ML20136J015

Notifies of Withdrawal of 850924 Schedule Extension Request for Environ Qualification Program.Approval of Revised Schedule Extension for Implementation of Program Beyond 851130 Under Provision of 10CFR50.49 Requested

ML20136J015
Person / Time
Site:	Fort Saint Vrain
Issue date:	11/22/1985
From:	Lee O PUBLIC SERVICE CO. OF COLORADO
To:	Chilk S NRC OFFICE OF THE SECRETARY (SECY)
References
P-85432, TAC-59787, NUDOCS 8511250258
Download: ML20136J015 (15)

Text

3 s

0rubiicservice-  : ": :.'

P.O. Box 840 Dene,co so20m November 22, 1985 Fort St. Vrain OSCAR R. LEE Un .1 VICE PRESIDENT Mr. Samuel J. Chilk, Secretary U.S. Nuclear Regulatory Commission 1717 H. Street N.W.

Washington, DC 20555 Docket No. 50-267

SUBJECT:

Revision of Request for Special Consideration and Schedule Extension, 10CFR50.49

REFERENCE:

PSC Letter Walker to Chilk dated September 24, 1985 (P-85334)

Gentlemen:

The Public Service Company of Colorado (PSC) withdraws the schedule extension request for the Fort St. Vrain (FSV) Environmental Qualification (EQ) ? mgram in the referenced letter, and hereby requests approval CY t revised schedule extension for implementation of the FSV Er F. #7- beyond November 30, 1985, under the provisions of paragraph c ' e,( .!FR50.49. This request contains a revised justification 'for continued operation which would restrict Fort St.

Vrain operation to 35 percent of rated power during a schedule extension period which would expire May 31, 1986. The exceptional circumstances that warrant this schedule extension to May 31, 1986 have not been revised, and are included herein for completeness.

Exceptional Circumstances

1) The FSV plant is a High Temperature Gas-Cooled Reactor (HTGR),

the only one of its kind in the USA. Because of this uniqueness, ,

the equipment qualification information sharing available to the i Light Water Reactor industry was in most cases not available or applicable to FSV because of the higher temperature harsh environments encountered during postulated FSV accident conditions.

g  ; g7 0 t I

t 2)';The EQ ' rule '(10CFR50.49) does not recognize the unique aspects

.., and design characteristics of an HTGR, and there is essentially no; HTGR EQ guidance available within the nuclear industry. This p imposed upon PSC the' added burden of determining the applicability of bulletins, NUREGs, and regulations on the EQ issue to the HTGR. In particular, the.NRC order for modification of the- FSV license concerning EQ, issued on October 27, 1980, required.PSC to determine and apply to the FSV plant the environmental qualification requirements in the D0R guidelines or NUREG 0588 "to the extent applicable to a gas-cooled reactor".

PSC did indeed make the required determinations and submitted to the NRC their understanding of how the EQ program would apply to an HTGR, and in particular to FSV. Attachment 1 to the referenced letter provides a summary of the chronology of significant events involving the development of the FSV EQ program which clearly documents that PSC has consistently applied its .best efforts in developing the FSV EQ p ogram. Since January, 1985, the FSV Environmental Qualification Program has greatly benefitted from written NRC staff guidance in applying the required regulations to the FSV HTGR.

3) Although the recent NRC staff guidance has been helpful, FSV is currently still in the unique and exceptional situation of never having received from the NRC a comprehensive Technical Evaluation Report (TER) or Safety Evaluation Report (SER), including required corrective actions, in response to the proposed FSV EQ program. These NRC evaluations would be of genuine assistance as evidenced :by the meaningful and constructive interactions with the NRC staff that have taken place since January 1985. As a result of these interactions, PSC has launched into an aggressive

. evaluation / corrective action program to resolve the various FSV EQ technical issues.

4) Since The early 1970's a basic assumption of the FSV EQ program has been-that plant operating personnel would be able to complete l

a number' of. manual actions to isolate a high energy line break during the first four minutes of an accident.* This position was  ;

reviewed and approved as part of Supplement No. 1 to the NRC SER for the FSV operating license in June 1973, and as part of Amendment No. 18 to the plant operating license in October 1977.

