                    UNIT 1
RN SYSTEM PRE-OPERATION FUNCTIONAL TEST COMPLETION DATES
7/25/79
-
UNIT 1
11/12/82
-
UNIT 2
i
i
                                  11/12/82                      -
*
                                                                    UNIT 2
NRC PRE-OPERATIONAL INSPECTION DATES COVERING RN SYSTEM
                        *    NRC PRE-OPERATIONAL INSPECTION DATES COVERING RN SYSTEM
TESTING
                            TESTING
11/03/78
                                  11/03/78                       -
UNIT 1
                                                                    UNIT 1               INSPECTION REPORT 369/78-33
INSPECTION REPORT 369/78-33
                                    8/16/83                       -
-
                                                                    UNIT 2               INSPECTION REPORT 370/82-19
8/16/83
                        *   SURVEILLANCE TESTING IMPLEMENTATION DATES
UNIT 2
INSPECTION REPORT 370/82-19
-
*
SURVEILLANCE TESTING IMPLEMENTATION DATES
a
a
                                    1/06/80                       -
1/06/80
                                                                    UNIT             1, TRAIN A
UNIT 1, TRAIN A
                                    2/06/80                       -
-
                                                                    UNIT             1, TRAIN B
2/06/80
,                                  2/22/83                       -
UNIT 1, TRAIN B
                                                                    UNIT             2, TRAIN B
-
                                    2/23/83                       -
2/22/83
                                                                    UNIT             2, TRAIN A
UNIT 2, TRAIN B
                            IWV AND IWP TESTING WOULD HAVE BEEN IMPLEMENTED DURING
-
                            THESE TIME FRAMES.
,
                        *    THE PRE-OPERATIONAL TESTS, lWP TESTS, IWV TESTS AND ESF
2/23/83
                            TESTS WERE OUR STANDARDS FOR ESTABLISHING AND
UNIT 2, TRAIN A
                            MAINTAINING RN SYSTEM OPERABILITY.
-
                  11. MAINTENANCE OF COMPONENTS BASED ON MONITORING OF OPERATIONAL
IWV AND IWP TESTING WOULD HAVE BEEN IMPLEMENTED DURING
                        PARAMETERS
THESE TIME FRAMES.
                        *
THE PRE-OPERATIONAL TESTS, lWP TESTS, IWV TESTS AND ESF
                            REFER TO LIST OF EQUlPMENT CLEANINGS
*
                        *
TESTS WERE OUR STANDARDS FOR ESTABLISHING AND
                            PERFORMANCE MONITORING PROGRAM BEGAN DEVELOPMENT IN
MAINTAINING RN SYSTEM OPERABILITY.
                            MARCH, 1984                                                                             l
11. MAINTENANCE OF COMPONENTS BASED ON MONITORING OF OPERATIONAL
                                                                                                                      l
PARAMETERS
                                                                                                                      l
REFER TO LIST OF EQUlPMENT CLEANINGS
      ________ _ -                 - _ _ _ _ _ _ _ _ _ _ _ _ _ _      - _ _ _ _ _ _
*
*
PERFORMANCE MONITORING PROGRAM BEGAN DEVELOPMENT IN
MARCH, 1984
l
-
-
-


  'l.
'l.
    t
t
  ''.
' ' .
      Ill. BEGAN EVALUATING RN SYSTEM HX'S FOR FOULING EVEN THOUGH
Ill. BEGAN EVALUATING RN SYSTEM HX'S FOR FOULING EVEN THOUGH
              THERE WERE NO INDICATIONS OF FOULING
THERE WERE NO INDICATIONS OF FOULING
              *
DATE WHEN UNIT 1 COMPONENT COOLING (KC) HX'S WERE
                  DATE WHEN UNIT 1 COMPONENT COOLING (KC) HX'S WERE
*
                    EVALUATED FOR FOULING WITHOUT INDICATIONS OF A FOULING
EVALUATED FOR FOULING WITHOUT INDICATIONS OF A FOULING
                    PROBLEM
PROBLEM
                          9/01/84
9/01/84
              *   DATE WHEN KC HX'S WERE CLEANED
*
                                11/84 - UNIT 1
DATE WHEN KC HX'S WERE CLEANED
                          6/85 - 7/85 - UNIT 2
11/84 - UNIT 1
            *     EVALUATION AND INSPECTION / CLEANING DID NOT DETERMINE
6/85 - 7/85 - UNIT 2
                  THAT KC HX'S WERE INOPERABLE
*
        IV. RN SYSTEM OPERABILITY REEVALUATION BASED ON 1A RN PUMP TEST
EVALUATION AND INSPECTION / CLEANING DID NOT DETERMINE
            RESULTS
THAT KC HX'S WERE INOPERABLE
            *
IV. RN SYSTEM OPERABILITY REEVALUATION BASED ON 1A RN PUMP TEST
                  DATE WHEN A FLOW MEASUREMENT PROBLEM ON 1A RN PUMP WAS
RESULTS
                  IDENTIFIED
DATE WHEN A FLOW MEASUREMENT PROBLEM ON 1A RN PUMP WAS
                          10/04/85
*
            *
IDENTIFIED
                  A REEVALUATION OF OPERABILITY CRITERIA WAS BEGUN TO
10/04/85
                  REFOCUS OPERABILITY CONCERNS FROM THE RN PUMP TO THE RN
*
                  SYSTEM AS A WHOLE
A REEVALUATION OF OPERABILITY CRITERIA WAS BEGUN TO
        V. ACTION ITEMS RESULTING FROM REEVALUATION OF RN SYSTEM
REFOCUS OPERABILITY CONCERNS FROM THE RN PUMP TO THE RN
            OPERABILITY CRITERIA
SYSTEM AS A WHOLE
            *
V. ACTION ITEMS RESULTING FROM REEVALUATION OF RN SYSTEM
                  BEGAN THE PERFORMANCE MONITORING PROGRAM ON
OPERABILITY CRITERIA
l                 RN HX'S ON 11/01/85
BEGAN THE PERFORMANCE MONITORING PROGRAM ON
            *
*
l
RN HX'S ON 11/01/85
*
THE RN SYSTEM TESTING PLAN WAS SUBMITTED TO
'
'
                  THE RN SYSTEM TESTING PLAN WAS SUBMITTED TO
REGION 11 ON 12/01/85
                  REGION 11 ON 12/01/85
THE UPDATED RN SYSTEM TESTING PLAN WAS SUBMITTED
            *
*
                  THE UPDATED RN SYSTEM TESTING PLAN WAS SUBMITTED
TO REGION ll TO INCLUDE TESTING OF ALL 62 RN
                  TO REGION ll TO INCLUDE TESTING OF ALL 62 RN
HX'S AND RESOLVE 1A RN PUMP FLOW INDICATION PROBLEM
                  HX'S AND RESOLVE 1A RN PUMP FLOW INDICATION PROBLEM
ON 12/18/85
                  ON 12/18/85
!
!           NOTE:   IN ALL THE TESTING AND ANALYSIS DONE IN 1985, WE
NOTE:
                      DID NOT DETERMINE THAT ANY OF THE HX'S EVALUATED WERE
IN ALL THE TESTING AND ANALYSIS DONE IN 1985, WE
                      INOPERABLE.
DID NOT DETERMINE THAT ANY OF THE HX'S EVALUATED WERE
INOPERABLE.
1
1
f
f


  _ _ _ _
_ _ _ _
, .,
, .,
    ,
,
    't
't
                                EQUIPMENT CLEANINGS                       j
EQUIPMENT CLEANINGS
                                                                            !
j
          *
!
            LOWER CONTAINMENT VENTILATION HX FOULING WAS IDENTIFIED AS
*
            ONE OF THE FACTORS IN THE LOWER CONTAINMENT COOLING PROBLEM
LOWER CONTAINMENT VENTILATION HX FOULING WAS IDENTIFIED AS
                      10/22/82
ONE OF THE FACTORS IN THE LOWER CONTAINMENT COOLING PROBLEM
            NOTE:     (A)   FOULING OCCURRED AT LAKE TURNOVER IN THE FALL.
10/22/82
                            ONLY TIME WE HAD TO CLEAN.
NOTE:
                      (B)   BIOFOULING WAS EVIDENT DUE TO HOT AIR ON SHELL
(A)
                            SIDE.
FOULING OCCURRED AT LAKE TURNOVER IN THE FALL.
          *                                                                '
ONLY TIME WE HAD TO CLEAN.
            CONTROL ROOM VENTILATION (SHARED BETWEEN UNITS 1 AND 2)
(B)
                      TRAIN A                   TRAIN B
BIOFOULING WAS EVIDENT DUE TO HOT AIR ON SHELL
                      11/19/82                       3/83                   {
SIDE.
                      10/03/83                   1/07/85
CONTROL ROOM VENTILATION (SHARED BETWEEN UNITS 1 AND 2)
                      12/19/83                   10/21/85
*
                        5/30/84                 11/05/85
'
                      10/31/84
TRAIN A
                        9/25/85
TRAIN B
                      10/24/85
11/19/82
                      10/31/85
3/83
          *
{
            PENETRANT / DISPERSANT ADDED TO THE RN SYSTEM IN ATTEMPT TO
10/03/83
            CLEAN LOWER CONTAINMENT COOLING UNITS
1/07/85
                      4/27/83
12/19/83
          *
10/21/85
            REACTOR COOLANT PUMP MOTOR COOLERS
5/30/84
                      UNIT 1                     UNIT 2
11/05/85
                      12/31/84                   8/10/84
10/31/84
                                                11/08/85
9/25/85
10/24/85
10/31/85
PENETRANT / DISPERSANT ADDED TO THE RN SYSTEM IN ATTEMPT TO
*
CLEAN LOWER CONTAINMENT COOLING UNITS
4/27/83
REACTOR COOLANT PUMP MOTOR COOLERS
*
UNIT 1
UNIT 2
12/31/84
8/10/84
11/08/85


                                                                      .
.
  ASSUMPTIONS
ASSUMPTIONS
  1.   ALL SAFETY RELATED EQUIPMENT REQUIRE FLOWS CONCURRENTLY         5
1.
                                                                        "
ALL SAFETY RELATED EQUIPMENT REQUIRE FLOWS CONCURRENTLY
        THROUGHOUT DESIGN BASIS EVENT.                                   g
5"
                                                                        5
THROUGHOUT DESIGN BASIS EVENT.
                                                                        s
g
                                                                        4
5s
                                                                        m
4
  2.   HEAT EXCHANGERS DESIGNED FOR MAXIMUM POND TEMPERATURE OF 95 F.
m
2.
HEAT EXCHANGERS DESIGNED FOR MAXIMUM POND TEMPERATURE OF 95 F.
:
:
4
4


                            FLOW AND FOULING DESIGN MARGIN
FLOW AND FOULING DESIGN MARGIN
                          AFFECTS ON CONTAINMENT PEAK PRESSURE
AFFECTS ON CONTAINMENT PEAK PRESSURE
                            (CONTAINMENT DESIGN = 14.9 PSIG)
(CONTAINMENT DESIGN = 14.9 PSIG)
              (x10bfuHROF)       (x10bIUHROF)         tbs     T0          OfE'
(x10bfuHROF)
    CLEAN
(x10bIUHROF)
  OfFFkbfEik       5.18               8.11               5000     8000   -
tbs
OfE'
T0
CLEAN
OfFFkbfEik
5.18
8.11
5000
8000
-
DESIGN
DESIGN
(FSAR)             2.94               5.00               5000     8000 12.36
(FSAR)
  75% NS
2.94
DEGRADED
5.00
FROM DESIGN       0.735               5.00               5000     8000 14.42
5000
8000
12.36
75% NS
DEGRADED
FROM DESIGN
0.735
5.00
5000
8000
14.42
REDUCED FLOWS
REDUCED FLOWS
DEGRADhDHXs       1.47               2.98               3800     6000 13.59
DEGRADhDHXs
                                    0 = U3 (LMTD)
1.47
                                    U = (-00 LING, FLOW)
2.98
3800
6000
13.59
0 = U3 (LMTD)
U = (-00 LING, FLOW)


                                      .                               ..
.
  l-
..
  .g
l-
    .
.g
                                        NUCLEAR SERVICE WATER
.
                        ESSENTIAL COMPONENT FLOW REQUIREMENTS                                           ;
NUCLEAR SERVICE WATER
,
ESSENTIAL COMPONENT FLOW REQUIREMENTS
,
4
4
                                  DESIGN F OWS                             !hkE
!hkE
      COMPONENT                     (FSAR                             AChl0WS
DESIGN F OWS
      KD Hx                             900                               900
AChl0WS
      KC Hx                           8000                               6000
COMPONENT
(FSAR
KD Hx
900
900
KC Hx
8000
6000
i
i
      NS Hx                           5000                               3800
NS Hx
                                                                                                        ,
5000
      VC/YC CONDENSER                   775                               775
3800
,
VC/YC CONDENSER
775
775
I
I
      KF ES COOLER                       20                                 15
KF ES COOLER
      NS ES COOLER                       45                                 20
20
      ND ES COOLER                       45                                 20
15
NS ES COOLER
45
20
ND ES COOLER
45
20
:
:
      NV PUMP COOLERS                   28                                 15
NV PUMP COOLERS
                                                                                                        ;
28
;    NI PUMP COOLER                     20                                 20
15
NI PUMP COOLER
20
20
;
;
i
i
i
i
!
!
<
<
      ,   + , . _ _ . -- n - - - - -     -       -- ,,     - - - - .       ,. , ,-.----.,,--,.-,.--y.
,
+ , . _ _ .
--
n
- - - - -
-
--
,,
-
- - - .
,.
,
,-.----.,,--,.-,.--y.


                                                                                                                                        - K ,*. ,,
K ,*. ,,
                                                                                                          ADDITIONAL DESIGN MARGINS
-
                                                                                                        1. LOWER SNSW POND TEMPERATURE
ADDITIONAL DESIGN MARGINS
                                                                                                        2. HIGHER ICE WEIGHT
1.
                                                                                                        3. LOWER RWST TEMPERATURE
LOWER SNSW POND TEMPERATURE
                                                                                                                              .
2.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ . - _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _____
HIGHER ICE WEIGHT
                                                                                                                                    w -
3.
LOWER RWST TEMPERATURE
.
. . .
. -
.
.
.
.
w
-


          .- -
,.
,.
    s.                                             ,
.-
  a
-
    ,,                  ATTACIGIENT 3
s.
                                                    i
,
    T. .                                           <
a
                  PERFORMANCE MONITORING PROGRAM
ATTACIGIENT 3
              * Reliability. Efficiency and
,,
                AvaiIabiiity
T. .
              * Monitors the overall health of
i
                equipment
<
              * Development begun in March. 1984
PERFORMANCE MONITORING PROGRAM
              * Tangible results already being
* Reliability. Efficiency and
                reaIized
AvaiIabiiity
                                                  .
* Monitors the overall health of
equipment
* Development begun in March. 1984
* Tangible results already being
reaIized
.


