ML20209F567
| ML20209F567 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/25/1987 |
| From: | Beck J TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20209F518 | List: |
| References | |
| NUDOCS 8704300320 | |
| Download: ML20209F567 (53) | |
Text
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O COMANCHE PEAK RESPONSE TEAM RESULTS REPORT ISAP: VI.a
Title:
Gap Between Reactor Pressure Vessel 9
Reflective Insulation and the Biological Shield Wall !
REVISION 1 L
O CG hsue Co rdinator si,sla, Date Rdylew Team Leader C 3l19lt7 Dath C & M ./?e= L Johyd, Beck, Chairman CPRT-SRT 3/zrl87 Date O
naamaa A
u8%e
Fevision: 1 Page 1 of 52 RESULTS REPORT h
\s / ISAP VI.a Gap Between Reactor Pressure Vessel Reflective Insulation (RPVRI) and the Biological Shield Wall
1.0 DESCRIPTION
OF ISSUE IDENTIFIED BY NRC Issue VI.a was identified in SSER 8 (Reference 9.1, page K-180) as follows:
"The TRT investigated an allegation that the Unit I reactor pressure vessel outer wall was touching the concrete biological shield wall. A TRT review of existing documentation and discussions with TUEC personnel indicated that this allegation was not factual. However, a significant construction deficiency report, submitted pursuant to 10CFR50.55(e), on August 25, 1983, documented that unacceptable cooling occurred in the annulus between the RPVRI and the shield wall during hot functional testing, apparently because of the existence of an inadequately sized annulus gap and possibly because the presence of construction debris in the annulus. TUEC corrected the situation by modifications to allow increased air flow for proper heat dissipation and by f"' removal of the construction debris. TUEC representatives
(
indicated that testing to verify the adequacy of the cooling flow will take place when additional hot functional testing is conducted. Information gathered by the TRT during the investigation indicated that a design change in the RPVRI support ring (i.e., locating the ring outside rather than inside the insulation) resulted in a limited clearance between the RPVRI and the shield wall. The TRT review of the 50.55(e) report revealed that TUEC failed to: (1) address the fundamental issue of the design change impact on annulus cooling flow, and (2) determine whether Unit 2 was similarly affected."
2.0 ACTION IDENTIFIED BY NRC
( The actions to be taken regarding Issue VI.a were identified in i SSER 8 (Reference 9.1, pages K-180 and K-181) as follows:
"Accordingly, TUEC shall:
- (1) Review their procedures for approval of design changes to non-nuclear safety-related equipment, such as the j RPVRI, and make revisions as necessary to assure that
- such design changes do not adversely affect safety-related systems.
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Rsvision: 1 Paga 2 of 52
-s RESULTS REPORT ISAP VI.a (Cont'd) 2.0 ACTION IDENTIFIED BY NRC (Cont'd)
, (2) Review procedures for reporting significant design and construction deficiencies, pursuant to 10CFR50.55(e),
and make changes as necessary to assure that complete evaluations are conducted.
(3) Provide an analysis which verifies that the cooling flow in the annulus between the RPVRI and the shield wall of Unit 2 is adequate for the as-built condition.
(4) Finally, verify during future Unit I hot functional testing that completed mo,difications to the RPVRI support ring now allow adequate cooling air flow.
The TRT noted that control of debris in critical spaces between components and/or structures was identified as an issue, both in the investigation of this allegation and the civil / structural area item II.c (Maintenance of Air Gap Between Concrete Structures), contained in Darrell G.
Eisenhut's September 18, 1984, letter to TUEC. Accordingly, TUEC shall also:
(1) Identify areas in the plant having critical spacing
- between components and/or structures that are necessary for proper functioning of safety-related components, systems or structures in which unwanted debris may collect and be undetected or be difficult to remove; (2) Prior to fuel load, inspect the areas and spaces identified and remove debris; and, (3) Subsequent to fuel load, institute a program to minimize the collection of debris in critical spaces and periodically inspect these spaces and remove any debris which may be present."
1
3.0 BACKGROUND
During the hot functional testing of cooling flow in the annulus between the reactor pressure vessel reflective insulation and biological shield wall performed during early 1983 for Unit 1, an j inability to meet the test procedure criteria was noted (i.e.,
measured temperatures were too high) and documented on test deficiency reports TDR-908 and 1221. A review of the test data led O
f Rsvision: 1 Peg 2 3 of 52
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- RESULTS REPORT ISAP VI.a (Cont'd)
3.0 BACKGROUND
(Cont'd) to a remote visual inspection of the area around the reactor vessel to determine the cause of the observed high temperatures and high heat load input to the neutron detector cooling system. The visual inspection revealed that the reactor pressure vessel reflective insulation support ring extended into the annulus area (air flow space) between the concrete biological shield wall and the reflective insulation. The flow path was further restricted by i debris resting on the support ring and due to an apparent eccentricity between the support ring and biological shield wall, which resulted in an uneven space. The Unit I support ring was i
modified to increase air flow and to improve air flow distribution.
This was accomplished by providing holes in the support ring. All construction debris was also removed from the reactor cavity area.
In addition, the neutron detector cooling system was modified to
, compensate for the higher heat load.
A visual inspection of the Unit 2 annulus was performed during this 4
~
time frame to determine if a similar condition existed. This inspection revealed a similar flow constraint; i.e., RPVRI support j ring extended into the annuits with debris resting on top of this 4
ring (the above-mentioned eccentricity between the support ring and j shield wall did not exist for Unit 2). Therefore, the annulus was j, first cleared of debris prior to testing and evaluation of the Unit 2 cooling system. The tests indicated that although additional heat removal capacity was required there was no need to modify the J
insulation support channel as had been done in Unit 1. As with Unit 1, this was accomplished by modifying the neutron detector cooling system.
l Initial startup test procedure ISU-282A, " Containment and Feedwater Penetration Room Temperature Survey", was conducted during the late 1984, early 1985 Unit 1 plant heatup to satisfy certain deferred
- preoperational testing requirements. This included repetition of
- portions of the HFT related to the neutron detectors performed in 1983 which gave rise to the cooling adequacy question. Testing was l performed to assess RCS pipe penetration concrete temperatures, i containment average air temperature, steam generator and l pressuriser compartment air temperatures, neutron detector well exhaust air temperatures, and reactor vessel support cooling air i supply temperatures under hot standby conditions (nominal 557'F RCS
! temperature). The results of these tests were not available for I
the TRT at the time SSER #8 was released. These results are i discussed in Section 5.0.
!O i
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R vicion: 1 Pago 4 of 52 RESULTS REPORT i
s/ ISAP VI.a (Cont'd) 4.0 CPRT ACTION PLAN The objectives of this action plan were to:
Evaluate the adequacy of the testing and analyses performed to confirm that sufficient air flow is provided in the annulus between the RPVRI and the shield wall.
- Provide additional assurance that design changes to non-nuclear safety (NNS) items have not adversely affected safety-related systems.
Identify areas in the plant that have critical spacing where debris could collect and be undetected or difficult to remove, confirm that these spaces are free of significant debris, and provide procedural controls to assure that they will remain clear of debris.
These objectives were addressed through the four tasks described below. Two additional tasks related to these objectives, i.e.,
/ reviews of the process for reporting significant deficiencies
(, , pursuant to 10CFR Part 50.55(e) and of current and past housekeeping methods, are included in ISAPs VII.a.2, "Non-conformance and Corra.itive Action Systems", and VII.a.7,
" Housekeeping and System Clesnliness", respectively.
4.1 Scope and Methodology 4.1.1 Review of Annulus Cooling: At the time of the TRT investigation, the annulus between the reactor vessel insulation and the shield wall in both units had been cleaned, inspected and closed to prevent debris intrusion. In addition, the insulation support ring in Unit I had been modified to provide a more uniform air flow distribution and the cooling capacity of the Neutron Detector Cooling System for both units had been increased. The modified systems had been tested and the results were under evaluation by the Project.
These measures were all conducted as a direct result of the hot functional test program.
To confirm that adequate air cooling is provided in the reactor vessel to shield wall annulus following removal of debris and modifications, these tests and evaluations performed by the Project for both units were reviewed rs by the third-party.
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R vision: 1 Pego 5 of 52 RESULTS REPORT ISAP VI.a (Cont'd) 4.0 CPRT ACTION PLAN (Cont'd) 4.1.2 Evaluation of NNS Design Change Process: The purpose of this evaluation was to assess the adequacy of the process by which changas in the design of NNS items are reviewed to identify and resolve adverse interactions between the NNS items and safety-related items. This process consists of a combination of multi-discipline reviews of the NNS design changes, programs such as the Damage Study
- to address interactions that are not readily identified by review of the NNS design changes, and the use of criteria such as electrical separation that are included in the design of both the NNS and safety-related items to preclude adverse interactions.
Since portions of this process were being reviewed in other elements of the CPRT Program (i.e., ISAPs I.c.
" Electrical Ccnduit Supports". II.d, " Seismic Design of Control Room Ceiling Elements", and the Design Adequacy Program) the evaluation performed under thi's action plan focused on the adequacy of the overall scope of the NNS design change process and review of those areas O not being addressed in other parts of the CPRT Program.
4.1.3 Review of Selected NNS Design Changes: Implementation of the NNS design change process was evaluated by reviewing design changes involving NNS items. Criteria for selecting the items to be reviewed were established based on a review of the history of the design change process, the applicable procedures and the scope of other CPRT activities which address portions of the NNS design change process.
4.1.4 Review, Inspection and Maintenance of Critical Spaces:
The concern for critical spaces where debris could collect, be undetected or difficult to remove, and affect the proper functioning of safety-related equipment was addressed through the identification and inspection of critical spaces under this action plan and a review of housekeeping methods (see ISAP VII.a.7,
" Housekeeping and System Cleanliness"). The results of the critical spaces review and reinspection were utilized to establish an ongoing Operations QC inspection program to assure that the identified critical spaces remain free of debris prior to and after fuel load.
