ML19274E521
| ML19274E521 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/15/1979 |
| From: | Justin Fuller PUBLIC SERVICE CO. OF COLORADO |
| To: | Gammill W Office of Nuclear Reactor Regulation |
| References | |
| P-79058, NUDOCS 7903290169 | |
| Download: ML19274E521 (10) | |
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public service company oe cdlende March 15,1979 Fort St. Vrain Unit No. 1 P-79058 Mr. William P. Gammill Assistant Director for Standardization and Advanced Reactors Division of Project Management U.S. Nuclear Regulatory Comission Washington, D.C.
20555 Docket #50-267
Subject:
In-Service Inspection Program For Fort St. Vra'n
References:
(1) NRC letter to PSC, dated January 15, 1979; In-Service Inspection and Testing Program For Fort St. Vrain (2) PSC letter to NRC, P-78001, dated January 5, 1978; Evaluation of Amendment No. 18 SER Licensee Required Actions (3) PSC letter to NRC, P-78169, dated October 19,1978; In-Service Inspection -
Fort St. Vrain Gentlemen:
This is in reply to your letter of January 15, 1979, reference (1), which requested preliminary written responses to specific inquiries described as pri-ority items in enclosure (2) of that letter. The January 15, 1979 letter con-tained references to a PSC letter dated January 8, 1978.
PSC assumes that the NRC wishes to address PSC letter P-78001 dated January 5,1978, reference (2),
instead.
Each NRC inquiry in reference (1) is reiterated below, followed by the associated PSC response.
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Mr. William P. Gamill March 15,1979 Page 2 NRC Inquiry No. I Structural Graphite "Your proposed surveillance program for the oxidation of Type PGX graphite is being handled by separate correspondence.
Following approval of this program, we plan to include its implementation and any further development of surveil-lance needs for structural graphite in the overall in-service inspection program."
PSC's Response to Item No. I No response is required in this transmittal.
NRC Inquiry No. IIA PCRV Relief Valves "We have determined that the PCRV relief valves are equivalent to Category A valves of the ASME Code in that seat leakage in the closed pcsition is re-quired to be limited for fulfillment of their safety function. Technical Specification SR5.2.1, 'PCRV Overpressure Safety Systems' does not address or identify testing procedures consistent with this requirement.
Please explain how seat leakage is controlled following safety valve action and describe the surveillance tests which confirm this."
PSC's Response to Item No. IIA PSC does not agree with the NRC's classification of the two PCRV safety valves as Category A valves or that seat leakage in the closed position is re-quired to be limited for fulfillment of their safety function. These valves are defined as Category C valves in Article IGV-2000,Section XI, Division 2 of the draft ASME Code. The function of the valves in the Fort St. Vrain design is to provide ultimate protection against primary coolant system pres-sure exceeding the PCRV reference pressure of 845 psig. The only credible means of substantially increasing the primary coolant pressure to the safety valve setpoints is defined in the Fort St. Vrain FSAR to be steam generator subherder rupture.
Instrumentation, however, is provided that will scram the reactor and shutdown one or both primary coolant loops in a manner which stops system pressure from rising to the rupture disc and safety valve setpoints.
Both moisture monitors and pressure switches perform this safety function. Thus, the probability of rupture disc and safety valve action to relieve PCRV over-pressurization is very remote. With regard to controlling valve seat leakage following the rupture of a rupture disc and subsequent opening and reclosing of a safety valve, the following points are given to justify our position:
Mr. William P. Garmiill March 15, 1979 Page 3 1.
The probability of the relief valves lifting is very remote on the basis of the above discussion.
Prior to their lifting, gas leakage is controlled by the rupture discs.
2.
The safety valves were specified, manufactured and tested for exceptional seat tightness (bubble tight).
3.
Purified helium, at a pressure slightly greater than primary cool-ant pressure at all reactor power levels, flows through the PCRV safety valve piping upstream of the rupture discs, providing a purified helium buffer seal which prevents primary coolant helium from reaching the rupture discs and safety valves. Subsequent to rupturing a rupture disc and opening and reclosing of a safety valve, any small leakage through the safety valve would be com-posed of purified helium.
4.
