ML20244D484
| ML20244D484 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 07/10/1985 |
| From: | Thompson H Office of Nuclear Reactor Regulation |
| To: | Denise R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| Shared Package | |
| ML20244D488 | List: |
| References | |
| TAC-42527, NUDOCS 8510080372 | |
| Download: ML20244D484 (3) | |
Text
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'4 9 Ofc UNITED STATES g
h NUCLEAR REGULATORY COMMISSION i
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r,E W ASHINGToN. D. C. 20555
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Q,M July 10, 1985 MEMORANDUM FOR:
Richard P. Denise, Director
)h Division of Reactor Safety and Projects Region IV J
FROM:
Hugh L. Thompson, Jr., Director l
Division of Licensing i
Office of Nuclear Reactor Regulation
SUBJECT:
STATUS OF EQUIPMENT QUALIFICATION AT FORT 'T.
VRAIN
Reference:
Memorandum from Hugh L. Thompson, Jr., to
)
Richard P. Denise dated June 14,1985, Fort St. Vrain-1 Recommendations Concerning Plant Restart" i
We are continuing our activities to resolve the open issues concerning Fort 1
St. Vrain's equipment qualification.
Our recommendation that these issues i
should be resolved prior to restart was included in our earlier memorandum concerning plant restart referenced above.
As you are aware, on July 2, 1985, members of the NRR and Region IV staff met 1
with the licensee in Bethesda to discuss the status of their equipment qualification program and the related issue of operator actions required to terminate a design basis reheat steam line break event.
A copy of.our meeting summary is enclosed.
NRR is currently in the process of completing its safety evaluation on the subject of Fort St. Vrain's current compliance with the EQ rule (10 CFR 50.49).
We anticipate that this evaluation will be completed during the week of July 8, 1985. The purpose of this memorandum is to provide you with early feedback on current NRR conclusions with regard to this matter and the question of operator action needed to terminate design basis ever,ts.
As stated in the enclosed meeting summary, both the licensee and the staff acknowledge that the Fort St. Vrain equipment qualification program and documentation files do not include established qualified lives or operability times for safety related electrical equipment.
The staff position is that without these items being included in the qualification program, we cannot conclude the program is in compliance with the EQ rule.
The licensee currently has under way activities to include these items in their qualification program.
As we discussed at the meeting, however, an EQ safety concen. Ns not arise provided the plant does not operate at a power level where the accide:c.+
l environment would be severe enough to challenge equipment operability.
from a safety standpoint, it would therefore be permissible to allow plant operation at appropriately limited power levels.
Such power levels could not be exceeded until the EQ program deficiencies have been resolved.
I CONTACT:
K. Heitner 27364
-J
1 1
lt Memorandum to R. P. Denise
- It is our understanding that the licensee desires to begin operation.in this mode as soon as possible to begin the dry out process net.esary to return to full power operation.
This phased approach to full power operation should allow the licensee Enough time to complete his EQ documentation in the areas where it is currently considered by NRR to be deficient.
Their current schedule would have the qualified equipment life established by August 30, 1985.
They should be requested to accelerate and define their schtdule for establishing j
operability times such that both the aging and operability times issues will be resolved prior to the plant's return to power operatirn.
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The question of operator actions necessary to terminctis a reheat steam line break is separate but related to the open EQ issues.
she design basis accidents which determine the environment for qualifying' equipment are breaks in the reheat steam lines.
The licensee originally analyzed the operttor actions needed to isolate j
these breaks and concluded that this could be done in four minutes.
The staff
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accepted this position in safety evaluations on equipment qualification in 1973 and 1977.
Therefore, staff questions in this area must be considered in light of current backfitting rules and procedures.
Nevertheless, the staff remains concerned about this issue.
More recent acceptance criteria for operator action are far more conservative. Operator response times now found acceptable are typically ten minutes or longer.
The licensee has partially addressed this problem by committing to install a steam lire break detection system, and considering automation of the isolation system.
Additionally, the licensee has made extensive submittals supporting his position.
In order to fully evaluate this situation, we have i
requested an opportunity to visit the site as soon as possible and review / observe the required operator actions.
Our final safety determination will be made after that visit.
Becaut.e of the backfitting aspect of this issue, NRR does not consider its resolution a restart requirement at this j
time. liowever, we would like to complete our site visit prior to the plant's return to power operation and the licensee has agreed to schedule the visit as soon as possible within the next two weeks.
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In summary, NRR believes that restart operations can begin without raising a L fety concern as long as these operations are restricted to power levels which do not produce steam conditions which would result in a design basis i
accident environment that could challenge the operability of safety related electrical equipment.
/
/t?/
i uh. Thompson, Jr.,
' rector Di i ion of Lice sin Of 'ce of Nuclear Re or Regulation cc: R. Vollmer i
C______.____.__
l MEETING
SUMMARY
DISTRIBUTION, 1
l Licensee: Public Service Company of Colorado l
i
- Copies also sent to those people on service (cc) list for subject plant (s).
