ML20050D358
| ML20050D358 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/29/1982 |
| From: | Brey H PUBLIC SERVICE CO. OF COLORADO |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| P-82061, NUDOCS 8204120153 | |
| Download: ML20050D358 (32) | |
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public Service Company Cf Onkndo 5
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Unit No.1
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P-82061 (Ji/-63 x's 3
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Mr. Robert Clark, Chief Operating Reactors Branch 3 Division of Licensing Office of Nuclear Reactor Regulation Nuclear Regulatory Commission Washing ton, 0.C.
20553 Docket No. 50-267
Subject:
Fort St. Vrain Inservice Inspection and Testing 5
References:
(1) Los Alamos Letter Q-13:82:5, G-82014 Charles A. Anderson to Robert Clark, dtd January 5, 1982 (2) P-80014 (3)P-80034 (4) P-80064 (5) P-80218 Gentlemen:
PSC has received the above referenced report reviewing the Public Service Company of Colorado's proposed Inservice Inspection and Testing Prog ram, prepared for the Nuclear Regulatory Commission by Los Alamos National Laboratory (LANL) and their consultant, Advanced Science and Technology Associates (ASTA).
It appears that LANL and ASTA generally agree with the modified program proposed by PSC in their submittals for priority category I oY7 systems and components.
There are, however, instances where the f
report concluded that specific information be provided to the NRC, or that PSC further investigate the applicability of additional jf examinations outlined in the report.
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r P-82061 Page 2 March 29, 1982 Encl osed are PSC's responses to the recommendations contained in the LANL/ASTA report.
PSC is now awaiting an official NRC response or comments on the referenced Category I submittals, the LANL Review report and this response to the LANL report in determining the adequacy and acceptability of the approach being used to develop the FSV ISI requirements.
PSC is also prepared to meet with the NRC staff to discuss the proposed ISI program and the responses to the LANL/ASTA review as mutually agreeable.
Please direct any questions you may have to Mr. Mike Holmes, (303) 571-6711.
Very truly yours,
_o_u.utww l6.a]
H. L. Brey, Manager Nuclear Engineering Division HLB /MAJ:pa Enclosure cc: C. A. Anderson, LANL
PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION TECHNICAL SPECIFICATION SURVEILLANCE PROGRAM l
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RESPONSE TO THE RECOMMENDATIONS OF LOS ALAMOS NATIONAL LABORATORY Report 0-13:82:5 I
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February 19, 1982
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GENERAL COMMENT
S i
COMMENTS ON THE SCOPE OF THE LANL/ASTA REVIEW A.
The introduction of the LANL/ASTA Report (reference Q-13:82:5 dated January 5, 1982, later referred to herein c
as the Report) states that three submittals of modified FSV Technical Specification Requirements have been trans-mitted to the NRC for review (February 8, March 3 and March 31, 1980).
It should be noted that PSC has submitted four review packages to the NRC for review.
The first three, identified above, cover all of the systems that were previously identified in PSC's letter of November 30, 1979 as Category I priority for review and these have been included in the LANL/ASTA review effort.
The fourth sub-mittal (July 16, 1980) covered all of the reactor auxiliary cooling water systems which were identified as Category II priority and was apparantly not' included in the scope of the LANL/ASTA review effort.
B.
Appendix A, Section A-4 of the Report provides comments and recommendations concerning the Helium Purification System and states that no SR covering this category I system was identified nor was it discussed by PSC in the category I submittals.
This system was identified as Category II priority for review in PSC's letter of November 30, 1979 to the NRC and therefore was not included in the Category II submittals.
Surveillance Requirements for this system have not been reviewed as yet but are scheduled as part of future Category II submittals.
During the normal course of the review effort, PSC will consider the comments and recommendations provided in section A-4 of the Report l
and, therefore, no specific response is included herein.
C.
The list of Surveillance Requirements on page iv of the Report includes SR5.2.19, IACM Diesel-Driven Pumps.
This SR was not included in the scope of the review and, therefore, t
I should not have been addressed in the LANL/ASTA report.
It had already been deleted by Amendment 21 to the Operating License.
