ML19305C341

Inservice Insp & Testing Program:Surveillance Requirements Review for Prestressed Concrete Reactor Vessel Internals, Revision 1

ML19305C341
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/27/1980
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML19305C332	List: ML19305C331 ML19305C333 ML19305C337 ML19305C339 ML19305C341
References
P-80034, PROC-800227-04, NUDOCS 8003260578
Download: ML19305C341 (22)

Text

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- - Enclosure (4) to P-80034-I i

5 FORT ST. VRAIN ,

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- - - -- INSERVICE INSPECTION AND TESTING PROGRAM j

- SURVEILLANCE REQUIREMENTS REVIEW ._

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FOR THE l

PCRV INTERNALS i

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, , EE-ll-0003 Rev. 1 p.1 REPORT EE-ll-0003 FORT ST. VRAIN INSERVICE INSPECTION AND TESTING PROGRAM SURVEILLANCE , REQUIREMENTS REVIEW FOR THE PCRV INTEPRALS A review was performed of the current surveillance re-quirements for the PCRV internals. As a result of this review, additional or modified surveillance requirements may have been recommended to meet the c::iteria established for the Fort St.

Vrain inservice inspection and testing prog' ram which has been presented to the Nuclear Regulatory Commission.

This report consists of two parts, each one dedicated to the following internals:

Part A: PCRV thermal barrier Part B: Core vertical and lateral support structures; circulator inlet plenum structure.

A list of the reference documents used in the course of this review is given individually for each part of this report.

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REPORT EE-11-0003 -

PART A

SURVEILLANCE REQUIREMENTS FOR -

THE FORT ST. VRAIN 6

9 PCRV THERMAL BARRIER I

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1. INTRODUCTION A review was performed to investigate the adequacy of the current surveillance requirements for the PCRV thermal barrier.

Additional or modified surveillance requirements may be re-commended as a result of this review, to satisfy the criteria established for the Fort St. Vrain inservice inspection and testing program as outlined in Ref. 1.

The review included the documents listed in section 5, and in particular the proposed ATME Code Section XI, Division 2, Draft (referred to in this report as the Code). The proposed Code requirements are identified, and an explanation is given where the current or recommended surveillance differs with these requirements. The review also included the operating experience with the thermal barrier and liner cooling system .

at the plant. I I

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2. SURVEILLANCE CLASSIFICATION l The PCRV thermal barrier and liner cooling system operate in conjunction tc limit and remove the heat load across the PCRV liner and yield temperatures and temperature gradients in the liner and concrete compatible with their design condi-tions. This assures that their integrity is maintained throughout the plant life, by not allowing unacceptable thermal i stresses to develop, and by maintaining liner temperature well above nil ductility transition temperature (taking into account the effect of integrated irradiation) to prevent its brittle ,

fracture. Therefore, in accordance with criteria 2.lc, 2.2c, i and 2.3b of Ref. 1, th'e thermal barrier, which is a passive i equipment item, is assigned to surveillance class S3. l The following criteria per Ref. 1 are considered when re-

'. viewing the surveillance program. For surveillance class S3, the operational readiness concept does not apply to passive components. There are no active comE onents nor instrumentation and controls associated with the thermal barrier. Structural integrity can be demonstrated by continuous leakage monitoring and/or alarm (criteria 3.3.la) unless the operating conditions are not expected to degrade the component integrity when com- l pared to the design conditions (criteria 3. 3. lc (i) ) , or unless a failure does not prevent the system from performing a safety function and does not lead to unacceptable release of radio- l l activity (criteria 3. 3. lc (ii) ) . Reactor internals, which l l cannot have their integrity directly monitored, will have i

their structural integrity verified, where feasible, by sur-vtillance inspections of material specimens exposed to condi-tions similar to their support components, at a frequency based on evaluations of the material properties (criteria 3.3.2b).

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LE-11-0003 R v.. L L.3 3.. GENERAL DESCRIPTION OF THE THERMAL BARRIER The thermal barrier designated Class A is used in regions where.the primary helium is at core inlet temperature.- The insulation material is Kaowool. The thermal barrier designated Class B is used in regions where the primary helsium is at core average iutlet temperature. The insulation material is pure silica felt, in the upper layers which are subject to elevated  !

temperatures, combined with Kaowool, in the lower layers ex-posed to temperatures compatible with the use of that material.