  . 3' .
. ' .
    l'
3
    .
l'
        NUCl. EAR SERVICE WATER PUMP (RN) 1A
.
        * RN Pump 1A did not meet its quarterly
NUCl. EAR SERVICE WATER PUMP (RN) 1A
            IWP acceptance criteria (10/4/85)
* RN Pump 1A did not meet its quarterly
        * Replaced impeller (10/5 - 10/6/85)
IWP acceptance criteria (10/4/85)
        * Performed new pump head curve /lWP
* Replaced impeller (10/5 - 10/6/85)
          baseline test (10/7/85)
* Performed new pump head curve /lWP
        * Troubleshooting
baseline test (10/7/85)
        * Evaluated the pump acceptance criteria
* Troubleshooting
          based on the actual system demand
* Evaluated the pump acceptance criteria
        * Conducted the pump head curve using
based on the actual system demand
          the 2A and 1A KC flow elements in
* Conducted the pump head curve using
          series with the 1A RN flow element
the 2A and 1A KC flow elements in
        * Using the most conservative head curve
series with the 1A RN flow element
          results 1A RN pump was declared
* Using the most conservative head curve
          operable (10/11/85)
results 1A RN pump was declared
        * Optimum replacement was a calibrated
operable (10/11/85)
          84" flanged spool-piece with a 0.831
* Optimum replacement was a calibrated
          beta ratio orifice
84" flanged spool-piece with a 0.831
        * Installation (February 26-28,1986)
beta ratio orifice
        * 1A RN Pump head curve conducted with
* Installation (February 26-28,1986)
          new flow element (March 3, 1986)
* 1A RN Pump head curve conducted with
        ** Summary - The pump was never
new flow element (March
            inoperable, fouling of the flow
3,
            element resulted in errors in the
1986)
            conservative direction.
** Summary - The pump was never
inoperable, fouling of the flow
element resulted in errors in the
conservative direction.
L
L


                                                                                                                                                                                                        I
I
            6
6
            8
8
              -             *                                                                                                                                                                           [
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    .
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                                                                                                                                                                        .
                                                                                                                                                                        ,h.
    n.                                                        ee                                  r
    .
    -                                                 're T T                                    g                      ,                                .\                                                  a
                                                        ree                                      g
                                                                                                                                                          4*
                                                        uvv
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                                                        t rr
                                                        cuu                                                                                        \                                                          w
                                                        aCC                                                                                      \                                                      '
            T                                          f                                                                                      n
            A                                                                                                                              s'
                                                                                                                                                                                                                u
                                                                                                                      .
            R                                            udd                                                        l
            W                                          naa                                                        l
                                                                                                                    l
                                                                                                                                          s
            O                                            aee
            L
                                                      f
                                                      F I I
                                                            I l
                                                                                                                                  's
            F                                                                                                                                                                                                    s
              .
                                                                                                                                                                                                              b
                                                                                                                                s
  C        S                                      ,
                                                                                                                              '
  U        V                                                                                                            i
  !
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            D
            A                                                                                            ,
                                                                                                                        N                                                                                        w
      .
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            l
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                                                                                                                      s
  U.
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                                                                                                                                                                                                              t
            C
            I
            M                -
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                                                                                                                ,-,                                                                                                w
        . AN      !l                        i;
            Y
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                  !p!  i                !
                                                                                              ,            i
                                                                                                            l
                                                                                                              s's                                                                                                sr
            A                                                                                      s
                                                                                                                                                                                                                  a
                                                                                          ,
                                                                                                                                                      ~
            1    l                                                                  L
                                                                                      !
            P
            f    I
                                                                                                                                                                                                                  e
            l
            U                                                                      .
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            N
                                                                                                                                                                                                            d
            R                                                                    s
                                                                                                                                    5                        m
                      '                                                      s'                                                    8          "            p      6      "          mmm
                                                                                                                                                                                        ppp                  i
                                                                            n'
                                                                            -                    j                                -        5 mmg                8      1
                                                                                                                                    8          1 pp                  -      2 mmggg
                                                                                                      -
                                                                                                                                                                    3          pp
                                                                                                                                              4    gg0                    4 gg000
.
.
.                l.
R
                                                                        .
vr
                                                                                                  3
N
                                                                                                    h
u
                                                                                                    L
l
                                                                                                        -
l
                                                                                                        .
C
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5
                                                                                                                                    1
s.%-
                                                                                                                                    f
l
                                                                                                                                      -
0
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68
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'
                                                                                                                                                    063
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d8 -
                                                                                                                                                                    3
e8
                                                                                                                                                                    f
q
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i 3 -
                                                                                                                                                                            4
l-
                                                                                                                                                                            1
q
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\\
                                                                                                                                                                                      000
f
                                                                                                                                                                                00 0 0,0,0,                !
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                                                                                                                                                                                                              I
_
                                                                _                                                                                                                                            I
l
                                                                -
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l
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'
                                                                                                                                                                                                            .m
h
          .
!
                    !
'
                    !
.
                                                            '
p
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                                                                                                                                                                                        p
.
                                                                                                                                                                                g
L
                                                                                                                                                                  .
J'
                                                    .                          L
.
                      i:ln                                                                g                                                                  J'                                           .I
#
                                                                #           f
g
                                  .
b,
                                              ?                              M            3.
;
                                                                                              .
g
                                                                                                            g                b,        ,
<"
                                                                                                                                        ;
g
                                                                                                                                                    g           <"
3
                                                                                                                                                                  .
?
                                                                                                                                                                                        3
i:ln
                                                                                                                                                                                        l
g
                                                                                                                                                                                                ?
f
                                                                                                                                                                                                l
,
                          !.
.
                                  .
M
                                                                                                                                                                                          .
3.
                                                                                                                                                                                                            L
l
                                                                                                                                                                                                                      .
l
                                                                                                                        P d Nv h x Q $ H $. m y m @
.
                                                                                                                                                                                                                    -
L
.
?
.
!.
P d Nv h x Q $ H $. m y m @
.
-


                                                                                                              .
.
'
'
      t ',
t ',
.   .
.
      l'
.
      .-
l'
                                    CONTAfNMENT SPRAY (NS) HEAT EXCHANGER
.-
                                    * 1A NS Heat Exchanger had a high
CONTAfNMENT SPRAY (NS) HEAT EXCHANGER
                                      differential pressure
* 1A NS Heat Exchanger had a high
                                    * Commission expressed concerns of
differential pressure
                                      biological attack of s t a i ti l e s s steel
* Commission expressed concerns of
                  '
'
                                      tubes
biological attack of s t a i ti l e s s steel
                                    * Testing Performed:
tubes
                                          1. Structural Integrity Test
* Testing Performed:
                                          2. Minute           Leakage Test
1.
                                          3.   Heat Balance Test
Structural Integrity Test
                                    * Structural Integrity and Minute
2.
                                      Leakage Test indicated insignificant
Minute
                                      leakage
Leakage Test
                                    * Visual Examination of the tubes
3.
                                    * Heat Balance Testing quantified the
Heat Balance Test
                                      extent fouling had occurred
* Structural Integrity and Minute
Leakage Test indicated insignificant
leakage
* Visual Examination of the tubes
* Heat Balance Testing quantified the
extent fouling had occurred
"
"
                                    * Peak Containment Accident Pressure
* Peak Containment Accident Pressure
                                      CLOTIC) calculations showed the heat
CLOTIC) calculations showed the heat
                                      exchanger could still perform its
exchanger could still perform its
                                      function
function
                                    * Cleaning iterations
Cleaning iterations
                                    * Tested and cleaned the other NS heat
*
                                      exchangers based on 1A experience
* Tested and cleaned the other NS heat
                                    * * Summary:
exchangers based on 1A experience
                                          1. NS Heat Exchangers are intact
* * Summary:
                                        2. The NS Heat Exchangers were
1.
                                                fouled; however reanalysis proved
NS Heat Exchangers are intact
                                                operability
2.
The NS Heat Exchangers were
fouled; however reanalysis proved
operability
j
j
  .
.
          ----gc*   - -
----gc*
                        y e,- ----     --e- -     m --pq a--n-- - me---w --e,--r 1 =ya---e- -~ ------e-- P
-
-
y
e,-
----
--e-
-
m
--pq
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-
me---w
--e,--r
1
=ya---e-
-~
------e--
P


      _   . . _ .                           _                 . _ .
_
                      _ _           _                                 _     _   _.       _ -_.
. . _ .
                                                                                                      *
_ _
                                                                                                  .*
_
                                                                                                    .   , , ,
_
  ._
. _ .
                              Containment Spray lleat Exchanger Testing
_
_
_.
_
-_.
*
.*
.
, , ,
._
Containment Spray lleat Exchanger Testing
i
i
t
t
i
i
.,
.,
                                          *                                lleat Exchanger
lleat Exchanger
                            .
*
                                                ,
.
                                                  ,
,
        ~
,
I
I
i
~
                              Temperature)
i
Temperature)
i
Y
m
:
:
                      m                    i                        Y                  a
a
                                        I     i                     'f
I
i
'f
,
,
l
l
,
,
!
!
          Containment
Containment
          Spray Pump                                                                Flow
Flow
f
f
Spray Pump
I
I
!
!
)/
>
>
                                                                          )/
FWST
                          FWST
(400,000 gal.)
3
3
                      (400,000 gal.)
                  '
i
i
'
l
l
.
.


  ",
",
,
"
McGuire Nuclear Station
*
.
'
4/,w
80 -
74.7
!!!!
f h
. . .
. . . .
....
/) N$
I;; p V !f V/
g
d d d ild $
As Found
C!ng 1
Cing 2
40
36
:: 7
.
[
27
27
! ::-
Fy F/ 9
/ /
/ p/)/ / /j/ '-9
..
'
-
:1s s s s 4.
.
.
.
.
_ _ - .
.-.
. _ .
-
_ _ - _ _ -
.
_
. - ,
,
,
  *
  "
  .                        McGuire Nuclear Station
      '
                                                                          4/,w
        80 -
                                                                  74.7
    !!!!
    I;; p V g!f V/
                              ...            ....      ....
                                                                  f h$
                  d d d ild $/) N
                  As Found  C!ng 1        Cing 2
        40
                      36
    .
    [
        :: 7                    27          27
    ! ::-                  Fy            F/      9
    '
          ..
                                          / / /j/
          -
                  /        p/)/ / '-9
          :1s s s s 4. .          .                            .  .
          _ _ - .            .-.    . _ .        - _ _ - _ _ -        . _  . - , ,


                                                  ___ _ ___ .
___ _ ___
  l*
.
    '
l*
'
s
s
  P
P
  e
e
      OTHER HEAT EXCHANGERS and FLOW BALANCE
OTHER HEAT EXCHANGERS and FLOW BALANCE
      * Began evaluation and procedure
* Began evaluation and procedure
        generation for testing essential heat
generation for testing essential heat
        exchangers (10/24/85)
exchangers (10/24/85)
      * 1A Train RN flow balance performed
* 1A Train RN flow balance performed
        aligned to low level intake (12/17/85)
aligned to low level intake (12/17/85)
      * 1A NV Pump Speed Reducer Oil Cooler
* 1A NV Pump Speed Reducer Oil Cooler
        cleaned (12/20/85)
cleaned (12/20/85)
      * Conducted test using NSWP as suction
* Conducted test using NSWP as suction
          (1/27/86)
(1/27/86)
      * Inadequate flow to some heat exchangers
* Inadequate flow to some heat exchangers
      * Reanalyzed the necessary flow rates to
* Reanalyzed the necessary flow rates to
        the KC and NS heat exchangers
the KC and NS heat exchangers
      * Design review of alignment
* Design review of alignment
        configuration to properly conduct flow
configuration to properly conduct flow
        balance to meet all design assumptions
balance to meet all design assumptions
      * Conducted the 1A Train flow balance
* Conducted the 1A Train flow balance
        throttling flow to the KC and NS heat
throttling flow to the KC and NS heat
        exchangers, aligned to NSWP with 6000
exchangers, aligned to NSWP with 6000
        gpm supplied to other unit (1/28/86)
gpm supplied to other unit (1/28/86)
      * Performed other flow balances
* Performed other flow balances
        (1/30 - 2/28/86)
(1/30 - 2/28/86)
      * Began extensive cleaning, testing and
* Began extensive cleaning, testing and
        inspections of all essential heat
inspections of all essential heat
        exchangers (2/3/86)
exchangers (2/3/86)
        Total Cleaned / Tested / Inspected:   54
Total Cleaned / Tested / Inspected:
        Total Number of Hea t -Ex change r s : 62
54
      ** Summary:
Total Number of Hea t -Ex change r s : 62
          Cleaning and testing of all essential
** Summary:
          RN system components is on schedule
Cleaning and testing of all essential
          to meet March 31, 1986 completion
RN system components is on schedule
to meet March 31, 1986 completion


                  .-                     . _ - _ -     . - .     .- - __   -     . - --   -   . _
.-
. _ - _ -
. - .
.-
- __
-
. - --
-
.
_
d
d
  .
.
    .
.
      .
.
    %
%
s
s
        -
-
.
.
l
l
i
i
i                                                     ATTACHMENT 4
i
!
ATTACHMENT 4
!
NRC Inspection Team Confirmatory VA Calculations
't
't
                            NRC Inspection Team Confirmatory VA Calculations
.
.
:          Calculations were performed to evaluate the containment spray heat exchanger UA
j          value used in containment pressure calculations performed by Westinghouse for
;          Duke Power on November 28, 1985 and January 17, 1986. The following nomenclature
          is used in the subsequent calculations:
:
:
.
Calculations were performed to evaluate the containment spray heat exchanger UA
                                                      Nomenclature
j
i
value used in containment pressure calculations performed by Westinghouse for
;
Duke Power on November 28, 1985 and January 17, 1986. The following nomenclature
is used in the subsequent calculations:
:
.
Nomenclature
i
A
-
heat exchanger area
,
,
          A          -
tube inside diameter
                          heat exchanger area
Di
-
;
;
          Di        -
i
                          tube inside diameter
Do
i          Do        -
-
                          tube outside diameter
tube outside diameter
i
.
.
i
1
1          Fi         -
Fi
                          tube inside fouling factor
-
          Fo         -
tube inside fouling factor
                          shell side fouling factor
Fo
-
shell side fouling factor
:
:
          F
F
            oAPP
-
                      -
appropriate shellside fouling factor
                          appropriate shellside fouling factor
oAPP
i
i
i         G         -
mass flux
                          mass flux
i
G
-
4
4
!          hi        -
tube inside heat transfer coefficient
                          tube inside heat transfer coefficient
!
          ho        -
hi
                          shell side heat transfer coefficient                                     i
-
!         K         -
shell side heat transfer coefficient
                          water thermal conductivity                                         ,
i
          K         -
ho
                          stainless steel thermal conductivity
-
            ss
!
;         Pr         -
K
                          water Prandtl number
water thermal conductivity
,
-
K
-
stainless steel thermal conductivity
ss
water Prandtl number
;
Pr
-
1
1
          Re        -
Reynolds Number GD
                          Reynolds Number GD
Re
              D
-
l                                                   y
D
                                                                                                    '
l
y
'
:
:
i         oFo         --
i
                          two standard deviation uncertainty in Fo
oFo
          u
two standard deviation uncertainty in Fo
                      -
--
                          liquid viscosity
u
liquid viscosity
-
!
!
t
t
!
!
f
f
'!
'!
l
li
i _ _ _ ..____._ _ _ _._. _. . _ _ _ . _ _ _ _ -_ ._ _ _ _ _ _ _ _ ._.. _ -_ _ _ ..- _ _ a
..
.
. .
. .
.
-
.
. ..
-
..-
a