I-page 1
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COMMENTS CONCERNING USE OF THE PROPOSED ASME CODE SECTION XI -
DIVISION 2 As part of the review of the Fort St. Vrain surveillance pro-gram, it was agreed that PSC would use the above referenced Code as guidance only and would compnre recommended surveil-lance requirements against proposed Code requirements.
Contrary to what may be understood from the Report, PSC did not request or invoke exemptions under the proposed Code as a justification for their recommended surveillance requirements.
However, when PSC considered that proposed Code exemptions were applicable, it was so stated in the comparison of recommended surveillance requirements and proposed Code requirements.
There appear to be differing views between LANL/ASTA and PSC concerning interpretation of Code exemption articles in the context of the specific features of the Fort St. Vrain Nuclear Generating Station.
Some of the most significant differences are discussed further in other sections of this response.
Throughout ehe LANL/ASTA Report, the " intent of the proposed Code" is 2 red to.
However, that " intent", as interpreted by LANL/AS' is not explicitly clarified for Fort St. Vrain.
It is PSC's asition that certain proposed Code requirements, e
which apply to large HTGR designs, do not adequately take into consideration unique features of the Fort St. Vrain design.
This position was previously stated in PSC's letter of October 13, 1978.
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page 2
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1 SECTION 1 - COMMENTS RELATIVE TO PCRV AUXILIARY SYSTEM 1
1.1 SR 5.1.2 - Reserve Shutdown System Surveillance Recommendation -
The Report recommends that NRC approve the changes proposed by PSC to SR 5.1.2.
PSC Resconse -
l None required.
i 1.2 SR 5.2.1 - PCRV and PCRV Penetration Overpressure Protection Surveillance Recommendation -
q The Report recommends that NRC approve the specific changes proposed by PSC to SR 5.2.1.
The Report fur-ther recommends that additional examination requirements be included in SR 5.2.1 for structural component welds and bolting.
Specifically, the Report recommends that the following additional requirements be included:
4 i
o PCRV safety piping primary boundary visual examination i
of all accessible welds, including integrally welded support attachments.
Examinations of bolting and 1
support components.
4 Safety tank pressure boundary visual examinations of o
3 all accessible welds, integral attachments and support components.
Tank boundary pressure testing.
Torque or tension testing of bolting not normally disturbed for access to tank internals.
PCRV penetration overpressure protection pipe and o
valve visual examinations including welded attach-1 ments and support components.
(
j PSC Response -
PSC rationale in support of the proposed surveillance i
requirements does not take the position stated in the i
Report that "the primary and secondary boundaries are page 3
c excluded from volumetric and surface weld examinations by the exempted conditions described by the proposed Code, subsection IGB and IGC-1220."
PSC's rationale is based on the ability to continuously monitor leakage of primary and secondary boundaries within predetermined and acceptable limits specified in the plant Technical Specifications.
Additional instrument functional and calibration tests have been recommended in PSC's pro-posal to enhance the effectiveness of that method.
Code exemptions are referred to by PSC in the comparison of recommended surveillance requirements and proposed Code requirements, as they are considered to apply to Fort St. Vrain.
The Report recommends additions to PSC's program in order to be " consistent with planned rules development for secondary containments."
It is PSC's position that the above referenced rules do not apply to Fort St.
Vrain for the following reason.
In large HTGR designs, a containment is required to prevent the release to the environment of unacceptable quantities of radioactivity in the event of a depressurization accident caused by the postulated failure of a single primary closure.
t For Fort St. Vrain, the same depressurization accident requires, as an initiating event, the postulated failure of both primary and secondary pressure boundaries.
Therefore, the secondary closures, unlike a containment, do not mitigate the consequences of a depressuri:ation accident.
Analyses demonstrate that the exposure, re-sulting from this design basis depressurization accident, is within the guidelines of 10CFR100 even with unfiltered i
release to the environment being postulated.
There do not appear to be identifiable technical concerns with respect to primary and secondary pressure boundaries of the PCRV safety valve system or the PCRV penetration overpressure protection system.
PSC does not consider that visual examination of accessible welds every 10 years would provide enhanced plant safety compared to that resulting from continuous leakage monitoring and periodic leakage testing.