The thermal barrier designated Class C is used only on the top of the core support floor which can be exposed to very high hot  ;

streak temperatures. The material used for these very high temperatures is fused silica blocks which cover a layer of Kaowool. The core support posts rest on alumina pads, which have an adequate strength at very high temperature, covering l fused silica pads.

Generally, the Class A and B thermal barriers consist of two layers, each one comprising metallic cover plates and thin seal sheets, and fibrous material blankets resting on insula-tion supports. The two layer arrangement provides two seals to prevent the flow of hot gas through the thermal barrier onto the liner in regions of high pressure gradients. This arrangement also provides redundancy with respect to thermal performance. A single thermal barrier layer is used in region of lesser pressure gradients, and where the loss of thermal performance is of lesser consequence.

The overall thermal barrier is assembled by nuts screwed on posts, themselves screwed on studs welded to the liner.

The posts are also welded to the liner. The attachments are designed to accomodate thermal movements of the thermal barrier.

There are generally six to eight such attachments for each cover

, plate. On the top head of the core cavity, there are four

, studs plus a welded connection to a cylindrical extension of each control rod penetration.

I Sliding metal surfaces have received a surface treatment, J the nature of which depends on the operating temperature (flame sprayed chromium carbide or nitrided surfaces).

The nature of the metallic material depends on its opera-ting temperature. Low carbon steel, Inconel 600 and Hastelloy X are used for increasing temperatures, thus providing the re-quired mechanical characteristics.

In most PCRV regions, the thermal barrier protects both the liner and the concrete, and the heat flux through the thermal barrier and liner is removed by the liner cooling system. The thermal barrier installed on the floor support columns limits the thermal stresses and the heat load is re-moved by the cooling tubes which also ccol the core support floor liner and concrete. In a few regions, the thermal barrier I

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reduces ~the heat losses through,- and protects the metal components.used to separate helium at reactor outlet temperature from-helium at reactor inlet temperature. This is essentially the case of the core barrel in the lower core plenum region, and of the steam generator inlet ducts below the core support floor.

4. STRUCTURAL INTEGRITY 4.1 MONITORING OF THERMAL BARRIER PERFORMANCE a) Current surveillance requirements:

Technical specifications SR 5.4.4 and SR 5.4.5 specify the surveillance requirements for monthly functional

tests and annual calibrations of reactor plant cooling water system instrumentation which monitors and alarms subheader flow and outlet water temperature and individual tube outlet temperature.

b) Recommended surveillance requirements:

By analogy with criteria 3.3.la of Ref. 1 which spe-cifies that the structural integrity of a fluid system can be demonstrated by continuous leakage monitoring and/or alarm, it is considered acceptable that structural inte-grity of the thermal barrier be demonstrated by continuous monitoring of its, thermal performance, in those regions where such monitoring is possible.

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The design of the liner cooling system is such that individual tubes are grouped in subheaders so that the tube and subheader instrumentation allows monitoring of the thermal barrier performance in specific regions.

Therefore, changes in thermal barrier thermal character-istics, due either to degradation of material properties or to loss of integrity causing a degradation of the sealing capability, are identified and alarmed by the related instrumentation in the reactor plant cooling water system.

Degradation of material properties will probably cause an overall and slow increase in the heat load removed by tne reactor plant cooling water system. Since several insulation materials are used, the effect may vary between the several thermal barrier regions depending on their design and class.

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Mechanical damage, considered here as random failure i of thermal barrier attachments, will probably cause local losses of compression of the thermal barrier, degraded sealing capability and subsequent local increase in the heat load removed by individual cooling tubes and in cooling tube outlet temperature,. Such an observed in-crease in tube cooling water outlet temperature, which cannot be related to a reduction in water flow, will be an indication of a possible structural defect of the thermal barrier somewhere along the tube path.

The whole thermal barrier related to the integrity of the PCRV and of the core support is monitored by the i liner cooling system instrumentation. The non-monitored l thermal barrier areas are two class B regions, namely l the core barrel thermal barrier in the lower core plenum l (area B1) and the steam generator inlet duct thermal barrier (area B3), and some parts of class A thermal barrier in the circulator inlet plenum, specifically the areas which protect the lower floor support posts (area A9),

the steam generator support skirts (area A10) and the bottom access penetration (areas All, A12 and A16).