Docket File NRC PDR L PDR ORB #3 Rdg KHeitner i
EButcher BGrimes OELD Edordan, IE ACRS-10 l
PMorriette NRC Meeting
Participants:
FMiraglia l
Glainas l
l JKnight DZiemann RKarsch RLaGrange WKennedy PShemanski AMasciantonic WShields TKing l
RIreland l
PWagner j
1 l
l l
l l
c
,Q p a:e UNITED STATES O
iQ NUCLEAR REGULATORY COMMISSION
- E WASWNGTON, D, C. 20555 h
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July 10, 1985 PPI#
..... /"
c MEMORANDUM FOR:
Edward J. Butcher, Acting Chief Operating Reactors Branch No. 3, DL i
FROM:
Kenneth L. Heitner, Project Manager Operating Reactors Branch No. 3. DL 8
SUMMARY
OF JULY 2, 1985 MEETING WITH PUBLIC SERVICE COMPANY.
SUBJECT:
0F COLORADO TO DISCUSS EQUIPMENT QUALIFICATION AT THE FORT ST. VRAIN NUCLEAR GENERATING STATION i
[I The purpose of this meeting was for Public Service Company (the licensee) to i
discuss with the staff the status of the Equipment Qualification (EQ) program The meeting was held by mutual agreement of the licensee at Fort St. Vrain.
and the staff with the principal objective being to clarify the status of
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Fort St. Vrain's compliance with the EQ rule (10 CFR 50.49). The attendees at this meeting are listed in Enclosure I.
The licensee restated his position that Fort St. Vrain is in compliance with 10 CFR 50.49 to the extent that it applies to a gas cooled reactor. The licensee stated that the plant would be ready to return to service in about a The approach of the licensee's EQ program had been to interpret weeks time.
the requirements as applied to a gas cooled reactor and base his submittals to the NRC on his interpretation.
If no statements to the contrary were made i
by the staff, he assumed his interpretation was correct. He presented a i
historical perspective on EQ at Fort St. Vrain (See Enclosure II).
l The following specific issues were discussed between the licensee and staff:
The licensee's commitment to a steam pipe rupture detection system (for 1.
reheat steam lines) i 2.
The licensee's commitment to consider an automatic isolation system (for reheat steam lines) l l
3.
The licensee's equipment aging program which is underway with a completion date of August 30, 1985.
4.
The licensee's study of equipment operability times.
5.
The scope of safety related equipment covered by the licensee's EQ program.
6.
The licensee's replacement program for the equipment identified in item 5 above.
7.
The temperature and pressure profiles actually used by the licensee to qualify the equipment.
8.
The status of documentation on file at the site for the EQ program.
u---_-_-______
Memo to Mr. E. Butcher,
l Following these discussions, the staff concludeo that items 5 through 8 did l
not require further discussion at this time.
l It was agreed that on items 1 and 2, further action to resolve the question l
of operator response time was necessary. The staff requested a meeting with the licensee at the plant site on this issue. The licensee agreed to arrange this meeting as soon as possible within the next two weeks.
The staff requested that the licensee continue its efforts to complete the equipment aging program (item 3) and the equipment operability time studies (item 4). The staff stated that until this work is completed it would be unable to conclude that the plant is is compliance with the EQ rule. The staff, however, observed that if these items were completed before the plant reached power levels where the design basis accident environment was credible, then the EQ rule compliance questions were moot. The licensee agreed to clarify by letter his plan and schedule for restart with an evaluation of the associated steam conditions that would affect EQ during each phase of the restart. The meeting was closed with the staff statement that it wished to be informed of the results of the licensee's studies for items 3 and 4 above prior to the plant returning to power operations.
/k Kenneth L. He tner, Project Manager Operating Reactors Branch No. 3, DL
Enclosures:
As stated cc w/ enclosures:
See next page
Enclosure I ATTENDANCE LIST FOR MEETING WITH PUBLIC SERVICE COMPANY OF COLORADO JULY 2, 1985 E0VIPMENT 00ALIFICATION ISSUES FOR FORT ST. VRAIN Phil C. Wagner NRC/ Region IV, Project Manager Ken L. Heitner NRC/NRR/DL, Oversight PM Richard E. Ireland NRC/ Region IV/ Chief, Special Projects Section Tom L. King NRC/NRR/ Chief, Advanced Reactors Group Bill Shields NRC Attorney A. S. Masciantonio NRC/EGN Sam Marquez PSC, Nuclear Projects Engineer James C. Selan PSC, Nuclear Licensing Dan Brown PSC, I&C Supervisor, Production Dave Toll PSC, Washington Representative James K. Tarpey PSC, Attorney Michael E. Niehoff PSC, Nuclear Design Manager Don Warembourg PSC, Manager, Nuclear Engineering
- 0. R. Lee PSC, Vice President, Elec. Prod.
Lawrence Brey PSC, Manager, Nuclear Licensing and Fuels Michael H. Holmes PSC, Manager, Nuclear Licensing John Reesy PSC, Staff Assistant Jack Kennedy G A Technologies Philip A. DiBenedetto Tenera/DBA Inc.