Similarly, PSC does not censider that pressure testing the safety valve tank every 10 years would provide enhanced plant safety ccmpared to the periodic leakage test performed every 2.5 years and considering that the tank does not experience any signi-ficant loading conditions during plant operation.
pace 4
With respect to equipment supports, it was stated in the PSC submittal that supports were not included in the review.
PSC takes note of the recommendations for visual examination of accessible support welds and torque or tension testing of accessible support b,olting, and will consider these recommendations when surveillance requirements for component supports are reviewed.
1.3 SR 5.2.15 - PCRV Penetration Interspace Pressure Surveillance Recommendation -
The Report recommends that NRC approve the PSC proposal that no changes be made to the existing SR.
t PSC Response -
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None required.
1.4 SR 5.2.16 - PCRV Closure Leakace Surveillance Recommendation -
The Report recommends that NRC approve the changes proposed by PSC to SR 5.2.16.
PSC Response -
None required.
pace 3 I
SECTION 2 - COMMENTS RELATIVE TO THE PRESTRESSED CONCRETE REACTOR VESSEL (PCRV) 2.1 SR 5.2.2 - Tendon Corrosion and Anchor Assemblies Surveillance 2.2 SR 5.2.3 - Tendon Load Cell Surveillance 2.3 SR 5.2.4 - PCRV Concrete Structure Surveillance 2.4 SR 5.2.5 - PCRV Liner Specimen Surveillance Recommendation -
The Report recommends that NRC approve the changes proposed by PSC to these SR's.
PSC Response -
None required.
2.5 SR 5.2.13 - PCRV Concrete Helium Permeability Surveillance 2.6 SR 5.2.14 - PCRV Liner Corrosion Surveillance i
l Recommendation -
l l
The Report recommends that NRC approve the PSC proposal l
that no chances are required to be made to these SR's.
1 PSC Response -
None required.
page 6
e 2.7 SR 5.2.14 - Refuelinc Penetration Holddown Plate Surveillance Recommendation -
The Report recommends that NRC approve this new SR proposed by PSC.
PSC Response -
None required.
2.8 PCRV Penetrations and Closures See comments in Section A-1 of this response.
page 7
J SECTION 3 - COMMENTS RELATIVE TO THE PCRV INTERNALS 3.1 SR 5.2.22 - PGX Graphite Surveillance Pecommendation -
The Report concludes that the existing SR, as clarified by PSC, responds meaningfully to a particular problem concerning the core support blocks and therefore should continue in effect.
The Report further recommends that requirements for additional graphite material specimens be included in SR 5.2.22.
Specifically, the Report recommends that PGX and ATJ material specimens, repre-sentative 'of each type of support component, be included in an expanded scope for core graphite surveillance specimen placement and testing.
PSC Response -
The Report notes that the PSC proposed requirements for SR 5.2.22 nodify the current requirements by the dele-tion of "non-destructive" examination from the range of tests on specimens removed from the core.
PSC considers that deletion of the term "non-destructive" is proposed by PSC only to reflect that the actual tests, as originally proposed and approved, are destructive in nature.
Thus the change is a clarification only and is not intended to modify existing requirenents.
The bases for the Report recommendation to expand the scope of core support graphite material surveillance specimens are the realirenents more recently adopted by the ASME in Subsection IGI of the Code.
These ASME requirements were adopted to allow potential variations in structural graphite properties to be monitored over the service life of the components in large HTGRs.
The existing PGX material specimens were installed to address a technically justified concern and are capable of being tested only to verify that oxidation profiles are as expected.
Although an expanded specimen program could provide additional assurance that no unforeseen degradation might be occurring, the Report did not identify specific technical concerns which would justify the cost of reactor nodifications to include additional PGX and ATJ graphite specimens.
Furthermore, the feasibility of ins talling specimens of a configuration page 8
suitable for mechanical testing in appropriate locations has not been demonstrated for the Fort St. Vrain reactor.
PSC does not consider that the rules and regulations of 10CFR50.55 for inservice inspection (even though they only apply to LWR's) would require a plant to be modified to accormodate requirements included in Code editions which were not applicable for design and construction.
More specifically, it is PSC's understanding that the ASME never intended that the requirements of Subsection IGI would apply to Fort St. Vrain because of the lack of design provisions to accmmodate these requirements.