! In January 1985 PSC was notified that as a matter of NRC policy, this four minute operator response time assumption may no longer be considered valid. This notification constituted a fundamental change in the previously- approved licensing basis of the FSV plant.

t' i

- D

'Since January 1985, it is PSC's understanding that, under current

.. NRC policy, operators cannot be relied upon to ir.itiate required manual actions for at least ten minutes following the onset of an accident. Because a ten minute operator response time would be technically unacceptable due -to FSV's high temperature steam conditions and the resulting FSV environmental accident conditions, PSC has recently committed to the installation of an automatic Steam Line Rupture Detection / Isolation System. While this new system will resolve a number of FSV EQ technical issues, the related FSV EQ programmatic changes will require an extended period of time to implement.

Justification for Continued Operation Prior to the exceptional circumstances that arose in early 1985, the redundant safe shutdown, forced circulation cooling water systems using firewater were considered qualified to respond to the design basis Fort St. Vrain harsh environment accident conditions. However, with the modified requirements that resulted from the recent interactions with the NRC staff, not all of this equipment can be considered fully qualified to the requirements of 10CFR50.49. PSC has determined that in the interim period until all of this equipment can be qualified, the health and safety of the public can be adequately protected by placing a 35 percent power level restriction on the operation of the Fort St. Vrain plant. The following is a summary of the safety aspects associated with operation at power levels not to exceed 35 percent of rated power.

During this proposed interim period of restricted operation, PSC would maximize reliance on non-electrical equipment and systems which can be manually actuated to mitigate the consequences of harsh environment accidents, and therefore, does not require environmental qualification. With the 35 percent power level restriction, only the PCRV liner cooling system, using firewater supplied by pumps located outside the harsh environment, would be required to provide core cooling. Due to the high heat capacity and slow thermal response of the reactor core, a period of many hours is available before manual actuation of PCRV liner cooling would be required.

, With the 35 percent restriction, no fuel particle coating failures would occur during a harsh environment accident beyond those already assumed and analyzed for nonnal full power operation in the FSV Final Safety Analysis Report (FSAR) and the bases for the FSV Technical Specifications. Relying only on the PCRV liner cooling system for core cooling, the fuel particle coatings would maintain their required integrity throughout the various postulated harsh environment accident conditions. Additionally, in the event that the PCRV could not be depressurized during a postulated accident from 35 percent of rated power, calculations have shown that peak fuel temperatures would be further reduced from the depressurized peak

. fuel temperature case, thereby providing additional margin for assuring fuel particle integrity.

The PCRV reactor coolant pressure boundary steel liner would retain its integrity following the various postulated high energy line breaks from 35 percent power, and the PCRV concrete fission product containment boundary would continue to perform its safety function.

Since there is no breach of either the reactor coolant pressure boundary or the containment boundary, there would be no release of abnormal quantities of fission products to the Reactor Building atmosphere. Offsite doses caused by a high energy line break and subsequent PCRV liner cooldown fron 35 percent power would be negligible.

For reactor operation at 35 percent of rated power, all safety features of the FSV Plant Protective System are in effect, as required by the Technical Specifications. PSC considers interim operation with these plant safety features fully in effect to be most prudent to protect against the complete spectrum of possible accident conditions.

PSC considers that operation of FS" at reactor power levels not to exceed 35 percent of rated will estatlish the necessary conditions to meet the intent of 10CFR50.49 for interim operation of FSV in that:

A. The integrity of the reactor coolant pressure boundary can be maintained.

B. The reactor can be shutdown and maintained in a shutdown condition.

C. Offsite exposures remain well below 10CFR100 guidelines.

.1 The above . justification for continued operation with a 35 percent

. power level restriction has been discussed with the NRC staff and the results of the various analyses have been presented by PSC. However, 1PSC acknowledges that formal written evaluations confirming the above justification and conclusions will have to be submitted to the NRC staff for final review prior to plant operation up to the 35 percent power restriction.. Attachment 1 to this letter lists the detailed technical evaluations which must be documented and submitted to the NRC' staff for.their final review regarding the: safety benefits of the 35 percent power restriction.

Necessity For Schedule Extension PSC has been aggressively pursuing the environmental qualification of Fort St. Vrain safe shutdown, forced circulation cooling systems and equipment since the inception of the environmental-qualification concern in.the early 1970's -(see Attachment 1 to the. referenced letter). Much of this earlier work involved actual environmental qualification testing at steam temperatures and conditions much more severe = than those required for light water reactor equipment qualification. This environmental qualification testing work -was performed throughout the 1970s and early 1980s on the Fort St. Vrain l equipment and-systems requiring qualification. This extensive EQ L testing program was required by the higher temperatures and pressures present in Fort St. Vrain steam piping systems at the onset of a high -

energy line break accident.