          .           _                   _   . .     __ _               .   .       ._ .                 .     . . - . _         -.
.
_
_
. .
__ _
.
.
._
.
.
. . - .
_
-.
;.
;.
i-
i-
!
!
:
:
1 -.       Attachment 4                                         2
1 -.
i
Attachment 4
2
i
!
!
J
J
j               The UA design value for this heat exchanger is 2.95 x 105 Btu /h/ F while
j
                Duke provided Westinghouse with a degraded value of 7.35 x 10' Btu /h/ F,
The UA design value for this heat exchanger is 2.95 x 105 Btu /h/ F while
                24.9% of the design value. Experimentally determined UA (11/22/85) values
Duke provided Westinghouse with a degraded value of 7.35 x 10' Btu /h/ F,
                  indicated that the actual degraded value was ~8.77 x 10' Btu /h/ F, 29.7% of
24.9% of the design value. Experimentally determined UA (11/22/85) values
indicated that the actual degraded value was ~8.77 x 10' Btu /h/ F, 29.7% of
f
the design value.
Confirmatory UA calculations were performed by initially determining a
design heat transfer coefficient for the shell side of the heat exchanger.
f
I
This was done by using design value fouling factors, and assuming that the
I
tube side heat transfer was correctly predicted by the McAdams equation at
j
the design conditions.
,
hiDi
'
k
=0.23 Re .8
1/3
(1)
.
D Pr
;
The UA for the heat exchanger is
!
UA =
A
(2)
p
1_ + F4 po + F
+D
In p
;
1_
+
g
o
g
g
h
Di
hi
Di
2K
UI
g
ss
For the design condition, all values (including UA) are known except h ,
;
g
[
which was determined to be 918 Btu /h/ F/ft2 using equation (2).
The information supplied to Westinghouse by Duke was acquired from experi-
mental testing of heat exchanger 1A on 11/22/85.
The data from this
;
e<periment was used to determine an appropriate value for the degraded UA by
'
j
first determining the as-tested fouling factor.
In order to do this,
experimental flow rates, temperatures, etc. had to be used to determine both
i
tube side (h ) and shell side (ho) heat transfer coefficients appropriate
4
!
for the test. Tube side heat transfer coefficients were determined using
equation (1) evaluated at the test conditions.
Shell side heat transfer
i
coefficients were assumed to scale as:
i
f
f
                the design value.
13
                Confirmatory UA calculations were performed by initially determining a
h po - Re
f                design heat transfer coefficient for the shell side of the heat exchanger.
Pr
I                This was done by using design value fouling factors, and assuming that the
*
I                tube side heat transfer was correctly predicted by the McAdams equation at
g
j                the design conditions.                                                                  ,
D
i
K
(3)
j
This equation is used frequently in determining shell side heat transfer for
shell and tube heat exchangers. Equation (3) was evaluated at both design
'
'
                        hiDi
and test conditions, and an h for the test was calculated from the design
                            k          =0.23 ReD.8Pr 1/3
g
                                            .                                                                        (1)
h, determined above.
;                The UA for the heat exchanger is
Equation (2) was then used to determine the fouling
!                       UA =                                  A                                                    (2)
factor appropriate for the shell side under as tested conditions assuming
;                              1_  +  pg      1_ + F4 po + F o +D g    In p g
:
                                hg        Di    hi          Di    2K
!
                                                                        ss
the tube side fouling factor is the design value of .0005 (this assumption
                                                                                UI
;
;
                  For the design condition, all values (including UA) are known except hg ,
actually has no impact on the final VA since the two fouling factors are not
[                which was determined to be 918 Btu /h/ F/ft2 using equation (2).
                  The information supplied to Westinghouse by Duke was acquired from experi-
;                mental testing of heat exchanger 1A on 11/22/85.                        The data from this
                  e<periment was used to determine an appropriate value for the degraded UA by
'
j                first determining the as-tested fouling factor.                      In order to do this,
i                experimental flow rates, temperatures, etc. had to be used to determine both
                  tube side (h4 ) and shell side (ho) heat transfer coefficients appropriate
!                for the test. Tube side heat transfer coefficients were determined using
i                equation (1) evaluated at the test conditions.                  Shell side heat transfer
i                coefficients were assumed to scale as:
                                              *
                                                  Pr
                                                    13
f                      hg po - Re D
i                            K                                                                              (3)
j                This equation is used frequently in determining shell side heat transfer for
  '
                  shell and tube heat exchangers. Equation (3) was evaluated at both design
                  and test conditions, and an hg for the test was calculated from the design
                  h, determined above. Equation (2) was then used to determine the fouling
:
                  factor appropriate for the shell side under as tested conditions assuming
!                  the tube side fouling factor is the design value of .0005 (this assumption
;                  actually has no impact on the final VA since the two fouling factors are not
1
1
  '
a function of flow and fluid conditions).
                  a function of flow and fluid conditions).                 The shell side factor was
The shell side factor was
                  determined to be
'
                        F
determined to be
                          g
.00912
                              =    .00912                                                                   (4)
(4)
                                                                                                                                              ,
F
    . - -
=
                Y                                                       -.         - _ -         - - - - , . , ,   -
g
                                                                                                                                  . . . , . ,
,
. - -
Y
-.
-
-
- - - - , . , ,
. . . ,
. ,
-


                    -_     -           . - -                                                         .     -         .     -     ___ - - _                           _   __                     -
-_
    .
-
    .
. - -
        .
.
                                                                                                                                                                                                        l
-
                    Attachment 4
.
                                                                                                                                                                                                        '
-
      .                                                                                                                   3
___ - - _
          ,
_
                                                                                                                                                                                                        l
__
                                                                                                                                                                                                        l
-
                          for the experiment vs. the 0.001 design value.                                                                         In addition to this
.
                        calculations, the experimental error associated with the testing equipment
.
                        and procedure was used to determine an uncertainty value for F . gThis
.
                        calculation was performed using propagation of errors (see for example
l
!                       Beers,1957, " Introduction to the Theory of Error") through the equation
Attachment 4
;                         (5), the energy balance on the NS side of the heat exchanger (only the NS
3
                          flow was used to determine overall heat flow).
'
  '
.
                                                            o
,
                                            Q = mCp (T out                                          -Tin)                                                                   (5)
for the experiment vs. the 0.001 design value.
                          The uncertainty in temperature measurements were given to the NRC team
In addition to this
                          by licensee representatives as                                                                 .4 F including both RTD, and signal
calculations, the experimental error associated with the testing equipment
                          conditioning equipment error. These RTD's were apparently calibrated before
and procedure was used to determine an uncertainty value for F . This
                          testing, which increases confidence in the temperature measurements.
g
                          Additionally, errors in the flow measurements were also included. Handbook
calculation was performed using propagation of errors (see for example
                          uncertainty values for uncalibrated orifice plates are typically 1%-2.5% of
!
                          measured flow. In addition to this, there are uncertainties associated with
Beers,1957, " Introduction to the Theory of Error") through the equation
                          the other instrumentation necessary to make the flow measurements (DP cells,
;
)                         readouts,etc.). The orifice plate was an uncalibrated process device so it
(5), the energy balance on the NS side of the heat exchanger (only the NS
                          was estimated the overall uncertainty was ~5% of the measured value. Each
flow was used to determine overall heat flow).
                          of the uncertainties stated above were treated as one standard deviation
'
.
o
                          (lo) uncertainties.                                                           It is believed that a two standard deviation (2o)
Q = mCp (T
i                         uncertainty bound should be applied in order to insure conservatism (two
-Tin)
                          standard deviations give a 95% certainty of the measurement). The 2a value
(5)
4                        for Q was found to be ~12%. Additionally, since design heat flow was based
out
i                         solely on calculations and not on tests. It was assumed that a 2.5% error
The uncertainty in temperature measurements were given to the NRC team
                          (lo value) was present in the design heat flow determination. It was also
by licensee representatives as
                          assumed that equations (1) and (3) could be used to correctly scale with
.4 F including both RTD, and signal
                          temperature level and flow rate (0 uncertainty was assigned to this
conditioning equipment error. These RTD's were apparently calibrated before
;                       process). The two errors above, experimental and design, were used to
testing, which increases confidence in the temperature measurements.
                          determine overall error in F by                                                         g
Additionally, errors in the flow measurements were also included. Handbook
                                                                                                                    propagating errors through the calcula-
uncertainty values for uncalibrated orifice plates are typically 1%-2.5% of
measured flow. In addition to this, there are uncertainties associated with
the other instrumentation necessary to make the flow measurements (DP cells,
)
readouts,etc.). The orifice plate was an uncalibrated process device so it
was estimated the overall uncertainty was ~5% of the measured value.
Each
of the uncertainties stated above were treated as one standard deviation
.
(lo) uncertainties.
It is believed that a two standard deviation (2o)
i
uncertainty bound should be applied in order to insure conservatism (two
standard deviations give a 95% certainty of the measurement). The 2a value
for Q was found to be ~12%. Additionally, since design heat flow was based
4
i
solely on calculations and not on tests.
It was assumed that a 2.5% error
(lo value) was present in the design heat flow determination. It was also
assumed that equations (1) and (3) could be used to correctly scale with
temperature level and flow rate (0 uncertainty was assigned to this
;
process).
The two errors above, experimental and design, were used to
determine overall error in F by propagating errors through the calcula-
g
'
'
                          tions described above.                                                               The two-standard deviation uncertainty in F was
tions described above.
                                                                                                                                                                            o
The two-standard deviation uncertainty in F was
o
determined to be:
4
4
                          determined to be:
oF
                                            oF g              = .00149                                                                                                     (6)
= .00149
i                         for the uncleaned case of heat exchanger 1-A. An appropriate UA value for
(6)
                          the Westinghouse calculations was then determined by using:
g
                                              FoAPP = F                           g + oF g                                                                                    (7)
i
i                       These values were determined for three cases:                                                                     unit 1-A before cleaning,
for the uncleaned case of heat exchanger 1-A.
                          unit 1-A as it existed after last cleaning, and unit 2-B. The table below
An appropriate UA value for
;                       summarizes these results (in all cases, RN flow was assumed to be 4800 gpm).
the Westinghouse calculations was then determined by using:
FoAPP = F + oF
(7)
g
g
i
These values were determined for three cases:
unit 1-A before cleaning,
unit 1-A as it existed after last cleaning, and unit 2-B.
The table below
;
summarizes these results (in all cases, RN flow was assumed to be 4800 gpm).
J
J
          -_ - -. .         - . - . - - _ _ . _ _ _ _ - - _ - , . , _ _ . - - _ _ - - - . _ - _ - .                                                 - - - _ . _ - , _ -         . - _ , . , _ _ - -
-
- -. .
- . - . - - _ _ . _ _ _ _ - - _ - , . , _ _ . - - _ _ - - - . _ - _ - .
- - - _ . _ - , _ -
. - _ , . , _ _ - -


.
.
  '.
'.
  *
Attachment 4
    ,
4
      Attachment 4                              4
*
                                      Summary of Calculations
,
            UNIT               STATUS                 F
Summary of Calculations
                                                          g        oFo           UA
UNIT
            1-A       uncleaned (11/22/85)           .009       .0015     8.18 X 105
STATUS
            1-A       cleaned (01/16/86)             .0033       .0007     1,63 X 105
F
            2-8       uncleaned (01/24/86)           .011       .0127     7.16 X 10'
oFo
                        Westinghouse input                                   7.35 X 105
UA
            The UA value calculated for the 2-B uncleaned case is slightly below that
g
            given to Westinghouse on 11/28/85 and 01/17/86. However, if the containment
1-A
            pressure calculations performed on 01/17/86 are used as a starting point,
uncleaned (11/22/85)
            and the containment pressure change with VA change is similar to that noted
.009
            in the 3 calculations performed on 11/28/85, the peak containment pressure
.0015
            can be estimated for a UA value of 7.16 X 105 These calculations estimate
8.18 X 105
            that the peak containment pressure for this UA value would be approximately
1-A
            P = 14.56 psig, still below the 15 psig limiting value.
cleaned (01/16/86)
            The calculational methods used to evaluate heat exchanger performance appear
.0033
            to be reasonable. However, when calculations are being performed to deter-
.0007
            mine heat exchanger performance at reduced flow, it is also necessary to
1,63 X 105
            apply appropriate fouling factors to heat exchangers which are suspected of
2-8
            being fouled. This has not been done in previous Duke calculations. As an
uncleaned (01/24/86)
            example, the inspection team looked at the charging pump speed reducer oil
.011
            cooler. Duke has found the oil inlet temperature to increase from 141'F to
.0127
            166 F when RN flow to the heat exchanger is reduced from 20 gpm to 10.7. _In
7.16 X 10'
            addition to the reduced water flow, the effect of fouling should also be
Westinghouse input
            considered. Confirmatory calculations were performed assuming both reduced
7.35 X 105
            flow and a fouling factor of ~.008 on the RN side and .001 on the oil side
The UA value calculated for the 2-B uncleaned case is slightly below that
            (design fouling factors were presented as a sum of Fg +F4 =.0025). The RN
given to Westinghouse on 11/28/85 and 01/17/86. However, if the containment
            fouling factor is an estimate based on findings in the uncleaned containment
pressure calculations performed on 01/17/86 are used as a starting point,
            spray heat exchanger (F, =.009) and recognizing that continuous water flow
and the containment pressure change with VA change is similar to that noted
            through the oil cooler might reduce fouling somewhat. A summary of the
in the 3 calculations performed on 11/28/85, the peak containment pressure
            maximum oil temperatures is presented in the following table.
can be estimated for a UA value of 7.16 X 105
            A calculation with the RN cooling water temperature reduced to 65 F is given
These calculations estimate
            to demonstrate the cooling water temperature effect on heat exchanger
that the peak containment pressure for this UA value would be approximately
            performance.   As can be seen in the below table, the reduction in RN
P = 14.56 psig, still below the 15 psig limiting value.
            temperature from 95 F to 65 F has a significant impact on oil temperature.
The calculational methods used to evaluate heat exchanger performance appear
            A similar effect would be seen in other heat exchangers in the train
to be reasonable. However, when calculations are being performed to deter-
            (although not exactly the same magnitude).
mine heat exchanger performance at reduced flow, it is also necessary to
                                Comparison of 011 Cooler Assumptions
apply appropriate fouling factors to heat exchangers which are suspected of
            Cooling Water Inlet Temp.       Flow (gpm)       F
being fouled. This has not been done in previous Duke calculations. As an
                                                                9
example, the inspection team looked at the charging pump speed reducer oil
                                                                          T oj) ( F)
cooler. Duke has found the oil inlet temperature to increase from 141'F to
                  95 F (Design)                   20         .0015           141
166 F when RN flow to the heat exchanger is reduced from 20 gpm to 10.7. _In
                  '95'F                           10.7       .0015           166
addition to the reduced water flow, the effect of fouling should also be
                  95*F                           10.7       .008           185
considered. Confirmatory calculations were performed assuming both reduced
                  65'F                           10.7       .008           155
flow and a fouling factor of ~.008 on the RN side and .001 on the oil side
(design fouling factors were presented as a sum of F +F4 =.0025). The RN
g
fouling factor is an estimate based on findings in the uncleaned containment
spray heat exchanger (F, =.009) and recognizing that continuous water flow
through the oil cooler might reduce fouling somewhat. A summary of the
maximum oil temperatures is presented in the following table.
A calculation with the RN cooling water temperature reduced to 65 F is given
to demonstrate the cooling water temperature effect on heat exchanger
performance.
As can be seen in the below table, the reduction in RN
temperature from 95 F to 65 F has a significant impact on oil temperature.
A similar effect would be seen in other heat exchangers in the train
(although not exactly the same magnitude).
Comparison of 011 Cooler Assumptions
Cooling Water Inlet Temp.
Flow (gpm)
F
T
( F)
9
oj)
95 F (Design)
20
.0015
141
'95'F
10.7
.0015
166
95*F
10.7
.008
185
65'F
10.7
.008
155
}}
}}