Design analyses demonstrating the adequacy of the core support component design and materials are referenced in the FSAR and in documents supporting Amendment 20 to the Operating License.
These analyses, the current specimen program, and other research efforts were deemed adequate by the NRC to ensure the structural integrity of graphite support components at Fort St. Vrain.
No recent development, either operating experience at the plant or further research efforts, has changed, or raised a concern about, the v.alidity of this conclusion.
PSC will, of course, continue to monitor any new findings concerning PGX and ATJ graphite, and would consider the use of, and investigate the feasibility of installing, other material specimens if there were actual technical concerns that the graphite core support structures might not be capable of performing their function for the entire life of the plant.
3.2 SR 5.2.25 - Core Support Block Surveillance Recommendation -
The Report recommends that NRC defer approval of this new SR proposed by PSC pending the results of further study to determine if additional areas can be included.
Specifically the Report recommends that, in addition to remote visual examination of selected core support block top surfaces as proposed by PSC, the use of this exami-nation method be investigated to include ccmponents located within the core cutlet plenum where practicable.
page 9
i PSC Response -
PSC has_previously reviewed the practicability of emote visual examination of the core outlet plenum.
l It was reported in PSC submittals that there was no practical access for examination without major. reactor i
i modi fi ca tions.
The only existing access would be through the coolant channels in the core support blocks.
Due to their configuration, a direct path from above is j
not available.
The risks involved in installing and removing remote equipment through the coolant channels is not considered to be worth the limited examination results possible within the plenum.
The large HTGR ASME Code requires special provisions on large HTGR designs to allow general viewing of core outlet plenum components.
Fort St. Vrain does not have any of these provisions in its design nor would it be feasible to backfit such provisions.
Furthermore, as previously stated, PSC does not consider that NRC rules and regulations and the ASME Code would require a plant to be modified to accommodate requirements in Code editions which were not applicable for design and construction.
PSC does not consider remote visual examination of the core o u tl e t plenun to be practicable, nor would it result in benefits that would justify the cost and risks; therefore, further investigation is not warranted.
1 3.3 SR 5.2.26 - Region Constraint Devices Surveillance A
Recommendation -
The report recommends that NRC approve this new SR j
proposed by PSC to cover inspection of these recently installed reactor components.
i i
PSC Response -
i None required.
3.4 PCRV Thermal Barrier See comments under Section A-2 of this response.
j 3.4 Core La teral Restraint I
i See comments under Section A-3 of this response.
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t page 10 i
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SECTION 4 - COMMENTS RELATIVE TO PRIMARY COOLANT SYSTEM CIRCULATORS 4.1 SR 5.2.17 - Helium Circulator Pelton Wheel Surveillance Recommendation -
The Report recommends that NRC approve the proposal by PSC to delete SR 5.2.17 from the Technical Speci-fications.
PSC Resconse -
None required.
4.2 SR 5.2.18 - Helium Circulators Surveillance Recommendation -
The Report reccmmends that NRC approve the changes proposed by PSC to SR 5.2.18 relative to inspection of circ alator operational components.
The Report also recommends that additional examination, require-ments be included in SR 5.2.18 for the helium circulator penetration interspace boundaries and structur':s.
Specifically, the Report recommends that the following examinations be included and performed at the same time a circulator is examined:
o Penetration primary boundaries forming part of circulator casings, or performing support functions or other functions vital to circulator structural integrity, be examined.to the maximum extent practicable.
This to include surface examination of accessible welds at structural discontinuities in representative areas of pressure-retaining boundaries, including support attachment welds, and visual examinations of bolting.
o As a surveillance requirement for PCRV structures, penetration secondary boundaries (shells and clo-sures) be examined in accordance with Report recommendaticns included in Appendi:: A, item A-1.
page 11
PSC Response -
Heliun Circulator PCRV Penetration Primary Boundary and Circulator support (Dwgs R1100-100 Sh5, C2101-300 Sh1 and 2, ll-R-3-84; see attached schematic)
The helium circulator machine consists of four major components assembled by bolting:
a support cone (C2101-301), a bearing assembly (C2101-500), a steam ducting assembly (C2101-430), and a helium ducting assembly (C2101-340).
