ML20214B231
| ML20214B231 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/14/1987 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20214B158 | List: |
| References | |
| FOIA-84-741, FOIA-84-742 NUDOCS 8705200194 | |
| Download: ML20214B231 (73) | |
See also: IR 05000275/1982042
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UNITED $ TATE $ OF AMERICA JJ
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NUCLEAk kEGULATORY Colff155 ION
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. REFORE THE ATOMIC SAFETY AND LICENSING _ APPEAL BOARD
In the Matter of )
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PACIFIC CA$ AND ELECTRIC COMPANY ) Docket No. 50-27$ -
Diablo Canyon Nuclear Plant. Unit No.1 )
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AFFIDAVfT OF FHII.Tp J. HD,R, ER
STATE OF CALIFORNIA )
COUNTY OF CONTRA COSTA ) SS .
1, Philip J. Morrill, being duly sworn do depose and say:- 1
1. I an employed by the U.S. Nuclear Regulatory Commissica in the Region V
office, Division of Resident. Reactor Projects and Engineeries Programs.
A statement of my professional qualifications is attached hereto as
Exhibit A and incorporated herein by reference.
- 2. I am a kesctor Project Inspector and have had responsibility for
i inspecLion of the Diablo Canyon preoperational testing program. 1 as
, also responsible for the inspection of power ascensions testing and plant
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operations. This inspection effort is in addition to the effort of the '
[x Resident inspectors and supplements their work. Since October 1981, I
have also had responsibility to conduct the Region V inspect.isa program
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related to the Diablo Canyon Nuclear Plant Independent Design
verification Program (IDVP).
3. I have personally conducted periodic inspections of the Diabla canyon
facility since the fall of 1978. As part of my inspection efforts during
. the period December 2,1982 .Tanuary 1,1983, I specifically examined
the allegations previously made to the Region's office by Mr. John
Cooper. The results of these eff orts are documented in paragraph "9" of
NRC Inspection Report No. 50-275/83-04, attached hereto as Esbibit 3 and
incorporated herein by reference.
I attest that the foregoing affidavit la true and correct to the heat of my
knowledge and belief.
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"
Philip J. Morrill ;
Subscribed and sworn to before == l
this day of Merch 1984 .
Notary Public .. .
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fly Commdsmion espires: [
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8705200194 870514 ~
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DEVINE84-741 PDR
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j Exhibit A ,
PHILIP J. NORRILL
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PROFESSIONAL QUALIFICATIONS
REGION V - WALNUI CREEK, CALIFORNIA
LSu}fUCMA_R REGULATORY COMMISSION
ny,.name is Philip J. Morrill . 1 am employed by the United States Nuclear
segulatory Commission as a Re. actor inspector in the Division of Resident,
Reactor Projects and Engineering Programs, Region V, Walnut Creek, California.
My primary responsibility in this position is the inspection of nuclear power
plants during the operating phase to determine compliance with NRC rules and
regulations and thereby Scrify safety of operations.
T received a Bachelor of Science degree from the U.S. Naval Academy in 1966.
I was empicyed by the U.S. Navy in the Naval Nuclear Poser Submarine program
from 1966 until 1971. During this time, I became qualified as Engineering
Officer of the Watch for the AIV pressurized water nuclear prepulsion plant
proLotype and was later qualified as Engineering Officer of the Watch on board
the USS John Marshall (SSBN 611 (G)), a nuclear powered polaris missile
submarine (1969 through 1971). T was also the ship's Main Propulsion
Assistant (responsible for maintenance and administration of the nuclear
reactor end power generation equipment) for two and uun-hall years of this
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time. In 1971, I joined the Bechtel Corporation in San Francisco, Unliforula
and was apsigned to the Susquehanna Steam Electric station project mechanical
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s. group. From August 1971 through September 1972, I was rwspuumible for the
design and development of the radioactive waste disposal system. From
September 1972 through January 1974, I was promoted to the position of project
licensing engineer. From January 1974 through March 1976, I was again
promoted to the position of Project Nuclear Group Leader, responsible for
managing and supervising the efforts of six to fourteen engineers.
In March 1976, I was hired by the U.S. Nuclear Regulatory Commission, Office
of Inspection end Enforcement, Region V, in Walnut Creek, California, as a
reactor inspector for the Reactor Construction and Engineering Support Branch.
In this position, 1 participated in construction inspections of the San Onofre
Nuclear Generating station and successfully completed a nondestructive
examination school al Convair Division of General Dynamics, (San Diego,
California). J also coupleted NRC sponsored schools in quality assurance and
InspecLion Techniques in Bethesda, Maryland. In January 1977, I transferred
l to the Reactor Operttions and Nuclear Support Branch of Region V, Office of
i Inspection and Enforcement and was assigned as back-up inspector for the
l Trojan Nuclear Plant. In succeeding months I participated in inspections of
the Rancho Seco. Humboldt, and Trojan nuclear plants in addition to completing
five weeks of pressurized water reactor systema and operating training. Tor
one year I was then assigned as Principal Inspector for the Trojan Plant. In
the fall of 1978, my assignment was changed to follow-up the preoperational
testing of the Diablo Canyon Nuclear plant and to conduct operations
inspections at the Rancho Seco Nuclear Plaat. In 1980 my duties were modified
to include conducting the portions of the NRC inspection pzogrAm_ conducted by
the regional office staff related to preoperational testing, power ascension
( j testing, and plant operations of the Diablo Canyon Plant. Atter the discovery
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k' of significant design deficiencies with the Diablo Canyon Plant la the Falt og
1981, I was tasked to monitor and inspect the activities of so ledependest '
DesignVergigationProgrammanagedbyTeledyneEnglaearingServices. .This
program i_ ...r. conducted to determine the adequacy of design and
construction of the Diablo Canyon Plant and to determine the adequacy of any
modifications which' result from that program. Although theaa have been my
principle assignments, I have also participated la a variety of power, and
research reactor inspections during the last five years.
I Is present.ly a registered Professional Mechanical Engineer and Nuclear
E5gineer in the State of California.
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U. 5. NUCLEAR RECUIATORY COMMisslott
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Report No. 50-275/82-42 ,
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t Docket No. 50-275 _ ticano. p., DPR-76 safeguards croup
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Licensee Pacific Gas and Electric Company -
P. O. Box 7442
San Francisco, California' 94106
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raettity wa= : Diablo Canyon Units 1 and 2
Inspection at: Diablo Canyon Site, San Luis Obispo County, California
Inspecti:n c ducted: December 2,1982 through January 1,1983
Inspectors: - ~f b*~b
d. D. Date Signed
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9,bCarlson,Sr.,gesidentInspector
M. M.JMendonca, R si(int Inspector
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Data signed
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Pi P. J
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. Porrtil, Reaqtor Inspector
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Date Signed
Approved by: l - ) k #83
0. F . K1rsch, Chief, l{f actor Projects Section No. 3 Doce signed
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Svenary:
Inspection from Dece:nber 2,1982_thrcuch January 1,1983 (Report No. 50-275/82-42)
Areas Inspected: Routine inspections of plant operations, surveillance testing,
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pbysical security, follow-up of allegations regarding the RHR system, maintenance,
the licensee's audit program and emergency preparedness activiti,es. The inspection
involved 128 inspector-hours by three NRC inspectors.
Results: No items of noncompliance or deviations wer* identified. l
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DETAILS
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1. P_ersons. Contacted
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a. Site -
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- R. C. Thornberry, Plant- Manager l
- R. Patterson, Plant Superintendent
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- J. M. Giscion, Pcwer Plant Engineer *
D. A. Backens, Supervisor of Maintenance
- J. A. Sexton, Supervisor of Operations
- J. V. Boots, Supervisor of Chemistry and Radiation Protection
- W. B. Kaefer, Technical Assistant to the Plant Manager
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- R. G. Todaro, Security Supervisor
- R. T. Twiddy, Supervisor of Quality Assurance
- R. M. Luckett, Interim Regulatory Compliance Engineer
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b. Corporate _
- J. O. Schuyler, Vice President Nuclear Power Generation
- W. A. Raymond, Manager Qua.lity Assurance ;
- T. G. de Uriarte, Senior Engineer (Audits) l
F. J. Dan, Supervisor Electrical Engineer
R. Otto, Electrical Engineer *
N T. Crawford, Senior Mechanical Engineer -
sj J. McCracken Senior Mechanical Engineer
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G. C. Wu, Licensing Engineer
The inspectors also interviewed a number of other licensee employees including
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shift supervisors, reactor and auxiliary operators, maintenance personnel,
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plant technicians and engineers, quality assurance personnel and members
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of General Construction.
l * Denotes those attending the exit interview of January 7,1983.
- Denotes those attending the exit interview of Deccinber 14, 1982.
2. Operational Safetyyrgication
During the inspection period, the inspectors observed and examined activities I
to verify the operational safety of the licensee's facility. The observations
and examinations of those activities were conducted on a daily, weekly or
- monthly basis.
On a daily basis, the inspectors observed control room activities to verify
compliance with limiting conditions for operation as prescribed in the facility
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Technical Specifications. Logs, instrumentation, recorder traces, and other-
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operational records were examined to obtain information on plant conditions,.
- trends, and compliance with regulations. Shift turnovers were observed
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on a sample basis to verif.y that all pertinent information * "'
on plant status
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7 During each week, the inspectors toured the accessible areas of the facility
to observe the following: :
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a. General plant and equipment conditions.
b. Maintenance activities and repairs (See Section 3).
c. Fire hazards and fire fighting equipment. -
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d. Ignition' sources and flarenable material control. r
e. Conduct of selective activities for compliance with the licensee's -l
administrative controls and approved procedures. '
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f. Interiors of electrical and control panels. -
g. Implementation of selected portions of the licensee's physical . security '
plan.
h. Plant housekeeping and cleanliness.
The inspectors talked with operators in the control room, and other plant
personnel. The discussions centered on pertinent topics of general' plant
,, conditions, procedures, security, training, and other aspects of the involved
work activities.
No items of noncompliance or deviation were identified.
3. Maintenance -
Maintenance activities on a safety injection accumulator isolation valve ~i
motor and a rod drive power supply motor-generator set were reviewed by _I'
the inspectors during the month. . Observations by the inspectors verified
/ that proper approvals were obtained and system clearance and tests of redundant.
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equipment were performed, as appropriate, prior to conducting maintenance i .
on safety related systems or components. The inspectors verified that qualified
personnel performed the maintenance and used appropriate maintenance procedures..
Replacement parts were examined to determine the proper certification of *
materials, workmanship and tests. During the actual performance of maintenance
activities, the inspectors verified proper fire protection controls and
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housekeeping. Upon completion of the maintenance activity, the component
was tested prior to return to service.
No items of noncompliance or deviation were identified.
4. Surveillance
Surveillance testing on 4 KV relays and contacts, and atmospheric steam
i dump instrument loops were reviewed by the inspectors. Observations by
the inspectors including verification that proper procedures were used,
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test instrumentation was calibrated, and that the tested systen or component
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was properly removed from service as required by the test procedure. L Upon
completion of the surveillance tests, the inspectors verified that the test
results met the acceptance criteria of the Technical Specifications and
were reviewed by the cognizant licensee personnel. The inspectors also
j verified that corrective action was initiated, if required, to determine .
the cause for any unacceptable test results and to restore the system or
A component to an operable ~ status consistent with the technical . specification .f
requirements. -
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No items of noncompliance or deviations were identified.
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i 5. Emergency _ Preparedness. i
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The inspectors reviewed and observed an emergency drill by a plant fire
4 brigade and industrial safety and fire protection training. -
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No_ itens of noncompliance or deviations were identified. -
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6. Audit Program and Implementation .
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The inspector reviewed the licensee's audit program and implementation thereof
to determine if the program conformed to ANSI N-18-7-1976 and ANSI N 45.2.12-
1977. In addition, the inspectors verif.ied that auditor qualifications
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were consistent with ANSI N 45.2.23-1978.
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The following procedure manuals that describe the licensee's Audit Program
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were' reviewed:
a. _ Quality Assurance Manual for Nuc1 gar _fogr Pl ants
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(1) Section SVIII - Audits'
(2) Section SVI - Corrective Action
(3) Procedure 10.1 - Nonconformance and Corrective Actions
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\ (4). Procedure 11.1 - Audits Performed by Company Departments
(5) Procedure 11.1, Supp. 1 - Open Items Report
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b. Nuclear Power Generation Manual - Quality Assurance
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t (1) Procedure 1.1 - QA Department Program and Organization
l (2) Procedure 2.2 - Training and Indoctrination
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(3) Procedure 15.1 - Nonconformance Reports
(4) Procedure 16.1 - Open Item Reports
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(5) Procedure 17.1 - Auditor Qualifications
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(6) Procedure 18.2 - QA Audits
(7) Procedure 18.6 - Planning / Scheduling of Audits
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c. _ Quality Auditor Handbook
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Based upon the review of the above noted procedures, the inspectors determined
that the licensce's QA Audit Program conforms to'the criteria of ANSI N
L . 18.7-1976 and ANSI N 45.2.12-1977
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1 The inspectors reviewed the licensee's auditor qualification program, tests,
and records to ensure audits were being conducted by properly qualified
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auditors. The inspectors determined the licensee's auditor qualifications
- were consistent with ANSI N 45.2.23-1978.
I Next, the inspectors reviewed the following audit reports to determine if :
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audit plans, checklists, findings and corrective actiorf'follosups were being
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i a. Audit #12300 " Criterion XVIII - Audits" -
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b. Audit f20400 " Criterion XV - Nonconformances and Criterion XVI li'
Corrective Action"
c. Audit #20416 " Criterion XV and XVI" -
d. Audit #20500 " Fire Protection"
e. Audit #21011 " Status of Open 0!R's"
f. Audit #20919 " Technical Specifications" .
g. Audit #21111 " Containment Annulus 5 teel"
The inspectors determined that the audits were being conducted properly
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using the prescribed audit plans and checklists; however, tracking of "Open
Items" was weak in- that audited organizations were not responding to adverse
audit findings in accordance with the criteria of Section 4.5 of ANSI N
45.2.12 1977 Specifically, estimated completion dates (ECO) were being
- ..excceded with no new (ECD's) being established. Additionally the audit
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findings were being tracked using the licensee's "Conraitment dontrol $ysten"
that assigns a noncontrolling priority to all adverse audit findings. The
inspectors identified to management that some of the findings would have
resulted in technical specificat, ton violations if fuel loading had comenced -
without correction of the identified problems from audit findings. During
, the exit interview, the licensee comitted to having revised ECD's for al
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outstanding Open Item Reports by February 1,1983. .and prioritizing all '
, outstanding Open Item Reports by February 28, 1983 (82-42.01).