O
- The Damage Study Program was renamed Systems Interaction Program during the course of this Action Plan.
Revision: 1
'Pago 6 of 52 RESULTS REPORT ISAP VI.a (Cont'd) 4.0 CPRT ACTION PLAN (Cont'd)
ISAP VII.a.2, "Non-conformance and Corrective Action Systems",
will evaluate the procedures for reporting significant design and construction deficiencies pursuant to 10CFR Part 50.55(e).
The action plan contained provisions for the evaluation and/or modification of procedures or equipment found to be deficient in accordance with the requirements of Appendices E and H of the CPRT Program Plan. The extent to which it was necessary-to implement these provisions is discussed in Section 5.0 of this Results Report.
4.2 Participants Roles and Responsibilities The organizations and personnel that participated in this effort are listed below with their respective scopes of work.
4.2.1 TUGC0 Nuclear Engineering (TNE) 4.2.1.1 Scope Confirmed history of design change process
- Selected NNS design change package sample and conducted change package review
- Identified potential critical clearances Developed inspection plan for critical clearances 4.2.1.2 Personnel Mr. C. Moehlman Mechanical Engineer (TUGC0 Issue Coordinator) 4.2.2 TUCCO Operations 4.2.2.1 Scope Developed operations program for maintaining critical spaces
R: vision: 1 Ptg2 7 of 52 RESULTS REPORT O
ISAP VI.a (Cont'd)
I. 0 CPRT ACTION PLAN (Cont'd)
I. 2.2.2 Personnel Mr. R. A. Jones Manager Plant Operations 4.2.3 TUGC0 Quality Assurance 4.2.3.1 Scope Developed critical spaces inspection procedure Inspected critical clearances 4.2.3.2 Personnel Mr. P. Halstead Quality Control Manager 4.2.4 Gibbs & Hill 4.2.4.1 Scope Analyzed Unit 2 cooling system performance 4.2.4.2 Personnel Mr. E. Nicolaysen Mechanical Engineer 4.2.5 Third-Party Activities 4.2.5.1 Scope Reviewed cooling system calculations and test data Reviewed design change process Reviewed NNS design change package sample selection process Reviewed Project's re-review of selected NNS design change packages O -
Reviewed development of critical spaces list
Rcvision: 1 Pcg3 8 of 52 O RESULTS REPORT ISAP VI.a (Cont'd) 4.0 CPRT ACTION PLAN (Cont'd)
Reviewed criteria and plan for critical spaces inspection Overviewed critical spaces inspection Reviewed Operations program for maintenance of critical spaces Reviewed modifications to procedures Reviewed design calculations (no new calculations were generated in this ISAP)
Prepared Results Report 4.2.4.2 Personnel Mr. H. A. Levin TERA Corporation - CPRT Mechanical Review Team Leader Dr. J. R. Honekamp TERA Corporation - TRT Technical Manager Mr. J. C. Miller TERA Corporation - TRT Issues Manager Mr. P. L. Turi TERA Corporation - Issue Coordinator 4.3 Personnel Qualification Requirement Third-party participants in the implementation of this Action Plan met the personnel qualification and objectivity requirements of the CPRT Program Plan and its implementing procedures.
Inspections performed as part of the critical spaces program l
were conducted by TUCCO QC, with Southwest Research Institute (SwRI) performing a third-party overview. TUCC0 QC personnel l
meeting the appropriate requirements of ANSI N45.2.6,
" Qualification of Inspection Examination, and Testing l Personnel at Nuclear Power Plants", were trained to TUCCO
R;vicion: 1 Pcga 9 of 52 PESULTS REPORT ISAP VI.a (Cont'd) 4.0 CPRT ACTION PLAN (Cont'd) procedure CP-QP-11.24, " Cleanliness of Critical Spaces". The SwRI third-party inspector was certified to the SwRI Quality Assurance Program. The SwRI inspector was also indoctrinated to TUGC0 procedure CP-QP-11.24 and the "SwRI Surveillance Plan for Verification of the Implementation of Inspection Procedure CP-QP-11.24 Cleanliness of Critical Spaces".
Other participants were qualified to the requirements of the CPSES Quality Assurance Program or to the specific requirements of the CPRT Program Plan. Activities performed by other than third-party personnel were governed by the applicable principles of Section III.K. " Assurance of CPRT Program Quality", of the CPRT Program Plan.
4.4 Procedures Construction QC conducted critical space inspections using TUGC0 procedure CP-QP-11.24 " Cleanliness of Critical Spaces",
(q
")
which was developed as part of this Action Plan. A third-party overview of the critical space inspections was done by Southwest Research Institute. This overview was conducted using the SwRI surveillance procedure developed for this task (Reference 9.2).
4.5 Standards / Acceptance Criteria The acceptance criterion for the review of annulus cooling flow was the CPSES Technical Specification limit for exit air temperature. A revision to this temperature limit has been proposed in a change request submitted to the NRC for their review (TUGC0 letter TX.t-4418, April 30, 1985). The modified temperature limit for the reactor cavity is 150*F (Technical Specification Table 3.7-6).
The acceptance criterion for the critical spaces undergoing inspection required the space to be free of debris or obstructions that would have an impact on the functionality of the associated safety-related item.
The acceptance criteria for the safety-related/NNS interface evaluations were determined by the specific safety-related/NNS interface being evaluated. In general, acceptance required that the NNS entity not adversely affect the ability of the safety-related entity to perform its function.
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Rsvisiont' 1 Pcs2 10 of 52 RESULTS REPORT ,
l ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS
.This section discusses the implementation of the tasks defined ,
in Section 4.0 (related sections shown in parenthesis) and the '
l associated results. The review of the reactor cavity annulus cooling flow (Section 4.1.1) is addressed in Section 5.1. A discussion is presented in Section 5.2 detailing the cooling problem identified by the Project during hot functional testing. Section 5.3 ccntains an evaluation of the CPSES non-nuclear 6afety design change process (Section 4.1.2). The review of selected NNS design changes (Section 4.1.3) is addressed in Section 5.4. Section 5.5 describes the critical Spaces Progr.m (Section 4.1.4). In addition to discussion of the completion of the tasks listed in Section 4.1, this section also includes a summary of Discrepancy / Issue .
Resolution Reports * (DIRs) (Section 5.6), an assessment of ,
safety significance (Section 5.7), and an evaluation of root cause and generic implications in accordance with the requirements of the Program Plan (Section 5.8). A summary of the acronyms used in this section is presented in Table 1.
5.1 Adequacy of Cooling Flow in Reactor Cavity ;
- 5.1.1 Statement of Probles The HVAC system providing cooling to the reactor cavity was originally designed for a heat load capacity of !
nearly 300,000 BTU /hr (Reference 9.3). The cooling ;
capacity provided prevents the occurrence of high i temperatures in the cavity which could have a detrimental effect on both neutron detector equipment and shield wall / vessel support concrete. The maximum ,
temperature for normal operations for the detectors is 135'F. However, equipment qualification testing has ,'
established that the detectors can withstand exposure to 200'T for five years (Reference 9.4). ;
During hot functional testing (HFT) performed in early 1983, temperatures higher than that allowed by Technical Specification limits were identified and recorded (as high as 314'T in the reactor vessel
- The Discrepancy / Issue Revolution Report (DIR) is a form used to track the status, classification and resolution of all open issues and of all discrepancies identified during the implementation of O this Action Plan as well as the Design Adernacy Program (DAP).
Pcvision: 1 Pcge 11 of 52
<-~3 RESULTS REPORT N' 'l ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS support and neutron detector well area - Reference 9.5). In investigating the deficiency, the Project determined that the cause involved two factors:
(1) HFT data showed that the heat load in the reactor cavity was approximately 600,000 BTU /hr (instead of the 300,000 BTU /hr design value).
(2) The RPVRI insulation support ring channel was found to intrude into the annular air space, which created a substantial flow restriction that was further compounded by the accumulation of construction debris on top of the channel.
It is believed that the second factor gave rise to the former in that cooling air was forced through seams at the base of the RPVRI because of the flow restriction I~')T
(_ (i.e., support ring and debris) (see Figures 1 and 2).
The air proceeded up the space between the reactor vessel outer wall and RPVRI (not intended to be a flow path), acquiring a significant quantity of heat energy.
The combination of the flow of air up the vessel vall (exhausting at the top of the cavity) and the reduced air flow in the intended flow path caused temperatures in excess of the Technical Specification limits. These high temperatures also exceed the temperature above which deterioration of concrete could occur (150*F.
Reference 9.6) and, if left uncorrected for a long period of time, could have led to degradation of concrete under the reactor vessel supports. The concrete reactor vessel supports are on the " List of Quality Asrured Items" (Table 17A-1 in FSAR Section 17A). Because the supports perform a safety-related function, the potential deterioration of the underlying concrete has safety significance.
The Project corrected the Unit I situation by first remeving the debris from the top of the support ring channel and then cutting holes in the support ring in order to increase air flow rates (see Section 5.2 for discussion of support ring). Modifications were made to the Neutron Detector Cooling System to accommodate
, a conservative heat load of 600,000 BTU /hr. In addition Westinghouse made modifications near the reactor veseel flanges to reduce the air flow between
R; vision: 1 Pcga 12 of 52 n RESULTS REPORT I \
C/ ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) the vessel and RPVRI (Reference 9.7). Westinghouse also undertook an effort to remove all debris from inside the cavity region (Reference 9.8).
The Project also examined Unit 2 to determine whether a similar problem could occur. Debris was detected and removed from the Unit 2 reactor cavity (Reference 9.9) and the same cooling system modifications were made.
Air flow tests performed after the modifications on Unit 2 demonstrated that adequate flow distribution exists around the circumference of the cavity, making modifications to the support ring unnecessary.