Latest revision as of 02:23, 24 May 2025

Insp Repts 50-369/85-38 & 50-370/85-39 on 851015-17 & 860127-31.Violations Noted:Failure to Adequately Perform Preoperational Test on Control Room Chiller & Implement & Maintain Procedures.Related Info Encl
ML20202B941
Person / Time
Site: McGuire, Mcguire  
Issue date: 06/02/1986
From: Debs B, Wilson B
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20202B938 List:
References
50-369-85-38, 50-370-85-39, IEB-81-03, IEB-81-3, NUDOCS 8607100458
Download: ML20202B941 (45)
v • d • e


See also: IR 05000369/1985038

Text

_ . . .

. . _

l

.

UNITED STATES

[WCEG

O

'o,$

NUCLEAR REGULATORY COMMISSION

[

REGION 11

$ .,

j

101 MARIETTA STREET.N.W.

2

ATLANTA. GEORGI A 30323

\\...../

Report Nos.:

50-369/85-38 and 50-370/85-39

Licensee:

Duke Power Company

-

422 South Church Street

Charlotte, NC 28242

Docket Nos.:

50-369 and 50-370

License Nos.:

NPF-9 and NPF-17

Facility Name: McGuire 1 and 2

Inspection Conducted: Oct er 15-17, 1985 and January 27-31, 1986

,

Inspectors:

[! M

'

B. T. ' Debs

Dite' Signed

F. McCoy

S. D. Stadler

W. Poertner

P. Moore

Accompanying Personnel: Gr y on L. Yoder, Ph.D. (ORNL)

Approved by

hw '. M

b

f?6

m

B.W)ionofReactorSafety

fson, Acting Section Chief

Da'te Signed

Divis

SUMMARY

Scope: This routine, unannounced inspection was in the area of Nuclear Service

Water System Operability.

Results:

Five violations were identified.

i

i

,

8607100458 860727

PDR

ADOCK 05000369

G

PDR

._

-

_

. _ ,

___

_.

._ .

_

REPORT DETAILS

1.

Persons Contacted

-

Licensee Employees

+G. Vaughn, General Manager, Nuclear Stations

  • +T. L. McConnell, McGuire Nuclear Station Manager
  • +R. L. Gill, McGuire Licensing
  • +E. O. McCraw, Compliance Engineer
  • +W. J. Kronenwetter, Design Engineer
  • +W. M. Suslick, Associate Engineer

Other licensee employees contacted included construction craftsmen,

engineers, technicians, operators, mechanics, security force members, and

office personnel.

NRC Resident Inspectors

  • +W. Orders, Senior Resident Inspector

R. Pierson, Resident Inspector

  • Attended exit interview on 10/17/85

+ Attended exit interview on 01/31/86

2.

Exit Interview

,

The inspection scope and findings were summarized on October 17, 1985, and

January 31, 1986, with those persons indicated in paragraph 1 above.

The

inspector described the areas inspected and discussed in detail the inspec-

tion findings.

No dissenting comments were received from the licensee.

The results of the inspection were discussed with utility management during

a meeting in Atlanta on March 14, 1986. The details of this meeting are

documented in Section 11 of this report.

During the exit interview the enforcement findings were presented as

preliminary and unresolved. Following NRC management review, the following

findings were determined:

369/85-38-01, 370/85-39-01 Violation - Failure to adequately perform

preoperational test on control room chiller - see paragraphs 7 and 8.

369/85-38-02, 370/85-39-02 Violation - Failure to implement and maintain

procedures - see paragraphs 7 and 8.

369/85-38-03, 370/85-39-03 Violation - Failure to meet Technical Specifica-

tion 3.7.4 for RN system operability - see paragraph 7.

369/85-38-04, 370/85-39-04 Violation - Failure to perform 10 CFR 50.59

evaluation on degraded equipment - see paragraph 8.

r

2

369/85-38-05, 370/85-39-05 Violation - Failure to identi fy and correct

conditions adverse to quality as * required by 10 CFR 50, Appendix B,

Criterion XVI - see paragraph 12.

369/85-38-06, 370/85-39-06 Unresolved Item - NRC followup of licensee

response of April 25, 1986 - see paragraph 11.

3.

Licensee Action on Previous Enforcement Matters

This subject was not addressed in the inspection.

4.

Unresolved Items

An Unresolved Item is a matter abcut which more information is required to

determine whether it is acceptable or may involve a violation or deviation.

A new unresolved item identified during this inspection is discussed in

Section 11.

5.

Nuclear Service Water System Description

The McGuire Final Safety Analysis Report (FSAR) states that the Nuclear

Service Water (RN) System provides assured cooling water for various

Auxiliary Building and Reactor Building heat exchangers during all phases

of station operations.

Each unit has two redundant " essential headers"

serving two trains of equipment necessary for safe station shutdown, and a

"non-essential header" serving equipment not required for safe shutdown. In

conjunction with the Ultimate Heat Sink, comprised of Lake Norman and the

Standby Nuclear Service Water Pond (SNSWP), the RN System is designed to

meet design flow rates and pressure heads for normal station operation and

also those flow rates and pressure heads required for safe station shutdown

normally or as the result of a postulated Loss of Coolant Accident (LOCA).

The system is further designed to tolerate a single failure following a

LOCA, and/or seismic event causing loss of Lake Norman, and/or loss of

station power plus offsite power (station blackout). Sufficient margin is

provided in the equipment design to accommodate anticipated corrosion and

fouling without degradation of system performance,

j

6.

Summary of NRC Findings

i

On October 4, 1985, the NRC Senior Resident Inspector reported to Region II

management that the 1A nuclear service water system, designated by the

j

licensee as the RN system, had failed to meet the acceptance criteria of

its quarterly inservice test. Although the Technical Specificttion Action

Statement period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> expired on October 7,1985, both units con-

tinued operation at full power based on the licensee's contention that

the 1A RN pump had been made operable by cross connecting it with the Unit 2

2A RN pump. On October 10, 1985, NRC informed Duke Power Company (DPC) that

operation in the unit shared mode was an unacceptable unanalyzed condition.

DPC restored unit separation and began justification for continuing opera-

tion with the apparently degraded pump.

7_

_ . _ _ _ _ _ _ _ _ _ _

_

.

. _

.

..

. . . .

..

..

.

.

3

Licensee representatives stated that they suspected that the pump was not

actually degraded, rather the pump discharge line flow orifice reading

was in error. One of the possible reasons stated was buildup of silt,

mud, or corrosion at the orifice.

Licensee representatives subsequently

stated several months later that the flow indication was erroneous and the

pump was not actually degraded.

The NRC became concerned that if system fouling was that bad at the pump

discharge, what was the status of the downstream components, especially heat

exchangers.

A reactive inspection was conducted October 15-17, 1985, to

review these matters. Numerous phone conferences and letters were exchanged

in ensuing months, and a followup inspection was conducted January 27-31,

1986.

l

A summary of the major NRC findings presented in this report are as

follows:

a.

Preop tests and subsequent surveillance tests performed in 1979 were

not adequate to ascertain operability of RN components.

Several test procedures did not contain acceptance criteria. For

example, a quarterly test of RN heat exchanger 1-A on October 7,

1985, indicated potential fouling but the test procedure contained

no acceptance criteria.

The potential fouling was apparently

pursued only because of questions from the NRC and not addressed

by the licensee until October 14 when it was attributed to a

faulty flow instrument.

The heat exchanger was assumed to be

operable during this period of evaluation.

Flow was not measured through control room air conditioner heat

exchangers.

Test results were recorded in units of differential pressure when

acceptance criteria were in units of flow rate.

Heat transfer characteristics of heat exchangers were not normally

determined.

Fouling factors or empirical tests could have been

used.

RN system was not originally preop tested in the most limiting

post-LOCA configuration in that both trains were not aligned to

simultaneously draw water from the Standby Nuclear Service Water

Pond.

l

b.

The positions of valves specified in preop test data were different

from the positions in operating procedures.

l

'

c.

The RN system had not been flow balanced since 1982 even though engi-

neering documents required it to be.

i

a

.--

-

- -

. . - .

.

. . _

4

2

d.

The following heat exchanger fouling problems had occurred:

Containment spray heat exchanger 1A tested per IEB 81-03 showed

increasing delta P from 20 psid- in 1983 to 29 psid in 1985.

In October 1982, a containment ventilation heat exchanger would

not function due to fouling.

Periodic cleaning of control room air conditioning heat exchangers

had been necessary since 1982 due to fouling.

RCP motor coolers required cleaning three times during the period

1984 - 1985.

Unit 1 component cooling water heat exchanger observed to be

fouled in September 1984.

e.

Inservice testing of the 1A RN pump indicated degraded flow on

i

October 4, 1985. Instead of entering a Technical Specification Action

Statement which would have required the operating unit to be brought

to the hot standby mode within six hours, the licensee inappropriately

cross-connected RN train A and train B and continued to operate.

f.

A flow balance test on RN train IA conducted on December 17, 1985,

revealed flow rates through several safety-related heat exchangers

to be below FSAR values. At the request of the NRC in January 1986,

the licensee evaluated these test results pursuant to 10 CFR 50.59.

This evaluation, which was based upon heat transfer tests by DPC and

calculations by Westinghouse, was completed and justified continued

operation on January 14, 1986. .The licensee apparently assumed the

system to be operable between December 17 and January 14.

I

Although it appears that RN heat exchangers were becoming progressively

'

more fouled with time, the licensee did not recognize the symptoms or place

priority consideration on the overall system operability and associated

safety concerns.

Rather fouled components required for continued operation

were cleaned as needed but no regard shown for the status of dormant safety

equipment, such as the containment spray heat exchangers.

When the concern was raised by the NRC, the licensee devoted significant

resources toward correcting the problem. As a result, during the months

of investigation, there were several instances when individual components

!

were found not to be capable of FSAR specified performance.

On these

occasions, the licensee revised their accident analysis supporting calcula-

tions to justify continued power operation. This mode of operation complies

with regulatory requirements but does not appear to represent to the NRC the

most conservative safety philosophy.

,

4

.

.

.

. .

-

-

- _ .

. .

-

-

-

- .

- - -

-

5

7.

McGuire Nuclear Service Water System History

1979

Preoperational functional testing was completed by the licensee on July 25,

1979, for the Unit 1 RN system and on November 12, 1982, for the Unit 2 RN

system.

In January 1986, NRC Region II inspectors reviewed selected areas

of preoperational test packages for both Units 1 and 2 RN systems.

,

l

It was noted that during the conduct of the Unit 1 preoperational tests

j

of nuclear service water, the safety evaluation section (8) of the major

procedure form was marked as not applicable.

Administrative Plant Manual,

1

Section 4.2.4.1(e) requires that prior to procedure use, a safety evaluation

of major changes to a procedure shall be performed.

Examples of the major

changes made to the preoperational procedures included changes to the

minimum acceptable RN flow criteria, initial RN system configuration at

test initiation, and the methods utilized to determine component flows.

>

The use of "not applicable" for safety evaluations was allowed by a licensee

internal memorandum dated September 14, 1979.

The memorandum deleted the

procedural requirement for a safety evaluation prior to fuel load.

The primary objective of the nuclear service water preoperational functional

,

test was to verify that the system could supply designed cooling water flow

j

to variou., components and to set each component throttle valve to provide

the proper flow rate. Adequate system and component flow was to be verified

j

for all modes of operation.

i

i

One of the safety related RN loads during post-LOCA conditions is the

l

control room air conditioner which requires a minimum flow of 789 GPM as

j

stated in McGuire FSAR Table 9.2.2-1(8).

During the RN preoperational

test for Unit 1, the flow to the control room air conditioning was unable

,

I

to be determined due to problems encountered with the installed instrumen-

l

tation. Subsequently, a major change to the preoperational test procedure,

!

TP/1/A/1400/01, was approved by the licensee to delete the requirement to

verify the minimum RN flow of 789 GPM.

The change to the preoperational

test was justified by the licensee on the basis that the flow control valve

is air operated and fails open during accident conditions.

This justifi-

cation assumed that there were no internal obstructions and that the wide

open valve flow would meet or exceed the FSAR~ required flow. Due to this

procedure revision, the subsequent RN preoperational test for Unit 2

also did not verify adequate flow to the control room air conditioning.

As stated later in this report, subsequent functional flow test data

obtained in late 1985 and early 1986 indicated that the required 789 GPM

was not being met. Failure to test the aforementioned component represents

,

1

a violation of 10 CFR 50, Appendix B, Criteria XI which requires a test

program to be established to assure that all testing required to demon-

strate system components perform satisfactorily in service (369/85-38-01,

a

370/85-39-01).

'

--

-

- - - -

. -

- - -

- -

-. .

.

_

-

_--

,

6

,

The inspector noted that in several instances during the conduct of the

preoperational tests of the RN system, the measured flows were stated

as differential pressure (psid) rather than flow (GPM). The engineers who

performed the tests and the preoperational logs indicated this was due to

[

problems experienced with the instaihd flow instrumentation. To continue

!

the tests with the inoperable flow instrumentation, the licensee utilized

i

temporary differential pressure instrumentation.

The conversion from

j

differential pressure to GPM was not made on test data enclosures.