The bearing assembly is inserted into the central hole of, and bolted to the support cone.
These bolts (C2101-300-40)" have a circulator support function.
However, they are not accessible for examination, unless the helium ducting assembly is removed, which is not required for the examinations recommended by PSC under SR 5.2.18.
The steam ducting assembly is also bolted to the support cone.
The bolts (C2101-300-39), however, have no structural function and the Report does not recommend that they be examined.
The overall helium circulator machine is supported by the helium penetration upper flange.
This is achieved by clamping the support cone outer flange and the steam ducting assembly upper flange between the underneath surface of the penetration upper flange and the extension stud sleeves (R1100-100-141).
The upper end of the studs (R1100-100-167) is screwed into threaded holes of the penetration flange.
The lower threaded end of the studs, l
located in an accessible area below the circulator header ring (91/90-M-19-8), is used to tighten the attachment by means of stud bearings (R1100-100-165) and nuts (R1100-100-166).
The studs and nuts are removed when disassembling the machine and, therefore, could be visually examined as recommended in the Report.
The practicability of visually examining the threaded holes in the penetration upper flange would have to be further investigated.
The helium circulator PCRV penetration primary boundary consists of the bearing assembly main housing (C2101-508),
support cone, and penetration upper flange.
Leak page 12
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l tightness between the three components is achieved by 0-rings.
The support cone and the bearing assembly main housing are fabricated from forgings and contain no welds.
The helium circulator penetration upper flange is also fabricated from a forging.
The only welds are those which connect the forging to the PCRV liner and to the penetration liner.
The design is such that there are lo structural discontinuities at the location of the welds.
The welds are backed by concrete, located inboard of the penetration main shear anchor and are not pressure retaining welds.
The liner cooling tubes, welded to the liner, also act as shear anchors.
The only accessible weld would be the one that connects the upper forging to the penetration liner.
- However, this weld does not meet the Report definition of pri-mary boundary welds for which surface examination.is recommended in the Report.
Therefore, the Report recommendation for surface examination of helium circu-lator PCRV penetration primary boundary pressure retaining welds does not apply to the Fort St. Vrain design.
PSC agrees to include in SR 5.2.18 a requirement to visually examine helium circulator PCRV penetration primary boundary bolting which is made accessible when performing the other examinations required by SR 5.2.18.
Helium Circulator PCRV Penetration Secondary Boundary The helium circulator penetration secondary boundary is discussed in section A-1 of this response, together with the other penetration secondary closures.
4.3 SR 5.2.19 - See General Comments 4.4 SR 5.2.27 - Helium Shutoff Valves Surveillance Recommendation -
The Report recommends that NRC approve this new SR proposed by PSC.
PSC Response -
None required.
page 14
SECTION 5 - CCMMENTS RELATIVE TO THE SECONDARY COOLING SYSTEM 5.1 SR 5.3.1 - Steam / Water Dump System Surveillance 5.2 SR 5.3.2 - Main and Hot Reheat Steam Stop Check Valves surveillance 5.3 SR 5.3.3 - Bypass and Pressure Relief Valves Surveillance Recommendation -
The Report recommends that NRC approve these existing SR's without changes as proposed by PSC with regard to operability testing.
PSC Response -
None required.
5.4 SR 5.3.4 - Safe Shutdown Cooling Valves Surveillance Recommendation -
The Report recommends that NRC approve the changes to SR 5.3.4 proposed by PSC with regard to operability testing.
PSC Response -
None required.
-,,,, 1:
5.5 SR 5.3.9 - Safety Valves Surveillance 5.6 SR 5.3.10 - Secondary Coolant System Instrumentation Surveillance Recommendation -
The Report recommends that NRC approve these new SR's as proposed by PSC with regard to operability testing.
PSC Response -
None required.
page 16
SECTION A COMMENTS RELATIVE TO PCRV PENETRATIONS AND CLOSURES Recommendation -
The Report recommends that additional examination requirements be included as new SR's for the secondary boundaries of the PCRV penetrations and closures.
Specifically the Report recommends the following:
1.