R. J. Burg Tenera D. D. Reiff ETA Technical Assoc.
Paul Shemanski NRC/NRR/DE/EQB Edward J. Butcher NRC/NRR/DL/0RB 3 W. G. Kenneriy NRC/NRR/DHFS Robert G. LaGrange NRC/NRR/DL/EQB Rudy 0. Karsch NRC/NRR/DL/0RAB Dennis L. Ziemann NRC/NRR/DHFS 2
Gus C. Laines NRC/NRR/DL l
James P. Knight NRC/NRR/DE Frank Miraglia NRC/NRR/DL l
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h)ib <..,- c. -j 2
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l ENVIRONMENTAL QUALIFICATION I
AT FORT ST. VRAIN Scope of Presentation l
Early E.Q. or E.Q. Related Programs IE-79-01B and DDR Guidelines 1
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e EARLY E.Q. AT FORT ST VRAIN NRC Question 6.1:
AEC Letter dated September 30,1970 (Morris to Walker)
"6.1 Question:
Identify all Reactor Protection and Engineered Safety Fetture equipment and components (e.g. motors, switchgear, cables, filters, pump, seal) located in the reactor or turbine buildings which are required to be operable during and subsequent to a loss-of-coolant (depressurization), feedwater line break, or a steam line break accident. Describe the qualification tests which have been or will be performed on each of these items to ensure their availability in the resulting environments (i.e.
helium temperature, pressure, humidity)."
PSC Responses were provided via:
- Amendment 18 to FSAR
j 2
EARLY E.Q. AT FORT ST. VRAIN FSAR AMENDMENT 25
- - Based on Gulf General Atomic Report Gulf-GA-A12045
" Qualification of Fort St. Vrain Safe Shutdown Equipment for Steam Environment Resulting from Pipe Ruptures" Steam Line Rupture Curves Were Developed Using the CONTEMPT-G Computer Code
- All Breaks Were Assumed to be Isolated Within 4 Minutes Amendment 25 was accepted by the AEC in Supplenent'l to the SER for the FSV FSAR June 12, 1973.
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l EARLY E.y. AT FORT ST. VRAIN EVALUATION OF THE CONSEQUENCES OF POSTULATED PIPE FAILURES OUTSIDE THE REACTOR BUILDING.
PSC Response provided by Amendment 26 to the FSAR Concluded that no single pipe failure outside the Reactor Building could prevent forced circulation cooling or PCRV liner cooling Concluded that no floodin: source could prevent safe shutdown cooling Pressure buildup was minimal (.5 psi maximum)
Amendment 26 was accepted by the AEC in Supplement 1 to the SER for the FSV FSAR June 12, 1973.
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i EARLY E.Q. AT FORT ST. VRAIN SEISMIC & ENVIRONMENTAL QUALIFICATIONS Fort St. Vrain had been restricted to 40% power while various commitments were being pursued.
(e.g. ACM, Fire Protection and Seismic & Environmental Qualification)
PSC submittal dated June 15, 1977 provided a report entitled i
" Seismic and Environmental Qualification Program Summary" and i
requested a release from the 40% power restriction. (P-77137)
This submittal provided the following information:
1 Described safety related instruments lists and safety related subtier list.
2 Defined 'the criteria for Class 1 equipment.
3 Defined the requirements for qualification of ' safe shutdown equipment.
4 Included Gulf General Atomic report Gulf-GA-A14212
" Environmental Temperatures in the Vicinity of the Rupture Point of Steam Lines for Fort St. Vrain. Equipment Qualification" which developed new curves-for qualification of equipment less than 20 feet from the point of rupture.
5 This submittal was based on the same assumptions as GA-
- A12045, i.e.
hot reheat line break in the turbine building and cold reheat line break in 'the reactor building were worst cases, and both were isolated within 4 minutes.
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_m__.-______m.
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EARLY E.Q. AT FORT ST. VRAIN AMENDMENT NUMBER 18 TO DPR-34 (FORT ST. VRAIN NUCLEAR GENERATING STATION) 1 An NRC letter dated October 28, 1977 provided Fort St. Vrain with (G-77076):
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(1) Amendment 18 to the Operating License
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(2) A Release to 70% Power (3) A Safety Evaluation Report
- THIS OCTOBER 28, 1977 SER INDICATED THAT:
PSC's definition of Class 1 items was acceptable.
The Office of Inspection and Enforcement had audited the detailed lists of safety related equipment and found them acceptable.
The Office of Inspection and Enforcement'had audited and accepted the environmental qualification documentation.
The time temperature conditions for items less than 20 feet from a steam line were based on a cold reheat line break in the reactor building and a hot reheat line break in the turbine building both lasting 4 minutes.
These assumptions were consistent with those found acceptable in the June 1973 safety evaluation.
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IE-79-01B & DDR GUIDELINES JANUARY 14, 1980 (G-80010)
- IE-79-01B 1
Requested a master list of all systems required to function under postulated HELB accident conditions.
Requested that Class 1E electrical equipment in the above systems, that is required to function, be listed in the format provided by Enclosure 2 of the Bulletin.
Requested that written evidence of environmental qualification to support the capability of each listed Class IE item identified above be provided in the format of Enclosure 3 of
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the Bulletin.
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Requested accident service condition profiles be provided as j
function of time.
]
Identify maximum flood level in containment.
Requested evaluation of the identified Class IE equipment against Enclosure 4 (DOR Guidelines).
I It is important to note that the D0R Guidelines specify that j
equipment outside containment should be qualified for the service conditions reviewed and approved in the HELB SER
)
(June,1973).