To

verify that the minimum FSAR flow results were achieved for the RN compo-

nents preoperationally tested, the inspector, in early 1986, requested that

the licensee convert the differential pressures to flows. In each case it

I

was verified, based on the licensee's calculations, that the minimum accept-

able flow rates had been achieved as stated in McGuire FSAR table 9.2.2-1.

The values from that table appear later in this report.

,

I

To assure minimum RN component flows, including adequate flow to the con-

tainment spray heat exchangers during design LOCA conditions, the normally

'

throttled valves associated with each RN component were required to be set

during p'reoperational testing of the RN system.

These throttled positions

,

established during preoperational testing were to be incorporated into

l

operating and surveillance procedures to protect these throttled settings

!

'

during future operations. The inspectors noted that, in some cases and

particularly for Unit 1,

the throttled valve positions listed in the

licensee's RN operating procedures and their locked valve verification

'

procedures were not consistent with earlier preoperational "as left" data.

It was noted, however, for those throttled valves reviewed, the operational

positions were further open than the "as left" preoperational test posi-

tions.

The licensee acknowledged these discrepancies and committed to

revise the operational procedures to meet those valve settings established

during recent 1985 and 1986 RN flow testing.

The inspection team noted that since 1976 the licensee has had a functional

r

system description for the RN system. Section 5 of this system description

(MCSD-0138.00) states that annually each essential RN train must be checked

!

for proper throttling. Also, after any throttle valve is repositioned, the

entire train must be checked for proper throttling. The system description.

A

then presents a detailed procedure to verify that the minimum flow condi-

tions for operability of the safety related portion of the system are met.

The licensee had decided not to adopt the aforementioned recommendations.

,

l

4

Consequently, no RN flow balances had been performed - beyond 1982 until

'

requested by the NRC in late 1985.

Functional system descriptions are

not used as procedures by licensees and, consequently, failure to follow

2

'

MCSD-0138.00 is not considered to be a violation.

However, compliance

with this document would have prevented the above violation.

However, The

requirements to verify proper throttling position should have been in plant

procedures.

Failure to measure flow through components and failure to specify post-

tions of throttled valves in procedures represent examples of inadequate

-

.

_

_

.

_.

.-

--.

,-

. - ,

7

procedural controls and are, therefore, a violation of McGuire Technical Specification 6.8.1 and 10 CFR 50, Appendix B, Criteria V which requires

that adequate written procedures be implemented and maintained (369/85-

38-02,370/85-39-02).

In addition to adding procedural requirements for RN throttled valve posi-

tions as addressed above, the licensee has implemented several other posi-

tive methods to control these valves. Currently, these valves are verified

locked every six months under the Locked Valve Verification Procedure

4700/23.

In addition, independent verification is utilized to ensure that

the valves are returned to the proper position following valve repositioning

for maintenance or other activities.

Despite these positive controls, the

inspectors noted the following recent deficiencies in the licensee's control

over these throttle valves:

-

The Locked Valve Verification Procedure requires that the operator

verify the valve to be locked. No verification of the actual throttle

position is required.

-

The valve locks utilized for RN throttled valves are chain locks.

These chain locks work well for wide open valves, but the slack in the

chains cannot ensure that a valve remains open 1/4 turn. A valve that

is required to be open 1/4 to 3/8 turn could be locked in the full-

closed position without detection.

One potential solution identified by the licensee for better control of

these throttle valves include the use of locking collars which are used on

throttle valves in other systems.

Since the locking collars can be sized

to ensure the exact valve opening desired, their use would provide positive

indication of valve position.

The licensee initiated a 10 CFR 50.72 notification to the NRC stating that

prior to January 27, 1986, the RN systems for Unit 1 and 2 had never been

tested under the requisite design basis accident configuration.

Specifi-

cally, the system valves had never been positioned to supply the required

flow to essential headers for Units 1 and 2 with the system taking suction

solely from the Nuclear Service Water Pond.

This issue is discussed in

Section 6. of this report.

1981

In response to IE Bulletin 81-03 which addressed the potential fouling of

safety related heat exchangers by clam and shell debris, the licensee com-

mitted to the NRC to monitor two RN supplied heat exchangers on a quarterly

basis. One of these heat exchangers is the 1A Containment Spray (NS) heat

exchanger.

Additionally in the licensee's response, it was stated that

"if significant fouling is detected on these heat exchangers, other heat

exchangers in the RN system will be inspected." The licensee performed

their monitoring under procedure PT/1/A/4403/04. This procedure for the 1A

NS heat exchanger requires that the test be performed for a FSAR accident

. _ .

- _ -

=

_ -.-

-_

!

!

8

4

i

RN flow of 5000 GPM to the NS heat exchanger and that the heat exchanger

'

differential pressure (D/P) be recorded.

In October 1985, the inspectors

,

reviewed the past test data which indicated the following:

DATE OF TEST

D/P (PSID)

l

6/20/83

20

!

9/22/83

Not Available

10/2/83

23.5

!

1/18/84

25

4/11/84

23

7/18/84

25

11/9/84

29.5

'

2/28/85

23.5

6/27/85

25

  • 10/7/85

29

  • RN flow was 4600 GPM

The test procedure did not specify criteria for determining "significant

fouling" and, tnus, other components were not inspected as a result of

these tests.

Further discussion of these findings appears later in this

'

report under the section titled 1985.

l

1982

On October 22, 1982, the licensee identified that fouling of the RN supplied

,

lower containment ventilation heat exchangers was a problem which was

causing unacceptable temperature increases in the lower containment areas.

This subsequently forced the units to operate at reduced reactor power

during certain seasonal conditions.

In April 1983, the licensee attempted

to add a penetrant / dispersant to the RN system in an attempt to clean lower

containment cooling units.

The attempt was ineffectual.

Eventually the

,

licensee modified the coolers with a self-cleaning mechanism which corrected

the problem.

As a result of a control room air conditioning trip due to fouling of the RN

supplied, safety related air conditioning chillers, the licensee established

a cleaning threshold based on increasing air conditioning condenser pres-

i

sures. On the following dates, these chillers have been rodded out to

maintain operability.

TRAIN A

TRAIN B

11/19/82

3/83

!

10/03/83

01/07/85

12/19/83

10/21/85

05/30/84

11/05/85

i

10/31/84

l

09/25/85

l

10/24/85

i

10/31/85

,

, - . - -

-

-

-- ----.

.,

, -

- - . - - ,

,,

, - . . - -

.

9

1984

In March 1984, the licensee began development of a heat exchanger perform-

ance monitoring program. At the time of this inspection, Duke Power Company

had not fully implemented this program at their nuclear plant.

The inspector reviewed the section of the program which requires monitoring

the performance of heat exchangers such as those in the RN system.

The

program appeared to be very comprehensive with provisions for monitoring

both flows and heat transfer capabilities, for increasing the frequency of

monitoring as warranted, and for initiating corrective actions as necessary.

Once fully implemented, this Performance Monitoring Program will be a major

improvement in the licensee's ability to monitor plant equipment perfor-

mance and to promptly identify degraded performance. A key to the relative

success of the program, however, will be the effectiveness and timeliness

+

of corrective actions taken in response to an identified deficiency.

The

inspector noted that this corporate monitoring program was scheduled to be

implemented in stages at the various plants. The RN heat exchangers were

scheduled for performance monitoring implementation during the second phase

of the program which will be several months into 1986.

As a result of

the fouling and degraded performance being experienced with the RN heat

exchangers and concerns expressed by the NRC, the licensee indicated this

phase of the program will be implemented on a priority basis.

Also in 1984, the licensee began to experience RN fouling problems in

their reactor coolant pump motor coolers.

The licensee has performed the

following cleanings of these coolers on the dates indicated:

'

UNIT 1

UNIT 2

12/31/84

8/10/84

11/08/85

In September 1984, the licensee evaluated the Unit 1 Component Cooling (KC)

heat exchanger for fouling, although, according to the licensee, there was

no indication of reduced heat transfer or high differential pressure. As

part of the evaluation, DPC engineering calculated a fouling factor for the

KC heat exchangers. These calculations were based on informal test data

which appeared to the cognizant engineer as nonrepresentative.

In November

1984 the Unit 1 KC heat exchanger was cleaned. In June and July 1985, the

Unit 2 KC heat exchanger was cleaned. Although visual inspection of the

,

heat exchanger by DPC engineering did not support the calculated fouling

factor (the calculated fouling factor appeared to be less conservative),

the licensee did not perform further evaluation of past operability of these

heat exchangers.

t

10

1985

On October 4,

1985, following in-service testing, the 1A Nuclear Service

Water pump performance was found to be degraded. The pump curve generated

from the test data deviated from the previously established base-line

curve. Delivered flow was estimated to be approximately 85 percent of that

required. Technical Specification (TS) 3.7.4 requires two loops of RN to

be operable. With only one loop operable, they must restore both loops to

operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and

cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The licensee performed a 10 CFR 50.59 analysis to justify cross connecting

the 1A and the 2A RN trains in an attempt to boost IA RN flow.

After

reviewing the 50.59 analysis and extensive interaction with the licensee,

the NRC Region II, on October 10, 1985, informed the licensee that the NRC

considered the licensee was not meeting the requirement of TS 3.7.4 which

requires two operable RN loops since the 1A train was inoperable due to

a degraded pump and that the cross connected configuration could not be

justified by a 50.59 analysis since it represented the possibility of an

unreviewed safety question and, in effect, changed the Technical Specifica-

tion.

The licensee's action to cross connect the 1A and 2A RN trains and to con-

tinue two unit operation for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> was contrary to TS 3.7.4

and, therefore, represents a violation (369/85-38-03, 370/85-39-03).

During this time period, the licensee discovered that one of the cross

connect valves had an erroneous position indication.

Thus, the valve

was actually closed when thought to be open.

This matter was discussed

previously in Region II Inspection Report 50-369/85-35, 50-370/85-36.

As a result of the interactions with the NRC, the licensee split the RN

trains and took compensatory measures to continue operation of the 1A

train under reduced flow conditions.

Further details of the apparent

degradation of the 1A RN pump are contained in NRC Region II Inspection

Report 50-369/85-37. As a result of the aforementioned event, during the

period of October 15-17, 1985, Region II inspectors reviewed the overall

'

RN system performance in light of the recent event.

The inspectors reviewed the licensee's quarterly performance, PT/1/A/

4403/04, data on the 1A NS heat exchanger which was tabulated earlier in

this report under 1981.

The following observations were made by the inspectors regarding PT/1/A/

4403/04:

The performance test lacked qualitative and quantitative acceptance

criteria.

The test results suggest an increasing D/P across the 1A NS heat

exchanger.

i

_

.

-

-_

_.

-_.

11

The pressure drop could not be measured at the required design basis

accident RN flow of 5000 GPM for the 1A NS heat exchanger because this

flow could not be achieved for the test performed on October 7,1985.

The measured flow was recorded as 4600 gpm.

The 1A NS heat exchanger outlet throttle valve was closed to the

extent that the as found flow through this heat exhanger was 800 gpm.

It appears doubtful that the required accident flow of 5000 gpm could

have been achieved with this as found valve position.

The licensee indicated that, at that time, a qualitative or quantitative

acceptance criteria had not been determined but that work had begun to

provide such criteria.

10 CFR 50 Appendix B Criteria V states that

procedures shall include appropriate quantitative or qualitative acceptance

criteria for determining that important activities have been satisfactorily

accomplished.

Contrary to this regulation, PT/1/A/4403/04 did not contain

an appropriate acceptance criteria.

This represents another example of

violation (319/85-38-02, 50-370/85-39-02).

Regarding the aforementioned increasing D/P across the 1A NS heat exchanger,

the licensee indicated that, although the test results suggest an increasing

D/P, some mathematical analysis should be performed to prove the apparent

trend of an increasing D/P.

Regarding the low 1A NS heat exchanger RN flow recorded on October 7, 1985,

the licensee indicated that the low reading could have been a result of a

calibration problem.

As a result of the inspector's questioning, the

licensee issued Work Request Number 65574 to check the calibration of the

flow instrument used to obtain the recorded 4600 GPM. On October 14, 1986,

the calibration results indicated that, at a flow of 5000 GPM, the instru-

ment indicated 4820 GPM. The licensee then took action to recalibrate the

instrument.

Based on the data reviewed and discussions with licensee personnel, the

inspector stated the following concerns:

Since the licensee did not have an acceptance criteria for the in-

creased D/P, could the apparently increasing D/P suggest heat exchanger

fouling which may have reduced heat exchange capacity to an unaccept-

able level? Could system flow reductions due to fouling affect other

RN system component performance? These concerns were discussed with

plant management on October 17, 1985. The inspector requested manage-

ment to consider the feasibility of performing a RN system integrated

flow test to provide confidence that all RN safety related loads could

be provided the requisite design basis flows.

Additionally, the

inspector discussed the feasibility of measuring the heat transfer

capability of the 1A NS heat exchanger.

_

- _ _ _ .

12

After growing concern by NRC Region II regarding the current ability of

the Unit 1 RN system to perform its safety function under accident condi-

tions, the licensee was requested, on October 18, 1985, to provide the NRC

Region II Office with a statement of operability for the RN system.

On

October 23, 1985, the operability statement was received from tne licensee.

This :tatement concluded that the RN system is operable and capable of

performing its intended safety function.

The statement of operability included an engineering evaluation by Duke

Power Company.

The evaluation summarized the results of a Westinghouse

computer calculation which utilizes the LOTIC code.

This code predicts

containment pressure response from inputs including the heat transfer

capability (UA) of the containment spray and component cooling water heat

exchangers.

The Duke Power engineering calculations used to determine

the UA for the 1A NS heat exchanger assumed the same fouling factor which

was calculated for the Component Cooling Water (KC) heat exchanger in

early 1985.

The inspectors expressed reservation over this assumption;

questioning the credibility of applying the existing fouling factor for a

single pass horizontal type heat exchanger (KC heat exchanger) to the NS

heat exchanger which is a vertical U-tube heat exchanger. Additionally,

RN flows through the tubes of the KC heat exchanger unlike the NS heat

exchanger where RN flow is on the shell side.

It, however, was agreed by

the NRC that for lack of any other available data this approach was accept-

able until specific empirical data could be obtained.

Based on the aforementioned assumptions and calculation as utilized in

the LOTIC program (WCAP-8282), a maximum containment pressure of 13.3 psig

was predicted during a design basis accident.

The McGuire containment

design pressure is 15.0 psig.

In response to NRC concerns over the potential fouling and degradation

of the 1A NS heat exchanger, the licensee developed a performance test

PT/0/A/4208/01, Containment Spray Heat Exchanger Performance Test.

The

purpose of the test was to:

Determine if a high flow flush reduces the heat exchanger differential

pressure.

Assure the structural integrity of the heat exchanger tubes.

Determine the overall heat transfer coefficient and fouling factor of

the NS heat exchangers.

The McGuire FSAR analysis utilized a containment spray heat exchanger UA

of 2.J4 x 10' BTU-Hr-Deg. F.