For the Steam Generator, Helium Circulator, Refueling, High Temperature Filter Absorber and Helium Purification Piping Penetrations, exami-nation of a representative number of secondary shells and closures as noted below:
o Surface examination of accessible pressure-retaining circumferential welds (primarily at structural discontinuities) in penetration shells and closures, of integral attachment welds not backed by concrete, and of circum-ferential welds (primarily at structural dis-continuities) in penetration shells backed by concrete but outboard of any shear anchor.
Examinations may be performed from the outside region where removal of the penetration closure is a regularly scheduled event (refueling, circulator changeout, etc.)
o Visual examination of bolting.
Additionally, tension testing of bolting not normally dis-assembled for maintenance or any other scheduled event.
o Visual examination of accessible-limit stops and structures identified as flow restrictors.
o Leak testing.
page 17
2.
For other PCRV penetrations, examination of secondary shells and closures as noted below:
o Visual examination of pressure retaining welds (primarily at structural discontinuities) in accessible areas of penetration shells and closures outboard of concrete.
o Visual examination of bolting, torque and ten-sion testing.
o Visual examination of accessible limit stops and structures identified as flow restrictors.
o Leak testing.
PSC Response -
The rationale for PSC's recommended surveillance require-ments is based on the unique penetration design concept, on conservative operating parameters, on the safety significance of a single failure, and on the ability to continuously monitor and periodically test for leakage from the penetration interspaces.
These considerations provide adequate assurance that the structural integrity of the penetrations and closures will be maintained.
The need for additional examinations, as recommended.
in the Report, is not supported by the results of opera-tional experience at the plant or by specific technically justified concerns.
Contrary to statements of the Report, PSC's position does not neglect the safety function of the helium circulators, steam generators, and control rod drive mechanisms.
I In reviewing the exemptions provided by the Code for the comparison of the recommended surveillance with proposed Code requirements, PSC took into account the safety significance of individual component failures considering that even a single failure is highly remote, would not prevent safe shutdown of the reactor, nor would it likely impair the safety function of the affected component.
PSC also took into consideration, when applicable, that normal operating conditions of the components were much less severe than their design page 1S
conditions, or that potential conditions adverse to structural integrity could not be postulated during normal plant operation.
PSC also disagrees that the chemical ingress control function assigned in the Report applies to the Helium Purification System.
This normal operating function is not required under accident conditions and is not a safe shutdown function on Fort St. Vrain.
Accidental chemical (water) ingress is controlled by the steam generator orificing, by the plant protective system moisture monitors, and by the steam / water dump valves.
The corresponding helium purification penetrations and wells should, therefore, be placed in the Report
" exempted" category.
PSC agrees that the occasions for scheduled penetration disassembly (such as refueling, circulator changeout under SR 5.2.18, or scheduled maintenance) provide the opportunity for additional examinations of components in that particular penetration.
However, it should be noted that,unlike other penetration closures which are of bolted design, the steam generator penetration secondary closures are welded and that no scheduled disassembly of these penetrations is anticipated.
Consequently, the Report recommendations for examination of components located within these penetrations is not practicable.
PSC notes that the Report recommendation for visual examination of PCRV penetration secondary boundary bolting applies to all bolts, while this requirement is limited to bolting larger than 2 inch diameter in Subsection IGC of the Code for Class 2 components.
PSC considers that the same bolting size limitation as pro-vided by the Code applies to Fort St. Vrain penetration secondary boundary bolting.
PSC also notes that the Report recommendation for leak testing is already accomplished as previously discussed.
PSC will further review the design and accessiblity of the PCRV penetration secondary boundaries in order to identify specifically which components and/or areas can be examined in accordance with the Report recommendations, page 19
i taking into account the above comments.
PSC will pre-sent the results of this review to the NRC, together with any appropriate recommendations for examinations to be included in the Fort St. Vrain Technical Specifi-cation Surveillance Program, that would provide additional
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a'surance of structural integrity beyond that provided s
by the current leakage monitoring and testing requirements.
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page 20
4 SECTION A COMMENTS RELATIVE TO THE THERMAL BARRIER 4
Recommendation -
The Report recommends that PSC investigate the applica-tion of visual examination techniques to thermal barrier installations located within the core outlet plenum and i
steam generator inlets.
PSC Response -
As previously stated in Section 3.2, PSC has already reviewed the practicability of remote visual examination i
of the core outlet plenum and this did not appear feasible.