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IE-79-01B & D0R GUIDELINES 1
PSC ATTENDED THE REGION IV N7,C TASK GROUP / LICENSEE WORKSHOP MEETING WITH OTHER REGION IV UTILITIES ON
{
FEBRUARY 1, 1980 l
Scope of Bulletin was presented by Region IV It was stated that an onsite audit of installed equipment would take place 4
i Many questions presented by the utilities I
NRC ISSUED SUPPLEMENT #1 TO IE-79-01B ON FEBRUARY 29,1980 (G-80034)
Question and Answer format providing guidance to major questions that resulted from regional meetings l
PSC'S INITIAL RESPONSE TO IE-79-01B (P-80037 DATED 3/4/80)
I Requested an additional 14 days to provide the requested infonnation PSC was in the middle of document turnover from General Atomic Problems were encountered with remotely accessing General Atomic's computers (computerized Seismic /
Environmental Data Base)
B
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IE-79-01B & DDR GUIDELINES
- PSC'S SECOND IE-79-01B SUBMITTAL (P-80051 DATED 3/18/80)
Provided information formatted similar to Enclosure 2 of the Bulletin Fomat was verbally accepted by the NRC and was utilized in the remainder of our submittals Documented areas that were not applicable to an HTGR 9
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IE-79-01B & D0R GUIDELINES l
AREAS NOT APPLICABLE TO AN HTGR
- PRESSURE No containment building thus no storage of blowdown Turbine and reactor buildings are vented Pressure transients will be short term and localized
- RELATIVE HUMIDITY For the reasons stated above humidity will not be a problem
- CHEMICAL SPRAYS No chemical sprays are utilized at FSV
- RADIATION High energy line process fluids are not contaminated 1
Postulated 180 day Total Accumulated Dose from DBA #1 is 400 REM I
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- SUBMERGENCE Primary coolant is gas not water Emergency shutdown accomplished by i
boron ball insertion
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Water is not used for emergency core cooling l
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IE-79-01B & D0R GUIDELINES PSC'S THIRD SUBMITTAL TO IE-79-01B (P-80078 DATED 4/11/80)
Provided component evaluation worksheets formatted similar to Enclosure 3 of the Bulletin Format was verbally accepted by the NRC and was utilized in the remainder of our submittals Provided notification as to items previously accepted by NRC, i.e.
(1) General Atomic Report Gulf-GA-A14212 (2) Calculations for post accident releases of radioactivity as required by THI lessons learned I
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I PSC'S FOURTH IE-79-01B SUBMITTAL (P-80090 DATED APRIL 18,1980) l
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l Provided information on " Generic" items that hE have been tested TF#dwa >
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Stated that generic items had never been found l
to fail during an environmental test and thus I
were not considered significant
.I Indicated that review of emergency procedures l
was complete and the procedures were found to be adequate i
i Provided additional inf,rmation regarding i
component evaluation worksheets l
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L-__--__-_______-_______________________-__-__________-
i MAY 23, 1980 MEMORANDUM AND ORDER 80-CLI-21 l
i THE MAY 23, 1980 COMMISSION MEMORANDUM AND ORDER ADDRESSING 80-l l
CLI-21 ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT l-DIRECTED BOTH THE LICENSEES AND THE STAFF TO TAKE CERTAIN l
i ACTIONS.
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Licensees were to qualify all safety related electrical equipment to the DDR Guidelines or NUREG-0588 by June 30, 1982.
Replacement' parts were to be purchased to IEEE 323 1974 unless there are sound reasons to the contrary.
The staff was to devise a sy' stem to check that qualified equipment is installed with its qualification.
The staff was alrected to complete its review of i
environmental qualification including the publication of j
i safety evaluation reports by February 1,1981.
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- JULY 17, 1980 MEETING ON ENVIRONMENTAL QUALIFICATION AND THE COMMISSION'S MAY 23, 1980 MEMORANDUM AND ORDER 80-CLI-21 Staff discussed memorandum and order l
Status of qualification reviews was presented Fort St. Vrain was discussed with the NRC E.Q. Staff l
This discussion covered:
l (1)
The problems involved with applying the D0R Guidelines to FSV.
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(2)
The exceptions we had taken to date.
1 (3)
NRC Staff stated that they were not q
exactly sure how to handle FSV other j
than the E.O. schedule must be maintained.
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(4)
PSC's concern regarding a timely NRC response
'l was presented, i.e., if we receive no feedback l
I until the last minute, the schedule may be in jeopardy.
(5)
NRC Staff indicated that the FSV situation would be pursued by NRC via the P.M. and I&E.
A call from NRC Staff was later received indicating that. steam line rupture and aging l
were the areas that PSC needed to address.
I am a stp l
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i NRC FORT ST. VRAIN ORDER FOR MODIFICATION OF LICENSE DATED AUGUST 29, 1980 AND REVISED ORDER FOR MODIFICATION OF LICENSE DATED Sr.PTEMBER 19, 1980 (G-80144 and G-80159)
Referenced Commission's May 23, 1980 memorandum and order Stated that equipment be fully qualified by June 30, 1982 Required submittal of all information to support a safety i
evaluation regarding Environmental Qualification of
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Equipment by November 1, 1980 l
Reiterated that NRC staff was to complete publication of Safety Evaluation Reports by February 1, 1981
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IE-79-01B & DDR GUIDELINES i
PSC'S FIFTH SUBMITTAL ON IE-79-01B AND THE AUGUST 29, 1980 ORDER l
(P-80350DATEDOCTOBER3,1980)
PSC's comprehensive response.