However, all of this EQ testing work was based on accident

. temperature profiles which assumed a four minute operator response time, which had previously been reviewed and approved by the NRC on two occasions. Upon being informed of the current NRC position in January 1985, i.e., that the NRC is no longer-convinced that Fort St.

.Vrain operators can respond to a high energy line break in less than ten . minutes, the fundamental basis of the original Fort St. Vrain environmental qualification program was disallowed. Consequently building temperature ' profiles resulting from the high temperature steam conditions at Fort St. Vrain made it impossible to requalify l the required equipment for the environment resulting from a ten minute high temperature steam leak.

PSC has now committed to the design and installation of a Steam Line Rupture Detection / Isolation System (SLRDIS). However, additional time will be required to complete the design, licensing and installation of this automatic system. The impacts and effects of this altered direction on the Fort St. Vrain EQ program are substantial:

I l

l 4

1. Revised Reactor Building and Turbine Building temperature profiles must be developed for the spectrum of line break sizes which will be either: (1) automatically detected and isolated by SLRDIS,(2) detected and alarmed by SLRDIS but isolated by operator action after ten plus minutes, or (3) be small leaks

-which are detected by operators making their normal rounds and  !

isolated by operator action.

2. NRC concerns with Fort St. Vrain equipment aging and operability times must be resolved. The above mentioned temperature profiles must be developed before the aging and operability time work can be finalized. .

3. SLRDIS must be designed and approved by the NRC (SLRDIS will require FSV Technical Specification changes and appears to involve an unreviewed safety question). The installation of SLRDIS must then be completed and tested.

4. The 10CFR50.49 environmental qualification work on the new equipment being added for the SLRDIS must be completed.

5. The EQ program documentation and equipment qualification files must be revised to reflect the resolution of the above issues.

PSC has been actively pursuing the resolution and completion of the above work since early-1985 when NRC's concerns with the Fort St.

Vrain EQ program were identified. Attachment 2 to this letter contains additional details on the environmental qualification work currently in progress. Completion of this work will, however, require a schedule extension to May 31, 1986, which is hereby requested.

Sumary

• PSC has acted in good faith and with due diligence with regard to the EQ program for FSV. PSC is convinced that the FSV EQ program has had to confront unique and exceptional circumstances, which are deserving of a schedule extension. ,

• Fort St. Vrain can be operated up to 35 percent of rated power without compromising the integrity of fuel particle coatings or other fission product barriers with cooling provided by the liner cooling flow paths using fire water in the event of a harsh environment accident. The confirmatory evaluations of Attachment I will be provided to the NRC staff.

l i

• .PSC will continue to aggressively pursue all efforts required to assure that SLRDIS and the redundant FSV safe shutdown, forced circulation cooling systems are qualified to the requirements of 10CFR50.49 on or before May 31, 1986.

PSC would be pleased to provide additional information or respond to any questions you may have regarding this request for extension of the Fort St. Vrain EQ program schedule.

Very truly yours, 0.R. Lee, Vice President Electric Production ORL/MHH:jmt Attachments @{M '

y cc: Director, NRR, NRC Director, IE, NRC

Attachment 1

. to P-85832 CONFIRMATORY ACTIONS IN SUPPORT OF THE FORT ST. VRAIN 35 PERCENT POWER RESTRICTION DURING THE ENVIRONMENTAL QUALIFICATION SCHEDULE EXTENSION PERIOD PSC will complete and document the following actions to confirm the acceptability of a 35 percent power restriction during the Fort St.

Vrain environmental qualification schedule extension period to May 31, 1986. PSC will submit the results of these actions to the NRC staff for review and approval prior to resuming Fort St. Vrain plant operation at power levels up to the 35 percent power restriction.

1. Complete an evaluation which confirms that PCRV liner cooling using fire water can be utilized to prevent significant damage to any of the fission product barriers, including fuel particle coatings, in the event of a high energy line break at power levels up to the 35 percent power restriction.

2. Evaluate the leak tightness and structural integrity of the PCRV during the heatup which would occur following an extended loss of forced circulation cooling resulting from a high energy line break from 35 percent of rated power. Consider the cold reheat helium interspace leak, PCRV penetrations and seals, and other portions of the PCRV where leakage may be a concern. Actual PCRV leakage experience should be considered (e.g., the recent LER on the PCRV penetration cold reheat helium leak).

3. Provide high energy line brean temperature profiles for accidents from reactor power levels up to 35 percent demonstrating access to plant areas where operators are required to take the necessary manual actions by specified times. Identify all assumptions made regarding operator response times to the high energy line break, Credit will be taken only for qualified systems. An evaluation will be made for the need for SLRDIS for 35 percent power operation.