Empirical data from the aforementioned test

indicated that an actual UA of 7.35 x 105 BTU-Hr-Deg. F existed under

current plant conditions.

This information was provided to Westinghouse

,

on November 27, 1985, to perform a LOTIC run utilizing this data. A con-

tainment response model which is less conservative than the one used in

the FSAR analysis was used by Westinghouse (WCAP-10325) for this run. Use

, _ .

-

,

13

of this model was accepted by the NRC since this WCAP had been reviewed

and found technically sound by the NRC staff, although the NRC's Safety

Evaluation Report had not been issued at that time.

This LOTIC run of

November 27, 1985, indicated that, for the aforementioned UA, a peak

containment pressure of 14.42 psig would be realized under design basis

accident conditions.

In addition to the heat transfer test, the licensee performed a heat

exchanger tube integrity test using the tritium activity of the Refueling

Water Storage tank (RWST) as a tracer source.

The test results indicated

insignificant leakage.

Several cleaning attempts using various chemical and hydraulic techniques

were employed by the licensee to clean the 1A NS heat exchanger. The latest

performance test results (January 28, 1986) indicate that a UA of 2.03 x

10' BTV/Hr-Deg. F had been achieved.

The inspectors viewed video tapes of the licensee's fiber optic inspection

of the 1A NS heat exchanger. Approximately the first seven feet of the

upper portion of -the tube bundle could be viewed. The tape indicated that

a fairly uniform silica deposit completely covered the tubes, prior to

cleaning.

Confirmatory UA calculations were performed by the inspection team. These

calculations appear as Attachment 4 to this report.

Those calculations

closely approximate those of the licensee.

'

8.

Review Of Flow Balance Testing

The inspectors conducted a review of the RN flow balance testing conducted

on December 17, 1985, January 27, 1986, and January 28, 1986, for Train IA

of the Nuclear Service Water System.

Additionally, flow balance testing

conducted on January 30, 1986, for Train 1B of the Nuclear Service Water

System was reviewed.

The Train 1A flow balance test conducted on December 17, 1985, was in

accordance with procedure TT/1/A/9100/105, Change 0 through Change 1.

The

test provided for:

-

Isolation of Train IA and 1B essential header.

-

The low level intake providing Train 1A suction.

-

Isolation of the Unit I non-essential header from Train 1A.

-

Control Room and Equipment Room A Train Cooling Chillers being supplied

by Nuclear Service Water Train 1A.

1

,

'

.

_

_

_ - , .

.

14

-

Securing of Nuclear Service Water Train 2A due to a condition of

Nuclear Service Water operability resultant from prior degraded pump

performance in Train IA when supplying Control and Equipment room

cooling.

-

Alignment of service water valves in accordance with a lineup that

was consistent with actual Safety Injection and Containment Spray

conditions.

The Train 1A flow balance test conducted on January 27, 1986, was per-

formed in this same manner with the exception that Change 2 of _ procedure

TT/1/A/9100/105 was also in effect which changed the Train IA suction from

the low level intake to the service water pond in order to duplicate the

most restrictive condition of operation for testing.

Flow rates through those essential heat exchangers required to mitigate

accident consequences during Safety Injection and Containment Spray were

measured during these tests and compared to target values which were

specified in the FSAR. Measurement results and comparisons for Train 1A

tests are delineated in Table 1.

The data for the December 17, 1985 test reflects that FSAR specified flow

rate values could not be attained for the containment spray heat exchanger

(4% degraded), control room chiller heat exchanger (10*s degraded), the

charging pump oil cooler (46% degraded), spent fuel pool pump room ai*

handling unit (27% degraded), and containment spray pump room air handlirg

unit (56% degraded).

Although the data from the December 17th test indicated multicomponent

degradation, the licensee performed an informal evaluation to support

continued operation. The results of this evaluation were not documented.

Not until requested by the NRC in January 1986, did the licensee perform

a detailed engineering evaluation as required by 10 CFR 50.59.

Failure

to perform this requisite evaluation is considered a violation of the

aforementioned 10 CFR 50.59 (369/85-38-04, 370/85-39-04).

In an operability statement dated January 14, 1986, the licensee performed

an engineering evaluation to demonstrate the adequacy of the tested perform-

ance of the charging pump oil cooler, the containment spray pump room air

handling unit and the spent fuel pool cooling pump room air handling unit

with the observed reduced flow rates.

In the operability statement the

licensee stated that the degraded containment spray heat exchanger flow

was adequate, and justified continued operation of Unit 1.

This operability

statement was based on the actual tested values of the thermal efficiency

for this particular heat exchanger and a containment pressure calculation

performed by Westinghouse and forwarded to Duke Power Company by letter

DAP-86-513 dated January 16, 1986. The Westinghouse calculation was based

on assumptions which included the following:

An active sump volume of 90,000 cubic feet.

15

A thermal efficiency heat transfer coefficient of VA=7.35 x 105

BTU-Hr-Deg. F for the containment spray heat exchanger and UA=1.64 x

10' BTU-HT-Hr-Deg. F for the RHR heat exchanger.

The licensee stated

that, for the containment spray heat exchanger, this represented a 75%

reduction in the UA coefficient. This value was a conservative selec-

tion by the licensee since the testing performed on December 17, 1985,

demonstrated the UA value to be nearly 58% degraded.

Under these assumptions the Westinghouse calculation demonstrated that

during a LOCA, containment pressure would remain below the containment

design pressure of 15 psig with service water flow through the containment

spray heat exchanger reduced to 4800 gpm.

The licensee, therefore, con-

sidered that the results of their evaluations and calculations justified

continued operation of Unit 1.

The basis for this conclusion was reviewed

and accepted by the NRC

Between December 17, 1985 and January 27, 1986, three cleaning cycles were

accomplished on the RN side of the 1A containment spray heat exchanger. The

licensee concluded that heat exchanger thermal ef ficiency increased from

42.1% to 74.7% as a result of these cleaning cycles. The affects of these

cleaning cycles is also demonstrated in the reduced RN header pressure

delineated in Table 1, for the flow balance test of January 27, 1986.

The data in Table 1 for the January 27, 1986 test reflects that, even af ter

the cleaning evolutions, FSAR specified flow rate values could again not be

attainec for the containment spray heat exchanger (2% degraded), control

room chiller heat exchanger (0.5% degraded), Spent Fuel Pool Pump Room Air

Handling Unit (30% degraded), containment spray pump room air handling

unit (56% degraded), diesel generator cooling water heat exchanger (8%

degraded), and safety injection pump motor air handling unit (15% degraded).

Degradation of the charging pump cooling flcws was attributed to faulty

flow indication which required instrument replacement.

The licensee stated that as a result of this test,' Train 1A of nuclear

service water v's declared inoperable pending resolution of the degraded

flow conditions and correction of the faulty flow indicator associated with

the charging pump oil cooler.

The inspectors noted that these flow balance tests were accomplished with

Unit 2 Train A secured which was not conservative with respect to the

design basis accident. Worst case conditions should assume Unit 2 Train A

providing unit coaldown loads during the operation of Unit 1 Train A to

mitigate accident conditions. This in effect would reduce the net positive

suction head for Unit 1 Train A.

The inspectors considered that testing

should reflect this condition.

The licensee stated that on January 28,

1986, another flow balarce would be performed and that Train 2A would

,

service necessary cooldown loads for Unit 2 during this test.

!

P

16

In conjunction with resolution of the degraded flow conditions reflected in

the service water Train 1A flow balance testing, the licensee had requested

that Westinghouse perform an analysis to determine new acceptable minimum

values of service water flow through containment spray and component cooling

water heat exchangers. The licensee was considering that a throttling back

of these two major heat exchangers would result in a higher RN header

pressure thus providing increased flow thru the smaller essential heat

exchangers. A Westinghouse calculation was forwarded to Duke Power Company

in January 1986 which demonstrated that, with service water flow through the

component cooling water heat exchanger reduced to 6000 gpm and service water

flow through the containment spray heat exchanger reduced to 3800 gpm, peak

containment pressure would remain below the containment design value of

15 psig during a LOCA.

On January 28, 1986, a third nuclear service water flow balance test was

accomplished on train IA. This test provided for reduced target flow values

of 6000 GPM through the component cooling water heat exchanger and 3800 gpm

through the containment spray heat exchanger which the licensae considered

to be acceptable target values based on the aforementioned Westinghouse

calculation. This flow balance test was performed under the same conditions

as the January 27, 1986 test with the exception that Train 2A was aligned

to provide a cooldown load of greater than or equal to 6000 gpm for Unit 2,

the flow instrument for the charging pump oil cooler had been replaced, and

the RN system took suction only from the SNSWP. The results of this test

are delineated in Table 2.

The result of this test demonstrated that flow

values through all heat exchangers were within the new acceptable values

established by the licensee within the operability statement of January 14,

1986. On March 11, 1986, the licensee made a 10 CFR 50.72 notification

to the NRC stating that, prior to January 27, 1986, the RN system for

both units had never been tested under the requisite accident conditions

with all RN being supplied by the SNSWP.

Apparently after addressing

both NRC and DPC engineering concerns regarding the desired RN flow test

system configuration, the licensee later realized that the preoperational

test configuration had not tested the system under the design basis

accident configuration.

The aforementioned event represents another

example of a violation of 10 CFR 50, Appendix B, Criterion XI (369/85-38-01,

370/85-39-01).

The NRC later learned from the licensee that during the establishment of the

flow test system configuration on January 28, 1986, the RN system entered

a pressure transient. While base loading the RN pumps (gradually placing

requisite heat exchangers on the line), a significant decrease in RN header

pressure was experienced. This event was not allowed to go full term and

was terminated by throttling down on large component flows. The test was

repeated with the throttled valve positions and acceptable results were

obtained. On March 12, 1986, NRC Region II learned of the January 28 flow

transient shortly after DPC management had been informed of it.

The NRC

expressed concern regarding the transient since it suggests that, under

actual accident conditions, the RN system's pumps could have lost net

positive suction head resulting in a loss of the ultimate heat sink for

both units. This concern is further addressed in Section 10 of this report.

-

-

~

. . _

17

The Nuclear Service Water Train IB flow balance test was conducted on

January 30, 1986, with the same test methodology utilized for the

,

January 28, 1986 flow balance test for Train 1A. The results of this test

are delineated in Table 3.

The results of this test demonstrated that

established operability values could not be obtained for the Spent Fuel Pool

Pump Room Air Handling Unit (5% degraded) and the Residual Heat Removal Pump

Room Air Handling Unit (1% degraded). The licensee was advised by the NRC

,

that prior to establishing Train IB as being fully operable, these degraded

conditions would require further evaluation and resolution.

4

o

1

i

i

I

i

i

-

-_

-

.,

--

_.

, . ,

. . - . . _

,

,

--

. . . .

. _ - _ .

_

_ . _ _ . _ . . .

_ _ . _ . _

_ _ . _ _ _ _ _ _ . - .

_m

= = . _ _ . _

_ . _ . . . _ _ , . _ . .

._. _ .

. __

-. _ . _ _ , _ _ , _ .

_

_

!

,

t

TABLE 1

Results of Heat Exchanger Flows and Comparison to FSAR Target Values During Nuclear Service Water

Train 1A Flow Balance Testing of December 17, 1985 and Janua ry 27, 1986.

December 17, 1985 Data

Janua ry 27, 1986

Ta rget Flow

flow Rate

Header

Flow Rate

Header Pressure

Heat Exchanger

Rate (CPM)

(CPM)

Pressure (psig)

(CPM)

(psig)

1.

Component Cooling Water

8000

0000

67.5

8000

56

'

2.

Conta inment Spray

5000

4800

67.5

4887

56

l

3.

Diesel Generator Cooling

900

900

67.5

830

56

Water

i

4.

Control Room Chiller

789

707

67.5

785

56

5.

Cha rg i ng Pump Oi l Coo l e r

28

15

67.5

3

56

i

6.

Safety injection Pump

20

21

67.5

17

56

Oil Cooler

c

7.

Spent fuel Pool Pump

20

14.7

67.5

14

56

'

Air Handling Unit

8.

Conta inment Spray Pump

45

20

67.5

20

56

Air Handling Unit

9.

Residual Heat Removal

45

51

67.5

52

56

Pump Air Handling Unit

,

i

1

$

1

1

-

,,

- , - -.

,

,

-

-_ - _ -

-

.,

. -

~ - ..

- -- - . - - - . .

. . ~ . . . - . . . . - . ~ . .

. ~ - . - . . _ .

- .

. - . . _ . . - - - - - . - - _ _ . - - _ - - - -

. - - - -

l

TABLE 2

Results of Heat Exchanger Flows and Comparison to FSAR Target Values and Licensee Established Operability

Values During Nuclear Service Water Train 1A Flow Balance Testing of January 28, 1986.

January 28, 1986 DATA

E

Licensee Established

Target flow

Ope ra b i l i ty Va l ue

Flow Rate

Header Pressure

Heat Exchanger

From FSAR (CPM)

' for Flow (GPM)

(GPM)

(psig)

,

1 Component Cooling Water

8000

6000+

6000

62.5

2.

Containment Spray

5000

3800+

3970

62.5

3.

Diesel Generator Cooling

900

900

950

62.5

Water

4.

Control Room Chiller

789

789

946

62.5

5.

Cha rg i ng Pump Oi l Coo le r .

28

15

22

62.5

6.

Safety injection Pump

20

20

23

62.5

Oil Cooler

,

7.

Spent fuel Pool Pump

20

14.7

19

62.5

Air Handling Unit

8.

Conta inment Spray Pump

45

20.0

23.5

62.5

Air Handling Unit

9.

Residual Heat Remova l

45

45.0

64.5

62.5

Pump Air Handling Unit

+ Based on assumption that Containment Spray Heat Exchanger Thermal Efficiency is greater than or equal to 70%.

The rma l performance data reflects it is currently 74.7% and past history indicates degradation will inc rea se

due to fouling.

. - _ . - , - .-

_ .. ._,____,_ _._ _._ _.

.

,,_

,_

- - - _ _ _ - - .

- _ _ . _ -

. .

- - . . _ - . _ _ - . ..

. - -

_ . _ -

_ . - - . - - - ~ . - _

_

. - . . _ . - _ . . . .

.

-- -..

.

I

TABLE 3

Results of Heat Exchangor Flows and Comparison to FSAR Target Values and Licensee Established Operability

Values During Nuclear Service Water Train IB Flow Balance Testing of January 30, 1986.

January 28, 1986 DATA

I

Licensee Established

j

-

Ta rge t flow

Ope ra b i l i ty Va l ue

Flow Rate

Header Pressure

j

Heat Exchanger

From FSAR (GPM)

for Flow (GPM)

(GPM)

(psig)

I

1 Component Cooling Water

8000

6000+

6900

52

2.

Conta inment Spray

5000

3800+

5000

52

3.

Diesel Generator Cooling

900

900

920

52

Water

j

4.

Control Room Chiller

789

789

912

52

i

5.

Cha rg ing Pump Oi l Coo le r

28

15

20

52

6.