Supplied information formatted similar to Enclosures 2 & 3 of the Bulletin l
l Reiterated our positions on areas previously I
presented as not applicable to an HTGR (i.e. pressure, relative humidity, chemical spray, radiationandsubmergence)-
l Requested that our response be evaluated with the design, operating, safety and HTGR and a LWR in mind g
g environmental differences between FSV's i
g Provided sound reasons to the contrary n
W for the procurement of replacement parts W
h Took exception to aging and provided a basis Q
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- REPLACEMENT PARTS " SOUND REASONS TO THE CONTRARY" IEEE 323 1974 was not based on FSV class HTGR Purchasing of spare parts needs to be consistent with the criteria established during the licensing k%
of the plant l
- AGING " BASIS FOR EXCEPTION" l
Radiation aging is definttely not a concern I
at Fort St. Vrain Since multiple redundant methods are available NW Q for maintaining cooling with forced helium circulation it is highly unlikely that a I
single line break could render all these methods inoperative l
s If for any reason forced circulation cooling i
could not be accomplished, cooling with Loss of Forced Circulation (LOFC) can be implemented with equipment outside the HELB accident environment l
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BASIS FOR EXCEPTION TO AGING j
FORCED CIRCULATION COOLING l
l AS A MINIMUM, FORCED CIRCULATION COOLING CONSISTS OF 1
Placing 1 Helium Circulator in Operation Providing Cooling to the EES or the. Reheater Section of One Steam Generator j
MULTIPLE EQUIPMENT COMBINATIONS AND FLOW PATHS EXIST FOR THE FORCED CIRCULATION COOLING METHOD ACRONYM DEFINITION
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Turbine Building HELB Environment j
TB2
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Reactor Building HELB Environment i
RX2
=
Outside of Plant OPL
=
d TABLE OF EQUIPMENT LOCATIONS STEAM GENERATOR EES SECTION OR MOTIVE POWER BEARING WATER I
REHEATER SECTION TO CIRCULATORS TO CIRCULATORS EQUIPMENT TB2 TB2 TB2 LOCATION OPL OPL/TB2 RX2 1
It is highly unlikely that a HELB could interrupt forced circulation cooling Multiple flow paths are available Valves (electric, pneumatic) are designed to fail open or to fail "As Is" 1
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BASIS FOR EXCEPTION TO AGING
SUMMARY
OF FORCED CIRCULATION COOLING REDUNDANCIES There are 2 Loops with 2 Circulators per Loop Only One Circulator is Required Source of Motive Power to Circulators (Each Circulator) location l
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1.
Feedwater via Emergency Feedwater (FW Pump)
TB2 Line 2.
Condensate via Emergency Condensate (4 Condensate Pumps)
TB2 Line 3.
Firewater via Emergency Feedwater (2 Fire Pumps)*
OPL Line 4.
Firewater via Emergency Condensate (2 Fire Pumps)*
OPL Line 1
l ACRONYM DEFINITION OPL Outside of Plant
=
TB2 Turbine Building HELB Environment RX2 Reactor Building HELB Environment
=
1 Diesel Driven Pump 1 Electric Pump
- EDG Backed and
- ACM Backed l
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BASIS FOR EXCEPTION TO AGING
SUMMARY
OF FORCED CIRCULAT10tl COOLING REDUNDANCIES J
Number of Cooling Loops 2
Number of Cooling Sections per Loop 2 (EES* & Reheater)
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Sources of Water to EES*
(Each Circulator)
- locatior,
(
1.
Nortnal Feedwater (1FWPump)
TB2 I
2.
Condensate via Emergency Feedwater (1FWPump)
TB2 Line 3.
Condensate via Emerge'ncy Feedwater (4CondensatePumps)
TB2 Line i
4.
Firewater via Emergency Feedwater (2FirePumps)**
OPL Line
.5.