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No items of noncompliance or deviations were identified.
7. Review of Stone and Webster Construc_t, ion Audit
As part of the Independent Design Verification Program (IDVP). Stone and
. Webster Engineering Corporation (SWEC) was tasked with the evaluation of
the construction quality assurance program at Diablo Canyon under the auspicies
of Teledyne Engineering Service (TES). The inspectors reviewed the following
documents and discussed the audit with PGI,E representatives to determine
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how open item reports were being generated and dispositioned.
a. Adjunct Program for Evaluation of Construction-Quality Assurance -
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Rev. I dated 10/1/82 (TES document).
I b. Construction Quality Assurance Evaluation (SWEC Project Procedure 4-2-1
dated 10/22/82).
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c. Diablo Canyon Verification Program (DCVP) Procedure #1 - Interface with
Consultants.
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d. DCVP Procedure #2 - Program Resolution Reports.
The inspectors determined the scope of the audit was to evaluate the as-
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built quality of two contractors: 1) Guy F. Atkinson Co. - Containment
Building Contractor, and 2) Wismer and Becker Co. -- installation of N555
piping. The above noted procedures described the auditintprocess to be f
x used and handling of audit findings. The inspectors ha*ve reviewed the program ,
for familiarization. At the present time, the SWEC onsite audit team has
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completedtheas-builtauditandhasgeneratedtwenty-nineOp[enItemReports
(OIR). So far, the licensee has dispositioned eighteen of the DIR's. The t.
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inspectors will complete'the review in this area when the remainder of the
- OIR's ar~e dispositioned (82-42-02). j
No items of noncompliance or deviations were identified.
8. Open Items Followup,
Plant Administrative procedures C451 and 0756 have been prepared to assure
reinstatement of Environmental Qualification conditions after maintenance
or surveillance testing. This closes open items 80-16-01 and TI-15-41.
9. Alic9ations Regarding the Diab,1o_ Canyon Residual Heat Removal System
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On Decmber 2,1982 the inspector met with licensee representatives to discuss
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allegations regarding the Diablo Canyon residual heat removal (RHR) system.
These allegations had also previously been examined at the jobsite and documented ,
in Region V inspection reports 50-275/82-26 and 50-323/82-13. The following
paragraphs paraphrase the allegations, summarize the inspection, .and state
j the findings of the inspector.
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(a) Allegedly there were no control and interlock circuit drawings for-
uj motor operated valves 8701 and 8702 (RHR hot leg suction isolation
- valves). The inspector examined PG&E drawings 437592 " Residual Heat
Removal Flow Control Valves", and 103058 " Circuit Schedule 480 Volt
, for Busses F, G. H" circuits H19P00 through H19P12 and G25P00 through
! G25P13. The inspector observed that these drawings describe the power. .
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control, and interlock circuits for the subject valves. The allegation
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was not substantiated.
(b) Allegedly no one knew how these circuits were routed in the plant.
Licensee project engineering personnel stated that in addition to the
, drawings described above, the raceway schedule depicts circuits in
a particular conduit, the conduit drawings show conduit locations in
the plant, and the circuit schedule itemizes the pull data for each
wire in the plant. They also stated that the drawings and schedules
. Were available to the plant staff-through the site document control-
! center if this material was not available in the control room. The
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inspector had previously verified that this type of documentation was
properly controlled and readily available to the plant staff. This
allegation was not substantiated.
! (c) It was alleged that the design was no good in that the control / interlock
l circuits are routed from the "hagen" racks via the solid state protection
- system to the relays which shut the valves. Licensee engineers explained
that this was a standard Westinghouse design and that-the. "hagen" racks
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( took low level analogue signals and (in this case) used bistables to
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generate signals in the milliamp range. The solid state protection
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- system completes the logic function and generates a larger output signal
(amps.) which in turn actuates relays in the auxiliary logic cabinet.
They explained that they were not in a position to change this arrangement
(since it is a Westinghouse design) and that they were unaware of any .-
problems with this arrangement. The inspector examined the location
- of the components of the RHR isolation valve control and interlock
circuits to verify the licensee's statements. The- allegation was substantistd
- to the extent that the circuits were as-alleged, however there was !
,' no apparent deviation from regulatory requirements or safety criteria.
! (d) It was alleged that a design change request (DCR) submitted about February
i 1981 to get " rid of that system" (i.e. RHR hot leg suction isolation
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interlocks) has never been acted upon by PG&E, The inspector verified
the system in question) and that none were originated from or arrived
at the Diablo Canyon project. The site Resident inspectors verified
that no DCRs were outstanding for this drawing at the jobsite. This
allegation could not be substantiated.
(e) It was alleged that the FSAR, Chapter 5, paragraph 5.7 pages 37b and
38 as well as Chapter 7, paragraph 6.2, pages 3 and 4 describe the
3 automatic high pressure /high temperature isolation of the RHR systes
! i from the reactor coolant system, and that this is inconsistent with -
a the technical specifications section 3.4.9.3 which requires AC to be
e removed from the associated valves (8701 and 8702) thereby disabling
the automatic isolation features. Therefore the FSAR should be amended.
Licensee representatives showed the inspector Table 6.3-10 of the FSAR
which shows that the valves are to be shut and racked out at power
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and open and racked out during shutdown cooling ' mode. This is in accordance
i with NRC direction. The licensee representatives also stated that
the entire FSAR would be updated (with inconsistencies removed) in
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September 1983 in accordance with 10 CFR 50. The allegation was partially
i substantiated, but no safety problem or.. noncompliance with regulatory
requirements was identified.
(f) The alleger stated that the FSAR section 3.1.3 states that spurious
j closure of normally open/ fail open valves'is not considered as either
a passive or active failure and is not analyzed for at all which is
. a problem. Licensee engineers explained that there were no reasonable
failure modes which would cause normally open/ fail open or normally
closed / fail closed valves to change state. The only possibility they
could imagine was a " copper octopus" which caused ~ selective shorting.
This issue had been dealt with in the Fire Protection Review and was
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- one reason that certain valve circuit breakers were racked out after
+ the valve was placed in the desired position. As far as control circuits
are concerned, any short with 120 volts or higher would cause the logic
circuits to go to a f ail- safe condition due to thei overwhelming signal '
[. ~. strength (normal signals are 4 to 20 milliamps). * Ths aT1egation could
j V not be substantiated. [
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- - - . . . . _.
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, . .
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! (g) It was alleged that there was no low flow alarm for the RHR system
'
,hj and that there should be one. The inspector verified that an RdR -
pump trip .is annunciated, that shut RHR suction valves are indicated,
and that the subcooling meter was available to ensure adequate core-
cooling. 1.icensee representatives pointed out that the*RHR pumps have
a miniflow recirculation to maintain some flow, and that the monitor light
box indicates valves or circuits in the incorrect state. The inspector '
concluded that the allegation was correct in that there was no " low v ,
flow" alann, but also concluded that there appeared to be no requirement
or necessity to have one.
'
. (h) It was alleged that an RHR pump ran without flow for 5 minutes in September
1981, and that this event was not reported as required by administrative '
procedure C-12 and 10 CFR 50.72 The site resident inspector verified '
- that a Nuclear Plant Problem Report (DCI-81-0P P1057) and the associated
corrective action was completed. The allegation was not substantiated.
~
!
(i) It was alleged that the RHR hot leg suction does not meet the single
! . failure criteria for function (suction from reactor coolant system
hot leg), that newer plants had this feature, and that this portion
of the system should be redundant to meet 10 CFR 50 Appendix A Design
Criteria. The inspector verified that this function was not safety-
j . related in the Diablo CanyoO plant design by examining the FSAR. The
1 - inspector observed that the suction from the containment sump and from .
J the refueling water storage tank were both safety related and arranged l
to meet regulatory requirements for redundancy. The inspector also
, observed that some other plants did have two RHR suction lines but- ,
4
'l that these plants 'used a different nuclear steam supply system vendor. '
The inspector concluded that the allegation was correct in that the
> RHR suction line was redundant only for the purpose of reactor coolant
- system isolation, but that there was no apparent safety problem or
! deviation from regulatory requirements associated with this design. ,
.
'
l .
(j) It was alleged that nuclear plant problem roports (NPPR) were not getting
, management review which is a violation of administrative procedure
. i C-12 and that NPPR DC 1-81-OP P1057 had-been signed off after this
-
shortcoming was identified to management. Other NPPRs should be examined.
The Resident Inspectors observed that other NPPRs were being given
' appropriate management review and resolution. The allegation was not
'
substantiated.
t
(k) It was alleged that NPPRs DC0 79 TI P0006 and 79 TI P0117 are still
, i open after three years and should be closed. The Resident Inspectors
- observed that response to NPPR P0006 was complete and that response
-
to P0117 was underway. The allegation was substantiated, but no particular
.i safety or regulatory significance could be attached to this situation.
reactor
(1) Itcoolant
was alleged that a change to the Plant Manual Volume 16,to-"lo-hi
>
!
pump "lo oil level alarm should hav.a been changed i
'
.,
oil level" but had not been corrected eight months after the correction
had been submitted. The Resident Inspectors identified this alle ;
.~ to the licensee. The -licensee initiated a NPPR (Df!-83-TN-P0001)gati -
.and f
the problem is to be resolved.
were not previously aware of thisThe licensee personnet th~atsubstantiated.,
were intervie
'
problem. The allegation was
, The inspector conc-luded that the allegations were partially correct but
! c that these had n6 apparent safety significance or deviations from regulatory .
j .. requirements. .@
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The inspectors met with licensee representatives (denoted in paragraph 1)
and discussed the scope and findings of the inspection,
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- January 19, 1983
g4gLg'
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Docket No. 50-275
Pacific Gas and Electric Company ,
l
!
P. O. Box 7442 .
i
San Francisco, California 94106 -l
l )
- Attention: Mr. Philip A. Crane Jr. *
i
Assistant General Counsel I
Gentlemen:
This refers to the routine, monthly inspection conducted by Messrs. J. D. Carlson
i and M. M. Mendonca of this office, during the period of December 5,1982 through
i
January 1,1983, and the inspection effort of Mr. P. J. Morrill of this office ,
j on December 2,1982 at your corporate offices, of activities authorized by NRC l
License No. OPR-76, and to the discussions of our findings held by Messrs. Carlson j
'
and Mendonca with Mr. Thornberry and other members of your staff at the conclusion '
of the inspection. j
, Areas examined during this inspectioil are described in the enclosed inspection
report. Within these areas, the inspection consisted of selective examinations
. of procedures and representative records, interviews with personnel, and observations
j by the inspectors.
4
(
'~
No items of noncompliance with NRC requirenents were identified within the scope
of this inspection.
. In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will
j i be placed in the NRC Public Document Room unless you notify this office, by telephone, ,
! ' . within ten days of the date of this letter and submit written application to .
withhold information contained herein within thirty days of the date of this
l . letter. Such application must be consistent with the requirements of 2.790(b)(1),
t
-
Should you have any questions about this inspection, we will be glad to discuss ,
5 them with you. l
- .
Sincerely,
- i
i .
ish , hief - '
i
'
- -
Reactor Pr s Branch No. 2
i* Enclosure:
l NRC Inspection Report
!
No. 50-275/82-42
- cc w/o enclosure:
i J. L. Schuyler, PG&E
! J. D. Shiffer, PG&E ~ ~
,
W. S. Raymond, PG&E t
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RHR PUMP COMMON MODE FAILURE :!
,
e SINGLE SUCTION LINE WITH TWO ISOLATION VALVES IN SERIES,
WITH AUTOMATIC CLOSURE FEATURE RESULTS'IN INCREASED :-
POTENTIAL FOR LOSS-0F SUCTION PRESSURE T0'RHR PUMPS (0N 1
-
SPURIOUS AUT0 CLOSURE) ,.
j
. .- . . . . . . .
, . . . . , , .
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4
.
e- 'MOST OTHER PWRs HAVE AE SIMILAR RHR. SUCTION LINE . DESIGN
-
.
,
. . . .
!
_-_#....
-
e 'MANY INDICATIONS .TO. ALERT OPE 3AT_0R .T0 A LOSS OF SUCTION
j
'
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PRESSURE .
'
-
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RHR FLOW INDICATION ,
.
-
RHR DISCHARGE PRESSURE j
,!
.
-
-
RHR ISOLATION VALVE POSITION .
3
'
-
RHR PUMP MOTOR CURRENT :
.
e LICENSEE STATES THAT 15 MINUTES AVAILABLE BEFORE RHR PUMP (S) i-
WOULD BE DAMAGED
1
e USI A 145 IS ASSESSING THE REL1 ABILITY OF RHR SYSTEMS.,
INCLUDING ASSESSMENTS OF SINGLE VERSUS DOUBLE SUCTION
'
LINES AND AUTOCLOSU.RE INTERLOCKS ,
-
>
'
e AUT0 CLOSURE INTERLOCK IS IMPORTANT FOR PROTECTION FROM
EVENT V
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RHR SUCTION LINE DESIGN
.
-
.
.
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E e RHR COMMON SUCTION LINES ARE DESIGNED WITH-ISOLATION VALVES
s
, 8701 AND-8702 IN SERIES .
[ e BOTH SUCTION ISOLATION VALVES ARE LOCATED-INSIDE CONTAINMENT
.
-
i
- ,
i
e BOTH SUCTION. ISOLATION VALVES ARE CONTROLLED BY INTERLOCKS' i 1
. . . .
y .,f. . 37, 1 . .
.
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j '(A)-AUTOMATICCLOSUREONRCSl.; PRESSURE-IS.GREATERTHAN- p 4+
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(s) OPEN PERMISSIVE. INTERLOCK.TO ENSURE. ISOLATION
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VALVES CANNOT BE OPENED BEFORE RCS PRESSURE IS j
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-
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BELOW RHR DESIGN PRESSURE - i j
- . .
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(d)
.
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-e CURRENTLY, POWER IS REMOVED FROM RHR ISOLATION VALVES DURING *i
4 i
,
!; AUT0 CLOSURE INTERLOCK TO PROTECT FROM EVENT V :
-
-
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e STAFF WILL REQUIRE INSTALLATION 0F' LOW FLOW ALARM FOR RHR
'
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. PUMP PROTECTION AND REINSTATEMENT OF POWER TO RHR IS0'LATION
"
- VALVES FOR EVENT V PROTECTION
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. UNITS I AND 2
nrSIDUAL llE AT !!EMOVAL SYSTEM . - DI ABLO. CANYON SITE
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Allegation No. 37
'
h.