After all modifications were completed, the HTT was reperformed for Unit I under initial Startup test procedure ISU-282A, " Containment and Feedwater Penetration Room Temperature Survey". The results of the testing demonstrated that the as-built condition is adequate. Calculations were developed for Unit 2 HFT Q conditions that demonstrated the adequacy of the cooling system for that unit. The third-party review of these tests and calculations are described in the following section. (It should be pointed out that, in addition to these calculations, the Unit 2 HFT will serve as a proof-test for Unit 2 cooling adequacy. The Unit 2 HFT will be conducted as part of the Unit 2 Startup Program.)
5.1.2 Third-Party Review The third-party review was conducted in two parts since cooling adequacy was demonstrated differently for each of the two units. The first part involved a review of the Unit 1 HFT data collected following modifications made to both the Neutron Detector Cooling System (increased capacity by adding another cooling cott) and the RPVRI support ring (removed debris and cut air flow holes in the channel). The second part involved both a review of Gibbs & Hill calculations that predict temperature behavior in the Unit 2 reactor cavity and a review of the results of air flow tests conducted after debris had been removed from the cavity annulus area.
The Unit 1 post-modification HTT results indicated that hs) v one of the eight exhaunt temperatures monitored (exhaust from the reactor cavity - see Figure 2)
R:vicicn; 1 Pego 13 of 52 RESULTS REPORT U ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) exceeded the allowable value (Reference 9.10). The Technical Specifications indicate a temperature limit of 135'F for the detector wells (see Technical Specification Table 3.7-6) but the test results indicated a temperature in excess of 140*F (recorded at the exhaust line temperature element located outside the biological shield wall). TUGC0 memo TCF-85103 (Reference 9.11) requested Engineering to evaluate the problem, while pointing out that temperature measurements (recorded in ISU-282A) using temporary instruments located in the neutron detector wells established that the temperatures at the neutron -
detectors were in the 80 to 88'F range (well below the 135'F Technical Specification limit for the neutron detectors) (see Figure 1). Engineering replied in TUGC0 meno TSG-7786 (Reference 9.12) that the temperatures recorded were acceptable and that the Technical Specification limit should be raised to 150'T A (A Technical Specification Change Request was attached Q to this memo and this request has been submitted to the NRC for their review in TUCCO letter TXX-4418. April 30, 1985 (Reference 9.13)]. It is noted that the increased temperature limit (150'F) is based on preventing concrete deterioration in the upper portion of the reactor cavity where the vessel supports are located. (As depicted in Figure 1. the neutron detector walls are well below the vessel support region.) It should further be noted that ACI allows local concrete areas to have increased temperatures, not to exceed 200'F.
The explanation as to the cause of the single " hot spot" identified in the HFT is derived from two physical conditions unique to the Unit I reactor cavity: 1) an apparent eccentricity between the support ring and inner surf ace of the shield wall, and 2) obstructions that prevented hole cutting at certain locations on the support ring. The non-uniform gap between the support ring and inner surface of the shield wall was observed by the Project using fiberoptic inspection techniques. This finding was reported in TUCCO memo CPP-13014 (Reference 9.14) as follows:
[b "The minimum clearance between the legs of the channel tron supporting the insulation and the cavity wall varied from approximately 7/8" at azimuth 180 to 1/8" or less at 0 azimuth."
R visicnt 1 Pcgt. 14 of 52 RESULTS REPORT
\b ISAP VI.a (Cont'd) 5.0 IMP 1.EMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
The small clearance at the 0 azimuth restricted cooling flow and consequently reduced heat removal from this area. It is this non-uniform gap that created the need to drill holes in the support ring, thereby increasing cooling flow rates in the narrow gap regions. Because the reactor vessel is concentric with the cavity (see SSER #8, pg K-101), the cause of the circumferential variance in gap width is suspected to be the cumulative effect of construction tolerances for both the biological shield wall and the insulation on the vessel.
TUCCO drawing 2323-M1-0550 (rev. CP-1) shows that the temperature element that recorded the high temperature .
(1-TE-5452) is located on the 12" diameter exhaust line \
at azimuth 0 (also 2323-M1-0551, rev. CP-2). Figure 2 depicts a cross-section of the upper reactor cavity region showing the exhaust pipe intake point. The p temperature element 1-TE-5452 is located in a Q downstream portion of this exhaust line outside the biological shield wall. TRANSCO (Westinghouse subcontractor for insulation) Drawing DR-4278C-12 shows the as-built condition of the support ring following modification (depicts hole locations). This drawing shows two areas, one on either side of the 0 azimuth, where holes in the support ring could not be made.
Between azimuth 0 and 25, the first 8.5 degrees of the ring contain holes, but the next 16 degrees do not (hardware obstructions made hole drilling impossible).
The position of obstructions and hence of the drilled holes is synusetric about the 0 azimuth. This situation left unchanged the reduced heat removal capacity for these specific locations. Further, the exhaust ports on either side of the one located at 0 azimuth (TE-5451, 315 azimuth and TE-5445, 45 artmuth) showed temperatures during the HFT that were essentially identical. These results support the supposition that the highest temperature (measured by TE-5452) occurred very close to the 0 azimuth.
From this information, it is concluded that the test results are acceptable, and that the cooling requirements for Unit I have been satisfied, provided that the Technical Specifications are changed to V
provide a limit of 150'T for the exit air temperature (per Ref erence 9.13).
R:visicn: t Pcg3 15 of 52 g RESULTS REPORT ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
Because the HTT had not yet been conducted on Unit 2, cooling adequacy for Unit 2 was demonstrated by the Project using a combination of Unit 2 reactor cavity air flow test data and Gibbs & Hill calculations.
TUGC0 memo SU-84350 (Reference 9.15) reported the results of the air flow tests which showed a uniform distribution of cooling flow at an adequate flow rate.
Gibbs & Hill used the Unit 2 flow rate data along with Westinghouse design information to predict the cooling flow temperatures at various points within the reactor cavity in calculation 2-0300-2, Revision 0.
Calculation 2-0300-2A, Revision 0, was developed to map the temperatures in the cooling loop using the heat loadings taken from the Unit 1 HTT data (pre-modification). The use of these data (i.e.,
higher heat loadings) assured conservative results because these heat leadings are considered higher than would be expected in l' nit 2 for the following reasons:
O C/ -
a unifora gap spacing between the support ring and biological shield wall will allow a more evenly distributed flow and less of a flow restriction (which caused the flow in Unit 1 to seep up between the vessel and insulation and acquire the heat) debris that accumulated at the top of the RPVR1 support channel has been removed.
The two calculations provide a range of results: the first being a design baseline and the second providing an approximation for the worst case as-built condition.
The third-party review of these calculations is documented in Reference 9.16. During this review, one discrepancy was identified. 01R D-0002 was issued to document the fact that the value of radiation heating from the biological shield wall in calculation 2-0300-2 (Reference 9.17) uses data from a superceded reference without supporting justification. This D1R has been classified as an observation since the discrepancy did not have a significant effect on the results of the calculation, i.e., no further corrective action was required. The calculations demonstrate the adequacy of V
R;visicnr 1 Pcgo 16 of 52
,s RESULTS REPORT I
V) ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESITLTS (Cont'd) the cooling system. Based on these calculations, the configuration of the Unit 2 cavity and the success of the Unit I ratest under hot functional conditions, it is expected that Unit 2 will also meet its HFT performance requirements.
5.2 Definition of RPVRI/ Cooling Problem At the core of the issue of cooling adequacy in the reactor cavity is the existence of the RPVRI support ring in the annular air space between the RPVR1 and the shield wall.* The Westinghouse Specification for the reflective insulation (Reference 9.18. Westinghouse Proprietary) states the following requirements:
" The reactor vessel coolant nozzles and/or reactor vessel supports shall support the insulation below the refueling seal ledge to include the shell, nozzle and O bottom head insulation. The support steel in this
( ,) region shall be contained within the insulation outline."
and, "The shell portion of the insulation and the nozzle insulation...shall not extend more than four (4) inches beyond the outside diameter of the reactor vessel at 90'F including canning and hardware..."
Gibbs & Hill used this specification in developing the structural design for the reactor cavity and for cooling considerations. The size of the gap between the RPVRI and the shield wall was fixed by Gibbs & Hill HVAC designers based on developing a certain air flow velocity. Gibbs & Hill received Revision 3 of the RPVRI Specification in January 1974 Comments / questions were transmitted to Westinghouse in April 1974 (GTN-1081). Westinghouse responded to G&H questions in August 1974 (VPT-0439), in which Westinghouse stated in regard to the RPVRIt "The insulation is supported from the primary nozzles.
Design specifics vary from supplier to supplier.
Fabrication drawings will be transmitted as they become available."
Reference 9.19 provides a tabulation of the related history and 1 tats the related documents.
Rwision: 1 Pcgo 17 of 52 RESULTS REPORT A
'V') ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
Gibbs & Hill received Revision 4 of the Westinghouse Specification in April 1976 as part of a transmittal of information related to the RPVRI (WPT-1064). This transmittal did not contain specific design information for the RPVRI. In July 1976, Gibbs & Hill issued for construction the structural d,rawings for the cavity area (drawings 2323-St-545 and 549).
Westinghouse issued purchase or ors to TRANSCO (subcontractor for insulation) in November 1976. TRANSCO completed drawings for Unit 1 RPVRI by late 1977, which Westinghouse subsequently
, approved and released for manufacturing. The insulation was sent to the site and installed in mid-1978. Up to this point in time Gibbs & Hill had not been sant any drawings depicting the support ring on the outside of the RPVRI. In addition, no correspondence or other records have been identified in which Westinghouse notified the Project that the support ring was not contained within the insulation outline as stated in the specification.
A)
' A design modification involving the support ring was authorized in June 1978 by Westinghouse. Site engineering identified a physical interference between the shield wall concrete at the nozzle penetration and the insulation support steel that is attached to the support ring on the RPVRI at one end and rests on the top of the vessel nozzles at the other I
(the interference was not the support ring itself). The modification involved changing the shape of the support steel i and its attachment to the ring. The support ring was not altered by this design change. Gibbs & Hill was notified of the modifications by Westinghouse's letter authorizing the change (WPT-2327).