Thermal barrier area designation is according to drawing R1104-100 sheet 2.

The temperature sensors permanently installed in the PCRV concrete could 7rovide an additional means of in-vestigation of the thermal barrier integrity. However, it is not recommended that such instrumentation be used )

for surveillance.since it cannot be functionally tested and calibrated.

The current surveillance is ther~efore generally con-

, sidered adequate. A further discussion of these sur-veillance requirements will be included in the report on the reactor plant cooling water system. Additional requirements under consideration include functional tests and calibration of the instruments which monitor the water inlet temperature, to assure that the heat loads can be determined accurately.

c'. Proposed AriE Code requirements:

The proposed Code does not address the monitoring of thermal barrier performance.

d) The current and recommended surveillance therefore exceeds the proposed Code requirements.

e . EE-11-0003 Rev. 1 A.6 4.2 SPECIMEN INSPECTION a) Current surveillance requirements:

Ther.e is no technical specification which explicitly-requires that the thermal barrier specimens (attached to the removable plug of the lower access penetration shield plug and to each of the plateout probe penetra-tion primary closure plugs) be visually examined. Para-graph 5.13.5 of the FSAR indicates that visual inspection of the liner-thermal barrier assemblies will be made whenever these plugs are removed or when results from the ultrasonic liner monitoring (per technical specifi-cation SR 5.2.14) indicate that a closer examination is required. Technical specification SR 5.2.6 specifies the time intervals at which a plateout probe is to be removed, namely at the first, third and fifth refueling, and at intervals not to exceed five refueling cycles thereafter.

b) Recommended surveillance requirements:

Surveillance procedure SR 5.2.6-X has been reviewed and does not specify that the liner-thermal barrier assembly be visually examined between removal and re- ,

installation of the plateout probe penetration primary i closure plugs. Surveillance procedure SR 5.2.14-X has j also been reviewed, but it addresses only the liner '

specimens in the above liner-thermal barrier assemblies,

1. Drawings R1103-313, for the plateout probe penetraticn

==</f shield plug, and R1103-407, for the lower access pene-tration removable plug, have been reviewed to determine if the thermal barrier specimens are representative. The thermal barrier specimen design differs essentially from the general design due to the fact that there are no

. attachments or seal sheets. Kaowool is enclosed in a metal box welded to the plug. Visual examination of such specimens would not provide any useful information.

However, should disassembly of the box be required to examine the liner specimen, the Kaowool could be removed for investigative testing of potential changes in its

. characteristics.

The thermal barrier attached to the upper access pene-tration was also reviewed to see if it would serve as a more representative specimen. However, its design also differs somewhat from the general thermal barrier design (drawing R1103-210) and the same conclusions are drawn as for the above specimens.

The removable part of thermal barrier most represen-tative of the general design is the one on the lower access shield plug. However, it would be a major task to remove, since it would require cutting the liner ex-tension on which the secondary closure is attached,

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i because the secondary closure dimension only allows disassembly of.the removable plug.- Therefore, considering also the potential. risks of contamination, such an exami-j . nation is not proposed as part of the surveillance program..

1. nelusion, it is not recommended that any thermal barrier specimens be examined for surveillance purposes.

c) Proposed ASME Code requirements:

The proposed Code has no requirements related to inspection of thermal barrier specimens.

d) The recommended surveillance and the proposed Code are identical with respect to examination of thermal barrier specimens.

4.3 VISUAL EXAMINATION a) Current surveillance requirements: None.

b) Recommended surveillance requirements:

As outlined in section 3 above, the thermal barrier essential to the.integriti of the PCRV and core support is designed in two redundant layers, so that total failure of the outer layer can be postulated without detrimental effects on the PCRV or core support. One would have to

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postulate the total failure of the thermal barrier, i.e., a common mode failure of the attachments over a significant

. area. Since the attachments have been designed and tested with adequate safety margin with respect to'their strength, no concern has been identified which would lead to such a common mode failure. In any case, such a failure would be sensed and alarmed by the reactor plant cooling water l system instrumentation before it became critical. Therefore, i visual examination of the thermal barrier is not expected to provide additional assurance of the integrity of the