RV 83A41 BN No.: 83-169 (10/20/83)
Characterization
-
The solid state protection system (SSPS) relays that initiate closure of RHR
,
'
letdown isolation valves 8701 and 8702 perforn no safety function, reduce the
reliability of the RHR system, and cause a potential for RHR pisnp damage.
. Therefore, these relays should be removed.
Implied Significance to Plant Design, Construction, or Operation
The RHR letdom line contains two isolation valves (8701 and 8702) in series
that are normally closed during power operation. These valves are opened when
entering Mode 4 (hot shutdown) to allow the RHR pumps to take suction from the
reactor coolant system (RCS) to the RHR heat exchangers for decay heat removal.
(N Both valves 8701 and 8702 are interlocked so that they will automatically close
-
)
to isolate the RHR system from the RCS if RCS pressure increases to a pre-de-
,
termined setpoint. This automatic isolation fu'nction (performed by the West-
4
j inghouse designed SSPS) is provided to protect the low pressure RHR system
piping from higher RCS pressures. Isolation is accomplished using a " fail
l {
'
safe" design (i.e., on a loss of SSPS power, valves 8701 and/or 8702 will
automaticallyclose). The concern here is that a loss of SSPS power will
cause an unwanted (spurious) isolation of the RHR letdom line causing event-
ual RHR pump damage assuming no operator action.
'- -
. .. _. .
1
2-85
Y . . . .
~~
'
F0' A -T 4 -N + ' 4 A .
p u' '
7-5
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-_. - -- _- _ . _ -. . _ _ _ .
l
( 11 ,
Assessmint of Safety Significance
[I Isolation of the low pressure RHR system from the high pressure RCS must be
'
i
'
provided to prevent RHR system overpressurization that could potentially re-
,
I
sult in a loss of coolant accident (LOCA) outside containment. Therefore,
.
RHR letdow1 line isolation is a safety function. The SSPS, including relays,1
I l;
which performs this function is safety related and designed to Class IE re- l
ll quirements.
Both valves 8701 and 8702 are provided with this automatic clo-
d. ,
IF ,
sure interlock on increasing RCS pressure so that a single failure will not
'
.
j prevent RHR letdown line isolation. Therefore, the relays used to initiate
li e
'
T.
closure of these valves are essential and should not be removed.
!
l ;'
i Diverse indications and alarms are provided in the control room (including a
[ RHR system low flow alarm to be installed during the first refueling outage)
'
j}
i
to allow the operator (s) to assess RHR system status and to alert them to
h
potential system degradation.
Technical Specification surveillance require-
L
ments at Diablo Canyon include periodic verification of RHR system flowrate
l , when using the RHR letdown line.
,
' In addition, diverse means of decay heat
- removal (i.e., reactor coolant loops) can be readily made available should
!l'
the RHR letdown line be inadvertently / spuriously isolated.
b
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.
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2-86
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. .
. _ _ _ _
i ,
.
Based on the above, the staff concludes that the existing SSPS design regard-
ing RHR letdown line isolation is acceptable.
Staff Position i
This allegation does not involve considerations that question plant readiness !
l
for power ascension testing or full power operation.
.l
'
Action Required
4 None.
1
, .
,
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. _ _ . . _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ _ _ _ _ _ _ . _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _.__
_ _ .
_
,
, , Task: Allegation or Concern No. 38
~j l -
! N
'
i
ATS No. RV83A47 BN No.83-169 (10/20/83)
t
i
i
Characterization
J .
1 : ,
,
r
l PG&E is ignoring evidence that the spurious closure of a motor operated valve
is not " impossible."
I
d
R
'
'
-l , Implied Significance to plant Design, Construction or Ooeration
l
1
-
'i '
l
The allegation suggests the licensee has not satisfactorily analyzed
il'
operational data.
{ t
-I ~
j ,
Assessment of Safety Significance
l! i'
,
J
J ,
'
The alleger has described operating events at the Otablo Canyon facility and
l i
other Westinghouse facilities during which motor operated valves in the
i
'
residual heat removal (RHR) system have, upon spurious initiation of their
l
l
automatic closure circuitry, moved from the normally open position (for RHR
operation) to the closed position, these presenting the potential for damage to
RHR eumps.
l .
l
.
The staff has examined in denth the licensee's actions in resoonse to an event ) i
l
invnivino the sourious initiation of RHR motor coerated valve closura as Wil as
the concerns expressed hv tha allener renardi,,g the potential for such event,
and concluded that timely evaluation and corrective measures yere. taken to preclude
\? '
.
i . -
lI
I :" 2-88
jt -
il
-l
- !
. _ _ - _ _
. _ . _ _ _ .. _ . _ _ .
.
repetition of such conditions. (See Allegation or Concern Nos.: 42 & 44).
-
Staff Position !
'
\
\
f ) The staff's position regarding the interlock cricurity which causes automatic closure l
(
of the RHR isolation valves is duscussed in Allegation or Concern No. 45. It
does appear that the licensee is giving proper attention to the spurious closure
of the valves in question.
l
l
I C .,
- / ,
I
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_
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i 2-89
-
.
!
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. .
. - - - -. __._ .
. , _ . . . - - . - _ - , . _ _ . , _ , - , _ . , , .___, y_ _ _ _ , _ _ _ _ _ - _ . , . . _ .
- - -- , - - -_ - . , -,. . - . , , . - . - - . . _ . ,
j' .
-
,' Task: Allegation #39
1
'
ATS No.: -
RV 83A47
BN No: 83-169 (10/20/83)
i~ -
) >
Characterization
! ! i
'
There is no control room annunciation provided to alert the operator (s) when
, the RHR letdowl line has been isolated during Modes 4, 5, and.6 (hot shutdown,
,; i cold shutdow1, and refueling respectively). .
I
!!
~
..,
. ,!
t
Implied Significance to Plant Design, Construction, or Operation
- r
..
.j
During modes 4, 5, and 6 the residual heat removal (RHR) system is aligned in
[ !
the shutdown cooling mode by taking suction from reactor coolant system (RCS)
!
? !
'
, ,
.
' l}
flow through the RHR heat exchangers for decay heat removal via the component
cooling water (CCW) system, and then back to the RCS cold legs. There are two
f(
'
'
s
isolation valves (8701 and 8702) in series located in the RHR letdow line.
I
! [' If one of these valves should inadvertently close, RHR pump suction would be
'
lost. The, concerns here are loss of decay heat removal capability and poten-
. tial damage to the RHR pumps. It has been estimated that pump damage could
[I! occur as soon as 10 to 15 minutes following a spurious isolation of the RHR
!
I
'
letdown line.
!
.
I !
+
6
,
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,
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_= - _ . ._ __ _ _ _.
. .. _ - . -. ______ __ __
-
q .. - .
.
- Assessmint of Safety Significance
For those modes of operation bere RHR shutdown cooling is used, only one RHR
train or one filled reactor coolant loop is necessary to provide sufficient
~ decay heat removal capability. The Diablo Canyon Technical Specifications
i require either two'RHR trains be operable and/or two filled reactor coolant
'
"
loops be available in order to allnw for single fat. lures. If both RHR trains
are being used and the RHR letdown line becomes isolated, the operator (s)- i
would have sufficient time to fill at least one coolant loop (as'suming no
loops are filled) for decay heat renoval. Control room indications of loss
.
I, of decay heat removal include RCS temperature RHR system flow, and RHR pump
- !
discharge pressure. With less than the required number of reactor coolant
jj
i
loops and/or RHR trains operable, the Technical Speci.fications require im- !
mediate corrective actions to return the required loop / train to operable sta-
%
i! . . ,
tus as soon as possible,
i' . 'i [\ ,/
Indication provided in the control room of RHR letdown line isolation in- l
- .
$i cludes position indication for valves 8701 and 8702 (red and green position
li
! status lights next to the valve control switches on the main control board)
as well as RHR system flow, pressure, and pump status information. Although
!
these features do provide a capability to assess RHR system status, the staff
has recognized the need for installation of a RHR low flow alarm. Accordingly,
1
- .
1
e e ne s .
o
-
- .
! ..
.
2-9.1
1 .
!
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.
e
,
'
, ,._.-,n--
f ,
I the licenseo is required to install a RHR low flow alarm during thO first re-
>!
.
l
p .
fueling. This requirement is documented in Supplement No.13 of NUREG-0675,
, .-
" Safety Evaluation Report related to the operation of Diablo Canyon Nuclear
'
{ Po w r Plant, Units 1 and 2." The staff has concluded that the existing con-
trol room indications and procedures are sufficient to assure adequate decay
'
heat removal in the interim.
t
- i
-
5 Staff position
6
s
This allegation does not involve considerations that question plant readiness
for power ascension testing or full powr operation.
l Action Required
e i
- I j None.
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Reactor Systems Branch
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Task: Allegation #40
ATS No.: RV83A 47 BN No.: 83-169 (10/20/83)
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1
1
Characterization
The question raised was with regard to whether or not the single RHR
> pump suction line from the RCS hot leg meets safety related standards.
.
l The newer PWRs are designed with redundant RHR pump suction lines from
I
the RCS hot legs.
Implied Significance.to Plant Design, Construction or Operation
The RHR suction line from the RCS hot leg in Diablo Canyon contains two
/
'
i isolation valves (8701 and 8702) in series that are normally closed
.
)
during power operation. When the RHR system is operated as a part of
the ECCS, the RHR pump suctions are aligned with either the RWST or the
containment emergency sumps. The RHR suction line from the RCS hot leg
is only used during modes 4 (hot shutdown while RCS temperature is less
than 323'F), 5 (cold shutdown) and 6 (refueling). A postulated failure
of either isolation valve (8701 or 8702) in the RHR suction line to open
during plant shutdown could prevent the plant from reaching a cold
shutdown condition. I
Assessment of Safety Significance
In the Diablo Canyon SER Supplement No. 7, the staff states that the
single RHR suction line from the RCS hot leg was acceptable. The staff
l
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conclusion was based on the followingi
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2-93 . l
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. ___ . . _ _ _ _ . . _ _ _ _ _ _ .. -.__ _.._ _ ___
.
j (1)
N' The Diablo Canyon design has a safe' ty related Auxiliary Fee
9 )
,1 '
System (AFWS).'
The condensate storage tank is the primary source
'
of AFW with about an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> water supply. - In order to ensure th
!
capability to remove heat via the steam generators for extended.
i
periods, provisions have been made to connect the raw water
' reservior to the suction line or the AFW pump. This will provide
!
enough AFW to allo. an additional 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of steam generato.r
operation for both units.
.
(2)
i
i The licensee has indicated that the combination of a
- failure of the
4 RHR isolation valves and an earthquake results in a
1
! risk of about 10% of the core melt risk from all caus d
i,
in the Reactor Safety Study. :
!
.
Branch Technical Position RSB 5-1 was not approved at .7time SSE
for Diablo Canyon was issued.
~ In accordance with the implementation
i
schedule of BTP R$8 5-1, the Diablo Canyon Units are conside
,
plants which are not required to fully implement this BTP.
Table 1 c?
[
BTP R58 5-1 shows what is necessary to be implemented for clas r
"
plants. !
A single RHR suction line from the RCS hot leg is considered i
i
acceptable for a class 2 plant as long as a single failure could be \
\ enrrected by manual actions
inside or outside of containment, or the
! >
,,
! plant could be returned to hot standby until manual actions ,
4
are accomplished. (page 5.4.7-16 of SRP 5.4.7). Also, BTP RS6' 5-1 for
i
- class 2 plants requires that the RHR isolation valves
i
!
diverse interlocks to protect against one or both valves being op
1
during an RCS pressure increase above the design pressure of th
9
9
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system. There was no assessment of the degree of compliance of the ,_
.
Diablo Canyon design against BTP RSB 5-1 documented in any staff SSER.
.'
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l
Based on the above facts, the staff evaluation of the subject allegation :
I
is as follows:
The RHR suction line from the RCS hot leg is not required for ECCS
f
! functionability. The RHR pumps take suction from RWST or containment
emergency sumps, and serve the ECCS function during a LOCA. The suction
i
line from RCS hot leg is used only for modes 4 ( 323'F),5and6. GDC 34 of Appendix A to 10CFR 50 requires that the decay heat removal safety
function should be accomplished assuming a single failure. THe Diablo
Canyon design complies with this requirement by having a RHR system plus
a safety related AFWs (with steam generators and atmospheric steam dump
valves). Based on the above, we conclude that the Diablo Canyon design
meets GDC 34 and the intent of BTP RSB 5-1. The current RHR design is
[ -
"
adequate for safe operation at Diablo Canyon.
The staff is currently conducting a reevaluation of the adequacy of the
decay heat removal system design of all LWRs. This work is being
i performed as an Unresolved Safety Issue (TAP-A-45), and the Task Action
Plan is projected to be complete within one year. Diablo Canyon, will
,
be subject to any new requirements that may result from the work of TAP
A-45.
Staff Position
This allegation does not involve considerations that question plant
s readiness for power ascension testing or full power operation,. , _,
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EMERGENCY CORE C00 LING SYSTEMS
1
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3/4.5.3 ECCS SUBSYSTEMS - T,y less than 350*F
'
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LIMITING CONDITION FOR OPERATION
3.5.3 As a minimum, one ECCS subsystem comprised,of the following shall be
. OPERABLE:
a. One OPERABLE centrifugal charging pump,#
' ~
b. One OPERABLE residual heat removal heat exchanger,
c. One OPERABLE residual heat removal pump, and .
,
d. An OPERABLE flow path capable of taking suction from the refueling
water storage tank upon being manually realigned and transferring
suction to the containment sump during the recirculation phase of
operation.
APPLICABILITY: MODE 4.'
ACTION:
a. With no ECCS subsystem OPERABLE because of the inoperability of
either the centrifugal charging pump or the flow path from the
I
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refueling water storage tank, restore at least one ECCS subsystem to
(
.
20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the next
,
b. With no ECCS subsystem OPERABLE because of the inoperability of
either the. residual heat removal heat exchanger or residual heat
removal pump, restore at least one ECCS subsystem to OPERABLE status
or maintain the Reactor Coolant System T"V9 less than 350*F by use-
of alternate heat removal methods.
c. In the event the ECCS is actuated and injects water into the Reactor
Coolant System, a Special Report shall be prepared and submitted to !
the Commission pursuant to Specification 6.9.2 within 90 days describing l
the circumstances of the actuation and the total accumulated actuation !
cycles to date. The current value of the usage factor for each -!
affected safety injection nozzle shall be provided in this Special
Report whenever its value exceeds 0.70.
l
IFA maximum of one centrifugal charging pump shall be OPERABLE'whenever-the
temperature of one or more of the RCS cold Tegs is less than or equal to 323*F.