4. Provide an evaluation of the estimated time and primary coolant temperature beyond which the PCRV should not be depressurized during the remainder of a PCRV liner c,oldown following a postulated high energy line break from the 35 percent power level.

1

Attachment 1 to P-85432

, 5. Provide an estimate of the maximum reactor power level at which it would be safe to perform a pressurized PCRV liner cooldown while protecting the integrity of the PCRV liner and liner cooling system.

6. Submit a plan and perform inservice inspections on several critical areas of Fort St. Vrain's high energy piping to verify the integrity of this piping, prior to returning the plant to operation.

7. Evaluate the need for actuating the Fort St. Vrain reserve shutdown system for postulated high energy line breaks from power levels up to 35 percent. Determine the latest time by which manual actuation must be accomplished, if needed, and evaluate the feasibility of taking the required manual actions by this time.

8. Confirm that procedures for taking the necessary operator actions in response to a high energy line break from 35 percent power will be in place prior to resuming Fort St. Vrain plant operation. Evaluate the need for any chances to the Fort St.

Vrain Technical Specifications to accommodate Fort St. Vrain operation up to the 35 percent power level restriction, including the possible use of a pressurized PCRV liner cooldown in the event of a high energy line break.

9. Document the extent of PCRV damage expected during a liner cooldown from 35 percent power following, a high energy line break.

10. Evaluate the effect of a PCRV liner cooldown from the 35 percent power level on the previously analyzed PCRV hot spots.

11. Evaluate the impacts on the integrity of the PCRV liner cooling system of re-establishing liner cooling after prolonged periods of core heating without liner cooling or forced circulation cooling. Verify that the impacts of re-establishing liner cooling following a postulated high energy line break from 35 percent power with a pressurized liner cooldown would be no worse than those for the depressurized liner cooldown after 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> from full power previously evaluated in the Fort St. Vrain FSAR.

Compare the associated heat fluxes to the maximum acceptable heat flux to the PCRV liner cooling system assuming single loop operation.

12. If Fort St. Vrain operators will be required to take manual actions in environments whose temperatures exceed normal power plant operating temperatures to respond to high energy line breaks from 35 percent power, provide the NRC with information on the ability of operators to work in these higher temperature environments and the need for operators to utilize cool suits.

L

Attachment 1 to P-85432

, 13. Provide a sumary listing of which systems and their associated equipment items are considered qualified for 35 percent power operation, and a confirmation that no unqualified equipment items are needed.

14. Submit an engineering evaluation that describes the systems and equipment,which will be utilized to respond to a high energy line break from 35 percent power.

o

Attachment 2 to P-85432

1. Safe Shutdown Cooling Path Reviews and Evaluations In order to ensure full compliance with 10CFR50.49, PSC is performing an evaluation of safe shutdown cooling modes at Fort St. Vrain. The purpose of these evaluations is to independently identify all components required for safe shutdown cooling and to ensure that the electrical equipment in these modes is included in Fort St. Vrain's Environmental Qualification Program.

As described in the FSAR, two basic modes of core cooling are available at Fort St. Vrain, redundant forced circulation safe shutdown cooling systems and redundant PCRV liner cooling flow paths. Although there are various means to accomplish each of these modes of core cooling, certain systems and components have been designated as required for safe shutdown cooling.

Both Fort St. Vrain Operation's personnel and Engineering personnel are participating in the evaluation of these core cooling mode evaluations. Existing plant operating procedures are being utilized to identify systems and flow paths for these cooling modes. P&I's are being reviewed in detail to identify all electrical equipment in these systems and flow paths.

Finally, the electrical schematic diagrams are being reviewed to ensure that all electrical components are identified.

2. Field Walkdown Efforts A walkdown of safe shutdown equipment is being performed to verify the installed condition of the equipment and to ensure that the equipment installed is the equipment analyzed in the EQ program. Thus far this walkdown has identified a number of concerns such as contaminated junction boxes, unanalyzed taped splices, equipment model number discrepancies, and unanalyzed equipment installation configurations. All of these concerns are presently being addressed as part of the overall EQ program.

Corrective action is presently underway. Site forces are presently being supplemented to mount an aggressive field work program to accomplish corrective action as soon as possible.

l

Attacliment 2 to P-85432

, 3. Aging and Operability Time Studies PSC has been involved in an extensive effort to provide full qualification of our safe shutdown equipment, including aging and operability times. The method used for qualification is based on current industry practice, and is based on conservative ambient temperatures and High Energy Line Break (HELB) profiles. Any age sensitive material whose qualified life does not extend into the next refueling outage will be replaced before May 31, 1986.