! PCRV or the support structures.

l Another detrimental effect attributable to the thermal barrier would be a gross failure with collapse of the thermal barrier (or parts of it) in such a manner that it would prevent adequate primary coolant flow. As out-lined above, no concern for such a gross failure has been identified. However, if such a condition were postulated, it would be analogous to a permanent loss of forced circulation accident (DBA1) for which adequate safety systems are provided. The likelihood of even a single coverplate to get loose is very remote, considering the

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- redundancy and diversity of the attachments. However, if some' flow restriction were induced by thermal barrier failure, it would be sensed by the instrumentation which-monitors reactor operation. Therefore, visual examina-tion of the thermal barrier is not expected to provide-additional assurance of core cooling capability. I Visual examination has been contemplated as a means of surveillance for those the mal barrier areas identified in section 4.1.b above, the tl.ermal pdrformance of which is not monitored. However, tie areas of interest (class B thermal barrier), are inaccessible for visual examination, and the other ones (class A thermal barrier) could only be accessible with major difficulties but are not con-sidered critical areas. Therefore, no visual examination can be recommended for the non monitored thermal barrier areas.  ;

The only thermal barrier areas readily accessible for visual examination are located in the upper core plenum (areas A4, A5 and upper part of A2) . It has been stated above that no concerns have been identified for this thermal barrier, and its thermal performance is continuously monitored. Therefore, visual examination could be recommended for investigative purposes, should the monitoring indicate an appreciable degradation of thermal preformance. However, visual examination is not recommended as a part of the surveillance program.

For the above reasons, no visual examination is recommended for the surveillance of the structural inte-grity of the thermal barrier.

c) Propo' sed ASME Code requirements:

Items I-D-1 in Tables. IGG-2500-1 and IGG-2600-1 require that a representative portion of exposed and accessible areas of thermal barrier in the upper and lower core plenums be visually examined during each refueling outage, so that at least 25 percent of the surface is

- examined during each inspection interval (ten years per IGA-2400). The proposed Code also provides exemption from examination for internals whose failure will not impair shutdown heat removal, nuclear reactivity control, or chemical ingress detection or control (IGG-1221) or for internals that have design redundancy with the ability to detect failure of each individual load path (IGG-1222) ,

or for those internals that are routinely monitored for I performance during normal plant operation to demonstrate l operability (IGG-1224).

EE-ll-0003 Rev. 1 A.9

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d) The- differences between the recommended surveillance - and the proposed Code requirements are considered justified, for the. reasons outlined in paragraph b above, taking also into consideration the fact that the Code was written for the large HTGR thermal barrier which had design features somewhat different from the one at Fort St.

Vrain (about 3 attachments per plate instead of 8 and larger coverplates). Further, the lower core plenum is not accessible for examination at Fort St. Vrain.

5. LIST OF REFERENCES Refc.ences :

1. PSC report EE-SR-0001: Surveillance inspection and test criteria for the Fort St. Vrain nuclear generating station.

2. FSV FSAR sections 5.8, 5.9, 5.13 .

3. FSV system descriptions: SD-ll-2, SD-ll-4, SD-46-1 1

4. FSV system diagrams: PI-46-1 through 10

5. FSV technical specifications: LCO4.5.2, LC04.2.13, 'I LC04.2.14, LC04.2.15, SRS.2.6, SRS.4.4, SR5.4.5, l SR5.4.ll

6. FSV surveillance procedures: S R5 . 2 '. 6 - X, SR5.4.4-M/

5.4.4-Al, SRS.4.4-A2, SR5.4.5-M, SR5.4.5-Al, SR5.4.5-A2, SRS.4.ll-M, SRS.4.ll-A

7. ASME Code Section XI, Division 2, Draft

8. FSV drawings: R1103-207, R1103-210, R1103-312, R1103-313, R1103-405, R1103-407, R1104-100 through -104 f

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PART B SURVEILLANCE REQUIREMENTS FOR THE FORT ST. VRAIN VERTICAL AND LATERAL CORE SUPPORT STRUCTURES AND CIRCULATOR INLET PLENUM STRUCTURE

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EE-ll-0003 Rev. 1 B. 2'

1. INTRODUCTION-A review was performed to investigate the adequacy of the current.. surveillance requirements for the PCRV internal structures

.which support the reactor and form the circulator inlet ple-nums. Additional or modified surveillance requirements, as well as additional investigations, may be recommended as a result of this review to satisfy the criteria established for the Fort St. Vrain inservice inspection and testing program as outlined in Ref. 1.