- ^~
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C DIABLO CANYON - UNIT 1 3/4 5-7
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' 3/4. 5 EMERGENCY CORE COOLING SYSTEMS
.
,
BASES
1
3/4.5.1 ACCUMULATORS _ 4
_
The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that
'
a sufficient volume of borated water will be immediately forced into the reactor
core through each of the cold legs in the event the RCS pressure falls below' '
the pressure of the accumulators. This initial surge of water into the . core
,
provides the initial cooling mechanism during large RCS pipe ruptures.
The limits on accumulator volume, boron concentration and pressure ensure that
the assumptions used for accumulator injection in the safety analysis are met.
- The accumulator power operated isolation valves are considered to be " operating
i-
bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses
of a protective function be removed automatically whenever permissive conditions
are not met. In addition, as these accumulator isolation valves fail to meet
single failure criteria, removal of power to the valves is required.
4
The limits for operation with an accumulator inoperable for any reason except
,
' .-- an isolation valve closed minimizes the time exposure of the plant to a LOCA ~
,
[ ) event occurring concurrent with failure of an additional accumulator which may
's
i
result in unacceptable peak cladding temperatures. If a closed isolation
-
l valve cannot be immediately opened, the full capability of one accumulator is
not available and prompt action is required to place the reactor in a mode
.
where this capability is not required.
3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS '
The OPERABILITY of two ECCS subsystems ensures that sufficient emergency core
cooling capability will be available in the event of a LOCA assuming the loss
of one subsystem through any single failure consideration. Either subsystem
operating in conjunction with the accumulators is capable of supplying sufficient
core cooling to limit the peak cladding temperatures within acceptable limits
for all postulated break sizes ranging from the double ended break of the
largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides
long term core cooling capability in the recirculation mode during the accident-
recovery period.
'
With the RCS temperature below 350*F, one OPERABLE ECCS- subsystem is acceptable j
without single failure consideration on the basis of the stable reactivity t
condition of the reactor and the limited core cooling requirements. l
..
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DIABLO CANYON - UNIT 1 - -- -
B 3/_4 5-1
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EMERGENCYCOREC0kINGSYSTEMS
7 .. .
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BASES
e
.
- ECCS' SUBSYSTEMS (Continued)
, .The Surveillance. Requirements provided >to ensure '0PERABILITY of each component- l
4
ensures that, at a minimum, the assumptions used in the safety analyses.are !
met and that subsystem OPERABILITY is-maintained. Surveillance requirements l
for throttle valve position stops and flow balance testing provide assurance . !
that proper ECCS flows will be maintained in the event of a LOCA. Maintenance 1
of proper flow resistance and pressure drop in.the piping system to each- -)
injection point is necessary to: (1) prevent total pump flow from exceeding .l
runout' conditions when.the system is in its minimum resistance configuration, '
l
(2) provide the proper flow split between injection points in accordance with !
-
t
'
the-assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable
level of total ECCS flow to all injection points. equal. to or above that assumed ;
j
. in~the-ECCS-LOCA analyses.
The requirement to maintain the RHR Suction Valves (8701 and 8702) in the.
!
locked closed condition in MODES 1, 2 and 3 provides assurance that a fire
could not cause inadvertent opening of these valves when the RCS is pressur -
ized to near operating pressure. 'These valves are not part of an ECCS subsystem.
~
'
The limitation for a maximum of one centrifugal charging pump.to be OPERABLE .
.-
and the Surveillance Requirement to verify all centrifugal charging pumps and
!
,
() ,
safety injection pumgs except the required 0PERABLE charg ng pump to be
inoperable below 323.F. provides assurance that.a mass add tion pressure transient
~
can be relieved by the operation of a single PORV.'
3/4.5.4 BORON INJECTION SYSTEM
The OPERABILITY of the boron injection system as part of the ECCS ensures that
' sufficient negative reactivity is injected into the core to counteract any
positive increase in reactivity caused by RCS system cooldown. RCS cooldown
can be caused by inadvertent depressurization, a loss-of-coolant accident or a-
steam line rupture.
>
The limits on injection tank minimum contained volume and boron concentration
,
ensure that the assumptions used in the steam line break analysis are met.
The contained water volume limit includes an allowance' for water not usable
i
because of tank discharge line location or other physical characteristics.
The OPERABILITY of the redundant heat tracing channels associated with the
boron injection system ensure that the solubility of the boron solution'will
be maintained above.the solubility limit of.135'F at 21,000 ppe boron.
,
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DIABLO CANYON - UNIT -1--- 8 3/4 5-2
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, HESIDUAL HEAT [
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NOTE:
, 1 Olt del AILS , REFER - TO , FIGURE ' 3. 2 - 10, PIPING SCllEMATIC, l'MTS IMD2
RESIDUAL llEAT IIEMOVAL SYSTEM ;
4 DW WW SITE
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RESIDUAL. IIEAT REMOVAI. SYSTEH
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RHR SUCTION LINE DESIGN
.
i
e RHR COMMON SUCTION LINES ARE DESIGNED WITH ISOLATION VALVES
8701 AND 8702 IN SERIES
e BOTH SUCTION ISOLATI0ft VALVES ARE LOCATED INSIDE C0f1TAINMENT
e BOTH SUCTION ISOLATION VALVES ARE CONTROLLED BY INTERLOCKS
(A) AUTOMATIC CLOSURE ON RCS PRESSURE IS GREATER THAN
600 PSIG TO PREVENT SBLOCA OUTSIDE CONTAINMENT
(EVENT V)
(s) OPEN PERMISSIVE INTERLOCK TO ENSURE ISOLATION
VALVES CANNOT BE OPENED "EFORE RCS PRESSURE IS
( ,'
, BELOW RHR DESIGli PRESSURE S d g M 4aK (fif
a CURRENTLY, POWER IS REMOVED FROM RHR ISOLATION VALVES DURING
LOW RCS TEMPERATURE / PRESSURE OPERATIONS. RESULT IS NO
AUT0 CLOSURE INTERLOCK TO PROTECT FROM EVENT V
e STAFF WILL REQUIRE INSTALLATION OF LOW FLOW ALARM.FOR RHR
PUMP PROTECTION AND REINSTATEMENT OF POWER TO RHR ISOLATION
VALVES FOR EVENT V PROTECTION
. . . - . . .
.
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._
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.
_
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..
.
(- RHR PUMP COMM0tl MODE FAILURE -
.
e SINGLE SUCTION LINE WITH TWO. ISOLATION VALVES IN SERIES,
WITH AUTOMATIC CLOSURE FEATURE RESULTS-IN INCREASED
POTENTIAL FOR LOSS OF SUCTION PRESSURE TO RHR PUMPS (0tl
SPURIOUS AUT0 CLOSURE)
e
MOST OTHER PWRs HAVE A SIMILAR RHR SUCTION LINE DESIGN
e
MANY INDICATIONS TO ALERT OPERATOR TO A LOSS-0F SUCTION
PRESSURE
-
-
RHR FLOW INDICATION
(: -
-
RHR DISCHARGE PRESSURE (ND(CAT 1od
RHR ISOLATI0ll VALVE POSITI0ll
-
RHR PUMP MOTOR CURRENT
e
LICENSEE STATES THAT 15 MINUTES AVAILABLE BEFORE RHR PUMP (
WOULD BE DAMAGED *'W~M ' D I ' . / [ " " * F " '" ' #
e USI A-45 IS ASSESSING THE RELIABILITY OF RHR SYSTEMS,
INCLUDING ASSESSMENTS OF SINGLE VERSUS DOUBLE SUCTION
LINES AND AUT0 CLOSURE INTERLOCKS
e
AUT0 CLOSURE INTERLOCK IS IMPORTANT FOR PROTECTION FROM
EVENT V
.
--
, . . _ . - . . .
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- TASK
- Allegation 45
(Previously addressed in SSER 21)
.
ATS NO.: RV 83A47 BN NO.: 83-169 (10/20/83)
l Characterization:
Section 5.5 of the Diablo Canyon FSAR describes the autoclosure interlock
for the RHR suction line isolation valves (8701 and 8702). Section 3.4.9.3.a
f of the Diablo Canyon Technical Specifications requries power to be removed
from these isolation valve operators during Mnde 4 (hot shutdown, RCS cold
.
/" leg temperature is less than 323*F), Mode 5 (cold shutdown) and Mode 6
('re fueli ng ) . This requirement defeats the function of autoclosure interlock
for the valves.
! Related Allegations: 37, 39, 40, 177
Implied Significance to Plant Desion, Construction or Operation
As stated in SSER 21, as the result of Technical Specification Section 3.4.9.3.a.
the isolation valves will be left in an open position with power removed during
L low p'ressure/ temperature operation of the plant. The automatic closure inter-
lock to these isolation valves causes them to lose their design fuction. This
, will result in a situation in which insufficient isolation capahiiity exists to' ,'
\ . s/ prevent an intersystem LOCA between high pressure RCS and the low pressure
"
RHR system.
Foi6-?4-14(F H W -
A.4-45.1 !
yLi--
l
--- _ _ _ - _ _
..
-
.
.
.
('
.
-2-
Assessment of Safety Significance
'
.
As stated in SSER 21, the staff concluded in Diablo Canyon SSER 13 that
the licensee should be required to provide an alarm to alert the operator to
a degradation in ECCS during long term recirculation. A low flow alarm was
stated to be an acceptable method to satisfy this concern and the staff
indicated that an alarm should be installed at the first refueling outage.
Until then, procedures and dedicated operators were to be implemented during
long term recirculation to manage and monitor ECCS performance.
.
/ ~
Staff Position
s
.
As stated previously in SSER 21, to implement the staff position stated in
SSER 13, the installation of a low flow alarm for RHR pump protection 'is
being considered as a license condition in the Diablo Canyon full power
license. Additionally, it is the staff position that power be available
to the RHR MOVs when in a shutdown condition. However, there is a question
as to when these requirements should be implemented. If the low flow alarm
,
were not installed until the first refueling outage, reinstating power to the
'
RHR MOVs in the meantime would res' ult in the autoclosure interlock being
anable to provide protection against intersystem LOCA.
~ ~
. . . .. - . .
..s'
1:- -
a
-
A.4-45.2
, _ _ _ . - - - . . -
7
-
. ..
. x, ,
-3- ,P
,.
- -
,t
h
i ; !.....
4 :*
Ir. a letter dated February 15, 1984 the licensee committed to install the RHR
low flow alarm prior to entering Mode 1, i.e. operation above 5 percent power.
.[
The licensee also.provided the administrative controls and procedures that are
now in effect. Based on the committment the staff finds these controls and
_
"
procedures acceptable for the interim, i.e. until installation of the alarm.
.
The staff concludes that this issue is resolved with regard to criticality and
.
low power operation.
Action Peouired
The staff requries that the low flow alarm be installed prior to entering Mode 1
C nd that the licensee advise the staff of the completion of the installation
i
prior to Pode 1.
.
. O ese e
\
.)
1 -~
_
,
A.4-45.3 .
. . _ . . - . . _ _e
r- -
- . . - _
.
. ..
'
Reactor Systems Branch
T- TASK: Allegation #45
4
ATS NO.: RV 83A47 BN NO.: 83-169 (10/20/83)
4
Characterization:
Section 5.5. of the Diablo Canyon FSAR describes- the autoclosure inter-
lock for the.RHR Suction line isolation ~ valves (8701 abd 8702).
Section 3.4.9.3.a of the Diablo Canyon Technical Specifications requires
power to be removed from these isolation valve operators during modes 4
(Hot shutdown when RCS cold leg temperature .is less than 323*F), 5 (cold
shutdown) and 6 (refueling). This requirement defeats the function of -
autoclosure interlock for the valves.
- Implied. Significance to Plant Design, Constru'ction or Operation
,
As the result of Technical Specification'Section 3.4.9.3.a, the iso-
O 1ation va'1ves (8701 and 8702) will be 1tft in.an open position with
'
,
power removed during low pressure / temperature operation of the plant.
[The automatic closure interlock to these isolation valves causes them to
s lose their design function. This will-result in a situation in which
there is inhufficient isolation capability '* ture, to' prevent an
intersystem LOCA between the high pressure RCS and the low pressure RHR
system.
.
.
Assessment of Safety Significance j
i Section 5.5 of the Diablo Canyon FSAR states that during low pres-
surk/ temperature operation, the isolation valves (8701 and 8702) between;
the RCS and the suction of the RHR pumps are interlocked with a pressure i
i
. signal to. automatically close the valves whenever the RCS pr: essure.. - l
9 i ,
d increase above approximately 600 psig. Section 3.4.9.3.a of the Diablo '
a . . . _
,,
Canyon Technical SpeUfication requi'res the RHR system isolation valves
, 2 109 "
.
-e.
---,, ,4- - , - , , - . , .-..,,...,-,.i,. ,, , . -
_
4
! . . .
.
.
.
(8701 and 8702) to-be open with power removed from the valve-operators
-
" '
' 'y - -
while the positive displacement charging pump is in operation. :The
'
applicability of the T.S. is during mode 4 when the temperature-of any
RCS cold leg is less than or equal to 323*F, mode 5, or mode 6 with the
reactor vessel head on this Technical ~ Specification requirement defeats
.
the automatic closure interlock function as designed.
..
_
Power removal from valves 8701 and 8702 while the RHR system is operat-
ing was required by the staff as the result of a meeting with the
licensees on RCS low temperature overpressure protection (LTOP) and RHR
pump protection concerns. Since the Diablo Canyon -iesign has only one
RHR suction line from the RCS, a spurious automatic closure of the
isolation valve would result in loss of RHR pump suction flow and would
,m result in a RCS pressurization as a result of the loss of letdown; flow.
-
q However, there was no documentation (SS'ER, letter or meeting minutes) of
the staff's basis for requiring power removal from those isolation
valves during modes 4, 5 and 6.
.
In the Diablo Canyon SER Supplement No.13, section 6.3. (ECCS), dated
April 2,1981, the staff concluded that the licensee should be required
to provide an alarm to alert the operator to a degradation in ECCS
(during long term recirculation).
A low flow alarm was stated to be an
acceptable method to satisf( this concern and ~ the staff indicated that
an alarm should be installed at the first refueling outage. Until then,
procedures and dedicated operators were to be implemented during long
-.
term recirculation to manage and monitor ECCS performance., There was no
.
'
._l documentation to indicate that the licensee committed to this
T'~. 2
. . .
. 2- 110
_
.
.