Numerous transmittals and meetings occurred in 1979 involving the design of non-crush insulation on the vessel nomsles. Of particular concern to Gibbs & Hill at this time were interferences between this non-crush insulation and orifice rings around the nosales. A sketch was prepared by Westinghouse in March 1979 (sketch #SK-GW-031679) that clearly showed the existence of the flow restriction caused by the presence of the support ring in the annular air space, in addition to non-crush insulation interferences. A meeting occurred in September 1979 in which this sketch was presented and discussed (personnel from TUCCO, Westinghouse and Gibbs &
Hill attended this meeting). No forest transmittal has h v occurred between Westinghouse and Gibbs & Hill involving this sketch, nor has a written description of the flow restriction caused by the channel been formally transmitted.
R: vision: 1 Pcg2 18 of 52 RESULTS REPORT ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
The above discussion clearly indicates thr.t the cause of the problem was a breakdown in cosununication between Westinghouse and Gibbs & Hill during the development of the original design. While Westinghouse internally approved TRANSCO's i design, they did not inform Gibbs & Hill in an appropriate or
'. timply manner. Accordingly, TUGCO's design change process lannot be linked to this particular discrepancy. This i
interface communication problem was documented in DIR D-0016.
1 No other instances of communication breakdowns between Westinghouse and other CPSES contractors were identified during the implementation of this Action Plan. The DAP is
, addressing the subject of Westinghouse - contractor interface as part of the self-initiated review scope.
In view of the fact that the support ring, an NNS feature. .
affected a safety-related function, and in response to the TRT l request, an evaluation was initiated to address NNS designs
} where the potential exists for adversely affecting I
safety-related systems, structures or components. This 3 subject is covered in the following section.
5.3 Evaluation of the NNS Design Change Process i
The process used at CPSES to identify and resolve adverse interactions between NNS and safety-related items was j evaluated by the third-party. The results of this evaluation
! are provided in this Section. The review indicated that the I
CPSES process includedt 1 -
criteria imposed on the design of both safety-related j and NNS items to preclude certain types of interactions, and i 1 -
programs that address standard industry-recognized i
interactions (see Table 2). ;
1 i
- multi-discipline reviews of the design and design changes for both the safety-related and NNS items.
From a programmatic perspective, the scope of this process was considered adequate in that the above program elements encompass all the methods necessary to address these interactions. However, review of individual program elements
. was necessary to determine if the process was adequate.
t 1
R vicicn 1 Pcgo 19 of 52 RESULTS REPORT b
V ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
The Design Adequacy Program includes extensive reviews of the design criteria and their implementation for safety-related systems and components. These reviews include the design criteria related to controlling adverse interactions with NNS components (e.g., electrical separation, structural and process isolation of piping at system boundaries).
All of the programs listed in Table 2 except for the seismic interaction portion of the Damage Study Program are included within the scope of the Design Adequacy Program. The seismic interaction portion of the Damage Study Program is evaluated under ISAP II.d (" Seismic Design of Control Roon Ceiling Elements"). Seismic interactions by NNS (train C) conduit are evaluated under ISAP I.c (" Electrical Conduit Supports").
The remaining program segment to be evaluated under this action plan was the proess for the multi-discipline review and approval of NNS design changes since the programs listed in Table 2 (and those aspects of the design of safety-related Os components that are intended to control interactions with NNS components) were covered in other CPRT reviews. An evaluation was made of the design change process implemented at CPSES through mid-1985. The results of this evaluation are discussed in the following sections (Reference 9.20).
5.3.1 History of Design Change Mechanisms at CPSES (through aid-1985)
Design work for CPSES began following award of the CPSES Architect / Engineer contract in 1972 to Gibbs &
Hill. The design was initially prescribed in drawings and specifications, and changed by revisions of the base documents or by using Design Engineering / Change Deviation (DE/CD) forms. During this initial phase, all design and change documents were controlled by the Gibbs & Hill New York office. Whereas Gibbs & Hill New York made use of the DE/CD, safety-related changes were initiated at the site using a Design Change / Design Deviation (DC/DD) form. Non-safety-related changes were initiated by the Field Probles Action Request (FPAR). Non-safety clarifications were made using mechanisms such as the Field Information Clarification Request (FICR). These clarifications, however, did not effect a design change.
R; vision: 1 Ptg3 20 of 52 n RESULTS REPORT ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
The first major evolution in the design change process occurred in 1977, when the TUSI Design Group was established on site. This group also had the authority to initiate design changes at the site. Design changes were approved by Gibbs & Hill-delegated representatives at the site and subsequently transmitted to Gibbs &
Hill New York for engineering review and design verification, as appropriate. Work called for by a design change package was initiated after Gibbs & Hill site approval.
Three (3) major design-related developments were instituted in the 1977/1978 timeframe. The first was the consolidation of site design engineering into one organization called Comanche Peak Project Engineering (CPPE). The second development was the initiation of two site-issued design change documents to replace the FPAR and DC/DD the Design Change Authorization (DCA)
O v
and the Component Modification Card (CMC). DE/CDs continued to be issued by Gibbs & Hill New York; but after mid-1978, they were assigned's DCA or CMC number for site disposition. The CMC was developed to facilitate changes specific to piping and supports.
The DCA was more suitable for addressing changes to specifications and other design documents. The third change was TUCCO's decision to exercise the option to waive prior Gibbs & Hill approval of design changes (DCAs and CMCs only - drawing revisions continued to be reviewed before issuance). The appropriate change document (e.g., DCA or CMC) was completed and transmitted to Gibbs & Hill for review and approval in parallel with the field work release. If the requested change was not approved, the already-effected field changes were removed and the necessary modifications were made in accordance with the approved Gibbs & Hill design documents.
The next major evolution in the design process occurred in 1979, when the Pipe Support Engineering Group was formed to more effectively address the small-bore pipe hanger design effort. Although this group us,ed the same design change mechanisms as the other engineering organizations, it was responsible for the ensuing review process. Pipe Support Engineering (PSE) became O a discipline at CPSES in 1981, assuming responsibility for nil pipe support activities.
R vision: 1 Pcg3 21 of 52 RESULTS REPORT I \
V ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
TUSI Nuclear Engineering (TNE) assumed responsibility for various aspects of the design process in late 1982.
Soon afterwards, they also gained custody of the original vendor drawings. In late 1985, TNE became the principal engineering body on site, absorbing CPPE and PSE.
5.3.2 Non-Safety Design Change Mechanisms 5.3.2.1 Design Engineering / Change Deviation (DE/CD)
The DE/CD mechanism has been used by Gibbs &
Hill New York from the beginning of the design effort to document both safety and non-safety-related design changes. From 1975 to 1977, these forms, in addition to drawing revisions, effected design changes. During this period of time, approximately 150 DE/CDs O were issued. Following the institution of CMCs and DCAs, Gibbs & Hill DE/CDs were assigned a corresponding CMC or DCA number for disposition.
5.3.2.2 Field Problem Action Request (FPAR)
The FPAR was used by the site to initiate a non-safety design change. This mechanism was used from the beginning of the design effort until it was replaced with the CMC /DCA.
During its effective timeframe, approximately 700 TPARs were issued.
5.3.2.3 Design Change Authorization (DCA) and Component Modification Card (CMC)
The DCA and CMC mechanisms replaced the FPARs and site-originated DC/DDs, and were still in use as of mid-1985. The CMC lends itself to depicting hardware changes easily, particularly in the piping and support areas.
The Project used these forms to effect both safety and non-safety design changes. As of June 1985, there were approximately 26,000 DCAs and over 100,000 CMCs that had been written for Units 1 and 2 and common areas.
Rcvision: 1 Ptss 22 of 52 p RESUI.TS REPORT ISAP VI.a (Cont'd) 5.0 IMPLLNENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) 5.3.2.4 Vendor Drawings (Revised by the Project)
Throughout the course of the project, some vendor supplied equipment and related vendor drawings required modifications. In some situations, the vendor made the necessary changes on the drawings. However, in many cases the Project (CPPE/TNE) made the changes themselves and issued the drawing with a "CP" prefix to the revision number. There are approximately 2,000 such drawings including I both safety and non-safety items, including incorporation of previously issued DCAs )-
and/or new design work. Before being issued to the field, all vendor drawings are approved by Engineering (CPPE/TNE).
5.3.2.5 Gibbs & Hill Drawings (Revised by Gibbs &
Hill New York)
Much of the design is defined through the use of drawings, the bulk of which have been originated by the Architect / Engineer, Gibbs &
Hill. Towards the end of the design work, control of drawing updates was transferred to CPPE/TNE. However, the majority of the drawing revisions were made by Gibbs & Hill in their New York office. The population of drawing revisions as of the end of 1985 was approximately 55,000 and includes both safety and non-safety changes.
5.3.2.6 Gibbs & Hill Drawings (Updated by CPPE/TNE)
As mentioned above CPPE/TNE assumed responsibility for drawing updates towards the end of design work. When CPPE had occasion to revise a drawing, they reissued it (following Engineering review) with a "CP" prefix to the revision number. There have been approximately 10,000 such updates for which DCAs, CMCs, or new design work have been incorporated over the course of the project as of the end of 1985. As in the O above two cases, these revisions receive engineering approval prior to field issuance.
R;visicn: 1 Page 23 of 52 RESULTS REPORT
~
ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) 5.3.2.7 Pipe Support Engineering CMCs Due to the large quantity of hanger sketches, it became necessary for PSE to track their own drawings (through HITS - see below) as well as modifications to these drawings through the CMC mechanism. All CMCs come from a common pool and are sequentially numbered and controlled. To date, there are approximately 40,000 CMCs originated by PSE.
These CMCs include both safety and non-safety design changes.