The review included the documents listed in section 7, and in particular the proposed ASME Code Section XI, Division 2, Draft. The proposed Code requirements are identified, and

, an explanation is given where the current or recommended sur-veillance differs with these requirements. The review also included the operating experience with the reactor internals 1 at the plant.  ;

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2. SURVEILLANCE CLASSIFICATIONS  ;

The following structural components.are considered in this report:

l a) the graphite core support blocks, the graphite support i posts and post seats, the lined and cooled reinforced concrete core supp, ort floor, and the steel floor support columns, which provide vertical support for the core; b) the permanent grap.hite side reflector and boronated side reflector spacers, the steel core barrel, the reflector to barrel keys, and the barrel to liner keys, which provide lateral support and restraint for the core; c) the top steel plenum elements which provide lateral re-straint between fuel columns to .. arm a fuel region, and the region constraint devices which provide lateral re-straint between fuel regions; d) the lower floor and seals which separate the inlet plenum from the outlet plenum of the circulators, and the loop divider plate and seals which separate one primary coolant loop from the other one.

The above structural components function in a passive manner to support and restrain fuel columns, fuel regions, and the core, and/or to provide a flow path forcing the reactor coolant through the core and the capability to isolate one l reactor coolant loop.

According to criteria 2.lc, 2,2c and 2.3b of Ref. 1, these l components are assigned to surveillance class S3.

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2. (cont.)

The following= criteria per Ref. 1 are considered when reviewing-the surveillance program. For surveillance class S3, the operational readiness concept does not apply to passive components. There are no active components nor instrumentation and controls associated with the PCRV internals. Structural integrity can be demonstrated by co'ntinuous leakage monitoring and/or alarm (criteria 3.3.la) unless the operating conditions are not expected to degrade the component integrity when com-j pared to the design conditions (criteria 3.3.lc(i)), or unless j

a failure does not prevent the system from performing a safety function and does not lead to unacceptable release of radio-activity (criteria 3. 3. lc (ii) ) . PCRV internal structures, which cannot have their integrity directly monitored, shall have their structural integrity verified, where feasible, by sur-veillance inspections of material specimens exposed to conditions similar to their support components, at a frequency based on evaluations of the material properties (criteria 3.3.2b). j i  !

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3. STRUCTURAL INTEGRITY OF THE VERTICAL CORE SUPPORT 3.1 CORE SUPPORT BLOCKS -

a) Current surveillance requirements:

Technical specification SR 5.2.22 requires that PGX graphite specimens be installed into 5 bottom transition reflector elements, and that specimens be removed at the 2nd, 4th, 6th, 9th, and 17th refuelings, and be subjected to exa,mination and compared with laboratory control specimens in evaluating oxidation rates, oxidation profiles and general dimensional characteristics. The specimen removal schedule can be adjusted to a faster rate should specimens show a significantly greater burn off than predicted.

b) Recommended surveillance requirements:

PGX graphite is the material of the core support blocks. Recent laboratory investigations have raised a concern with respect to the burnoff of PGX graphite which has been found much higher than originally taken into account in the design of the core support blocks. The laboratory tests have also demonstrated that burnoff rate and profile are strongly dependent upon operating conditions, but actual reactor conditions could not be duplicated at this time in laboratories, making it dif-ficult to accurately predict the effect of actual burn-off on the strength of the core support blocks. The current surveillance meets criteria 3.3.2b of Ref. 1 and 1 .- . , , _

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is considered an adequate means of monitoring the actual burnoff which takes place under actual operating conditions and thereby of supporting conservative calculations of ex-pected core support block life. However, direct moni'oring of core support blocks would provide additional assurance of integrity, in particular if the calculated life is shorter than or close to the design life.

A program is underway, sponsored by DOE, to develop eddy current and ultrasonic techniques which may allow direct non destructive examination of the cor'e support blocks (Ref. 7 through 9). The ultrasonic technique is being de-veloped in case the implementation and evaluation prove the eddy current technique inadequate. If these techniques prove to be successful, their use would be considered for surveillance in the Fort St. Vrain reactor.