'
staff position, nor was this staff position included in the Diablo
Canyon low power license. SRP 5.4.7 (BTP RSB 5-1) requires an
autoclosure interlock on the RHR suction line isolation valves. Without
cwer to the valve operators, the autoclosure function is defeated.
Based on the a'bove facts, the staff evaluation of the subiecLallecatinn
-
-
-n-as 414cwsr
udtt
3, UE tt 2-l $ 5%2: S.hb $.k
Without power to the isolation valve operators, the plant design does
not conform to BTP RSB 5-1, Position B.1.C, for the requirement of f <]
1
'
autoclosure interlock. By having power available to the isolation
valves during shutdowns ensures an event V (intersystem LOCA) will not -
cccur as a result of the operator failing to close both isolatien valves
during a return to power.
A
,
(. 3
QJ
With power on the isolation valves, a spurious closure of the isolation
valves would result in a loss of suction flow to the RHR pumps. Howev-
er, the, low flow alarm #ccu:::J m 333 lw._19 would enable rapid
operator detection and mitigation. ' 'i;er.;cc % 44rmally indicat-
.ed that : minimu of- 10 minutes- without-adequate.. suction _ pressure would -
be availahlo44-thout pump-damager--Also,-there are numerous indications
ayaMabic tc alert the ope"ter +n imnenner op ::he eliginumt ( A list-
i-6cevided .. de f f cvah tions-to--a4+egat4en-%-37-and-39).
i
Staff Position
To implement the staff position stated in SSER No. 13, the installation
of a low flow alarm for RHR pum[ protection is being considered 'ss a ~
~
J l
.-- ;
1
2-111 __; l
-
' ' " -, _m
. .. . .. . - . -
, . . a.
.
,
,
license condition' in' the Diablo Canyon full power license.
( Additionally, it is the staff position that power be available to the
RHR MOVs when in a shutdown condition. However, there is a question as '
to when these requirements should be. implemented. 'If the low flow-
-
alarm were not installed until the first refueling outage, reinstating
'
power to the RHR MOVs in the meantime would result in the autoclosure
interlock being enabled to provide protection against intersystem LOCA.
. However, 'the chances of spurious autoclosure ~and-consequent loss
of RHR suction pressure (without the low flow alarm) and of an
overpressure event would be increased. If power restoration to the RHR
MOVs were not implemented until the low flow alarm is installed at the
first refueling outage, the' chance of loss of RHR suction in.the interim' -
is reduced but there is a possibility of an intersystem LOCA. To
,s determine which option results in the safest operation of the plant, the -
staff considered the following:
4
1.
During the first cycle of operation, plants operate more frequently
on the RHR system as a result of maintenance, testing and training
requirements for a new plant. Thus,~the period of. vulnerability.to
a
spurious RHR suction MOV closure may be greater than in subse-
quent cycles.
2. The RHR relief valve would open to relieve pressure if a plant =
startup were attempted with both RHR MOVs open. ~ It is not, in the -
staff's judgment, credible to postulate plant startups with both
MOVs left open. The operator would have to shut at least one MOV
to continue the plant startup.
3.
Failing to close the second RHR suction MOV would n,ot, in itself,
, result in an intersystem LOCA. The first MOV must also fail . The
-
c
'
'
L- 1
2 112
i-
- - - . . . .
.,-m,.-,- m e yry y - - - e.--.- -i -. -- ---#-- -.
.. . _ - _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _
. .
,
first MOV can fail in either of two ways by- either the "open ~
-
pemissive" interlock failing along with the operator reinstating
/ power to the valve, (it is required to be de-energized) then
attempting to open the valve. The second mode.cf failure would be
for the valve to rupture in .such a way that flow between the two-
l systems' occurred. Both of these failure modes are judged to
1
i have an extremely low probability. However, the consequences of an
! intersystem LOCA could be severe.
- 4. There have been many occasions ;of spurious RHR suction valve
closures on operating plants. This has resulted in not only a loss
- of decay heat removal, but also an overpressure event due to the
loss of the letdown flowpath.
ACTION REOUIRED
f) '
, , .
Based on the above factors, the staff believes. the_ best course of action
is to continue the current technical specification for power to be
removed from the RHR MOVs during Modes 4,_5 and 6 until the low flow
alarm is installed. However, the staff position that would permit the
!
licensee to wait until the first refueling outage before installing the
low flow alarm was taken over two years ago. Staff will puruse with
the licensee a comitment to a schedule for accomplishing this
installation at the earliest possible time. In the interim, until the-
low, flow alarm is installed, the staff believes that strict
administrative controls should be developed and implemented to ensure
that MOVS 8701 and 8702 are closed with power renoveo during plant-
startups when RCS pressure is above the RHR design pressure. -
= = -
"~
2 113 _ :: -
,
e
6 .ome 4 o-o w- w = emww m~ w _
. _ . . - . .. _ _ . .
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SPRAY HEADER n o
- ,, , i 8 *
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r an I e
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.' COLD LEG ' '
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REslDUAL HEAT i
u ng - ~1 RESIDUAL HEAT "
'
2101
EXCHANGERS FCVs )b
,
l REMOVAL PUMPS a Q
- -
= ,
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NOTE: '
, ,
FOR DETAILS , REFER TO FIGURE 3. 2 - 10, PIPING S CHEMATIC, UNITS I AND 2
, % RESIDUAL HEAT REMOVAL SYSTEM . DI ABLO CANYON SITE
, - - -
-
FIGURE 5.5-6
RESIDUAI. IIEAT REMOVAL, SYSTEtt
'); '
i
... .. .
_ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - - _ _ . _ _ _ _ _ _ _ _ _ - _-- -- - __ __ -
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TABLE-5.5-8
DESIGN BASES FOR RESIDUAL HEAT REMOVAL SYSTEM OPERATION (BOTH UNITS)
Residual Heat Removal System start up. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after
Reactor shutdown
Reactor Coolant System initial pressure, psig s425
Reactor Coolant System initial temperature, 'F N350
"
Isolation valve interlock setpoint, psig 425
Component cooling water design temperature. *F 95
Cooldown time, hours af ter initiation of RHRS operation - sl6
-
%
-
} Reactor Coolant System temperature at end of
s ..- cooldown, 'F 140
.
Decay heat generation at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after Reactor 70.6 x 106 (Unit'l)
shutdown, BTU /hr 72.1 x 100 (Unit 2)
,
a
~ -
. . . .. . . .
s
.
- - :
l
- *
5*
2
+w -ma n= = on.* = m m w . -e --m.mm.w.._, .. ,_, ,
-
, _ m _ _ . _ . . . __ __ _. - _ _ _ _ _
- . .- -- .. -..--- _ .. . . _ ..
,-- -
..
. . !
l
l
j. - 7.6.2 RESIDUALHEATREMOVAk,ISOLATIONVALVES
-'- ;
-t ,
-:
Description
3
There are two motor operated gate valves in series in the inlet line from the
Reactor Coolant System to the Residual Heat Removal System. They are normally -
closed and are only opened for residual heat renoval after system pressure is
reduced below approximately 400 psig and system temperature has been reduced to -1
approximately 350*F. (See Chapter 5 for details of the Residual' Heat Removal
System). They are the same type of valve and motor operator as those used for-
accumulator isolation, but they differ in their controls and indications in the
1
following respect:
'
l. One isolation valve, that nearest the Reactor Coolant System, is inter-
'
locked with a pressure signal to prevent its being opened whenever the
system pressure is greater than 425 psig. The valve will also be closed '
!
automatically whenever the system pressure increases above approximately
l 600 psig. This interlock and automatic. closing action is derived from
( ,
one process control channel.
.
2. The other valve, that nearest the Residual Heat Removal System, is
similarly interlocked and automatically controlled. Control signals are
derived from a second process control channel. In order to comply with
IEE-279 and to provide diversity, this valve will also be prevented from
4
opening when the pressurizer vapor space temperature exceeds approximately
'
455'F and automatically closed when the pressurizer vapor space temperature
exceeds approximately 490*F. This temperature control signal is derived
from one process instrumentation protection channel.
1
' Analysis -
1
l Based on the scope definitions presented in Reference 2-(IEEE-279),1971) and
j Reference 3 (IEEE-338, 1971), these criteria do not apply to the residual heat-
removal isolation valve interlocks; however, in order to meet AEC requirements,
and because of the possible severity of the consequences. of los~f oT"fhnetion,
'
(, / the requirements of IEEE-279 vill be applied with the following comments.
, . ..
(February 1974)' ,
Amendment 4 a
!
-
.
- emoe - anw o . oew . * *** esse =+=* * * * * = - .===*4
e --- -- H 4 cc-t-% v m- -n- ----my-+---< e-e+ -- v- e--> ---- - v -ww i i-vyw w wwrw--w --
-+eer-4-
-
-vvv'i --**--1-w-----w+ y--- w--
_ _ _ _ _ ._. .
_ _.
4
.
,
.
- .
~
p
' 1.- For the purpose of applying IEEE-279,1971,- to this circuit, the
'
following definitions will be used.
a.- Protection System
. The two valves in series in each line and all components of
- their interlocking and closure circuits.
1 *
~
b. Protective Action
'
a
The automatic initiation and maintenance of' Residual Heat
Removal System isolation from the Reactor Coolant System
pressures above redidual heat removal design pressure.
.
2. IEEE-279,-Paragraph 4.10: The requirement for on-line test and '
calibration capability is applicable only to the actuation signal
[ and not to the isolation valves, which are required to remain
closed during power operation. .
.
3. IEEE-279, Paragraph 4.15: This requirement does not apply, as
4 the setpoints are independent of mode of operation and are not
changed.
'
Environmental qualification of the valves and wiring are discussed in
Section 3.11.
7.6.3 REFUELING INTERMCKS
-
Electrical interlocks (i.e., limit switches) aye1rovided for minimi-
i
sing the possibility of damage to the fuel during fuel handling
operations. Mechanical stops are- ded as the primary means of ,
preventing fuel handlin a ts . For example, safety aspects of the
'
manipulator crane on the use of electrical interlocks 'and " " '
'
s . .
-
, =- :
-.
i
I
'
- 7.6-4
.
- &&* * * s **40 w - +p q....*1we - ,m4
y -- -
+-- ,-n-vw> -
9' v' y
-. -. . . . . . . .- .-- -- .- -.- __
_
~
' .
.
f) '
- ijQ
.
e
.
.
A 5.5.6 REACTOR CORE ISO'LATION COOLING SYSTEM
}.
This system, which is referred to as anj.xem61e in the_AEC's " Standard Format
and Content of Safety Analysi orts for Nuclear Power Plants", dated
i February, 1972, is apolicable to Westinghouse pressurized water reactor
,
desi
, .
4
' 5 . 5 '. 7 . RESIDUAL HEAT REMOVAL SYSTEM
'
A' separate Residual Heat Removal System is provided for each unit. This
section describes one system with the second being identical unless noted
otherwise. There is no sharing of this system between the two units.
'
'The Residual Heat Removal System transfers heat from the Reactor Coolant
'
. System to the Component Cooling Water System to reduce the temoerature
,
Kh .:D y
)
iof the reactor coolant to the cold shutdown temperature at a controlled
rate during the latter part of normal plant cooldown and maintains this
]
-
i temperature until the plant is started up again.
(x-
-
j
As a secondary function, the Residual Heat Removal-System also serves as
4
part of the T.mergency Core Cooling System during the injection and recirculation
I phases of a loss of coolant accident.
i
1
The Residual Heat Removal System also is used to transfer refueling water
- between the refueling water storage tank and the refueling cavity before
i
and after the refueling operations.
4
Design Bases
4
.
, Residual Heat Removal System design parameters ere listed in Table 5.5-8.
!
i The Residual Heat Removal System is designed to remove heat from the core
and reduce the temperature of the Reactor Coolant System during the second
-
phase of plant cooldown. During "the first phase of cooldown, the temperature
'
} >
i (d .
u* ' : ,
.
5.5-29 - ~ -
-
..
1
.,
I
j.
. . . . . . ._ _ _ _ . , , . _ ,,.
__ . .
. _ . - - - ;
_ _ - - _ _ _ _ - . _ . . _ - - _ , _ _ _ . . _ _ _ _ - _ _ , _ _ . _ _ . ._ ..
__ __ .__ _ ._ . _ _ _ . _ _ _ . . ._ _ _ . . _ . . _ _.
--- - "
-.4
,
,,-
' -
.
.
..
6 of .the- Reactor Coolant System i's reduced by transferring heat from the
' Reactor Coolant System to the Steam and Power Conversion System through
the use of the steam generators. .
l
The Residual Hea't-Removal System is placed in operation approximately'four-
hours after reactor shutdown when the temperature and pressure of.the Reactor
Coolant. System are approximately 350*F and 425 psig, respectively.. Assuming
that two heat exchangers and two pumps are in service and that each heat
- exchanger is supplied with component cooling water at design flow and temperature,
- the Residual Heat Removal System is designed to reduce the temperature
of the reactor coolant from 350*F .to 140*F within 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. The heat load
- handled by the Residual Heat Removal System during the cooldown transient
4.
j includes residual and decay heat from the core and reactor coolant pump
. heat. The design heat load is based on the decay heat fraction that exists
-
_
at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following reactor shutdown from an extended run at full power. i
!
\
i System Description
'
The Residual Heat Removal System as shown in Figure 5.5-6 consists of two
4
residual heat exchangers, two residual heat removal pumps, an~d the associa'ted '
- piping, valves, and instrumentation necessary for operational control.
1 The inlet line to the Residual Heat Removal-System is connected to the
I
hot leg of reactor coolant loop 4, while the return lines are connected
.
to the cold legs of each of the reactor coolant loops. These normal return
l lines are also the Emergency Core Cooling System low head injection lines
i
i (see Figure 6'.3-1) .
!
$
\ [The Residual Heat Removal System suction line is isolated from the Reactor
' '
Coolant System by two motor operated valves in series while the discharge
lines are isolated by two check' valves in ,each line. These check valves are
4 not a part of the Residual Heat Removal System; they are shown as part of
the Emergency Core Cooling System (see Figure 6.3-1) . The isolation valves
! inlet line pressure relief valve, and associated piping are located inside
,3 the containment. The reasinder of'the system is located outside the containnent.
'h
i r- :
i
'
5.5-30
t
!
- ,
- :...:- _ _ . _. ._-- ._ . . . _ .. . _ _ _ . _ _ . _ . . _ . ~ . . _ _ . . _ . - _ .
. . .
__ ___ _ . . _ _ . _ , . _ _
. .- - . - . ~ - - - - - . . .
.. _ .. _ ,
..
.
.