Firewater via Emergency Condensate (2FirePumps)**
OPL l
ACRONYM DEFINITION q
Outside of Plant OPL
=
Turbine Building HELB Environment TB2
=
Reactor Building HELB Environment RX2
=
Economizer - Evaporator - Superheater 1 Diesel Driven Pump 1 Electric Motor Driven Pump
- EDG Backed and
- ACM Backed 20
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BASIS FOR EXCEPTION TO AGING
SUMMARY
OF FORCED CIRCULATION COOLING REDUNDANCIES Sources of Water to Reheater (Each Loop) location l
- 1. Condensate via Emergency (4 Condensate Pumps)
TB2 Condensate Line 1
- 2. Firewater via Emergency (2 Fire Pumps)*
OPL Condensate Line ACRONYM DEFINITION
)
OPL Outside of Plant
=
TB2 Turbine Building HELB Environment
=
RX2 Reactor Building HELB Environment
=
1 Diesel Driven Pump l
1 Electric Driven Pump EDG Backed and ACM Backed l
1 1
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l BASIS FOR EXCEPTION TO AGING l
l-COOLING WITH LOFC - PCRV LINER COOLING Redundant Two Loop System Requires Depressurization of PCRV Requires cooling water to the PCRV liner cooling tubes DEPRESSU.RIZATION Depressurization must be initiated within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the onset of an LOFC incident LINER COOLING Cooling water sources are available that are outside of the HELB environment (1) FireWaterPumps(1 Diesel,1 Electric)
(2) PCRV Cooling Water Pumps i
Bothfirewater&PCRVcooling(waterpumps have normal essential power, EDGbacked) and ACM power i
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- Main depressurization valves have both romal essential (EDG backed), ACM power and manual operators 1
Cooldown using the PCRV liner cooling system must be initiated within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> following the onset of an LOFC incident l
LOFC COOLING IS EXTENSIVELY DISCUSSED l
IN APPENDIX D OF THE FSAR 22 l
PCRV LINER COOLING SERVICE WATER COOLING Electrically Operated Location Relative Cooling Equipment to Accident Environment ACM Backup Required (HELB)
Power PCRV Cooling P-4601 Not in Accident Water Pumps Loop 1 or Environment Yes P-46015 P-4602 Not in Accident Loop 2 or Environment Yes P-46025 Service Water P-4201 Not in Accident Yes Pumps or Environment P-4202 Yes or P-4202S No Service Water C-4201X Not in Accident Cooling Tower or Environment Yes Fans C-4202X Circulating Water P-4118 Not in Accident Yes Makeup Pumps or Environment P-41185 l
One Loop is Sufficient for PCRV Liner Cooling i
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PCRV LINER COOLING CIRCULATING WATER COOLING i
Electrically Operated Location Relative Cooling Equipment to Accident Environment ACM Backup Required (HELB)
Power l
PCRV Cooling P-4601 Not in Accident Water Pumps Loop 1 or Environment Yes P-4601S P-4602 Not in Accident Loop 2 or Environment Yes P-4602S l
Circulating P-4101 Not in Accident Water Pumps or Environment No
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P-4102 or P-4103 or l
C-4104 Circulating Water P-4118 Not in Accident
.Yes Makeup Pumps or Environment P-4118S One Loop is Sufficient for PCRV Liner Cooling 24
PCRV LINER COOLING FIREWATER COOLING Electrically Operated Location Relative Cooling Equipment to Accident Environment ACM Backup Required (HELB) SEE NOTE 1 Power Firenater Pumps P-4501 Not in Accident Yes 1 Electric Driven or Environment 1 Diesel Driven P-45015 Circulating Water P-4138 Not in Accident
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Makeup Pumps or Environment Yes l-P-4118S One Loop is Sufficient for Liner Cooling l
NOTES:
1.
Includes Equipment, power supply, and cables 25 j
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I IE-79-01B & D0R GUIDELINES l
SUPPLEMENT #2 TO IEB 79-01B SEPTEMBER 30.1980(G-80161)
Provided question and answers' on' topics that arose during regional meetings held in' July 1980. -
Clearly stated that reactors in operation as of May 23, 1980 would be evaluated against the D0R Guidelines.
Clarified the definition-of' a
" central location". for qualification documentation.
SUPPLEMENT #3 TO IEB 70-01B OCTOBER 24,1980 (G-80175) j Additional clarification of Supplement #2 answers.
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FINAL NRC EQ ORDER TO PSC NRC FORT ST. VRAIN ORDER FOR MODIFICATION OF LICENSE CONCERNING ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT DATED OCTOBER 27, 1980 (G-80191)
(a)"By no later than June 20, 1982 all safety related electrical equipment in the facility shall be qualified in accordance with the provisions of:
Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DORGuidelines),or,NUREG-0588," Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," December 1979, to the extent appliceble to a gas-cooled reactor.
Copies of these documents are attached to order for modification of license No.
DPR-34 dated October 27, 1980.
b) "By no later than December 1, 1980 complete and auditable records must be available and maintained at a central location which describes the environmental qualification method used for all
{
safety-related electrical equipment in sufficient detail to j
document the degree of compliance with DOR Guidelines or NUREG-j 0588, to the extent applicable to a gas-cooled reactor.
i Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified."
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~IE-79-01B & DDR GUIDELINES J
In June 1981 PSC received a phone call from NRC' indicating 1
that a review of our. IE-79-01B, submittals was getting under way and that some additional information was required.'
On June 17, 1981, PSC provided a written response to the above mentioned questions about IE-79-01B (P-81168) that provided information. on the. following:
FSAR references to design basis events DBA1 & DBA2 j;
1-Response to DRL question 6.1 Gulf General Atomic Report GA-A12045' Evaluation of postulated pipe failures outside of the-ReactorBuilding(Amendment 26ofFSAR) 1 i
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IE-79-01B & THE D0R GUIDELINES' NRC Environmental Qualification Meetings July 7-10, 1981 The status of the Ft. St. Vrain SER was discussed with NRC staff
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it was stated that Ft. St. Vrain was a special case and was being handled on a separate schedule Ft. St. Vrain should receive their SER sometime in August 1981 29 J
10 CFR 50.49 10 CFR 50.49 WAS EFFECTIVE FEBRUARY 23, 1983 AND IT WAS PSC'S UNDERSTANDING THAT FT. ST. VRAIN FELL UNDER THE PROVISIONS OF PARAGRAPH'S K & L (k) Applicants for and holders of operating licenses are not required to requalify electric equipment important to safety in accordance with the provisions of this section if the Commission has previously required qualification of that equipment in accordance with " Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors," November 1979 (D0R Guidelines), or NUREG-0588 (For coment version), " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."