The core support blocks are not normally visible during refueling of the corresponding core region.

However, they can be visually examined when removing the two lower layers of permanent lower reflector elements. i Since the PGX graphite specimens are not directly rep-resentative of the core support block geometry, and since l the fuel regions in which PGX graphite specimens have been 1 installed have been selected due to their higher potential  ;

for PGX graphite burnoff, it is recommended that the top  ;

surface of the core support block be examined by remote TV  !

when the PGX graphite specimens are removed from the core i in that region. Such visual examination should provide i indications that no cracks occured, in particular in areas )

where the analysis shows the highest tensile stresses, i.e.

on the upper surface of the core support block at the key- I ways and webs between reactor coolar.t channels. l

, c) Proposed ASME Code requirementF:

1 Item I-C-1 of table IGG-2500-1 specifies that areas subject to examination include representative surfaces of components serving as vertical supports and also material surveillance coupons on the core support posts. The proposed  ;

. Code rr. quires that 25 percent of exposed and accessible areas of support posts and blocks be inspected during each inspection interval (10 years per IGA-2400). Item I-C-1 of table IGG-2600-1 specifies that inspection is to be per-formed by visual examination, and that newly developed surveillance techniques are to be used, in addition to visual examination, as they become available.

d) Coupons of the support block material are not required by the proposed Code, and such coupons are used for surveil-lance at Fort St. Vrain because of the concern of potential faster burnoff than predicted for this material. This ex-ceeds the proposed Code requirements. The difference in I

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visual _ examination between the recommended surveillance and the proposed Code' requirements is-justified since the' core support blocks to be examined have been identified as those with a higher potential for burnoff. The development pro-gram being pursued is also consistent with the proposed Code requirements. -

3.2 CORE SUPPORT POSTS AND SEATS a) Current surveillance requirements: None.

b) Recommended surveillance requirements:

Tests and calculations have demonstrated that there is an adequate safety factor in the design of the core support posts for the life of the reactor. The core support posts and seats are made of Union Carbide ATJ graphite which has not been demonstrated to be more prone to burnoff than expected and taken into account in the~ design.

There are no specimens of ATJ graphite which can be tested, and the core support posts and seats are not accessible for visual examination. Further, it is not expected that the rupture of one core support post would result in the collapse of. the core.

Therefore, no surve i llance is recommended for the core support posts and seats.

c) Proposed ASME Code requirements:

The proposed Code requirements are outlined in para-graph 3.lc above.

, d) The proposed Code requirements were written for the large HTGR where a removable plug was provided in the core support blocks which allowed visual examination of the core support posts. These plugs could also contain spe-cimens of the core support posts material. None of these features is applicable at Fort St. Vrain. This and the reasons outlined in paragraph b above justify the dif-ference between the recommended surveillance and the proposed Code requirements.

3.3 CORE SUPPORT FLOOR AND FLOOR SUPPORT COLUMNS a) Current surveillance requirements: None.

b) Recommended surveillance requirements:

. The core support floor is a reinforced concrete structure, covered with a steel liner cooled on the

, , inside by welded tubes of the liner cooling system and in-sulated on the outside with thermal barrier. Its structural integrity is assured by these design features. _

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EE-ll-0003 Rev. 1 B. 6-3.3b (cont. ). .

The floor support columns are vertical steel pipes enclosed in thermal barrier. The core support floor cooling, tubes run through the pipes and remove the heat load in the cc.umns. The core support floor rests on the top of twelve such columns. The bottom part of the columns is anchored in the concrete of the PCRV.

The surveillance requirements discussed in part A of this report are adequate to assure the integrity of the core support floor and floor support columns.

Further, instrumentation in the PCRV auxiliary systems  ;

(see report ref. EE-ll-0001 part C paragraph 3.3.3) monitors helium inleakage.in the core support floor and floor support columns, as well as water leakage from the tubes. The surveillance requirements for this in-strumentation are adequate to provide additional l assurance of structural integrity. I 1

The core support floor and floor support columns  ;

are otherwise not accessible for visual examination. No i further surveillance is recommended over'that speci-fied above.

c) Proposed ASME Code requirements:

The proposed Code has no requirements for the core support floor and columns since it was written for the large HTGR which has separate cavities for the core and "Leam generators.