4- During system operation, reactor coolant flows from the Reactor Coolant
( System to the residual heat removal! pumps, through the tube side of the
i
residual heat exchangers, and back to the Reactor Coolant System. The
heat is transferred in the residual heat exchangers to the -component cooling .
'
wate'r circulating through the shell side of the heat exchangers.
.
Coincident with residual heat removal operations, a portion of the reactor
!
coolant flow may be diverted from downstream of the residual heat exchangers
, to the Chemical and Volume Control System low pressure letdown line for
cleanup and/or pressure control. By regulating the diverted flow rate
and the charging flow, the Reactor Coolant System pressure may be controlled.
Pressure regulation is necessary to maintain the pressure range dictated
by the fracture prevention criteria requirements of the reactor vessel-
.
l and by the number 1 seal differential pressure and net positive suction
l head requirements of the reactor coolant pumps.
l
The Reactor Coolant System cooldown rate is manually controlled by regulating-
I
~ ') the reactor coolant flow through the tube side of the residual heat exchangers.
- * .
]
' '
A line containing a flow control valve bypasses the residual heat exchangers
and is used to maintain a constant return flow to the Reactor Coolant System.
Instrumentation is provided to monitor system pressure, temperature and
total flow.
4
The Residual Heat Removal System is also used for filling the refueling
cavity before refueling. After refueling operations, water is pumped back
to the refueling water storage tank until the water. level _ is brought down
to the flange of the reactor vessel. The remainder is removed via a drain ,
connection at the bottom of the refuelina canal.
i
i
When the Residual Heat Removal System is in operation, the water chemistry
1 is the same as that of the reactor coolant. Provision is made for the
Sampling System to extract samples from the flow of reactor coolant downstream
of the residual heat exchangers. A local sampling point is also provided
~
on each residual heat removal train between the pump and hest exchange-r. -
l s_ l .
i
( . ..
--
- .
- t
5.5-31
i
-
_ - . . . _ _ _ _ . _ _ _ . _
- -*,m -p yyn --w.- m.-- , . ,7 -..
-m.-w- m , m .%,, ,, ., --m - . - -
--g
. -- . . - . . . - . .. - - -
. . . . . .
-
.
.
-
.
a
..%
l
^ The Residual Heat Removal System functions in conjunction with the high i-
'
head portion of the Emergency Core Cooling System to provide injection
,
_of borated water from the refueling water storage tank into the Reactor
Coolant System cold legs during the injection phase following a loss .of
I, coolant accident. During normal operation, the Residual Heat Removal System
is lined up to' perform this emergency function.
!
i ,
In its capacity as the low head portion of the Emergency Core Cooling System,
'
The Residual Heat Removal System provides long term recirculation capability
for core cooling following the injection phase'~of the. loss of- coolant accident.
This function is accomplished by aligning the Residual Heat Removal' System
to take suction from the containment sump.
!
"
For a more complete discussion of the use of the Residual Heat. Removal System
- as part of the Emergency Core Cooling System, see Section 6.3.
I
I Component Description
'
i
\
)
/
' - '
The materials used to fabricate Residual Heat Removal System components
'
l are in accordance with the applicable code requirements. All parts of ,
components in contact with borated water are fabricated or clad with austenitic
stainless steel or equivalent corrosion resistant material.
Component codes and classifications are given in Table 5.5-9 and component
parameters are listed in Table 5.5-10.
Residual Heat Removal Pumps - Two pumps are installed in the Residual Heat
Removal System. The pumps are sized to deliver reactor coolant flow through
i - the residual heat exchangers to meet the plant cooldown requirements.. - The
l use of two pumps assures that cooling capacity is only partially lost should.
one pump become inoperative. One. 9 ump doop) is sdzwd (p s.s-36)
i
The residual heat removal pumps are protected from overheating and loss ,
i -
of suction flow by miniflow by-pass lines that provide flow to the pomo -
'
j suction at all times. A control valve located in each'miniflow line is.
- 5.5-32
a _-- :: :. .:: . 2. - - , , _ _ . - -. . -
.. .. . . . - - . .
_
-
.
.
,
~
,
.m regulated by a signal from the flow transmitters located in each pump discharee
'( header. ". ; ,
'
.nr~ .; ~ , .m m s.s ec: .ee .w ni sm
A pressure sensor in each pump discharge header provides a signal for an
indicator'in the control room.
'
by the pressure sensor.
The two pumps are vertical, centrifugal units with techanical shaf t seals.
All pump surf aces in contact with reactor coolant are austenitic stainless
steel or equivalent corrosion resistant caterial.
Residual Heat Exchangers - Two residual heat exchangers are installed in
the system. The heat exchanger design is based on heat load and temperature
differences between reactor coolant and component cooling vater existing
twenty hours after reactor shutdown when the temperature difference between
the two systems is small.
-' The installation of two heat exchangers assures that the heat removal capacity
of the system is only partially lost if cne heat exchanger becomes inoperative.
The residual heat exchangers are of the shell and U-tube type. Reactor
coolant circulates through the tubes, while component cooling water circulates
through the shell. The tubes are welded to the tube sheet to prevent leakage
of reactor coolant.
Residual Heat Removal System Valves - Valves that perform a medulating
function are equipped with two sets of packings and an intennediate leakoff
connection that discharges to the drain header.
Manual and motor operated valves have backseats to facilitate repacking
and to limit stem leakage when the valves are open. Leak off connections
are provided where required by valve size and fluid conditions.
. . _ . . . . .
,
k . ..
.
'
5.5-33
. _ _ _ . . _ _ . .
_ , . . - - - - ,,_
. - .
,
-
. . l
i
. I
.
I
m System Operation
!
Reactor Startup - Generally, while at cold shutdown condition, residual
heat from the reactor core is being removed by the Residual Heat Removal
System. The number of pumps and heat exthangers in service depends upon
the residual heat removal load at the time.
__- ,
'
lA " ' ' .~.
'
'
"~
.s . . , . . -. .. . ., , .b' , ';
.. . 14 %. h,
g. ~,1 w .. =, + . : . ,. , . # p ,3 ,
g ,, , . . : .
u.pp.' c. . c .e , w w MG E( , however, the discharge is directed to the
Chemical and Volume Control System via a line which is connected to the
common header downstream of the residual heat exchanger. The normal residual
return line is isolated. Failure of any of the valves to operate in the
line from the Residual Heat Removal System to the Chemical and Volu=e Control
System has no safety implication, either during startup or cooldown.
Af ter the reactor coolant pumps are started, the residual heat removal
N pumps are stopped, but pressure control via the Residual Heat Removal System
\s
- and the low pressure letdown line is continued until the pressurizer steam
bubble is formed. Indication of steam bubble formation is provided in *
.
the control room by the darping out of the Reactor Coolant System pressure
fluctuations, and by pressurizer level indication. The Residual Heat Removal
System is then isolated from the Reactor Coolant System and the system
pressure is controlled by normal letdown and the pressurizer spray and
pressurizer heaters.
Power Generation and Hot Standby Ooeration - During power generation and
hot standby operation, the Residual Heat Removal System is not in service
but is aligned for operation as part of the Emergency Core Cooling System.
Reactor Shutdown - The initial phase of reactor cooldown is accomplished
by transferring heat from the Reactor Coolant System to the Steam and Power
Conversion System through the use of the steam generators.
~ ~
. . -.. .
,
/
. ,
1:- - :
'
5.5-34
_ . . _ . _ . _. . _ . . _ __
- - .-- ._.
- .-. .. - . - .- -. - - - - - .
m _
-
. .
- .
4 .
.
.
g-ss. When the ; reactor coolant temperature and pressure are reduced to approximately
- I L :350*F and 425 psig,-approximately four hours after reactor shutdown, the
'
. second phase of cooldown starts with the Residual Heat Removal System being -
placed in operation.
Startup of the Residual Heat Removal System includes a warmup period during
which time reactor' coolant flow through the heat exchangers is limited
to minimize thermal shock. The rate of heat removal from ~ the reactor coolant
'
is manually controlled by-regulating the coolant. flow through the residual
heat exchangers. By adjusting the control valves downstream'of the residual
'
heat exchangers the mixed mean temperature of -the return flows is controlled.
Coincident with the manual adjustment, the heat ' exchanger bypass valve
contained in the common bypass line is regulated to give the required total
4
flow.
4
The reactor cooldown rate is limited by Reactor Coolant System equipment
cooling rates based on allovable stress limits, as well as the operating .
' ' '; temperature limits of the Component Cooling System. As the reactor coolant
l temperature decreases, the reactor coolant flow through the residual heat
'( \ s -
exchangers is increased.
As cooldown continues, the . pressurizer is filled with water and the Reactor
Coolant System is operated in the water solid condition.
At this stage, pressure control is accomplished by regulating the charging
flow rate and the alternate letdown rate to the Chemical and Volume. Control
System from the Residual Heat Rcmoval System.
Af ter the reactor coolant pressure is reduced and the temperature is 140'F ' -
or lower, the Reactor Coolant System may be opened for refueling or maintenance..
Refueling - Both residual heat removal pumps are utilized-during refueling :
to pump borated water from the refueling water storage tank to the refueling-
. cavi.ty. During this operatiod, the isolation valves in the inlet.line. .
~_si
!L - ---
- .
~"
5.5-35
c
,
.n.n. . . - . , , . . , . . , , , ,.n -- - - _ - - . , , , ., ~,, , . - . w
.._ - . .
,
-
. .
.
cf the Residual Heat Removal-System are closed, and the isolation valves to
m
-
the refueling water storage tank are opened.
The reactor vessel head is lifted slightly. The refueling water is then
pumped into the reactor vessel through the normal Residual Heat Removal
System return lines and into the refueling cavity through the open reactor
vessel. The reactor vessel head is gradually raised as the water level
in the refueling cavity increases. After the water level reaches normal
refueling level, the inlet isolation valves are opened, the refueling
water storage tank supply valves are closed, and residual heat removal
is resumed.
During refueling, the Residual Heat Removal System is maintained in service
with the number of pumps and heat exchangers in operation as required
by the heat load.
Following refueling, the residual heat removal pumps are used to drain
the refueling cavity to dhe top of the reactor vessel flange by pumping
I water from the Reactor Coolant System to the refueling water storage tank.
, ,
.
Design Evaluation -
System Availability and Reliability
The system is provided with two residual heat removal pumps and two residual
heat exchangers arranged in separate flow paths. If one of the two pumps
or one of the two heat exchangers is not operable, safe cooldown of the
plant is not compromised; however, the time required for cooldown is extended.
The two separate flow paths provide redundant capability of meeting the engineer-
ed safety function of the Residual Heat Removal System. The loss of
one Residual Heat Removal System flow path would not negate the capability
of the Emergency Core Cooling System since the two flow paths provide
full redundancy for engineered safety- requirements.- , , _, ,
'
- .i
- r-
- 5.5-36
-
, .. - ._ .._ _ __ _ - - _ _ _ - _ _ - _
~
..
- ..
..
.
.
m
j To assure reliability, the two residual heat removal pumps are connected.to-
two separate electrical buses so that each pump receives power from a
different source. If a total _ loss of off-site power occurs while the system
-is in service, each bus is. automatically transferred 'to a separate emergency
e
diesel power supply.
Leakage Provisions and Flooding' Protection
$
In the event of a loss of coolant' accident, fission products may be recircu-
lated through the Residual Heat Removal System pumps that are exterior to the
Containment. If a residual heat removal pump seal should fail, water would
spill-onto the floor of the pump compartment. Each RER pump is in a separate,
shielded compartment that drains to a sump containing two 30 gpm pumps that 3'
can-pump the spillage to the Waste Disposal System. Each sump pump is capable
of removing the spillage that would result from the failure of one RHR pump
seal.
/
( 's ,i If flooding occurred, overflow from one pump compartment would drain.throu'gh a
14 inch line to the pipe trench rather than flood the adjacent compartment.
Added sump pump reliability is achieved by elevating .the drive motors above
the compartment overflow drain so that the pump motors would not be flooded.
Gross leakage from the RER system can be accommodated in the pump compartments,
each of which has a capacity of 9450 gallons.
The RHR heat exchangers and pumps can siso be isolated'in the event of-gross
leakage through appropriate isolation valves. The isolation valves are opera-
ted manually by means of remote valve reach rod operators located in a
shielded valve gallery. The radiation shielding design criterion for this
~
valve gallery is the same as for manual contahent isolation valves. Post-
accident radiation levels in the vicinity of the recirculation loop are-
discussed in Chapter 12.
Leakage of a recirculatibn loop component is detected by,means_of,a radiation
1 : monitor that samples the air in the ventilation exhaust ducts from each com-
partment. Supplemental radiation monitoring is provided by the plant vent gas
_
j monitoring systein. Alarmsjath'scontrolroomalerttheoperatorwhenthe
i activity exceeds a ~preset level, and the capability exists to detect small leaksk l
1
4
'(July 1975 ~5.5-37 Amendment 30 i
. .. .. 1
i_, . __
. _ _
,
- -
.
,
.
'
.
.
m
'
within a short period of time. Operation of the su=p pumps is a less
sensitive indication of leakage. The various components of the recirculation
loop that are potential sources of leaks are described in Table 5.5-10A. 'The
table lists conservative estimates of the maximum leakage expected from each
leak source during normal operation. However, the design basis for sizing
Auxiliary Building sump pumps that will be required to dispose of this leakage
employs a conservative value of 35 gpm, as described above. 3C
The consequences of a leak through a RHR heat exchanger to the Component
Cooling Water System are discussed in Subsection 9.2.2.
Overpressurization Protection
The inlet line to the Residual Heat Re= oval System is equipped with a pressure
- wnm g. ,or. ..wi% ,,%. 33 . .po . 4 .a m , . . . m . . . . ._ .
. .
N
(' Each discharge line to the Reactor Coolant System is equipped with a pressure
relief valve located in the Emergency Core Cooling System (see Figure 6.3-1).'
They relieve the maximum possible back-leakage through the valves separating
the Residual Heat Removal S'ystem from the Reactor Coolant System.
The design of the Residual Heat Removal System includes the following features
for valves on the inlet line between the high pressure Reactor Coolant System
and the lower pressure Residual Heat Removal System:
1. The isolation valve adjoining the Reactor Coolant System is interlocked
with a pressure signal to prevent its being opened whenever the Reactor
~ ~
. . _. .
,
'
( . ..
.
'-
Amendment 30 5.5-37b (July 1975)
~ ~ ~'~ : __ - _ _
. _ . _ . . _ _ _ _ . _ _ . _. __
_
.. - . . _ . . . . . . . . _ _ _ _ , . .