5.3.2.8 Pipe Support Drawings The Hanger Inventory Tracking System (HITS) contains most of the drawings within the PSE scope of design. Not included are approximately 10,000 balance-of-plant v' typicals, which are in the DCC tracking system. Within HITS, the population of non-safety drawings is approximately 16,200, with each drawing having an average of one revision. Revisions to hanger sketches are usually performed when a significant change is required (otherwise, a CMC is used).
5.3.3 Evaluation of NNS Design Change Procedures The process for the review and approval of DCAs and CMCs in effect at the time the third-party investigation began (June 1985) is shown in Figure 3.
This process includes a verification of the NNS classification and any effects on safety-related equipment. The procedures that control this process are listed under each of the activities in Figure 3.
The overall process and the procedures involved are the same for both safety-related and NNS design changes except for the Design Verification activity. Design verification (same as 10CFR50, Appendix B design review) is performed for safety-related items and certain NNS items such as the fire protection nystem.
Design verification is not required for the majority of p the NNS design changes.
O
Rcvision! 1 PKg2 24 of 52
~_
RESULTS REPORT
\- /
ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
The key activity in the review of NNS design changes for identification and resolution of adverse interaction with safety-related items is the Engineering Review performed under procedure TNE-DC-8-1. This is a multi-discipline review where the affected disciplines are indicated on the Change Verification Checklist (CVC) by the reviewing engineer. ,
The use of a multi-discipline review for this purpose is a standard industry practice.
The third-party has reviewed the historical Gibbs &
Hill and TNE procedures that controlled the design process and concluded that the implementation of multidiscipline engineering reviews has been a basic function within the CPSES design process since the beginning (Reference 9.21).
5.4 Review <j, Selected NNS Design Changes b)
\,, 5.4.1 Selection of the Sample As discussed in Section 5.3 changes to the design of NNS items during the early stages of construction were made primarily by revisions to drawings and specifications. Later, CMCs and DCAs, or their predecessor change documents, were used to initiate many of these design changes. It was concluded that review of a representative sample of DCAs and CMCs would provide an effective test of the implementation of the NNS design change process. The basis for this conclusion was:
CMCs and DCAs have been in use throughout most of the design process (mid-1977 to present).
They were used by all disciplines and therefore include a broad spectrum of design activities.
+
I
- .. . - _ _. . - . . _ _ ~ .. . - - . .. - - . .-
Rsvision: 1 Pags 25 of 52 i
RESULTS REPORT ISAP VI.a (Cont'd)
- 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) 4 Both drawing / specification revisions and DCA/ CMC-initiated changes were subject to the same basic review process. However, because CMCs and DCAs were not always immediately incorporated in the base document at the time of their design review, the Project review process could be more complex than the review of an equivalent drawing or specification change, especially where there are multiple
- unincorporated changes outstanding against the base documents at the time of the review.
The review of drawing or specification
!. changes was performed prior to their p issuance, whereas review of CMCs/DCAs was.
sometimes performed after they were issued
! for construction. While this approach is an
! acceptable practice that has been used i elsewhere in the nuclear power industry, it
} increases the potential for adversely
< influencing the review process.
1
! The computerized drawing control system data base was j used to identify the population of CMCs and DCAs from
! which to sample. This data base'does not uniquely
! distinguish between NNS and safety-related CMCs and l' DCAs, but it has the capability to identify those CMCs and DCAs that do not require design verification. As shown in Figure 3, the CMCs and DCAs that do not i require design verification are NNS changes. However.
l some NNS design changes (e.g., fire protection system)
- are subject to design verification. In addition, some
{ CMCs and DCAs that were initially designated as ,
safety-related were later determined (during the design verification review) to be NNS changes. Thus, while i the group of CMCs and DCAs identified in the data base l as not requiring design verification are NNS changes, this segment does not include all of the NNS CMCs and DCAs in the population.
i iO l
l t
-, -----. --_____.---.--____.._.--,.,,m~-,--. . _ . ~ . _ . , , . _ , _ - _ , - _ _ - , - .
Rsvision: 1 Pags 26 of 52 RESULTS REPORT
\~/ ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) k'hile this limitation in the structure of the data base precluded sampling from the entire population of NNS CMCs and DCAs, the use of this data base was considered representative from an engineering viewpoint since all of the NNS CMCs and DCAs were subject to the same multi-discipline review process regardless of whether they subsequently went through part or all of the design verification process.
To provide a critical test of the effectiveness of the NNS review process in identifying potential interactions with safety-related items, screening criteria were established to eliminate those CMCs and DCAs which had little chance of leading to such interactions. In addition, it was anticipated that a common interaction resulting from NNS design changes would be seismic interactions of the type addressed by the Damage Study Program. Since this type of
()
( ,/
interaction was the subject of extensive investigations under ISAPs I.c, " Electrical Conduit Supports", and II.d. " Seismic Design of Control Room Ceiling Elements", it was concluded that the investigations under this action plan should focus on non-seismic types of interactions.
The screening criteria used in selecting the sample were:
The change must involve some physical alteration in the plant as opposed to changes only in the documentation.
- The change must be reflective of the currently approved plant configuration.
The change must be in a Category I structure.
The change should have the potential for interactions other than seismic, i
The NNS DCAs and CMCs reviewed were selected randomly and evaluated against these screening criteria.
- Sampling continued until 60 change packages were l 7s obtained. This sampling process required the review of l (' nearly 3800 NNS change packages to obtain 60 that met the above screening crite'ria (approximately 1~ of the
, - e . . _ , ... -
.- . _. -- ~,.-
Ravision: 1 Pags 27 of 52 s RESULTS REPORT ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) control population could meet the above screening criteria). As samples met these screening-criteria, j they entered the review phase.
A review of the sampling program conducted under this Action Plan was performed by the CPRT Statistical i
Advisor. The review consisted of evaluating the soundness of the sampling program and the computational i
~ accuracy of the selection process. The findings of this review identified errors made by the Project in the randem selection and associated documentation process. These errors appeared not to be systematic or i
biased in nature (e.g., errors of transposition) and were few in number compared to the total number of items checked. The findings were addressed by the third-party and corrections made to all identified documents. There was no impact to the 60 that entered the design change review portion of the investigation (Reference 9.37).
Therefore, it was determined by the third-party that the possible occurrence of other such errors in the sampling program would have no significant effect on the conclusions of the review effort for the following reasons:
The review of the sampling program conducted
- by the CPRT Statistical Advisor found no evidence of bias in the errors identified i
and, thus, it appears that the randomness of the selection process was not affected, The number-of errors identified by the CPRT Statistical Advisor were small in relation to the number of items checked (i.e., low error
$ rate),
1 Over 25% of the control population was screened in the process of identifying the 60 samples to be evaluated, and None of the errors identified and subsequently corrected had any effect on the
! sample examined (the 60 DCAs/CMCs), i.e.,
none passed through the screen such that further evaluation was required.
. . . _ . _ . _ , . - _ _ . _ , , - - , - - _ _ . _ , _ - - - _ . . . _ . _ . _ . _ _ _ _ , . _ . _ . . . . . . . . - - - _ _ ~ ~ ~ . . . . . _ _
Ravision: 1 Page 28 of 52 I,-~T RESULTS REPORT
\~- ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) 5.4.2 Results of the Review of Selected NNS Change Packages The review of the sample of NNS design change packages was performed in two steps. First, the change packages were re-reviewed by the Project using the process shown in Figure 3. After this review was completed the third-party conducted an independent review of each change package using a multi-discipline team (Reference 9.22).
The Project review of the sample of 60 NNS change packages did not identify any adverse interactions with safety-related systems, structures or components. The third-party review identified one case (DCA 21,181, Revision 0) in which an affected discipline was not included in the review of the change. This change, which authorized a generic increase in the thickness of insulation for a group of pipe sizes including
/T (m,) safety-related lines, had not been reviewed by the piping discipline to assess the effect of the change on pipe support or pipe stress levels. Further investigation by the Project and the third-party confirmed that the increased weight of insulation had not been accounted for in the pipe support and pipe stress analysis. This discrepancy was documented as DIR D-0053.
The Project initiated immediate corrective action (Reference 9.23) to assure that the piping re-analysis i
I being performed by Stone & Webster (SWEC) utilized the correct insulation sizes.
The Project conducted analytical studies to determine if the oversight had safety significance. The third-party reviewed these studies and concurred with l the Project conclusion that the oversight did not have safety-significance because the increase in stresses i
associated with piping, piping components, and supports l
were shown to be below code allowables (Reference 9.24). Additional assurance that the insulation changes will not adversely affect safety-related piping is provided by SWEC's decision to include insulation as a walkdown attribute in the as-built verification b
V l
l
Ravision: 1 Page 29 of 52 RESULTS REPORT f%
() ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) effort (Reference 9.25 and 9.26). Consistent with the above, DIR D-0053 was determined by the third-party to not be safety-significant (see Section 5.6). Three other DIRs were written by the third-party in reviewing the Project studies. DIR D-2372 documents the identification of certain Gibbs & Hill piping analyses that failed to include the weight of insulation. DIRs C-0200 and C-0201 documented discrepancies identified in the SWEC studies related to nozzle loadings. These DIRs do not affect the conclusions drawn regarding the outcome of the study and have been transferred to the DAP action plan addressing piping issues (DSAP IX).
Since no other discrepancies were encountered in the review of the 60 change packages, and since the discrepancy in DCA 21,181 was not safety-significant, no further investigation was necessary.
Two weaknesses were identified by the third-party in (s
A their investigation of the NNS DCA/ CMC design change review process as controlled by and documented on the CVC:
Documentation accompanying the DCA/ CMC and CVC was considered insufficient in some cases. The missing information (e.g.,
referenced documents, a complete discussion explaining the nature of the change, or the basis of the justification) made it difficult to arrive at a conclusive decision regarding the possible existence of adverse interactions. The information not included within the DCA/ CMC and CVC was subsequently found to exist (following third-party requests for specific documents, or by third-party contacting individuals knowledgeable of the design change and the associated equipment / structure). Once acquired, the additional information made determination of the existence of adverse interactions possible for the third-party reviewers. Although it is recognized that obtaining information by directly contacting the individual responsible for the design change may be a necessary part of the
P4 vision: 1 Pegs 30 of 52 r RESULTS REPORT
('
ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) process, the absence of any written bases for decisions made by affected parties encumbers the review for adverse interactions and, therefore, increases the risk for such interactions remaining undetected.