- 4. STRUCTURAL INTEGRITY OF THE LATERAL. CORE SUPPORT AND I

, RESTRAINT This section includes the lateral core support and re-straint permanently installed in the reactor, i.e., the per-manent s. '.e reflector, the core barrel and the several keys which connect the reflector to the core barrel and the core

, barrel to the liner.

a) Current surveillance requirements: None.

b) Recommended surveillance requirements:

1 No specific areas of concern have been identified with respect to the structural integrity of the lateral core support and restraint. Further, most of the

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components are not accessible for examination.. Therefore, no surveillance is recommended for these equipment items.

c) Proposed ASME Codc requirements:

The only Code requirements for the lateral core support address the spring packs of.the large HTGR. There are no such devices at Fort St. Vrain. Therefore, there are no applicable Code requirements.

d) The recommended surveillance meets the proposed Code requirements.

5. FUEL REGION AND COLUMN RESTRAINT l

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5.1 REGION CONSTRAINT DEVICES  !

Region constraint devices have been installed recently in i

t. ;w reactor to prevent excessive movements of one fuel region '

Mf. 2: elation to the adjacent fuel regions. The region constraint 45.n.As are metallic triangular plates with three centering el,34 phich are inserted in the handling hole of three adjacent metallic plenum elements of three different fuel regions. The top of each region constraint device is equipped with three dowels and a grappling hole which allow handling and positioning with the fuel handling machine. 1 a) Current surveillance requirements:

Proposed surveillance requirements for the region con-straint devices were outlined.in PSC letter P-79279 dated

' November 26, 1979, and they are based upon visual observa-tions using the facilities of the fuel handling machine, in those regions undergoing refueling, as follows:

at each refueling perform a visual TV examination of the upper core plenum to verify that the region constraint devices are in place on top of the core; compare as installed vs. as found fuel handling machine coordinates (after correction to take into account changes in fuel column height due to irradiation of graphite and coordinate changes which will occur when '

a RCD is removed from a different refueling penetration than the one from which it was installed) so that unan-1 ticipated changes in region constraint device location

, can be determined; check the lifting force requirea-to remove each region constraint device with the fuel handling machine; re-moval and reinstallation will act as go/no-go dimensional test of the region constraint devices;

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visually examine and photograph selected. region con -

straint devices in the fuel handling machine during their removal at refueling.

b) Recommended surveillance requirements:

Each region constraint device spans three fuel regions of different ages (in the equilibrium cycle). The different ages imply different fuel column heights due to graphite shrinkage, so that the centering pins will not be equally engaged in the several fuel columns, with the least engaged pin being in the oldest region. The m'aximum shrinkage is about 4 inches for a six year age fuel column. However, the maximum differential shrinkage of two adjacent fuel columns occurs between a new fuel column and a five year age column and it amounts to about 3.5 inches which, added to the length of the inlet chamfer of the handling hole, is about two thirds of the length of the centering pin. The shrinkage rate of new fuel is about twice as high as the one of 5 year age fuel. This results in longer portions of the centering pins to be engaged in the older fuel columns as the new fuel is being irradiated. Since the forces ceuld be applied by a fuel region on part of the length of the corresponding centering pins,. bending of the pins could result, which can be detected, if significant, by the lifting force required to remove the region con-straint device with the fuel handling machine.

The proposed surveillance requirements have been re-viewed and are co'r.sidered to meet criteria 3.3.2b of

, Ref. 1 for monitoring of structural integrity.

1 The visual TV' examination of che upper core plenum is adequate to assure' that there has been no phenomenon which has caused a region constraint device to migrate completely out of some fuel columns. However, it is recommended that such a visual examination be limited to the region constraint devices which can be observed from the penetrations which are scheduled to be opened during any particular refueling, since visual examination cf the whole upper core plenum would require that other penetrations be opened and the fuel handling machine be placed on them, which would con-siderably increase the duration of the refueling shutdown.

Only if a region constraint device were found out of place, should the upper core plenum be visually examined to assure that no other region constraint device is out of place.

i The comparison of as installed /as found fuel handling machine coordinates, particularly in the vertical direction, once corrected for expected shrinkegc, is adequate to provide an additional indication that the region constraint devices are engaged in the fuel columns.