Since access to the core outlet plenum is j
required to gain access to the steam generator inlet ducts, it follows that remote visual examination of thermal barrier inside these ducts is also not prac-ticable.
The Report did not identify any specific area of concern which would justify the need for examining these in-stallations.
PSC considers, therefore, that no addi-tional investigation is warranted for the thermal barrier in the core outlet plenum or steam generator inlet ducts.
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SECTION A COMMENTS RELATIVE TO THE CORE LATERAL RESTRAINTS Recommendation -
The Report recommends that PSC develop a material spe-cimen surveillance program for core lateral restraint components, or document analysis data which demonstrate that plant safety will not be compromised by lack of surveillance for these components.
PSC Response -
Design analyses of core lateral restraint components are documented in the FSAR, and it was ascertained that the components had an adequate life.
Specifically', FSAR Section 3.3.2.2 states that_ during operation of the reactor, all of the structural metal components are under elastic loading conditions below their creep range, and at a temperature above the FTP (fracture transition for plastic loading) temperature for the material after irradiation corresponding to 30 years of reactor opera-tion.
This FSAR section also. states that the effects of neutron irradiation and nuclear heating have been taken into account in the design and have been found to be negligible.
Recent fatigue analysis, to take into account actual operating experience of Fort St. Vrain, did not reveal any concern relative to cycle life for reactor internals.
Since these analyses demonstrate that unacceptable de-gradation will not occur under prevailing service condi-tions, PSC considers that development of a material specimen surveillance program is not warranted.
page 22
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SECTION A COMMENTS RELATIVE TO THE HELIUM PURIFICATION SYSTEM Recommendation -
l The Report recommends that PSC include surveillance requirements for the Helium Purification System.
Specifically, the Report recommends that requirements be developed based on the following:
o For primary boundary components, including vessels and piping, when not contained within a monitored secondary boundary designed to withstand primary pressure:
- Surface examination of accessible pressure retaining welds at structural discontinuities and integral attachments.
- Visual examination of bolting and, additionally, tension testing of bolting where not disassembled for maintenance or other reasons.
- Visual examination of support members.
- Leak testing.
o For piping boundaries acting as secondary containment:
I
- Visual examination of accessible pressure retaining welds and support members.
- Leak testing.
o For system isolation valves:
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- Verification of valve operability.
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- Leak testing.
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- Other tests determined appropriate to valve type and function.
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cace 23
e PSC Response -
As stated in the General Comments, PSC has included the Helium Purification System as a review Category II system.
PSC has not reviewed the requirements for this system as yet, but will consider the recorrendations cited above when this review is performed.
9 e
page 24
SECTION A COMMENTS RELATIVE TO THE STEAM GENERATOR TUBING Recommendation -
The Report concludes that Steam Generator tubing inspection for Fort St. Vrain is not necessary to assure the long term heat removal and water ingress protection of the primary system.
The Report does, however, recommend that data be supplied to the NRC which describe material tests performed to date to assure the long term thermal performance of the bi-metallic welds located at the top of the steam generators.
PSC Response -
In describing the steam generator tubing configura-tion, the Report states that helical tubing in the main section is subheadered and connected to the superheat section at the top thrcugh a series of bimetallic welds.
PSC notes that the tubing in the crossover region between the EESHI bundle and SHII bundle is not subheadered and each tube has one bimetallic weld in the vertical portion of the crossover region.
The Report also states that the superheat section is located inside the helical bundle.
PSC notes that both the EESHI bundle and the SHII sections are helical bundles and SHII is located above EESHI and not inside as stated.
The Report further states that bimetallic welds are located in the top of the steam generators between the main and superheat sections which are subject to severe thermal conditions and that one failure has occurred in this region.
PSC notes that the bimetallic welds are located in a straight vertical section of the crossover region between EESHI outlet and SHII inlet, while the tube failure which occurred was determined to be located in an outlet tube at the bottom of a SHII bundle (subheader F of module 3-1-1, as identified in PSC letter P-78007 dated January 13, 1978).