(1) Replacement equipment must be qualified in accordance with the provisions of this section unless there are sound reasons to the contrary.
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I 10 CFR 50.49 THE NRC LETTER DATED APRIL 13, 1983 (G-83161) REQUESTED THAT PSC SUBMIT:
1 A list of previous correspondence A discussion of methods used to identify equipment in paragraphb(2)
Also stated that a plant specific review of the previous Ft.
i St. Vrain submittals had recently begun i
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10 CFR 50.49 PSC FIRST RESPONSE TO 10 CFR 50.49 DATED MAY 17, 1983 (P-83178)
Summarized previous submittals Reiterated PSC's position on aging, radiation, pressure, etc.
Discussed various methods used to identify safety related electrical equipment.
Asserted that only Class 1 equipment needed to b'e considered.
Provided schedules for qualification of RG 1-97 equipment, testing of several items, & ! highlighting of safe shutdown electrical circuitry.
l FOLLOWING THIS INITIAL SUBM1TTAL HUMEROUS 10 CFR 50.49 STATUS SUBMITTALS WERE MADE:
PSC Letter dated August 15,1983(P-83280)
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PSC Letter dated January 9, 1984 (P-84012)
PSC Letter dated June 18, 1984 (P-84175)
PSC Letter dated December 18, 1984 (P-84527)
(1) Existing RG 1.97 equipment to be qualified by March 31, 1985 (2)NewRG1.97equipmenttot'equalifiedinaccordance with 10 CFR 50.49 as it applies to'a HTGR 32 u_-________________-______
10 CFR 50.49 A NRC/PSC E.Q. MEETING WAS HELD FEBRUARY 29, 1984 The NRC indicated that they were meeting with all licensees It was stated that the'SER on Fort St. Vrain submittals should be issued by September or October 1984 E1 It was also stated that onsite audits would take place VARIOUS TECHNICAL ISSUES WERE ALSO DISCUSSED The FSV E.Q. program was discussed in detail including assumptions made in applying 10 CFR 50.49 to Fort St. Vrain Fort St. Vrain operational characteristics and design basis events were presented Sample E.Q. records were presented and left with the Staff Operator response time and aging were discussed N: ' ;tated that our time / temperature vs distance approach was very conservative. The rest of the industry uses a bulk temperature approach
- NRC SITE VISITS NRC Site Visit April 3, 1984 PSC provided status update of PSC E.Q. Program Conducted tour of FSV NRC Contractor Site Visit June 19, 1984 PSC provided a status update of PSC E.Q. Program Conducted tour of FSV 33 L:
s.
10 CFR 50.49 THE NEXT CORRESPONDENCE FROM THE NRC REGARDING 10 CFR 50.49 WAS A REQUEST FOR CERTIFICATION OF COMPLIANCE TO 10 CFR 50.49 (GENERIC LETTER 84-24) (G-85022)
This letter was written December 27, 1984 but not received by i
PSC until January 21, 1985 A response was requested within 30 days of the date of the letter CERTIFICATIONS WERE REQUESTED AS FOLLOWS:
The utility either has inplace or will have an E.Q. program that will satisfy the requirements of 10 CFR 50.49 within the currently approved schedule without the need for further extension l
The plant has either one path to safe shutdown using qualified equipment or has submitted a JC0 pending full qualification of equipment not fully qualified That all other equipment within the scope of 10 CFR 50.49 is j
either fully qualified or a JC0 has been submitted The certifications were to address various IE Bulletins and Notices ONE WEEK LATER PSC RECEIVED A LETTER FROM THE NRC REGION IV, DATED JANUARY 28,1985(G-85041)
]
This letter requested a response within 60 days The letter indicated that PSC may fail to comply with the implementation deadlines imposed by 10 CFR 50.49 The letter contained requests for additional information:
34
4 10 CFR 50.49 January 28, 1985 Requests for Additional Infomation (G-85041) l l
- 1) Operator response time to steam line ruptures 1
- 2) The staff took the position that the aging requirc.ments of the D0R Guidelines do apply to FSV
- 3) Requested operability times of: equipment following an accident j
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- 4) The previous justification for not using replacement part qualified to 10 CFR 50.49 was inadequate 1
- 5) Requested JCOs
- 6) Requested confirmation of methodology utilized to identify equipment within the scope of 10 CFR 50.49 l.