-
i- . ,. .
.
.
.
.
'
m-
/
v Coolant System pressure is greater than approximately M
The valve is also automatically closed whenever the system pressure
increases above approximately 600'psig. ' This interlock and . automatic
closing action is derived from one process control channel.
.
,
2. The second isolation valve, the one adjoining the Residual Heat
Removal System, is similarly interlocked and automatically closed .(
with the actions being derived from a second process control channel.
.
See Section 7.6 for a more complete discussion of the interlocks on these
\
isolation valves.
i
- .
Shared Function
, The safety function performed by the Residual Heat Removal System is not
compromised by its normal function during plant cooldown. - The valves
% associated with the Residual Heat Removal System are normally aligned
i ( sto allow immediate use of this system in its engineered safety feature mode
of operation. The system has been designed in such a manner that two
redundant flow circuits are available, assuring the availability of at
least one train for safety purposes.
1
The normal plant cooldown functf.on of the Residual Heat Removal System
is accomplished through a suction line arrangement which is independent
of any safety function. The normal cooldown return lines are arranged
3 in parallel redundant circuits and are utilized also as the low head safety
injection lines to the Reactor Coolant System. Utilization of~the same
return circuits for the safety function as well as for normal cooldown
lends assurance to the proper functioning of these lines for safety purposes.
Radiological Considerations
The highest radiation levels experienced by the Residual Heat Removal
System are those which would result from a loss of coolant accident. Following
'
!
J
-
a loss of coolant accident, the Residual Heat Removal System is used as
,,
.
i
!
'
A.
s.
1
5.5-38
^l
. . . . . . . - .
-
. - - , . . - - . . - - - - .. ~
. _ _ _
-
.
,
-
l
. .
.
.
, ,s part of the. Emergency Core Cooling System. ' During the recirculation _ phase i
.
ti .>mergency core cooling, the Residual Heat Removal System:is designed-
~
{: (
to operate for.up to a year pumping' water from the containment sump, cooling
lit, and returning it to the containment 1to cool the core.
.
Since, except for some valves and. piping, the Residual Heat Removal System
is located outside the containment, most of the system is not subjected
to the high levels of radioactivity in the containmentf post-accident environment.
To assure continued operation of the Residual Heat Removal System components,
the valve motor oper.ators. the residual heat removal pump motors, and
the residual heat removal pump seals are of the same grade as the equipment
qualified by environmental tests to withstand the containment post-accident
environment (1).
The operation of the Residual Heat Removal System does not involve a radiation
hazard for the operators since the system is controlled remotely from
the control room. If maintenance of the system is necessary, the portion
,- of system requiring maintenance is isolated by remotely operated valves
{ \, and/or manual valves with stem extensions which allow operation of the
valves from a shielded location. The isolated piping is ' drained and flushed - -
before maintenance is performed. '
Tests and Inspections
Periodic visual inspections and preventive maintenance are conducted during
! plant operation according to normal industrial practiec.
The instrumentation channels for the residual heat removal pump flow instru-
mentation devices are calibrated on a nominal 18-month frequency. 56
4
.
The residual heat removal pumps are periodically tested by starting them.
- -
.
. . . . . .
<
.
)
( - - -
.
~~ #
(February 1978) - 5.5-39 Amendment 58
._. _ _- - ._ . _ _ _ _ _ . . . _ . _ _ . _ . _ _, . . . _ _ ._.
-
__ _ _
-_
. .
,
-
.
The water in the tank is borated to a neration of 2000 ppm boron (nominal)
which assures reactor shutdo' approximately 10% Ak/k when all RCC j
f ' assemblies are in and when the reactor is cooled down for refueling. .
\t ,
l
i
Pumps
Residual Heat Removal Pumps - Residual heat removal pumps are provided
to deliver water from the refueling water storage tank to the Reactor
<
Coolant System should the Reactor Coolant System pressure fall below their !
shut-off head. Each residual heat removal pump is a single stage, vertical,
centrif ugal pump. It has an integral motor pump shaft, driven by an induction !
motor. The unit has a self contained mechanical seal, which is cooled
by component cooling water. The performance curves for the RHR pumps are
given in Figure 6.3-1. The lower curve is the curve used in the analyses
of Chapter 15. The upper curve is the performance curve determined from
]
'
shop tests.
l
l
s,, _ , ,, .
v. .. em : r r - w-<1%m a .,.. . .
, . . , , ..n.
.
S . The changeover from injection mode to recir- l
culation mode (described later in this Subsection and in Table 6.3-4) is ini- !
1
tiated by low level in the RWST, which results in an automatic trip of the
RHR pumps. This trip is initiated by two-out-of-three logic using f.nformation !
provided by three independent RWST level instrumentation channels. This trip )
function is designed to meet the requirements of paragraph 4.2 of IEEE 279-1971. The changeover is thus automatically initiated after water is avail-
able in the containment sump and before water is exhausted from the RWST.
Adequate net positive suction head (NPSH) is always available to the RHR pumps in
both the injection phase and the recirculation phase. Table 6.3-9 lists
available and required NPSH. Phase 1 of the preoperational system test (see
'
7l
Section 6.3.4) verified that the required RHR pump NPSH will be available for
all conditions.
!
'
,
Once
'
( (,Urll Y10
_ _ _
eyh lc5$ Oh @ C- W (*e> clost cb5C1My
vobe. k {o en1 recEcu kt tm M bve b su chen ress )f
(October 1978) s
-
6.3-7 Amendment 71
a __. _ - . . .. _. .
.- . . . _. . - . - _ _ . . - . .- _- -.
_ _ _ - - - . - _ _
__
-
.
-
,
.
.
4
.,f(ov is established to the Reactor Coolant System, the bypass line is
I
omatically closed.
,
The residual heat removal pumps are
also discussed in Subsection'5.5.7.
4
4
4
0
\
.
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!
i
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\
g e e. . .
.- :
- 6.3-7a
.. . . -
-- _ - -. . . -- -- -. - . ..-
, - . . . - - - . _. . - .-. . . ..-. - - _ _ .
_
-.?
,
-
. .
,
. ..
,
.
.
'
'
- 7.. A flow indicator is provided in the ' safety inje on pump header,
- r
'
,
and in the residual heat; removal pump heade . Pressure instru-
mentation is also provided-in these lin .
.
8. An" integrated system test'can performed when the plant is cooled
down and the Residual Heat moval System is in operation. This
test does not introdu flow into the Reactor Coolant System but.
,
does demonstrate e operation. of the valves, pump circuit breakers,
and automati circuitry including diesel starting and the automatic
-loadin ,
Eazrgency. Core Cooling System components off the~ diesels
(b simultaneously simulating a loss of .off-site power to the vital
electrical buses) .
,
6.3.5 INSTRUMENTATION APPLICATION
.
Instrumentation and associated analog and logic channels employed for
, initiation of Emergency Core Cooling System operation is discussed-in.
.
-
Section 7.3. This section describes the instrumentation ~ employed for
monitoring Emergency Core Cooling System components during normal plant
operation and also Emergency Core Cooling System post-accident operation.
All alarms are annunciated in the control room.
Temperature Indication
&
Boron Injection Tank Temperature
Duplicate temperature control channels are provided for the boron injection
tank electric strip heaters. Both actuate high and low temperature alarms.
One channel provides temperature indication in the control room, the other
provides local indication.
Residual Heat Exchanger Outlet Temperature
. .
. ..; -. .
j The fluid temperature at the outlet of each residual heat exchanger is ',
recorded in the control room. ;
- .
e
- '
6.3-38.
.. , - _ _ _ . . - _ _ . - . _ _ _ . - _ _ _ . . _ ... ,___ - ,-. _ _ _
.- . - . . - .. -
, . . . . .
_
-
. .
.
.-
' Heat Tracing Temperature -
.[
Separate. thermostatic controls' are provided for each section of the heat
tracing in the boron recirculation loop to maintain the. temperature within
the speci,fied range. High and low temperature alarms are provided to warn
'of . failure to maintain the temperature within the control band.'
Pressure. Indication
Boron Injection Tank Pressure
- ' Boron injection tank pressure is indicated in the control room. . A high
pressure alarm is provided.
S afe ty Injection Hesder Pressure
. Safety injection pump discharge header pressure is indicated in the control
I
(, \
room..
. . r
.
Accumulator Pressure ,
,
Duplicate pressure channels are installed on each accumulator. l Pressure
indication in the control room and high and low pressure alarms are provided
by each channel.
,
Test Line Pressure
A local pressure indicator used to check for proper seating of the accumulator
check valves between the injection lines and the Reactor Coolant System is
'
installed on the leakage test line.
.
Residual Heat Removal Pump Discharge Pressure
Residual heat removal pump discharge pressura for each puinp 'is 'i'ndicated in -
(*j the control room. A high pressure alarm is actuated by each channel.
..
- t
.
,
a
6.3-39
i
_ . _ , . _ - . , - . - ..,.-v.-mme -, -~.4 ------e * -ym--n- r- <- mm" m -- e ---- - r"--v< ==vt- " *-* r "" -r7 * -
,. . . . .. - - .. .
'
'
, .
,
..
.
.
[ v
Flow Indication
Eoric Acid Recirculation Flow .
'
Boric acid recirculatien flow through the boron injection tank is indicated ~
locally.
Charging Pump Injection Flow
Injection flow through the common header to the reactor cold legs is indicated
in the control room.
Safety Injection Pump Header Flow
Flav through the safety injection pump headers is indicated in the control
room.
_.
'
i.
-
Residual Heat Removal Pump Iniection Flow
Flow through each residual heat removal injection and recirculation header
leading to the reactor cold or hot legs is indicated in the control room. ,
Test Line Flow
Local indication of the leakage test line flow is provided to check for
proper seating of the accumulator check valves between the injection lines
and the Reactor Coolant System.
.
Safety Injection Pump Minimum Flow
A local flow indicator is installed in the safety injection pump minimum
flow line.
.. . . .
' * e
.
,
u- :
i
- *
.
l
'
6.3-40
i
-
~ . - - , . _ _ . _ . . . - . . . , _ , , _ . . _ _ . _ , . . , , _ . . , .._.m..._,.. . ..
_ _ __ . _ _ _ .
- - -
. . .
_,
- .
. .
.-
.
. . Residual Heat Removal Pump Minimum Flow
3 ~
A flowmeter installed in each residual heat removal pump discharge header
l provides control for the valve locatedLin the pump minimum flow line.
.
- Level Indication
. Refueling Water Storage Tank Level
4
,
Three water level instrumentation channels are provided for the RWST. Each
channel provides independent. indication on the main control board,. thus -
satisfying the ' requirements of paragraph 4.20 of IEEE 279-1971. Two-out-of-
-
three logic is provided for RHR pump trip and low level' alarm initiation. One
channel provides-low-low water level alarm initiation. ,
i
Accumulator Water Level
Duplicate water level channels are provided for each accumulator. Both
channels provide indication in the control room and actuate high and low
-s water level alarms.
'
Containment' Sump Water Level
.
.
Two containment sump water level indicator channels are provided. Both
! indicate in the control room.
i
Valve Position Indication
'
!
4 1
{ Valve positions which are indicated on the control board are done so by a
] " normal off" system; 1.e. , should the valve not be in its proper position,
- a bright white light will give a highly visible indication to the operator.
Accumulator Isolation Valve Position Indication .
~
The accumulator motor operated valves are provided with red (open) and
~
.,
- green (closed) position indicating lights located at the control switch for
j aach valve. These lights are energized from valve control power and actuated
^
by valve motor operator limit switches.
. . . . . . .
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. 6.3-41 l
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- -
-
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A monitor light that. is on when the valve is' not fully open is provided in
~
.-
.
an array of monitor lights that are all off when their respective valves
are in proper position enabling safety features operation. This light is
energized from a separate monitor light supply and actuated by a valve
motor operated limit switch.
.
An alarm annunciator point is activated by a valve motor operator limit
switch or a stem travel limit switch whenever an accumulator valve is not
.
fully open with the system at pressure (the pressure at which the safety .
injection block is unblocked). The alarm is reinstated once an hour. A
separate annunciator point is used for each accumulator valve.
_de. fueling Water Storage Ta-k Isolatien Valves
These valves are manually controlled and normally sealed in the open position.
27
1
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q *
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i
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i
6.3.6 REFERENCE
i
- 1. Igne, E. G. and J. Locante, ." Environmental Testing of Ens,ineered
!- , Safety Features Related Equipment (NSSS-Standard Scope)",* WCAP T '
l j 7744, Volume I, August, 1971.
,
,
C'~ :
Amendment 27 ,
. 6.3-42 (April 1975) --- -
1
.
. - . _ -, , _ , , - - - -,
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- -
, --- ,,,r--,--i-- --% w- ,r,-- - -
. _ _ _ _ _ - _ _ - .
-
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, REACTOR COOLANT SYSTEM - -
f
O'!E72RESSURE PROTECTION SYSTEMS
.L*u: TING CONDITION FOR OPERATION
3.4.9.3 The following overpressure protection systems shall be OPERABLE:
a. RHR system isolation valves 8701 and 8702 open with power removed
from the valve operators when the oositive disolacement charging _
pump is in operation, and
b. Two power cperated. relief valves (PORVs) with a lift setting of less
than or equal to 450 psig, or
c. The Reactor Coolant System (RCS) depressurized with an RCS vent of
greater than or equal to 2.07 square inches.
ADLICABILITY: MODE 4 when the temperature of any RCS coId leg is less than
o equal to 323'F, MODE 5 and MODE 6 with the reactor vessel * head on.
ACTION:
a. With the positive displacement charging. pump in operation with the
,
, RHR isolation valves closed, within one hour either open the RHR
isolation valves or secure the positive displacement charging pump.
f;-
'
\s:
b. With one PORV inoperable, restore the inoperable PORV to OPERABLE *
status within 7 days or depressurize and vent the RCS through a 2.07 *
square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.- ,
c. With both PORVs inoperable, depressurize and vent the RCS through a
2.07 square inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />,
d. In the event either the PORVs or the RCS vent (s) are used to mitigate
an RCS pressure transient, a Special Report shall be prepared and
submitted to the Commission pursuant to Specification 6.9.2 within
30 days. The report shall describe the circumstances initiating the '
transient, the effect of the PORVs or vent (s) on the transient,tand
any corrective action necessary to prevent recurrence,
e. The provisions of Specification 3.0.4 are not applicable.
.