Instances were identified where the reviewer did not send the change package to all affected disciplines in the Interdisciplinary Review phase. The procedure (TNE-DC-8-1) did not provide any guidance to the reviewer on how this assignment was to be made, nor did this procedure require any check that the selection of affected disciplines made by the reviewer was complete.
Since initiating the NNS change process review, TUGC0 has established a new set of policies and procedures s
i j organized within the Nuclear Engineering and Operation
( Policies and Procedures manual. In addition to specific changes to the implementing procedures related to DCAs and CVCs, the establishment of a cohesive set of policies and procedures would result in a strengthening of documentation practices within the engineering organization. Confirmation of improvements in this area will be made by SRT in its overview of the i Project's response to programmatic issues related to design control.
l In the review process for design changes, the Project has committed (Reference 9.27) to modifying the procedure that controls review of DCAs (ECE-DC-8-1) by the inclusion of the following:
The Supervising Engineer will be responsible for the Interdisciplinary Review distribution, l -
The Supervising Engineer will use the review l matrix provided in ECE-DC-7 for guidance in assigning the Interdisciplinary Review distribution.for DCAs against drawings. For other documents (e.g., specifications), he
()
will use the original review distribution for the affected document.
l
Rsvision: 1 Pacs 31 of 52 RESULTS REPORT tm ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
The procedure now includes instructions which give disciplines charged to perform the Interdisciplinary Review the option of adding additional disciplines to those already indicated on the Interdisciplinary Review.
Reference 9.27 also states that this approach will be i adopted for other design change mechanisms used at CPSES. The third-party has concluded that these procedural changes will enhance the Interdisciplinary Review process in its ability to detect and prevent adverse interactions that may be caused by NNS design changes (Reference 9.22).
5.4.3, closure The sampling effort described in Sections 5.4.1 and
- g. 5.4.2 identified no occurrences of a NNS design change causing an adverse interaction. The third-party has concluded that there is reasonable assurance that the review process related to DCAs/CMCs to detect and correct such interactions was effective. Although this conclusion is based on sampling from the control population, it is considered to be applicable to the entire DCA/ CMC population because the control population was considered to be representative of the entire population (see Section 5.4.1). Because of the reasons cited at the beginning of Section 5.4.1 (which support the selection of the DCA/ CMC as an effective test of the overall NNS design change process), it is also possible to extend the conclusion reached for the DCA/ CMC review process to the entire NNS design change review process. This result from the sampling investigation, in concert with the thoroughness of the scope of hardware and design evaluations provided by the CPRT yields reasonable assurance that NNS design changes have not adversely affected safety-related j
systems.
.O
Rsvision: 1 Paga 32 of 52 RESULTS REPORT ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) 5.5 Critical Spaces Program This program had three objectives:
Identify spaces in the plant where debris could accumulate and be undetected or difficult to remove.
, These spaces were associated with safety-related equipment or structures such that the presence of debris in the space (e.g., flow path, seismic gap, thermal expansion slot, etc.) could jeopardize the item's functioning (hence, the term " critical spaces"),
Verify the cleanliness of the identified critical spaces through QC inspections, and Provide procedural controls to assure that the spaces remain clear.
The following sections further describe these objectives as well as the processes and results developed to achieve them.
5.5.1 Development of Critical Spaces List The first step in the investigation of critical spaces was the development of a general list of areas where debris could accumulate and impair the proper functioning of a safety-related item. The list was compiled with input from each Project discipline, Gibbs
& Hill, and the third-party. The resulting list consisted of 25 items as shown in Table 3. .
Following completion of this list, each item was considered for inspection. Some of the 25 items had already been inspected through various start-up test programs (i.e., not part of CPRT) or had been verified to be free of debris by other Project activities.
Table 4 lists those items exempted from further inspection and the reason for the exemption. The remaining items were subjected to further review to obtain a complete listing (population) for each item to facilitate the subsequent inspection. This final listing was also reviewed by third-party in addition to the justifications provided for the exclusion of items from the inspection program.
O
Ravision: 1 Page 33 of 52 RESULTS REPORT ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) 5.5.2 Critical Spaces Inspection The completed critical spaces list was then forwarded to QC for inspection. TUGC0 procedure CP-QP-11.24,
" Cleanliness of Critical Spaces", was developed for these pre-fuel load inspections with concurrence on the approach given by third-party. TUGC0 QC performed the inspections in accordance with this procedure for all completed plant areas. Some items on the critical spaces list, mostly in Unit 2, could not be inspected at this time due to construction activities that are on-going. These items will be inspected prior to transfer of responsibility to Operations (see Section 7.0). In addition, some items could not be inspected because the items were inaccessible. These were identified on NCRs for subsequent: disposition by TNE.
The Project will resolve the NCRs by either completing inspection of previously inaccessible items, or by O providing appropriate technical justification for h "use-as-is" dispositions. Since the corrective action is driven by NCRs, the Project's NCR technical review will verify the adequacy of NCRs containing "use-as-is" dispositions (Reference 9.28).
Southwest Research Institute (SwRI) performed a third-party overview of the QC inspections in accordance with the SwRI procedure (Reference 9.2). A summary of the SwRI overview is presented in their report (Reference 9.29). SwRI concluded that both the procedure and the QC personnel were effective in assuring that all identified items were properly inspected.
Af ter completing these inspections, the Project inspectors were requested to list the kinds of debris that were removed from the various spaces. This list showed that some critical spaces had accumulated various construction debris (Reference 9.30). These findings confirmed the TRT concern regarding debris in critical spaces [ debris in seismic gap (ISAP II.c) and debris found in the Reactor Cavity (References 9.8 and 9.9)]. The original concern was documented in DIR E-0268. The occurrence of debris in various critical spaces has been categorized as an unclassified (n) deviation. Root cause and generic implications associated with debris in critical spaces are addressed in Section 5.8.
. . . ~. - _ _ . - . - __ _ _ , - - _ . - __
-R3 vision: 1 Page 34 of 52-S RESULTS REPORT ISAP VI.a i
(Cont'd) i 5 . 0' IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) 5.5.3 Operations Program for Critical Spaces The TRT had required that TUGC0 define a program that specifically addresses the maintenance of critical spaces subsequent to fuel load (see Section 2.0). From a logistics standpoint, this program is most effectively initiated at the time of room turnover, i.e., the time at which Operations begins to accept responsibility for plant equipment. To assure the adequate maintenance of critical spaces at CPSES, it was agreed between Project Construction and Operations 1
organizations, and the third-party, that the Operations critical spaces program needed to be fully implemented at the time Operations responsibility for plant
, equipment began.
i The inspections described in Section 5.5.2 provided a
} " baseline" for establishing the cleanliness of most i
critical spaces. As the critical spaces in a rcom, I
d group of rooms, or an area of a building are verified as being satisfactory, Construction will inform Operations so that the spaces are incorporated into the Operations' surveillance plan (Reference 9.31).
4 The Operations program for critical spaces was reviewed
- by third-party (Reference 9.32). The elements of the program are detailed in TUGC0 memo QIM-86125. The key elements are the training program and procedural controls that include periodic inspection. The TUGC0 i memo includes a discussion of the training program
its content, to whom it applies, and how it will be implemented. Two procedures have been revised to include provisions for ensuring critical spaces are
' maintained. Station procedure STA-607 (" Housekeeping Control", Revision 7), now includes a definition of critical spaces as well as inspection requirements, i
Surveillance instruction QAI-001 (" Plant Housekeeping and Equipment Inspection Plan") also includes critical space definitions as well as instructions to verify the effectiveness of STA-607 in maintaining the cleanliness of critical spaces.
QIM-86125 includes as an attachment a Generic Critical
) .
Space List / Action Response summary. This summary
- 3 incorporates information from Construction and the i
results of the critical space " baseline" inspection to i
_ - - _ - - . . - . . _ , . . _ _ - - - - _ . . . ~ - . , _ , . _ . , _ _ . - . _ _
R; vision: 1 Pags 35 of 52 RESULTS REPORT
[T ks/ ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) define the scope of the Operations program. Each of the twenty-five critical space types is addressed and given a disposition as to the need for inspections under the Operations program. Some items have been excluded from the inspection scope due to their being addressed under an existing program / procedure (e.g.,
pneumatic instruments are periodically checked by I&C calibration, polar crane gaps are addressed by maintenance instruction MMI-317 Rev. 2). Each critical space type is adequately addressed within this summary in terms of those needing subsequent inspection as well as those that do not need to be in the Operations program. The summary is consistent with the surveillance instruction QAI-001 in that the list of Generic Critical Space items in QAI-001 includes all the items marked for inclusion in the Operations Critical Space Program with QIM-86125.
(,) The Operations' program as defined in QIM-86125 is i
(~'s considered to be adequate to maintain critical spaces free of debris subsequent to fuel load. In the interim between initial inspections under the new procedures and the time the facility has been completely turned ,
over to Operations' custody, the impact of having a i significant work force still in the plant may place a strain on housekeeping activities.
However, trending of inspection findings will allow TUGC0 to monitor housekeeping efforts. Should trending results indicate an adverse trend, TUGCO will conduct a comprehensive reinspection / reverification of critical spaces to provide assurance that all critical spaces are free of debris prior to fuel load (References 9.33 and 9.34).