(b)(2)
- 7) Requested confirmation that flooding was considered
- 8) Requested confirmation that all electrical equipment within the scope of 10 CFR 50.49(b)(3) is RG 1.97, Category 1 and 2 equipment or that a JC0 has been written i
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l 10 CFR 50.49 PSC LETTER DATED FEBRUARY 4,1985 (P-85033)
Requested extension on deadline to Generic Letter 64-24 PSC LETTER DATED FEBRUARY 28,1985(P-85065) l Identified our RG 1.97 equipment on schedule per a PSC commitment to NUREG-0737 Supplement 1 PSC LETTER DATED MARCH 25, 1985 (P-85103)
Response to Generic Letter 84-24
'i Summarized previous submittals Described the FSV E.Q. program i
Identified areas of concern not applicable to a HTGR j
including aging, radiation, etc.
1 Described methods used to identify equipment requiring E.Q.
All safe shutdown and RG 1.97 equipment was qualified in accordance with the FSV E.Q. program Responded to all IE Bulletins & Notices identified in Generic l
Letter 84-24 i
Certified that PSC is in compliance with 10 CFR 50.49 as it n
applies to a HTGR L
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- NRC/PSC E.Q. MEETING MARCH 26, 1985 Met to discuss compliance with 10 CFR 50.49 PSC reiterated our position of compliance g'
p It was agreed that. we were not in a position to file an g
extension request per the rules in 10 CFR 50.49 Technical issues were raised by the staff It was agreed to handle technical issues separately from the matter of compliance with 10 CFR 50.49 37 t.
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t PSC LETTER DATED MARCH 28,1985 (P-85112).-
Responded to each'of the staff technical concerns on the FSV E.Q. Program Reiterated that PSC was in compliance with 10 CFR 50.49 as it Gml applies to a HTGR If!S Provided a schedule for resolution of any technical issues.
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IN REGARDS TO OPERATOR RESPONSE TIME, PSC PROVIDED:
A detailed report "Four Minute Isolation of Postulated Steam Line Breaks at the Ft. St. Vrain Nuclear. Generating Stati.on A report from a human factors specialist that verified correct operator response within a 4 minute response time based on scenario walk throughs at the Ft. St. Vrain control room mockup A report from General Atomic on 10 minute temperature i
l profiles Copies of the Emergency procedures for steam line ruptures IN REGARDS TO AGING, PSC:
Stated that we still believed that our basis in regards to aging was sound Would resolve the NRC's aging concerns by establishing an aging qualification program IN REGARDS TO OPERABILITY TIMES, PSC:
Stated that operability times' would be established by November 30, 1985 IN REGARDS TO REPLACEMENT PARTS, PSC:
Provided justifications considered adequate by R.G.1.89 Stated that not all of the requirements-of 10 CFR 50.49 are applicable to the Ft. St. Vrain E.Q. Program IN REGARDS TO JUSTIFICATIONS FOR CONTINUED OPERATION, PSC:
Stated that we were in full compliance to 10 CFR 50.49 as it applies to the Ft. St. Vrain HTGR Stated that although there are certain technical issues that require resolution, that there were not items that required JCO's per the provisions of 10 CFR 50.49 39
.o IN REGARDS TO THE EQUIPMENT IDENTIFICATION METHODOLOGY OF 10 CFR 50.49(B)(2),PSC:
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' Stated that our safety related lists identifes our safe shutdown equipment Stated that this. list was based on the'FSAR,. Design Basis Events, P&I Diagrams'and Electrical Schematic Diagrams Stated that' an audit had been recently completed to insure that equipment required to mitigate HELB's was included in our safety related lists Stated that investigations.regarding power circuitry were underway and would be complete by November 30, 1985
'Electr.ical equipment required _ for proper operation of safe shutdown mechanical systems and equipment has been. qualified in accordance with our E.Q. Program Protective coordination studies exist for 480V switchgear breakers and protective relays Stated that a selective sampling of MCC breakers and fuses will be reviewed to confirm proper coordination by July 26, 1985 l
L IN REGARDS TO FLOODING, PSC:-
Stated submergence from in. plant water sources is not a problem
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Stated that Plant is above the 500 year flood level.
IN REGARDS TO R.G. 1.97, CATEGORY 1 & 2 EQUIPMENT, PSC:
l Stated that our March 25, 1985 letter (P-85103) confinned that our R.G. 1.97 equipment is qualified in accordance with our E.Q. Program.
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NRC/PSC E.Q. MEETING APRIL 3, 1985 Purpose of meeting was to discuss PSC's response (P-85112datedMarch 28.1985) 'to the January 28, 1985 letter (G-85041) including:
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1 4 minute operator response time 1
l Operability times
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1 Aging Request confirmation with methodology l
used to define 10 CFR 50.49' equipment 41 l
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10 CFR'50,49 NRC LETTER DATED MAY 7,
1985 RAISED QUESTIONS / CONCERNS IN THE FOLLOWING AREAS (G-85178):
- 1) Operator Response Time
- 2) Consideration of Automatic Isolation Systems
- 3) Requested the Expediting of PSC's Aging Program
- 4) Equipment Operability Times PSC LETTER DATED JUNE 11,1985(P-85197)
Responded to the NRC questions / concerns:
(1) Committed to a Pipe Rupture Detection System (2) Stated that Automatic Isolation would be considered (3) Indicated that Aging Program has been expedited (4) Stated that a study to determine operability times has begun Maintained the position that PSC is in compliance to 10 CFR 50.49 as it applies to a HTGR 42 L