In the event that a PCRV safety valve opens and fails to reclose, operator action to close the manual isolation valve (6-inch) up-stream of the failed rupture disc and safety valve will reduce the rate of PCRV depressurization. With one safety valve iso-lated, the second valve will continue to provide the safety func-tion. Leakage of primary coolant helium, if any, will be limited to the flow through a 1-inch crossover line and by the purified helium seal described under Item 3 above. All primary coolant helium escaping will then be filtered by the PCRV relief filter that will remove particulates at an efficiency of 99.99% in the 0.3 to 3.0 micron range.
In addition, an activity monitor that alarms in the plant control room is installed on the relief fil-ter outlet piping.
5.
Primary coolant leakage through the failed safety valve could hypo-thetically result in a reduction of reactor pressure. Unantici-pated reactor pressure reduction, however, would initicte an auto-matic reactor scram to shutdown the reactor.
Should the PCRV de-pressurize to atmospheric pressure, the offsite dose at the exclu-sion area boundary would be below 10CFR100 limits, and adequate core cooling could be maintained, even with primary coolant circu-lation provided by only one helium circulator.
6.
Primary coolant overpressure of a magnitude that lifts a PCRV safety valve must be the result of a major system upset and double failure of the Plant Protective System (PPS). Consequently, safety valve lifting would necessitate depressurization of the PCRV at which time the failed relief valve rupture disc would be replaced.
Mr. William P. Ganmill March 15,1979 Page 4 PSC believes that Technical Specification SR5.2.1 provides adequate testing to confirm the valves safety function and does not believe that periodic testing of these valves for seat leakage is necessary.
NRC Inquiry No. iib Loop Isolation Valves
" Capability for automatic closure and controlled leakage is required for the loop isolation valves in order to fulfill the ECCS functional requirements in the event flow from a helium circulator is lost. While the Technical Speci-fications do not describe provisions for the formal testing of these valves, their operability may be demonstrated at every occurrence of circulator shut-down by pressure indications from taps in the circulator ducts.
It is neces-sary to formalize the testing procedure and the reporting of any abnormal valve behavior or condition that may be observed.
Please propose a surveillance requirement that meets these objectives."
PSC's Response to Item No. iib PSC agrees that automatic closure of the helium circulator primary cool-ant isolation valves is necessary for single loop operation. Leakage through the isolation valves is required, however, to maintain a backflow condition in the shutdown steam generator to prevent steam generator damage resulting from contact with hot helium exiting from the core. Position of the isola-tion valves (open or closed) rather than closure tightness indication is im-portant to reactor operation.
the four (4)g differential pressure measuring instruments connected acrosscir Existin the position of the primary coolant isolation valves.
PSC plans to investi-gate the benefit of collecting this information for each valve when its asso-ciated circulator is shut down and will determine if incorporation into the Fort St. Vrain Surveillance Program is justified.
NRC Inquiry No. III Steam Generators "In a letter of January 8,1978 (J.K. Fuller to R.P. Denise), ' Evaluation of Amendment 18, SER License Required Actions', three types of inspections for the steam generators were listed.
Please provide a short summary of the status of each with inclusion of any available results.
Do the plans for tube removal and examination include inspection in regions where welds have been made between dissimilar metals? As stated in your letter, these programs will be reported under a DOE-funded program. However, we should be kept currently informed on thn e inspections. Furthermore, an outline for a Technical Specification or
Mr. William P. Gammill March 15,1979 Page 5 licensing-type document for surveillance of the steam generators should be proposed."
PSC's Response to Item No. III The following replies reference information in General Atomic GA-A series documents. As these documents were published under contract with the U.S.
DOE, it is assumed that the NRC has access to these documents. Thus, copies of the reference documents are not enclosed with this letter.
A.
Steam Generator Performance and Corrosion Surveillance:
AP instrumentation was installed in FY-76 as a result of a DOE-funded program through FY-78.
Funding for the surveillance work was disconti-nued in FY-79 as a result of a shift in emphasis to the HTGR Direct Cycle power plant. During the period when funding was available, the results of the surveillance were as follows:
1.
AP Instrumentation and Data:
Final evaluation of data in Section 2.3.2 of GA-A15066 showed a correspondence to calculated predictions.
Results of data analy-sis are available in Section 2.1.2 of GA-A14804 and Section 2.1.3 of GA-A14519.
2.