The results of numerous preliminary aging and operability time studies have been submitted to PSC by our consultants and have been reviewed. The results of these studies are presented below in tabular form.

TABLE 1: STATUS OF AGING AND OPERABILITY TIME STUDIES TOTAL NUMBER TOTAL NUMBER OF ITEMS OF COMPONENTS l EQUIPMENT REQUIRED 156 1246 FOR SAFE SHUTDOWN EQUIPMENT REVIEWED 74 742 TO DATE EQUIPMENT FOUND TO 56 594 BE ACCEPTABLE NOTE: Several components may constitute one item; for example, if there are six valves of the same type, the valve type would be considered one item and the six individual valves would be considered six components.

The acceptability of the components identified in Table 1 are based purely on their ability to withstand a high energy line break. This number does not account for any additional field work required (e.g. field splices, cleaned junction boxes, etc.).

o Attachment 2 to P-85432

. The results are presented in preliminary form due to the preliminary nature of -the existing HELB profiles. Once the profiles are finalized, the results will be submitted to PSC in their final form. The qualification packages will be included in the FSV auditable files.

The aging and operability time studies have been impacted by the recent decision to install a steam line rupture detection / isolation system. These studies are being conducted utilizing the new preliminary steam line rupture detection / isolation system profiles.

4. Temperature Profiles Originally, the FSV Environmental Qualification Program was based on Reactor Building and Turbine Building temperature profiles

-that assumed a 4 minute isolation time based on operator action.

In order to resolve the 4 = minute versus 10 minute operator response time issue, PSC committed to install an automatic steam line rupture detection / isolation system. As a result the building HELB temperature profiles must be reanalyzed. Due to rapid isolation of the systems, preliminary scoping studies indicate that the new temperature profiles have a significantly reduced impact on the environmental qualification of electrical equipment. However, these profiles must be first evaluated and then submitted to NRC. The reanalysis of the temperature profiles is very critical to many of the programs and evaluations presently underway. The reanalyses, however, is time consuming.

PSC is proceeding as rapidly as possible to finalize these effects.

l i

3-4

~.

Attachment 2 to P-85432

, 5. Steam Line Rupture Detection / Isolation System The detection portion of the steam line rupture detection / isolation system is currently scheduled for delivery and installation by mid-December. Following installation, PSC will proceed with an extensive functional test to ensure the that system will not activate spuriously. PSC expects an in service date of May 31, 1986.

The steam line rupture detection / isolation system should resolve the 4 minute versus 10 minute operator response time issue for isolation of an HELB. This will greatly decrease the amount of energy that will be released into the reactor or turbine building, resulting in environmental qualification HELB profiles that have significantly reduced impact on the environmental qualification of electrical equipment.

6. Electrical Coordination Study An engineering evaluation was conducted to ensure selective fault clearing on the essential AC and DC load centers to confirm the safe operation of the plant in the event of a steam line rupture.

This review identified 2 breakers requiring setpoint changes, 1 circuit requiring rewire, 3 breakers requiring changeout and 5 valve breaker sections requiring additional fuse installation.

7. Auditable Files With the addition of aging and operability times, it is necessary to establish a new auditable file system for our EQ documentation. The new file system will be based on the criteria identified in IE Information Notice 85-39. Auditable files for the safe shutdown cooling equipment will be complete by May 31, 1986.

. Attachment 2 to P-85432

. 8. Other Reviews To complement the other systems reviews, the following activities are being performed.

(a) Identification of the electrical state (energized versus de-energized) of the electrical components required to achieve safe shutdown, both in normal plant operation and during the post steam line break state.

(b) Review and delineation of the function of the components required to achieve safe shutdown, both in normal plant operation and during the post steam line break state.

(c) Review of all the electrical circuitry for the components in Item (b) to assure that any subcomponent or circuitry involved is: (1) in the EQ Program; (2) that an analysis has been performed to determine that its EQ induced . failure will not compromise the components safe shutdown function; or (3) being added to the EQ program.

(d) Detennination of the anticipated cyclic operation of electricol components to include past and future operation, testing, calibration, and safe shutdown functions. This is being used as part of the aging evaluation and overall qualification review of each component.

All of the above evaluations, reviews or studies will be included as part of the overall EQ documentation by May 31, 1986.