Wirh respect Oc the Report recuest tha: ?SC provide the NRC with material tes data relative to -he cross-cver bi-metallic welds, PSC notes that such data are included in Appendix A of the FSAR (section A.l.12.8, page 25
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l A. l.12.14, A. l.12.16, A. l.12.17, A. l.12. 21).
Additional information concerning the design and analyses of these welds, as well as an evaluation of the welds in the Fort St. Vrain configuration and environment,was provided by PSC letter P-77084 dated March 7, 1977 to answer NRC concerns which were raised by bi-metallic weld failures of a design used in fossil-fired boilers.
The differences in bi-metallic weld design between Fort St. Vrain and fossil-fired boilers were such that no specific areas of concern were identified.
PSC is not aware of any new test data that might be made available to the NRC and considers that no further investigation is warranted.
It should be noted that the structural integrity of the i
cross-over region in general, which includes the referenced bi-metallic welds, is further assured by operational temperature limitations which are more conservative than those established from ASME Code material characteristics.
Correlations have been es-tablished between the crossover region temperature and measured plant operating parameters, to include in particular the main steam temperature which is con-trolled as a function of reactor power to prevent the crossover temperature limit from being exceeded.
PSC considers that there is currently no technical justi-fication for concerns about the crossover bi-metallic welds.
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SECTION A COMMENTS RELATIVE TO HIGH ENERGY PIPING Recommendation -
The Report recommends that PSC review requirements pertaining to protection against the effects of high energy piping failures and that the following method be used:
1.
Define systems and components important to safety as follows:
(a) Required to assure the long term integrity of the reactor coolant pressure boundary.
(b) Required to assure safe shutdown and core heat removal following a shutdown.
(c) Required to prevent the release of contaminants which would result in offsite doses above acceptable limits.
2.
Identify high energy piping systems which are physically located in critical areas (as defined in Sections A-6.b and A-6.c of the Report).
3.
Identify systems important to safety that could suffer loss of function as a result of a high energy piping failure.
4.
Describe a program of augmented ISI to mitigate the potential effects of high energy piping failures in critical areas.
Augmented ISI may include the following:
(a) Nondestructive examination of welds, (b) Visual inspection during hydrostatic testing, (c) Examination of pipe supports both during operating and shutdown conditions.
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PSC Response -
The requirements pertaining to protection against the effects of postulated high energy piping failures have been thoroughly analyzed for Fort St. Vrain.
PSC notes that the requirements for analyzing high energy piping for Fort St. Vrain were provided separately by the NRC in a letter dated December 18, 1972.
These requirements and the resulting analysis are presented in Amendment No. 26 to the FSAR.
Items 1, 2,
and 3 of the Report recommendations above were previously defined and identified in the detailed analysis presented in this FSAR Amendment.
The analysis did not reveal.
critical areas with regard to postulated piping failure and safe shutdown equipment locations.
Nevertheless, further extensive analysis was performed assuming that failures occurred in the worst locations.
The results of this work clearly show that there is a high degree of assurance provided by the conservatism and the pro-tective features included in the design of the affected systems.
No concerns were identified in this analysis which would justify " augmented ISI" for Fort St. Vrain.
PSC does not consider another review to be warranted.
It is PSC's position that the existing surveillance moni-toring provisions as described in the submittal are adequate to assure the long term structural integrity of the high energy piping systems.
The operating ex-perience to date and the results of detailed analysis of the potential effects of postulated high energy piping failures clearly do not warrant that augmented examination or test requirements be imposed.
Further-more, PSC does not consider the LWR requirements or experience in this particular area to be directly applicable to the Fort St. Vrain design.
PSC has, however, been monitoring the effects of plant transients on the service life of plant components.
Recently, an analysis was performed to determine the effect on component fatigue life due to the increased number of thermal cycles experienced since the plant went into operation.
This analysis showed that all components of the PCRV, PCRV internals and the primary coolant system had adequate service life.
- However, several piping components in the main steam and hot reheat steam systems were shown to have potentially less service life remaining than required to operate cace ?.3
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the plant over its 30 year design life.
PSC is currently reviewing the impact of these identified areas of concern.
Further, more detailed analysis may demonstrate that the components are adequate.
If this cannot be done, PSC will recommend appropriate examination and test requirements that address the specific concerns raised by the recent analysis.
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