Temperature Instrumentation and Data:
Temperature readings taken during the period of surveillance demonstrated close agreement between measured and calculated thermal performance.
Results of data analysis are shown in:
Section 2.1.2 of GA-A14626 Section 2.1.2 of GA-A14386
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Section 2.1.2 of GA-A14246 Section 2.1.2 of GA-A14154 3.
Feedwater Chemistry:
With the exceptions of reactor shutdown, low level power and a few brief periods at higher power levels, the feedwater chemistry was within specification limits.
Results of data analysis are shown in:
Section 2.1.2 of GA-A14962 Section 2.1.3 of GA-A14804 Section 2.1.3 of GA-A14626 Section 2.1.4 of GA-A14519 Section 2.1.3 of GA-A14386
Mr. William P. Gamill March 15,1979 Page 6 Section 2.1.3 of GA-A14246 Section 2.1.3 of GA-A14154 B.
Steam Generator Tube Removal and Physical Examination:
Under a DOE-funded program sections of one or more subheader tubes were to be removed and physically examined after approximately, but not exceed-ing, five years of service. DOE funds supporting this program have been withdrawn, preventing any further tube removal for examination purposes.
Although not part of the DOE-funded program, in November,1977, sections of tubing leading to and from a leaking steam generator tube were removed from Module B-1-1, as part of the tube plugging operation. The removed tube sections were cut into samples and inspected.
A comparison between the Sanicco 31 tubing (superheater lead-out) removed from steam generator between the secondary closure and super-heater outlet ringheader, and virgin material, revealed no damage.
Slight pitting was observed on the SA-210-Al material (feedwater lead-in) when it was examined. Since the tubing had been in service for a short period of time, and the pitting was slight, a conclusion could not be drawn as to the significance of the pitting.
C.
Steam Generator Ring Header Valve Disassembly and Examination:
To date, two ring header valves and a strainer have been disassembled and examined.
During June,1977, feedwater inlet valves of Module B-1-1 and the feedwater ringheader strainer of Module B-1-5 were disassembled and inspected.
Examination of the strainer assembly yielded the following observa-tions:
(1) No particulate matter of any detectable size was found in the strainer, (2) no evidence of scale deposits were found, (3) no indication of wear, (4) all interior wall s-faces were coated with a surface layer of magnetite, and (5) small spotty areas of rust were found on either side of the feedwater inlet nozzle to the ringheader.
Sulzer valves A and T of Module B-1-1 were dissassembled and inspected.
The results of that inspection were:
(1) Both valves were clean with a uniform surface layer of magnetite, (2) the interior walls of the inserts showed no indication of wear, (3) seat lines were clean and continuous, (4) there was no evidence of bypass flow, (5) no signs of wear or corrosion.
Additional information is supplied in Section 2.1.5 of GA-A14591 and Section 2.1.4 of GA-A14804.
Mr. William P. Gammill March 15,1979 Page 7 D.
Dissimilar Metal Weld Inspection:
No dissimilar metal welds in the neat affected zone will be examined as dissimilar metal welds in this zone are not accessible for removai.
E.
Proposed Technical Specifications Surveillance Outline:
No changes to the Fort St. Vrain Technical Specifications are proposed by PSC. Surveillance of the steam generator heat transfer surfaces is per-formed presently on a continuous basis.
Primary coolant moisture monitors and activity monitors in the hot reheat section of the steam generators continuously monitot for tube deterioration.
PSC is currently developing an In-Service Inspection Program which will assess the desirability and feasibility for additional steam genera-tor in-service inspection. Any additional ISI inspection will be con-ducted under this Program rather than as a Technical Specification re-quirement.
NRC Inquiry No. IV Prestressed Concrete Reactor Vessel "Also in the January 8 letter, mention was made that strain gage, tendon load cell and deflection data were being obtained for the PCRV, and that the liner was being monitored for corrosion. Please provide a short summary of the status of these surveillance activities with the inclusion of any available results."
PSC Response to Item No. IV PCRV structural response surveillance has been funded by the DOE from FY-76 through FY-79. Data acquisition and analysis is continuing and PCRV deflection measurements are planned during the FSV rise to power after the first refueling. The results of the PCRV surveillance to date are:
A.
Strain Gage and Tendon Load Cell Data:
PCRV sensor data collected during reactor rise to 70% power level with vessel pressure to 659 psia have been reduced and analyzed.