Safety injection Pump

20

20

28.2

52

,

-

Oil Cooler

7.

Spent ruel Pool Pump

20

14.7

14

52

Air Handling Unit

f

8.

Containment Spray Pump

45

20.0

46

52

,

'

Air Handling Unit

9.

Residual Heat Removal

45

45.0

44.5

52

Pump Ai r Handling Unit

+ Based on assumption that Containment Spray Heat Exchanger Thermal Erriciency is greater than or equal to 70%.

Thermal performance data reflects it is currently 74.7% and past history indicates dearadation will inc rea se

due to rouling.

I

l

,

. - - .

7

18

Following performance of the nuclear service water train IA flow balance

test of January 28, 1986, the inspector observed a train IA Diesel Generator

operability test.

During performance of this test, the inspectors noted

that the flow indicator for service water flow through the diesel generator

cooling water heat exchanger was off scale high (greater than 1000 gallons

per minute) rather than indicating an expected value of 900 gallons per

minute.

Interviews with licensee personnel who had performed the earlier

Train 1A flow balance test reflected that, during test restoration, valve

IRN73A was left in the test position rather than being returned to the

normal position.

The test position for this valve is " throttled to 900

gallons per minute in the test lineup configuration". The normal position

for this valve is " throttled to 900 gallons per minute in the normal lineup

configuration." Since the normal RN system lineup configuration isolates

the large engineered safety feature loads, RN header pressure was increased

which resulted in greater flow through those valves which were not throttled

back from the test position. Failure to restore valve IRN73A to its normal

position is contrary to step 12.8 of procedure TT/1/A/9100/105 and is a

third example of a violation for failure to properly implement procedures

(369/85-38-02,370/85-39-02).

The inspectors noted that restoration of the 1A train service water diesel

generator heat exchanger outlet isolation valve (IRN73A) and the 1A train

service water containment spray heat exchanger outlet isolation valve

(IRN137A) to their normal throttled positions could result in insufficient

nuclear service water flow being supplied to the diesel generator heat

exchanger and containment spray heat exchanger when the containment spray

heat exchanger is placed on line during transfer to cold leg recirculation

unless specific operator actions were taken to ensure proper flow through

these heat exchangers. A review of the emergency operating procedures for

safety injection (EP/1/A/5000/01, EP/2/A/5000/01) and for transfer to cold

leg recirculation (EP/1/A/5000/2.3, EP/2/A/5000/2.3) reflected that provi-

sions were not established to assure proper service flow through the diesel

generator cooling water heat exchanger and containment spray heat exchanger

when these component were required during accident conditions.

These

inadequacies in the emergency operating procedures are considered a fourth

example of violation 369/85-38-02, 370/85-39-02, failure to properly

establish and implement procedures.

During the course of this inspection, test procedure TT/1/A/9100/105, RN

Train 1A Flow Verification, was revised, and test procedure TT/1/A/9100/107,

RN Train IB Flow Verification, was written to leave the service water outlet

isolation valve to the containment spray heat exchangers in the tested

throttle position.

Additionally, licensee actions were initiated to

revise emergency operating procedures EP/1/A/5000/01,

EP/2/A/5000/01,

EP/1/A/5000/2.3, EP/2/A/5000/2.3 in order to establish adequate service

water flow through the diesel generator heat exchanger and containment

spray heat exchanger, during safety injection and transfer to cold leg

recirculation.

-

.

_ _

!

19

.

9.

Changes to the McGuire Containment Pressure Response Model

During the course of the licensee's engineering evaluations to justify

the apparent RN system degradation, many changes were made to the input

parameters used in the McGuire containment pressure response model.

The following parameters have significant effect on peak containment

pressure:

ice mass

NS and KC heat exchanger UAs

NS and KC heat exchanger tube and shell flows

mass and energy releases into containment

auxiliary containment spray flow

auxiliary containment spray actuation time

active containment sump volume

Table 4 provides a chronology of these parameters and when each parameter

was changed by Duke. Some values such as active containment sump are based

on engineering judgement by Duke since calculations have not been completed

to justify the value.

TABLE 4

McGuire Containment Pressure Response Model Changes

Parameter

10/31

11/28

1/17

1/28

Ice Mass

2.220

2.220

2.220

2.220

(millions of LBM)

NS HX UA

1.86

0.735

0.735

2.03

(millions of BTU /HR- F)

KC HX UA

5.00

5.00

5.00

2.98

(millions of BTU /HR- F)

NS/RN Flow (GPM)

5000

5000

4800

3800

KC/RN Flow (GPM)

8000

8000

8000

6000

Mass and Energy Release

1974

1979

1979

1979

Model (year)

ND Containment Spray

1623

1623

1841

1841

'

,

Flow (GPM)

NO Containment Spray

3000

3000

3000

3000

Actuation Time (SEC)

_

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _

20

Active Containment

46,500

46,500

90,000

90,000

Sump Volume (FT3)

Peak Pressure (Psig)

13.3

14.42

14.45

12.7

10. RN System Walkdown

The inspectors conducted a detailed walkdown of portions of the Unit 1

Nuclear Service Water System. The inspectors reviewed the system operating

procedures, the valve checklist procedure and the system piping drawings.

The inspection was conducted to confirm that procedural valve lineups and

drawings matched the as-built configurations, to verify that equipment

conditions were satisfactory and items that might degrade performance were

identified and evaluated, to verify that valves were in proper positions and

locked if appropriate, and to verify that instrumentation was properly

valved in.

The inspectors made the following observations. Valves 1RN 893 and 1RN 894,

the inlets to the 1A1 and 1A2 Diesel generator Air Dryer and af ter dryer

respectively, were mislabeled.

Valve 1RN894 was labeled as IRN893.

The

Nuclear Service Water System valve checklist correctly described these

valves and the licensee made arrangements to correct the label plates on the

valves prior to the inspector leaving the site.

The inspector noted slight inaccuracies in the system piping diagrams, in

that relief valve 1RN-295 is located upstream of flow element 5360 as

opposed to downstream as indicated on DWG MC-1574-2.0 and vent valve IRN141

is located upstream of flow element 5930 as opposed to downstream of the

flow element as indicated on DWG MC-1574-2.0.

The licensee made arrangements

to correct these inaccuracies prior to the inspectors leaving the site.

11. Details of NRC/DPC Management Meeting Held on March 14, 1986

a.

Attendance at the Duke - NRC Management Conference on March 14, 1986,

held at DPC's request at the NRC's Region II Office included:

Duke power Company

G. Vaughn, General Manager, Nuclear Stations

T. L. McConnell, McGuire Nuclear Station Manager

,

W. A. Haller, Manager, Technical Services

'

R. L. Gill, McGuire Licensing

,

B. H. Hamilton, McGuire Superintendent of Technical Services

J. E. Snyder, Supervising Engineer

E. O. McCraw, Compliance Engineer

W. J. Kronenwetter, Design Engineer

R. W. Revels, Design Engineer

W. M. Suslick, Associate Engineer

.

-

21

Nuclear Regulatory Commission

,

R. D. Walker, Deputy Regional Administrator

A. F. Gibson, Director, Division of Reactor Safety

C. A. Julian, Chief, Operations Branch

B. T. Debs, Acting, Chief, Operational Programs Section

i

M. V. Sinkule, Chief, Reactor Projects Section

F. R. McCoy, Reactor Engineer

W. T. Orders, Senior Resident Inspector, McGuire

'

,

C. W. Burger, Project Inspector

C. L. Vanderniet, Reactor Engineer

i

b.

Members of the Duke Power Company staff met with members of the NRC

Region II staff to discuss the status of the McGuire Units 1 and 2

Nuclear Service Water System. A copy of the meeting agenda and DPC

handouts appear as Attachments 1, 2, and 3 to this inspection report.

DPC representatives stated that, from the information available to the

DPC staff, the Nuclear Service Water System had been and is currently

.

operable. The NRC staff acknowledged that, once the NRC had surfaced

concerns regarding the Nuclear Service Water System, the licensee has

placed extensive resources on solving the problem.

As a result of the aforementioned meeting, NRC representatives

contacted DPC staff on March 24, 1986, to request additional informa-

tion.

DPC staff agreed to formally submit a response by April 25,

1986, regarding the following seven requested items.

,

-

Provide the as-found and as-left RN flow balance test results for

all RN trains.

Provide the as-found and as-left VA test results for all

-

containment spray heat exchangers.

-

Provide an RN operability determination for early October 1985

when RN flow was recorded as 800 GPM to the 1A containment spray

heat exchangers.

Provide safety evaluation of the January 28, 1986 RN header

-

pressure transient.

,

-

Provide an RN operability determination based on the 1A contain-

ment spray heat exchanger throttle valve setting which existed

just prior to the first heat transfer test and based on expected

'

,

flow under accident conditions prior to heat exchanger cleaning

)

cycles.

)

-

Provide the final parameters for use in the LOTIC program and

their engineering basis.

l

Provide DPC plans to prevent a recurrence of these events.

-

-.

-.

-- --

-.

. _ .

_-

.

-

-

-

,

22

By memo of April 25, 1986, Duke Power Company responded to these

requests.

The responses contend that the RN system was continuously

operable.

Inspectors will follow up on this information during a

future inspection.

-

The resolution of these matters represents unresolved item (369/85-38-06,

370/85-39-06).

12. General Conclusions

During the operating history of the McGuire plant, the licensee has experi-

1

enced an increasing degradation of the RN system.

It is apparent that

the licensee has dealt with this situation on a case-by-case basis. Until

prompted by the NRC, the licensee had not determined the full extent of

the RN system degradation or taken adequate corrective action to preclude

i

repetition.

Although the licensee has recently dedicated significant

resources to addressing the problem, serious doubt exists regarding the

past operability of the RN system and those safety related systems, such as

containment spray, for which RN is an ancillary system.

This doubt is

fostered as a result of aggregate observations of significantly reduced

heat transfer capability of various safety related heat exchangers, reduced

RN flows, improper throttle valve settings, increased corrosion, and lack of

,

adequate preoperational testing.

This situation is contrary to 10 CFR 50,

'

Anpendix B, Criterion XVI which requires that measures shall be established

tr assure that conditions adverse to quality, such as failures, malfunc-

.

tions, deficiencies, deviations, defective material and equipment, and

'

nonconformances are promptly identified and corrected.

In the case of

'

significant conditions adverse to quality, the measures shall assure that

the cause of the condition is determined and corrective action taken to

'i

preclude repetition.

The identification of the significant condition

adverse to quality, the cause of the condition, and the corrective action

taken shall be documented and reported to appropriate levels of management.

The licensee's failure to meet these requirements, in the case of the RN

system, is a violation (369/85-38-05, 370/85-39-05).

,

)

,

1

i

,

'

l

i

.

!

.

.

_

-

_ . .

' . ' .

ATTACHMENT 1

!

DUKE POWER /NRC REGION 11

i

MEETING TO DISCUSS McGUIRE NUCLEAR STATION

NUCLEAR SERVICE WATER SYSTEM PERFORMANCE

MARCH 14, 1986

AGENDA

OPENING REMARKS

GERALD VAUGHN

OVERVIEW OF NUCLEAR SERVICE WATER SYSTEM

NEAL McCRAW

i

'

NUCLEAR SERVICE WATER SYSTEM EXPERIENCE

TONY McCONNELL

RECENT OPERATIONAL EXPERIENCE

BILL SUSLICK

(10/04/85 TO PRESENT)

DESIGN CONSERVATISMS

BILL KRONENWETTER

!

CLOSING REMARKS

GERALD VAUGHN

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^*

4

NUCLEAR SERVICE WATER SYSTEM EXPERIENCE

l.

ESTABLISHMENT OF BASIS FOR RN SYSTEM OPERABILITY

RN SYSTEM PRE-OPERATION FUNCTIONAL TEST COMPLETION DATES

7/25/79

-

UNIT 1

11/12/82

-

UNIT 2

i

NRC PRE-OPERATIONAL INSPECTION DATES COVERING RN SYSTEM

TESTING

11/03/78

UNIT 1

INSPECTION REPORT 369/78-33

-

8/16/83

UNIT 2

INSPECTION REPORT 370/82-19

-

SURVEILLANCE TESTING IMPLEMENTATION DATES

a

1/06/80

UNIT 1, TRAIN A

-

2/06/80

UNIT 1, TRAIN B

-

2/22/83

UNIT 2, TRAIN B

-

,

2/23/83

UNIT 2, TRAIN A

-

IWV AND IWP TESTING WOULD HAVE BEEN IMPLEMENTED DURING

THESE TIME FRAMES.

THE PRE-OPERATIONAL TESTS, lWP TESTS, IWV TESTS AND ESF

TESTS WERE OUR STANDARDS FOR ESTABLISHING AND

MAINTAINING RN SYSTEM OPERABILITY.

11. MAINTENANCE OF COMPONENTS BASED ON MONITORING OF OPERATIONAL

PARAMETERS

REFER TO LIST OF EQUlPMENT CLEANINGS

PERFORMANCE MONITORING PROGRAM BEGAN DEVELOPMENT IN

MARCH, 1984

l

-

-

-

'l.

t

' ' .

Ill. BEGAN EVALUATING RN SYSTEM HX'S FOR FOULING EVEN THOUGH

THERE WERE NO INDICATIONS OF FOULING

DATE WHEN UNIT 1 COMPONENT COOLING (KC) HX'S WERE

EVALUATED FOR FOULING WITHOUT INDICATIONS OF A FOULING

PROBLEM

9/01/84

DATE WHEN KC HX'S WERE CLEANED

11/84 - UNIT 1

6/85 - 7/85 - UNIT 2

EVALUATION AND INSPECTION / CLEANING DID NOT DETERMINE

THAT KC HX'S WERE INOPERABLE

IV. RN SYSTEM OPERABILITY REEVALUATION BASED ON 1A RN PUMP TEST

RESULTS

DATE WHEN A FLOW MEASUREMENT PROBLEM ON 1A RN PUMP WAS

IDENTIFIED

10/04/85

A REEVALUATION OF OPERABILITY CRITERIA WAS BEGUN TO

REFOCUS OPERABILITY CONCERNS FROM THE RN PUMP TO THE RN

SYSTEM AS A WHOLE

V. ACTION ITEMS RESULTING FROM REEVALUATION OF RN SYSTEM

OPERABILITY CRITERIA

BEGAN THE PERFORMANCE MONITORING PROGRAM ON

l

RN HX'S ON 11/01/85

THE RN SYSTEM TESTING PLAN WAS SUBMITTED TO

'

REGION 11 ON 12/01/85

THE UPDATED RN SYSTEM TESTING PLAN WAS SUBMITTED

TO REGION ll TO INCLUDE TESTING OF ALL 62 RN

HX'S AND RESOLVE 1A RN PUMP FLOW INDICATION PROBLEM

ON 12/18/85

!

NOTE:

IN ALL THE TESTING AND ANALYSIS DONE IN 1985, WE

DID NOT DETERMINE THAT ANY OF THE HX'S EVALUATED WERE

INOPERABLE.