5.6 Summary of DIRs Eleven (11) DIRs have been written relative to the VI.a Action Plan. They are summarized in Table 5. Four of the eleven l
DIRs were identified through review of external source documents. One of these DIRs (E-0985) documents the TRT
< uncertainty regarding the results of the Unit 1 post-modification HFT (test results were not available for TRT review at the time SSER 8 was published). This DIR has been O closed with the particular issue being unsubstantiated based on a third-party review concluding that test results were
~ . _ ~ _ _ _. _ _ - _..__ _ _ ___.._.. _ _ . _ _ _ _ . . _ . _ _ _ ___ _ __ _ _ _ _ _ _ _
Rsvision: 1 Pags 36 of 52
,, RESULTS REPORT k_- ISAP VI.a (Cont'd) 1 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) adequate. DIR E-1273 has also been closed with a classification of " unsubstantiated". This DIR related to the concern that non-nuclear safety design changes could have an adverse impact on safety-related equipment. The investigation described in Section 5.4 concluded that a basis for this concern does not exist.
Another external source DIR (E-0268) documents the TRT concern that debris may exist in critical spaces. This DIR has been categorized as an unclassified deviation as a result of the stated concern (debris in critical spaces) being confirmed through inspections. Root cause and generic implications is addressed for this DIR in Section 5.8.
The fourth external source DIR (E-0666) has been categorized as unsubstantiated based on the conclusions reached by the investigations described in Sections 5.3 and 5.4.
"'g Of the remaining seven DIRs, two (D-0002 and D-0016) have been y ,/ closed as observations, i.e., the identified discrepancies have been determined not to be violations of design criteria, commitments, or specifications. Another DIR (D-0001) has been found to be unsubstantiated as a result of further investigation into the related issue (control of design input).
DIR D-0053 has been categorized as a deviation. The investigation described in Section 5.4.2 concluded that the deviation did not have safety significance.
, The remaining three DIRs (D-2372, C-0200, and C-0201) were initiated within ISAP VI.a. but have been transferred to the action plan addressing piping issues within DAP (DSAP IX).
5.7 Safety Significance Assessment 1
In the course of implementing this Action Plan two deviations have been identified. The first was documented on DIR E-0268
- and involves confirming the existence of debris in critical spaces. Because the Action Plan specifically required the inspection of all critical spaces to verify cleanliness, and since final corrective action is being taken in response to all identified cleanliness infractions, an assessment of safety significance is not required by Appendix E of the i
I Program Plan. The DIR has been categorized as an unclassified deviation. Root cause and generic implications are assessed in Section 5.8.
--.n- .-c,. -w - , , - . - - - . _ , - - - - . , , . . . . - - . - , ,
Rsvision: 1 Page 37 of 52 RESULTS REPORT
\~ / ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
The second deviation was identified in conducting the review of NNS design changes and their potential for adversely affecting safety-related items. The selected design change (DCA 21,181) increases the thickness of insulation associated with piping 6" in diameter and smaller, by altering a table in the insulation specification (2323-MS-30). The associated DIR (D-0053) was classified a deviation because the engineering review for this NNS DCA failed to identify the safety-related piping analysis specification (2323-MS-200) as an affected document (MS-200 contained a copy of the table in MS-30 that was revised by DCA 21,181) which resulted in the Gibbs & Hill New York piping analysis group being unaware of the change.
The piping analysis for 6" and under lines after this change did not accurately account for the additional mass associated with the increased insulation thickness. As described in Section 5.4.2, the Project performed studies to show that the increase in weight would not result in a safety-significant impact to piping or supports. The third-party reviewed and
(N
,\s,)
concurred with these studies. Accordingly, the associated DIR has been classified as a deviation.
5.8 Root Cause and Generic Implication Assessments 5.8.1 Debris in Critical Spaces The circumstances that lead to debris existing in the various critical spaces inspected as part of this ISAP were examined from two perspectives: (1) why was debris allowed to enter the space, and (2) why was the debris not detected and removed prior to this ISAP?
This section summarizes the potential causes that were evaluated and the conclusions reached.
The source of debris that accumulated in critical spaces cannot be confined to any one particular phase of plant construction or to any one organization.
Construction and modification activities extend into the startup testing program and, following turnover, continue into the operations phase. Various areas and systems at CPSES have gone through the entire turnover process (e.g., Construction has turned a system over to Startup who, in turn, has turned the system over to Operations). At the time of conducting the critical spaces inspection, the Fuel Building was the only area O in the plant in the custody of Operations (other areas
Rovicion: 1 Page 38 of 52 a RESULTS REPORT D ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd) that had previously been turned-over to Operations were turned-back to the Construction organization to facilitate the reinspection required as part of the CPRT effort.)
Since debris has been found in critical spaces that are Operations' responsibility (i.e., debris found in '
turned-over equipment), and since it cannot be ascertained from the critical space inspection results at what point in time the identified debris entered the space, it is reasonable to assume each organization shares some responsibility for allowing debris to collect and remain undete'cted in a critical space.
This is supported by the fact that the work activities themselves in general do not change from one phase of plant status to another (e.g., the work done in an electrical cabinet and the kinds of debris associated with this work is independent of whether the cabinet is
,Q under Construction, Startup or Operations control).
O The source of debris is also generally difficult to link with a specific construction activity. For example, a block of wood in a pipe sleeve may have been placed as a temporary support for the pipe during l
erection of the line early in the construction phase.
Alternatively, the wood block may have been inadvertently forgotten in the sleeve during the
, removal of temporary scaffolding completed recently.
l l
Typically, emphasis on verification of cleanliness (and identifying and correcting nonconformances) increases
! as the plant evolves from a construction site to an operating plant, with personnel and material access rigidly controlled (personnel access is controlled in the Fuel Building but material access control had not yet been instituted at the time of conducting the critical spaces inspection). Moreover, the question of why debris had not been detected and removed prior to the critical spaces inspections is more significant from the standpoint of root cause and generic implications than how debris entered the space initially.
O
Rsvision: 1 Paga 39 of 52 RESULTS REPORT V ISAP VI.a (Cont'd) 5.0 IMPLEMENTATION OF ACTION PLAN AND DISCUSSION OF RESULTS (Cont'd)
The control of debris is typically addressed through programs for housekeeping and housekeeping surveillance. The QA/QC Review Team was requested to evaluate these programs in light of the findings from the critical spaces inspection (Reference 9.35). The QA/QC Review Team response (Reference 9.36) made three observations:
- 1. The historical housekeeping and housekeeping surveillance programs were typical of normal industry practice,
- 2. Housekeeping and housekeeping surveillance procedures did not address the cleanliness of critical spaces, and
- 3. Design specifications did not identify the critical nature of these spaces or define requirements for cleanliness and protection.
Based on the above, the QA/QC Review Team postulated that the root cause was the failure of design specifications to adequately define cleanliness requirements for critical spaces (documented within DIR E-0268).
The corrective actions implemented as part of this Action Plan comply with the requirements imposed upon TUGC0 by the TRT. The corrective actions described in Section 5.5 conservatively anticipate inadequacies in both procedural requirements and implementation, and comprise an aggressive approach to ensuring the cleanliness of critical spaces. Because of the comprehensive nature of the critical spaces program and l the corrective actions implemented, generic implications considerations are considered to be completely addressed through the corrective actions implemented.
6.0 CONCLUSION
S Based on the third-party reviews of hot functional testing results for Unit 1, and review of flow tests and calculations for Unit 2, cooling flow in the annulus between the RPVRI and biological shield wall for both units is adequate. The reviews were based on tests and calculations that consider the current as-built configuration
Ravicion: L Pcgs 40 of 52 RESULTS REPORT ISAP VI.a (Cont'd) j
6.0 CONCLUSION
S (Cont'd)
(i.e., post-modification). Although not within the review scope of this Action Plan, it is noteworthy that further proof of Unit 2 cooling adequacy will be demonstrated by the hot functional testing required as part of the Unit 2 Startup Program.
A review of the circumstances that gave rise to the issue (RPVRI support ring) was conducted by the third-party. It was concluded that the cause of the problem was a breakdown in communication between Westinghouse and Gibbs & Hill during the development of the original insulation design. No other instances of communication
, breakdowns between Westinghouse and other CPSES contractors weru i
identified during the implementation of this Action Plan. The DAP is addressing the subject of Westinghouse - contractor interface as part of the self-initiated review scope.
The investigation of the potential for NNS design changes to have an adverse impact on safety-related systems, structures, or components caused the third-party to conclude that there is no basis p to suspect that this has occurred. No safety-significant V interactions were identified during the course of the review.
This, and the thoroughness of the scope of hardware and design
- evaluations provided by the CPRT, yields reasonable assurance that
! NNS design changes have not adversely affected safety-related j systems. Weaknesses in the design change review procedures have been identified by both the third-party and the Project, with the Project committing to strengthen these procedures. The procedural changes proposed by the Project were determined to be acceptable by i the third-party.
I The Critical Spaces Program implemented by this Action Plan achieved the objectives established to satisfy the TRT l requirements, namely:
l -
Spaces in the plant where debris may collect and be j undetected or difficult to remove were identified.
The identified spaces are now verified to be free of debris by existing records or by inspections done as j part of this Action Plan. Nonconformances were
) identified and the appropriate corrective action taken (i.e., debris removal or development of a technical
, justification for not inspecting and/or removing debris j in specific cases), and iO
Rsvision: I Pago 41 of 52 ,
)
i I
RESULTS REPORT O- ISAP VI.a (Cont'd)
6.0 CONCLUSION
S (Cont'd)
) -
A program (i.e., procedures and training) to minimize the collection of debris in critical spaces following turnover.co Operations has been developed. *
,4 i.
Each phase of the Critical Spaces Program listed above has been '
reviewed or overviewed by third-party and has been was determined to be adequate.
i j The third-party has evaluated the unclassified deviation associated' i
with debris found in critical spaces to identify root causes and !
generic implications. This evaluation found that the failure of "
, design specifications to define cleanliness requirements for -
I critical spaces was the probable root cause. It was concluded j that, due to the comprehensive nature of the corrective actions 1
implemented, generic implication considerations have been
- completely addressed.