Changes in concrete strains and tendon loads due to vessel pressurizations were found acceptable. The recorded strains were in general agreement with predicted values obtained from elastic three-dimensional finite element analysis.
f
Mr. William P. Gammill March 15,1979 Page 8 The PCRV time-dependent structural analysis was extended to cover the period from initial proof test pressure test to pressurizations during reactor rise to power. Assessment of analytical results indicates that the creep analysis over-estimated creep recovery resulting from pressuri-zations.
Additional information is provided in:
Section 2.4 of GA-A1A154 Section 2.2 of GA-A14246 Section 2.2.1 of GA-A14386 Section 2.2 of GA-A14519 Section 2.2 of GA-A14804 Section 2.2 of GA-A14962 Section 2.4 of GA-A15066 B.
PCRV Deflection Data:
PCRV optical deflection measurements were performed after the first year of reactor operation.
Deflection measurements were taken at vessel pressures of 35,191, 375, and 600 psig. The measured deflections were compared with analytical results and found acceptable. The acceptance criteria are based on calculated deflections from elastic stress analy-sis with error bands for both calculational and measurement inaccuracies as established for the initial proof test pressure tests. The magnitude of creep was consistent with that observed during the initial proof test pressure tests.
Additional information is provided in Section 2.4 of GA-A14042 and Section 2.2.2 of GA-A14626.
C.
Liner Cooling Tube Corrosion Data:
PSC assumes that the NRC inquiry regarding the PCRV liner being moni-tored for corrosion refers to liner cooling tube corrosion, since that subject was mentioned in the letter referenced by the NRC, reference (2).
During the period of 1976-1977, the corrosion rate of the liner cooling tubes as measured by a corrosometer was found to be 0.3 mils / year or 1/3 of the maximum allowable rate of 0.9 mils / year. This should be considered an interim PSC response since more recent data will be avail-able shortly.
PSC will provide the NRC an updated report on liner cool-ing system monitoring by April 25, 1979.
Mr. William P. Gammill March 15,1979 Page 9 NRC Inquirv No. V Valves Required for Safe Reactor Shutdown "We are investigating the potential benefits of revising the in-service testing requirements for safety related valves in the secondary cooling system and other systems required for safe shutdown of the reactor.
In this regard, our recomendation would be that you conform with the ASME Code in this area insofar as practical. Justified exceptions would be recognized for cases where full conformance could not be reasonably achieved. Please coment on any diffi-culties you would foresee in this approach."
PSC Response to Item No. V PSC is also investigating the in-service testing requirements for safety related valves required for safe shutdown of the reactor as part of our overall ISI program development.
PSC outlined in reference (3) the approach being taken in our current development program, which includes the development of safety categories and boundaries that are realistically related to the safety role of systems and components and which consider the differences in design philosophy between Fort St. Vrain and Light Water Reactor Plants. This approach will categorize each valve as to its importance to plant safety.
For those valves which may need testing, based on their importance to safety, PSC plans to meet the inspection requirements of the ASME Code, insofar as practi-cal.
PSC will be available to meet with the NRC to discuss the above responses.
PSC is continuing to develop an in-service inspection and testing pro-gram for Fort St. Vrain in that Section XI, Division 2 of the ASME Code is in most cases not applicable to Fort St. Vrain.
The program is focusing on devel-oping criteria to define safety classifications that will categorize specific system elements and boundaries relative to their safety role. The classifica-tion of each system element will be based upon its function in mitigating the consequence of design basis accidents and other analyzed plant accident cases as defined in the FSAR. The safety criteria, inspection / testing requirements and system elements assigned to each safety class will be defined. The program will be based on, developed from, and enhance the Fort St. Vrain Technical Specification Surveillance Testing Program.
PSC will be prepared to meet with the NRC by the end of July to report the results of this work currently in progress.
There are uncertainties that make it difficult to estimate a completion date for our program at this time. PSC plans to provide the NRC with a program
Mr. William P. Gamill March 15,1979 Page 10 implementation date for new inspection / testing requirements subsequent to NRC concurrence with our proposed program, which will be defined by the end of July.
Very truly yours,
.. Fuller, Vice President Engineering and Planning JKF/MLP/ RAS:ler
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