- _ . - ~v - --

i. .

P*'*

• EE-ll-0003 R v. 1 B. 9-5.lb (cont.)

The check of the lifting force required to remove-the

, region constraint devices is adequate to provide early indicati'ons, should a phenomenon occur over time-which might ev,entually prevent them from moving with the fuel columns'or prevent their removal from the reactor.

Visual examination and photographing selected region constraint devices in the fuel handling machine is adequate to assure that there are no unacceptable deformations, loc 9e or missing pa cs, or other visible' defects.

c) Proposed ASME Code requirements: Not applicable.

5.2 TOP KEYED PLENUM ELEMENTS a) Current surveillance requirements: None, b) Recommended surveillance requirements:

One central control rod element and six outer plenum elements, are keyed to form an individual fuel region in which the primary coolant flow can be controlled by ad-justing the position of the corresponding orifice valve.

These elements are of a welded steel construction with flame sprayed areas,where metal to metal contact is expected,to prevent wear and gallinb The lower part of the elements between granules.

the coolant channels is filled with moderator During normal operation, the plenum elements experience only small loads. The load applied on the elements by the

' region constraint devices is not likely to damage the plenum elements.

Therefore, no surveillance is recommended for these components.

c) Proposed ASME Code requirements:

The proposed Code does not address the plenum elements.

. . . .. -- ~. . . - . . . .

q

p e 2

- EE-11-0003 Rcv 1

. B.10-

6. CIRCULATOR INLET PLENUM: STRUCTURE, a) Current surveillance: requirements: None.

b) Recommended surveillance requirements:

The lower floor is a steel membrane which separates the circulator inlet plenum from the outlet plenum. The steel plate is welded to the liner extension of the lower access penetration and supported. by flexible columns.

Leak tightness between the plenums is assured by two re-dundant flexible seals on the periphery of the lower i ficor, and by sliding seals around the circulator and ,

steam generator shrouds. The lower floor is free to move i so that it does not experience high thermal stresses. It is only subject to the pressure differential across the i circulators which results in low mechanical stresses. l Therefore no concern has been identified which would re-quire that the structural integrity of the lower floor be monitored. l The loop divider plate is a vertical steel plate which 1 separates the circulator inlet plenum in two parts so that )

one loop can operate without drawing hot helium on the j shutdown steam generator of the other loop, thus preventing i damage.to that steam generator. The divider plate is i attached by its upper end to the lower floor. The lower part of the divider plate is free to slide between two flexible seals which assure leak tightness between the two loops. As the lower floor, the divider plate is not subject to stresses which would raise a concern with respect to its integrity.

. Further, access to both the lower floor and the divider

, plate is obtainable only with major difficulties.

Therefore, no surveillance is recommended for either i the lower floor or the loop divider plate.

c) Proposed ASME Code requirements: Not applicable.

7. LIST OF REFERENCES

References:

1

1. PSC report EE-SR-0001. Surveillance inspection and ,

test criteria for the Fort St. Vrain nuclear generating  !

station. l

. 2. FSV FSAR Section 3 l

l

C e's k w.'

EE-ll-0003'Rev. 1 l B.ll  ;

1

7. (cont.) l

3. FSV technical specification: LCO 4.5.2, SR 5.2.22

4. FSV surveillance-procedure SR 5.2.22-X l

5. FSV system description SD-ll-5 l l

6. FSV drawings R1100-100, R1100-700, R1701 series

7. BNWL-B-359 Inservice monitoring of the strength of HTGR core sq ort structure by W.C. Morgan and F.L. Becker

8. BNWL-2239 Feasibility of monitoring the strength of HTGR core support graphite - Part I by W.C. Morgan and F.L. Becker

9. .,JREG/CR-0995 Feasibility of monitoring the strength of HTGR core support graphite - Part II by W.C. Morgan and F.L. Becker

10. FSV safety evaluation report: Docket No. 50-267 dated April 20, 1979 Amendment 20 to the operating license (PSC Ref. G-79070)

11. Draft ASME Code,Section XI, Division 2

12. PSC letter P-79279, F.E. Swart to S.A. Varga, dated November 26, 1979

13. 'PSC. letter P-78177, J. K. Fuller to W. P. Gammill, dated October 26, 1978 e

4

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