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,-
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- ~~Lu
DIABLO CANYON - UNIT 1 3/4 4-32
i. . if C*'Z E!*'!.J- . : _ _ . - . . . - . - . - . . . - - - - ,_. . . _ . . - , _ . _ _ , _ - -
-
_ _ , . . - - . - -
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EMERGENCY CORE COOLING SYSTEMS
3/4.5.2 ECCS SUBSYSTEMS - T,y greater than or equal to'350*F
). LIMITING CONDITION FOR OPERATION
i
,
'
3.5.2 Two' Emergency Core Cooling System (ECCS) subsystems shall.be OPERABLE
with each subsystem comprised of: ~
,
'
,
a. One OPERA 8LE centrifugal charging pump,
'
b. One OPERABLE safety injection pump,
c. One OPERABLE residual heat removal heat exchanger,
d. One OPERABLE residual heat removal pump, and
I e. An OPERABLE flow path capable of taking- suction *from the refueling
water storage tank on a safety injection signal and manuall
'
transferring suction to the containment sump during the recirculation
phase of operation.
1
^ ~
~ APPLICABILITY: MODES 1, 2 and 3.
'
,
ACTION:
a. With one ECCS subsystem inoperable, restore the inoperable subsystem
to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least 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 in at least HOT SHUTDOWN within the
following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b. In the event the ECCS is actuated and injects water'into the Reactor
.
Coolant System, a Special Report shall be prepared and submitted to
j the Commission pursuant to Specification 6.9.2 within 90 days' describing
' the circumstances of the actuation and the total accumulated actuation
cycles to date. The current value of the usage factor for each
affected safety injection nozzle shall be provided in this
. Special Report whenever its value exceeds 0.70.
!
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,
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DIABLO CANYON - UNIT 1 3/4 5-3
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3/4.5.3 ECCS SUBSYSTEMS - T,yg less than 350'F
.
'
LIMITING CONDITION FOR OPERATION
-
,
l
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. 3.5.3 As a minimum, one ECCS subsystem' comprised of the following shall be .
-OPERABLE: '
,
a. One OPERABLE centrifugal charging pump,#
'b.
' -
One OPERABLE residual heat removal heat exchanger,
c. One OPERABLE residual heat removal pump, and
'
d. An OPERABLE flow path capable of taking suction f rom the refueling
.
water storage tank upon being manually realigned and transferring
suction to the containment sump during the recirculation phase of
- - operation.
L
i
APPLICABILITY: MODE 4.
3
.
m ACTION:
-'
'
\
{. .- a. With no ECCS subsystem OPERABLE because of the inoperability of -
either the centrifugal charging pump or the flow path from the .. ;
refueling' water storage tank, restore at least one ECCS subsystem to .
i
OPERABLE status within I hour or be in COLD SHUTDOWN within the next
20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
b. With no ECCS subsystem OPERABLE because of the inoperability of
i;
' either the residual heat removal heat exchanger or residual heat
removal pump, restore at least one ECCS subsystem to OPERABLE status
or maintain the Reactor Coolant System T"V9 less than 350'F by use
l of alternate heat removal methods. I
l
-
c. In the event the ECCS is actuated and injects water into.the Reactor l
4
Coolant System, a Special Report shall be prepared and submitted to
' the Commission pursuant to' Specification 6.9.2 within 90 days describing
the circumstances of the actuation and the total accumulated actuation
1 cycles to date. The current value of the usage factor for each
.
affected safety injection nozzle shall be provided in this Special
I Report whenever its value exceeds 0.70.
l
l
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. ., . . . .!
' '~
1
y #A maximum of one centrifugal charging pump shall be OPERABLE whenever the
temperature of one or more of the RCS cold legs is less than or equal to 323'F. '
.
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l DIABLO CANYON - UNIT 1 3/4 5-7
-- . . _ . _ . . - - - . . _ - - . . -
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- - -
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, 1REFUEEING OPERATIONS .
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3/4.9.8 ' RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION I
i~
f;)JIGH WATER LEVEL
.
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- , , , , ,
..
,,
LIMITING CONDITION FOR OPERATION
. . .
. .
,
,
i
3.9.8.1 At least one residual heat removal (RHR) train shall be OPERABLE-and
,
-
in operation."" *
!
! APPLICABILITY: MODE 6 when. the water level 'above the'. top of the reactor,
i
vessel-flange is at least 23 feet.
'
. . . . .. ,
\ -
'
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ACTION: -
i a. With no residual heat removal train OPERABLE and in operation,
i ' suspend all opera' ions involving an increase in the reactor decay
heat load or a reduction in baron concentration of- the Reactor.
Coolant System and immediately ini'tiate corrective action to return
the required RHR train to OPERABLE and ' operating status as soon as
i possible. Close all containment penetrations providing direct -
l
access from the containment atmosphere to the outside atmosphere .
.I .
within 4-hours.
'
i h
!ks .
-
SURVEILLANCE REOUIREMENTS
-
, .
4.9.8.1.1 The required RHR train shall be demonstrated OPERABLE pursuant to
- Specification 4.0.5.
4.9.8.1.2 At least one residual heat removal train shall be verified to be in
.
operation and circulating reactor coolant at a flow rate of greater than or
l equal to 3000 gpm at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. .
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"The residual heat removal train may be removed from operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the performance of CORE ALTERATIONS in the vicinity
of the reactor pressure vessel hot legs.
i **The residual heat removal train Imay- be removed from operation.a.nd.02ERABLE
status' for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period for the performance of leak testing
-\, ,-) the RHR suction isolation valves.
>
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n'.$310 CANYON - UNIT 1. 3/4 9-8 AMENDMENT NO. 5 -
. - .- . - - .,. _ - - -..._- . - . - - -. -. -_-.-,-,- -- - . - . . - _
._ _ _ . _ _ _ _ __-- _ _
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REFUELING OPERATIONS
FT' )
-LOW WATER' LEVEL
4
~ '
'
LIMITING CONDITION FOR OPERATION-
! . .
2.S.3.2 Two' independent residual heat removal (RHR) trains shall be OPERABLE
and at least one RHR train shall b,e in operation.
APPLICABILITY: MODE 6, when the water level above the top of the reactor
i pressure vessel flange is less than 23 feet.
. .
,
ACTION:
l a. With less than the required RHR trains OPERABLE,'immediately initiate
1 corrective action to return the required RHR trains to OPERABLE.
status, or to establish at least 23 feet of wa er above the reactor
pressure vessel flange, as soon as possible.
b. With no RHR train in operation, suspend all operations invo'lving a
reduction in boron concentration of the-Reactor Coolant System and
immediately initiate corrective action to return the required RHR
,-s train to operation. Close all . containment penetrations providing
- % direct access from the containment atmosphere to the outside
-k. .I atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />..
. s_.
. .
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- SURVEILLANCE REQUIRMENTS
4.9.8.2 The required residual heat removal trains shall be determined OPERA 8LE
j per Specification 4.0.5.
.
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DIABLO CANYON - UNIT 1 3/4 9-9
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- t.icense It.w
Low Power Full Power
(Step 2) (Step 3)
1
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_
l. IDVP & ITP SER llant S./DE Kni tihl. '
'
,
3.
Horong1irc (Pump Trip) T.S. .1 Wilson /DSI Ilouston
- Full Power Lic.
g
4 T. S.
- 4 Emergency Facilities lipgrade (Amend) BBuckley/ Chandler low Power Amend.
(Sitoller in OELD review)
- 5 PASS-Procedures llans S./DSI Muller
. 6.
Full' Power Cond.
Breakwater T.S. Hans/0 ELD Full Power T.S.
'7. * CCW T.S. Wilson /0 ELD Full Power T.S.
lH. Appendix R Buckley/DSI Houston Full' Power Lic.
'
'
Condition
1 9. R g. Guide ,1.97 . Buckley/DSI Houston ... Full Power Lic.
Condition
10 FSAR update. Hans S./0 ELD EGD letter
11. Centrol lleavy Loads (Phase II) llans S/DSI Rubenstein Full Power I.ic.
Condition
- 12. sligh Point Vents (Generic) -
llans/DSI llouston Full Power Lic.
Condition
-13. R21ief Valve Testing (Generic) llans/DE Knight Full Power Lic.
' Cordition
.
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!
I TABLE OF ALLEGATION ASSIGIMENTS '
!
} All:: cation Assigned Division Branch Support Branch-
, 1. Attempted sabotage OI '
i 2. Explosives /Antinuclear blockage OI
,
i 4. Single Failure Capability CCE NRR . g'/7.
DSI/Rubenstein AS8/Parr
y#" sid
'
j 5. Heat Removal Capability CCW NRR DSI/Rubenstein '
,-s
6 1&O Design Classification -AS8/Parr y
' NRR DSI/ Houston ICS8/ Rosa
-
Seismic Category I/ Category II Interface
-
7 NRR
,
'
DST /Rowsome PRA8/Thadant AS8/SGE8'
, 8. ' Seismic Design of Diesel Gen. I & Exh. NRR DE/ Knight
' 9. USI-17 Systems Interaction Generic EQ8/Noonan SGES
NRR DST /Schroeder GIB/Kniel
10 Seismic TiHing of Containment NRR DE/Xnight
.
SGE8/ Lear-
l 11. Classification of Platform . MRR DST /Rubenstein ASR/Parr SGE8
'
- (Category I/ Category II . -
! 12. HELBA did not meet FSAR, RG 1.46 NRR DSI/Rubenstein AS8/Parr
i 13. Inadequate Seismic Systems NRR .DSI/Rubenstein ASB/Parr
4 14. Loads on Annulus Structural-Steel NRR DE/ Knight SGE8/ Lear
! not Calculated Properly i
4 15. Inadequate Tornado Load Analysis NRR DE/ Knight SGE8/ Lear
1
i of Turbine Building *
!16. _High Energy Pipe Break Restraint NRR De/ Knight.
Inadequate ME8/80snak
i ,
l 17. NSSS SSE Load Inadequate NRR -
DE/ Knight
l 18 0A/0C Allegations SGE8/ Lear
DI
,
! 19. Safeguards - Sensitive OI .
l 20. . Health Physics personnel do not meet Region V . -
- 21. ANSI ALARA requi,rements
Program - Paper Tiger
e
l Region V '1
- 22. Radiation Monitors lack sensitive -
Region V
j23 Iaimidation of OC Inspectors 01
124 HPFoley NCR's rejected without Region V
! good cause*.
125. Deficiency in use of " Red Head" Region.V DE/ Knight
j anchors for raceway support ME8/Bosnak
!26 Feley not document NCR's issued. Region V-
- by field inspectors
i
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,
-
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.- - _ _ .. . - . . - .. -. .- . -- ._ _- - -_______ _ __
_ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ . . _ . . - - _ _. _ _ _ _ _ ._ _ . _ _ _ . ,
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-
.
'
iAllegation
~
Assigned : Division Branch Support Branch
j 1
j 27 Welding and QA deficiency in " Super Region V DE/ Knight .SGE8/ Lear
' 28. Annulus Structure Reverification .
. ;
'
Program inadequate '
- 29. Pipe Restraints Design Inadequate . NRR DE/ Knight MEB/Bosnak
!30
AES/-
Inadequate Documentation of Safety NRR De/ Knight EQB/Noonan
! R21sted Equipment -
131. QA Procedures for Struct. Analysis NRR DE/ Knight. SGE8/ Lear IE/QAB.
!32.'. Seismic Analysis Containment NRR . DE/ Knight SGES/ Lear-
,
l33. Tcrbine Building (Class 2) Contains NRR DE/ Knight SGES/ Lear DSI/AE8
3 Class 1 .
l34. Incomplete as-built drugs Region V -
i35. Lcck of support cales for fluorescent NRR DE/ Knight EQB/Noonan
1
l light fixtures SGE8/ Lear ;
i36 R:s'o lution of Fluorescent Light NRR DE/ Knight EQB/Noonan Region V
i
'
} fixture interaction .
j37. Solid State Protection System Relays NRR DSI/ Houston ICS8/ Rosa
138. PG&E ignoring spurious closure.of RS8/Sharon ,
Region V ' '
l me valve
j39 No control Rna annunciation of closed NRR DSI/ Houston ICS8/ Rosa' RS8/Sharon -
! RHR suction valve ,
240. RHR hot leg suction not single failure NRR
-
DSI/ Houston RS8/Sharon - ICS8/ Rosa
141. Drug inadequate . Region V
- 42. License management unresponsive to Region
{ problems
- 43. License Reporting Failure Region V e.
!
- 44. License Improp. Assessment of Design Region.V. -
i Change ..
l45. Desion inconsistency in FSAR RHR valves NRR DSI/ Houston ~ RS8/Sharon
l
ICS8/ Rosa-
'
! . incorrect -
47 Plcnt Amancing System Region V
48. SI Study and associated Mods NRR DST /Rowsome
4g. Emergency Sirens not seismic RRAS/Thadant. '
ISE - -
1
.
.
.
_ _ _ _ _ . _ .
-
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Allegation '
Assigned Division Branch Support Rranch
50 Plant Security should have been MNSS '
retained
51. Risk of job action to allegers OELD
,
52 Construction & hrgs in progress - '
after fuel load inappropriate
53 Welder Qualification could be purchased 01
54! Wire _ traceability not evident work by
PG4E and Foley Region V.
-
55 Bechtel approved analys~is of small bore .
NRR DE/ Knight ME8/Bosnak
pipe by altering failed analysis
56 Pitting of Main steam and feeduster piping NRR DE/ Knight
57 Foley used* uncertain and unqualified 0.C. MER/8osnak
Region V
Jr.spection prior to 1983
58. ,Foley allows " Red Head". anchors Region V
studs reported improprely installed .
59. Foley lost cable traceability . Region V' ' .
60 Foley purchased material things Region V
inapproved vendors
61. Lack of Document control Region V
62. Foley lacks adequate sampling of cable Region V
pull activities .
63 Foley has' lost material tracability through Region V .-
upgrade Class 2
i64. Grout test sampling based on special tests : Region V e
! rather than field
!65. Foley documents prior to 1980 questioned Region V
I no review, required prior to 9/1981 license
l issuance date
66. Defective weld reports rejected by Foley Region V -
.
Q7. Negligence by PG&E flooding at 55 system _R egion V
! olevation pipe tunnel .
68. NCS pullman Kellog. audit ISE
/. -
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! '
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.
,
_. _ _ - . . . _ - . _ _ _ _ _ _ . . . _ _ . _ _ _ _ _ . _ _ . _ . _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _. _ _ . - . .
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Ollegation Assigned Division Branch Support ilranch -
!
09. Revision of Draft Case Study "C" I&E
FO. Falsification of response to NRC not
,
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.
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cf Viol i
- 1. Alledged use and sale of drugs 01 - .
@2. Region V i
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