1

f

_ _ _ _

, .,

,

't

EQUIPMENT CLEANINGS

j

!

LOWER CONTAINMENT VENTILATION HX FOULING WAS IDENTIFIED AS

ONE OF THE FACTORS IN THE LOWER CONTAINMENT COOLING PROBLEM

10/22/82

NOTE:

(A)

FOULING OCCURRED AT LAKE TURNOVER IN THE FALL.

ONLY TIME WE HAD TO CLEAN.

(B)

BIOFOULING WAS EVIDENT DUE TO HOT AIR ON SHELL

SIDE.

CONTROL ROOM VENTILATION (SHARED BETWEEN UNITS 1 AND 2)

'

TRAIN A

TRAIN B

11/19/82

3/83

{

10/03/83

1/07/85

12/19/83

10/21/85

5/30/84

11/05/85

10/31/84

9/25/85

10/24/85

10/31/85

PENETRANT / DISPERSANT ADDED TO THE RN SYSTEM IN ATTEMPT TO

CLEAN LOWER CONTAINMENT COOLING UNITS

4/27/83

REACTOR COOLANT PUMP MOTOR COOLERS

UNIT 1

UNIT 2

12/31/84

8/10/84

11/08/85

.

ASSUMPTIONS

1.

ALL SAFETY RELATED EQUIPMENT REQUIRE FLOWS CONCURRENTLY

5"

THROUGHOUT DESIGN BASIS EVENT.

g

5s

4

m

2.

HEAT EXCHANGERS DESIGNED FOR MAXIMUM POND TEMPERATURE OF 95 F.

4

FLOW AND FOULING DESIGN MARGIN

AFFECTS ON CONTAINMENT PEAK PRESSURE

(CONTAINMENT DESIGN = 14.9 PSIG)

(x10bfuHROF)

(x10bIUHROF)

tbs

OfE'

T0

CLEAN

OfFFkbfEik

5.18

8.11

5000

8000

-

DESIGN

(FSAR)

2.94

5.00

5000

8000

12.36

75% NS

DEGRADED

FROM DESIGN

0.735

5.00

5000

8000

14.42

REDUCED FLOWS

DEGRADhDHXs

1.47

2.98

3800

6000

13.59

0 = U3 (LMTD)

U = (-00 LING, FLOW)

.

..

l-

.g

.

NUCLEAR SERVICE WATER

ESSENTIAL COMPONENT FLOW REQUIREMENTS

,

4

!hkE

DESIGN F OWS

AChl0WS

COMPONENT

(FSAR

KD Hx

900

900

KC Hx

8000

6000

i

NS Hx

5000

3800

,

VC/YC CONDENSER

775

775

I

KF ES COOLER

20

15

NS ES COOLER

45

20

ND ES COOLER

45

20

NV PUMP COOLERS

28

15

NI PUMP COOLER

20

20

i

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-

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K ,*. ,,

-

ADDITIONAL DESIGN MARGINS

1.

LOWER SNSW POND TEMPERATURE

2.

HIGHER ICE WEIGHT

3.

LOWER RWST TEMPERATURE

.

. . .

. -

.

.

.

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ATTACIGIENT 3

,,

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PERFORMANCE MONITORING PROGRAM

  • Reliability. Efficiency and

AvaiIabiiity

  • Monitors the overall health of

equipment

  • Development begun in March. 1984
  • Tangible results already being

reaIized

.

. ' .

3

l'

.

NUCl. EAR SERVICE WATER PUMP (RN) 1A

  • RN Pump 1A did not meet its quarterly

IWP acceptance criteria (10/4/85)

  • Replaced impeller (10/5 - 10/6/85)
  • Performed new pump head curve /lWP

baseline test (10/7/85)

  • Troubleshooting
  • Evaluated the pump acceptance criteria

based on the actual system demand

  • Conducted the pump head curve using

the 2A and 1A KC flow elements in

series with the 1A RN flow element

  • Using the most conservative head curve

results 1A RN pump was declared

operable (10/11/85)

  • Optimum replacement was a calibrated

84" flanged spool-piece with a 0.831

beta ratio orifice

  • Installation (February 26-28,1986)
  • 1A RN Pump head curve conducted with

new flow element (March

3,

1986)

    • Summary - The pump was never

inoperable, fouling of the flow

element resulted in errors in the

conservative direction.

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CONTAfNMENT SPRAY (NS) HEAT EXCHANGER

  • 1A NS Heat Exchanger had a high

differential pressure

  • Commission expressed concerns of

'

biological attack of s t a i ti l e s s steel

tubes

  • Testing Performed:

1.

Structural Integrity Test

2.

Minute

Leakage Test

3.

Heat Balance Test

  • Structural Integrity and Minute

Leakage Test indicated insignificant

leakage

  • Visual Examination of the tubes
  • Heat Balance Testing quantified the

extent fouling had occurred

"

  • Peak Containment Accident Pressure

CLOTIC) calculations showed the heat

exchanger could still perform its

function

Cleaning iterations

  • Tested and cleaned the other NS heat

exchangers based on 1A experience

  • * Summary:

1.

NS Heat Exchangers are intact

2.

The NS Heat Exchangers were

fouled; however reanalysis proved

operability

j

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Containment Spray lleat Exchanger Testing

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lleat Exchanger

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Temperature)

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Containment

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Spray Pump

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FWST

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As Found

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OTHER HEAT EXCHANGERS and FLOW BALANCE

  • Began evaluation and procedure

generation for testing essential heat

exchangers (10/24/85)

  • 1A Train RN flow balance performed

aligned to low level intake (12/17/85)

  • 1A NV Pump Speed Reducer Oil Cooler

cleaned (12/20/85)

  • Conducted test using NSWP as suction

(1/27/86)

  • Inadequate flow to some heat exchangers
  • Reanalyzed the necessary flow rates to

the KC and NS heat exchangers

  • Design review of alignment

configuration to properly conduct flow

balance to meet all design assumptions

  • Conducted the 1A Train flow balance

throttling flow to the KC and NS heat

exchangers, aligned to NSWP with 6000

gpm supplied to other unit (1/28/86)

  • Performed other flow balances

(1/30 - 2/28/86)

  • Began extensive cleaning, testing and

inspections of all essential heat

exchangers (2/3/86)

Total Cleaned / Tested / Inspected:

54

Total Number of Hea t -Ex change r s : 62

    • Summary:

Cleaning and testing of all essential

RN system components is on schedule

to meet March 31, 1986 completion

.-

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ATTACHMENT 4

!

NRC Inspection Team Confirmatory VA Calculations

't

.

Calculations were performed to evaluate the containment spray heat exchanger UA

j

value used in containment pressure calculations performed by Westinghouse for

Duke Power on November 28, 1985 and January 17, 1986. The following nomenclature

is used in the subsequent calculations:

.

Nomenclature

i

A

-

heat exchanger area

,

tube inside diameter

Di

-

i

Do

-

tube outside diameter

i

.

1

Fi

-

tube inside fouling factor

Fo

-

shell side fouling factor

F

-

appropriate shellside fouling factor

oAPP

i

mass flux

i

G

-

4

tube inside heat transfer coefficient

!

hi

-

shell side heat transfer coefficient

i

ho

-

!

K

water thermal conductivity

,

-

K

-

stainless steel thermal conductivity

ss

water Prandtl number

Pr

-

1

Reynolds Number GD

Re

-

D

l

y

'

i

oFo

two standard deviation uncertainty in Fo

--

u

liquid viscosity

-

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t

!

f

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li

..

.

. .

. .

.

-

.

. ..

-

..-

a

.

_

_

. .

__ _

.

.

._

.

.

. . - .

_

-.

.

i-

!

1 -.

Attachment 4

2

i

!

J

j

The UA design value for this heat exchanger is 2.95 x 105 Btu /h/ F while

Duke provided Westinghouse with a degraded value of 7.35 x 10' Btu /h/ F,

24.9% of the design value. Experimentally determined UA (11/22/85) values

indicated that the actual degraded value was ~8.77 x 10' Btu /h/ F, 29.7% of

f

the design value.

Confirmatory UA calculations were performed by initially determining a

design heat transfer coefficient for the shell side of the heat exchanger.

f

I

This was done by using design value fouling factors, and assuming that the

I

tube side heat transfer was correctly predicted by the McAdams equation at

j

the design conditions.

,

hiDi

'

k

=0.23 Re .8

1/3

(1)

.

D Pr

The UA for the heat exchanger is

!

UA =

A

(2)

p

1_ + F4 po + F

+D

In p

1_

+

g

o

g

g

h

Di

hi

Di

2K

UI

g

ss

For the design condition, all values (including UA) are known except h ,

g

[

which was determined to be 918 Btu /h/ F/ft2 using equation (2).

The information supplied to Westinghouse by Duke was acquired from experi-

mental testing of heat exchanger 1A on 11/22/85.

The data from this

e<periment was used to determine an appropriate value for the degraded UA by

'

j

first determining the as-tested fouling factor.

In order to do this,

experimental flow rates, temperatures, etc. had to be used to determine both

i

tube side (h ) and shell side (ho) heat transfer coefficients appropriate

4

!

for the test. Tube side heat transfer coefficients were determined using

equation (1) evaluated at the test conditions.

Shell side heat transfer

i

coefficients were assumed to scale as:

i

f

13

h po - Re

Pr

g

D

i

K

(3)

j

This equation is used frequently in determining shell side heat transfer for

shell and tube heat exchangers. Equation (3) was evaluated at both design

'

and test conditions, and an h for the test was calculated from the design

g

h, determined above.

Equation (2) was then used to determine the fouling

factor appropriate for the shell side under as tested conditions assuming

!

the tube side fouling factor is the design value of .0005 (this assumption

actually has no impact on the final VA since the two fouling factors are not

1

a function of flow and fluid conditions).

The shell side factor was

'

determined to be

.00912

(4)

F

=

g

,

. - -

Y

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-

-

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. . . ,

. ,

-

-_

-

. - -

.

-

.

-

___ - - _

_

__

-

.

.

.

l

Attachment 4

3

'

.

,

for the experiment vs. the 0.001 design value.

In addition to this

calculations, the experimental error associated with the testing equipment

and procedure was used to determine an uncertainty value for F . This

g

calculation was performed using propagation of errors (see for example

!

Beers,1957, " Introduction to the Theory of Error") through the equation

(5), the energy balance on the NS side of the heat exchanger (only the NS

flow was used to determine overall heat flow).

'

o

Q = mCp (T

-Tin)

(5)

out

The uncertainty in temperature measurements were given to the NRC team

by licensee representatives as

.4 F including both RTD, and signal

conditioning equipment error. These RTD's were apparently calibrated before

testing, which increases confidence in the temperature measurements.

Additionally, errors in the flow measurements were also included. Handbook

uncertainty values for uncalibrated orifice plates are typically 1%-2.5% of

measured flow. In addition to this, there are uncertainties associated with

the other instrumentation necessary to make the flow measurements (DP cells,

)

readouts,etc.). The orifice plate was an uncalibrated process device so it

was estimated the overall uncertainty was ~5% of the measured value.

Each

of the uncertainties stated above were treated as one standard deviation

.

(lo) uncertainties.

It is believed that a two standard deviation (2o)

i

uncertainty bound should be applied in order to insure conservatism (two

standard deviations give a 95% certainty of the measurement). The 2a value

for Q was found to be ~12%. Additionally, since design heat flow was based

4

i

solely on calculations and not on tests.

It was assumed that a 2.5% error

(lo value) was present in the design heat flow determination. It was also

assumed that equations (1) and (3) could be used to correctly scale with

temperature level and flow rate (0 uncertainty was assigned to this

process).

The two errors above, experimental and design, were used to

determine overall error in F by propagating errors through the calcula-

g

'

tions described above.

The two-standard deviation uncertainty in F was

o

determined to be:

4

oF

= .00149

(6)

g

i

for the uncleaned case of heat exchanger 1-A.

An appropriate UA value for

the Westinghouse calculations was then determined by using:

FoAPP = F + oF

(7)

g

g

i

These values were determined for three cases:

unit 1-A before cleaning,

unit 1-A as it existed after last cleaning, and unit 2-B.

The table below

summarizes these results (in all cases, RN flow was assumed to be 4800 gpm).

J

-

- -. .

- . - . - - _ _ . _ _ _ _ - - _ - , . , _ _ . - - _ _ - - - . _ - _ - .

- - - _ . _ - , _ -

. - _ , . , _ _ - -

.

'.

Attachment 4

4

,

Summary of Calculations

UNIT

STATUS

F

oFo

UA

g

1-A

uncleaned (11/22/85)

.009

.0015

8.18 X 105

1-A

cleaned (01/16/86)

.0033

.0007

1,63 X 105

2-8

uncleaned (01/24/86)

.011

.0127

7.16 X 10'

Westinghouse input

7.35 X 105

The UA value calculated for the 2-B uncleaned case is slightly below that

given to Westinghouse on 11/28/85 and 01/17/86. However, if the containment

pressure calculations performed on 01/17/86 are used as a starting point,

and the containment pressure change with VA change is similar to that noted

in the 3 calculations performed on 11/28/85, the peak containment pressure

can be estimated for a UA value of 7.16 X 105

These calculations estimate

that the peak containment pressure for this UA value would be approximately

P = 14.56 psig, still below the 15 psig limiting value.

The calculational methods used to evaluate heat exchanger performance appear

to be reasonable. However, when calculations are being performed to deter-

mine heat exchanger performance at reduced flow, it is also necessary to

apply appropriate fouling factors to heat exchangers which are suspected of

being fouled. This has not been done in previous Duke calculations. As an

example, the inspection team looked at the charging pump speed reducer oil

cooler. Duke has found the oil inlet temperature to increase from 141'F to

166 F when RN flow to the heat exchanger is reduced from 20 gpm to 10.7. _In

addition to the reduced water flow, the effect of fouling should also be

considered. Confirmatory calculations were performed assuming both reduced

flow and a fouling factor of ~.008 on the RN side and .001 on the oil side

(design fouling factors were presented as a sum of F +F4 =.0025). The RN

g

fouling factor is an estimate based on findings in the uncleaned containment

spray heat exchanger (F, =.009) and recognizing that continuous water flow

through the oil cooler might reduce fouling somewhat. A summary of the

maximum oil temperatures is presented in the following table.

A calculation with the RN cooling water temperature reduced to 65 F is given

to demonstrate the cooling water temperature effect on heat exchanger

performance.

As can be seen in the below table, the reduction in RN

temperature from 95 F to 65 F has a significant impact on oil temperature.

A similar effect would be seen in other heat exchangers in the train

(although not exactly the same magnitude).

Comparison of 011 Cooler Assumptions

Cooling Water Inlet Temp.

Flow (gpm)

F

T

( F)

9

oj)

95 F (Design)

20

.0015

141

'95'F

10.7

.0015

166

95*F

10.7

.008

185

65'F

10.7

.008

155

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