- 7.0 ONGOING ACTIVITIES 4 .
i j
As noted above, the adequacy of the Unit 2 reactor cavity cooling '
i flow will.be further verified (i.e., in addition to the
- calculations prepared and reviewed as part of the scope of this l Action Plan) by hot functional testing required as part of the Unit i 2 start-up program. This activity will add to the assurance that ,
i the cooling system for the reactor cavity meets FSAR/ Technical l Specification requirements.
l Based on the satisfactory results of the NNS desige change i investigation, there are no ongoing activities related to this l l area. '
4 4
4 I
Inspections of identified critical spaces will continue as the associated equipment / structures become available through the room turnover process. Nonconformances identified as part of the completed inspections are being resolved through NCR dispositions.
. The NCRs with dispositions that exempt an item (or items) from '
! inspections are subject to a technical review being conducted by j the Project.
8.0 ACTION TO PRECLUDE OCCURRENCE IN THE FUTURE No actions are required relative to the reactor cavity cooling '
problem. In regard to the NNS design change process, the Project has identified all design change mechanisms and is strengthening the corresponding procedures relating to the Interdisciplinary ,
Review. The improvement of documentation practices has been a ,
l i- !
i
Revision: 1 Pags 42 of 52 RESULTS REPORT b
d ISAP VI.a (Cont'd) 8.0 ACTION TO PRECLUDE OCCURRENCE IN THE FUTURE (Cont'd) principle consideration in both the restructuring of the CPSES Engineering organization as well as the development of a new set of plant engineering procedures.
As part of the Critical Spaces Program defined in this Action Plan, the third-party has reviewed procedures established by Operations designed to prevent the accumulation of debris in critical spaces.
The third-party has determined the procedures are adequate in this regard and satisfy the TRT requirement communicated to TUGCO.
9.0 REFERENCES
9.1 Safety Evaluation Report, Supplement 8 NUREG 0797, Related to the Operation of Comanche Peak Steam Electric Station Units 1 and 2, Docket Number 50-445 and 50-446, February, 1985.
9.2 SWRI Surveillance Procedure, "SWRI Surveillance Plan for Verification of the Implementation of Inspection Procedure y CP-QP-11.24, ' Cleanliness of Critical Spaces,' CPRT File Number VI.a.Sa.
9.3 Gibbs & Hill Calculation 2323-0300-2, Revision 2 " Containment Neutron Detector Well Ventilation System".
9.4 Impell memo, IMT-0509, 9/29/86, from J. Everett to Peter Turi,
" Neutron Detector Qualification Status".
9.5 TUCCO letter, TXX-4054, 9/26/83, from R. J. Gary (TUGCO) to G.
L. Madsen (NRC), re: SDAR-118.
9.6 American Concrete Institute Standard 349-85, " Code Requirements for Nuclear Safety Related Concrete Structures" Section 6.3.8.
9.7 Westinghouse letter WPT-6789, 11/18/83, from T. R. Puryear to J. T. Merritt, re: Reactor Cavity Cooling Problem.
9.8 Westinghouse Memo, MP-84-223, 8/8/84, from D. M. Trombola to Ray Moller, re: Debris Removed form RV Annulus - Unit #1.
9.9 Westinghouse Memo, MP-85-057, 1/28/85, from D. M. Trombola to R. L. Moller, re: Reactor Vessel Annulus Search and Retrieval.
Rsvicion: 1 Pcgs 43 of 52 RESULTS REPORT ISAP VI.a (Cont'd)
9.0 REFERENCES
(Cont'd) 9.10 TUGC0 Startup Test Summary Report, ISU-282A, Revision 2
" Containment and Feedwater Penetration Room Temperature Survey" 4
9.11 TUCCO Memo, TCF-85103, 1/8/85, " Neutron Detector Well Exhaust Air Temperatures", from R. A. Jones to R. D. Calder.
9.12 TUGC0 Memo TSG-7786, 1/8/85. " Neutron Detector Well Exhaust Air Temperatures", from F. W. Madden to R. A. Jones.
9.13 TUGC0 Letter, TXX-4418, 4/30/85, from J. W. Beck to V. S.
Noonan.
9.14 TUSI Memo, CPP-13014, 7/22/83, " Reactor Vessel Cavity High Temperature", from C. K. Moehlman to M. R. McBay.
9.15 TUGC0 Memo SU-84350, 4/23/84, " Air Flow Data for Unit II Neutron Detector Well Cooling System", from T. P. Miller to C.
K. Moehlman.
9.16 TENERA Memo, CPRT-054, 9/16/85, from P. Turi to C. Moehlman, CPRT File Number VI.a.4f.
9.17 Gibbs & Hill Calculation 2323-2-0300-2, Revision 0,
" Containment Neutron Detector Well Ventilation System".
9.18 Westinghouse Equipment Specification #676449, Revision 4,
' 1/30/76, " General Reactor Vessel Insulation Specification",
Westinghouse Proprietary.
9.19 TUGC0 Meno, CPP-18006, 3/14/85, from C. K. Moehlman to C.
, Mortgat.
9.20 TENERA Memo, 9/19/86, "ISAP VI.a Results Report Sections 5.3.1.
and 5.3.2", from J. Miller to R. Calder.
9.21 TENERA Memo, 8/15/86, " Design Control Procedure Review", from D. Timmins to P. Turi.
9.22 TENERA Memo, 2/27/87, " Evaluation of NNS Design Change Packages", from P. Turi to file.
l 9.23 TUGC0 Memo, CPPA-48988, 4/1/86, " Piping Insulation", from R.
P. Baker to J. C. Hicks.
O
R2 vision: 1 Pega 44 of 52 RESULTS REPORT ISAP VI.a (Cont'd) d
9.0 REFERENCES
(Cont'd) l 9.24 CPRT File Number VI.a.4e, review package associated with DCA i 21,181.
9.25 SWEC Calculation GENX-019. Revision 3, pg. 8A, " Impact Study -
Effect of Change in Insulation Type Thickness / Weight on Pipe Stress".
I
, 9.26 SWEC Calculation GENX-063, Revision 3, pg. 6C, " Generic ,
Calculation of Evaluation of Effects of Calcium Silicate Insulation Due to Increase in Thickness". !
9.27 TU Electric Letter, NE 4555, 2/19/87, from J. Krechting to J.
j Miller, " Interdisciplinary Reviews of Design Changes".
! 9.28 SWEC Project Procedure PP-041, "Nonconformance Evaluation r Procedure".
9.29 SWRI Report, " Surveillance Report for the Verification of the j Implementation of Inspection Procedure CP-QP-11.24 4
Cleanliness of Critical Spaces at CPSES, Unit 1", 8/86, CPRT i File Number VI.a.5C.
9.30 TUGC0 Memo TUQ-4361, 9/23/86, " Critical Spaces Program", from i J. D. Hicks to J. C. Keller.
l l 9.31 TUGC0 Meso. NP-0991, 9/29/86, " Maintenance of Critical Spaces", from R. E. Camp to R. A. Jones.
l' 9.32 TENERA Memo, 10/14/86, " Review of Operations Critical Spaces Progras", CPRT File Number VI.a.4f.
l 9.33 TUCCO Memo, 9/15/86, " Maintenance of Critical Spaces", froa l D.C. Snyder to S. Ali.
9.34 TUGC0 Memo, 9/22/86, " Maintenance of Critical Spaces", from D. ,
- C. Snyder to S. Ali.
l 9.35 QAP/DAP Document Interface Transmittal Form. Number S-1081, l 10/7/86, to M. Obert from J. Miller.
{ 9.36 ERC Memo. QA/QC-RT-4642, 11/24/86, " Evaluation of j
Housekeeping / Surveillance Relative to Critical Spaces", to J.
! Miller from G. W. Ross.
9.37 Memo, CPRT-848, 03/18/87, " Evaluation of Errors in Sampling Program", to VI.a File from F. Webster.
Revision: 1 Page 45 of 52 RESULTS REPORT
\.
ISAP VI.a (Cont'd)
Figure 1 Cross-Section View of Reactor Cavity
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R0 vision: 1 Page 46 of 52 RESULTS REPORT l ISAP VI.a
-Q (Cont'd)
Figure 2 Diagram Showing Relationship of Components Near RPVRI Support Ring
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R vicion: 1 Page 48 of 52 f- RESULTS REPORT k
ISAP VI.a (Cont'd)
Table 1 Summary of Acronyms / Abbreviations CMC Component Modification Card CPPE Comanche Peak Project Engineering CVC Change Verification Checklist DAP Design Adequacy Program DCA Design Change Authorization DCC Document Control Center DC/DD Design Change / Design Deviation DE/CD Design Engineering / Change Deviation DIR Discrepancy / Issue Resolution Reports DM Design Modification DSAP Discipline Specific Action Plan FICR Field Information Clarification Request FPAR Field Problem Action Request HFT Hot Functional Testing HVAC Heating. Ventilation, Air Conditioning ISAP Issue Specific Action Plan NCR Nonconformance Report NNS Non-Nuclear Safety PSE Pipe Support Engineering QC Quality Control RCS Reactor Coolant System RPVRI Reactor Pressure Vessel Reflective Insulation SwRI Scuchwest Research Institute TDR Test Deficiency Report TNE TUCCO (or TUEC or TUSI) Nuclear Engineering i
I l
O i
, . -. - _ _ . - - . _ _ . .- _ ~. -- - _.- . _. --
Revision: 1 Pags 49 of 52 RESULTS REPORT O
(m- ISAP VI.a (Cont'd)
- Table 2 Programs that Address Interactions Between Safety-Related and Non-Safety-Related Items
) CPSES Program Description CPRT Review Area 3 Damage Study Program
, Seismic Category 2/ Category 1 (General) ISAP II.d Train C Conduit Seismic Interactions ISAP I.c Moderate & High Energy Line Breaks DSAP X Internal Plant Flooding DSAP X Internal Plant Missiles DSAP X 4
Fire Protection Program DSAP XI Environmental Qualification Program DSAP XI Single Failure / Failure Modes and O Effects Analysis DSAP XI l
P 4
i i
.l I
t l
O l
- - -_ -- . . . - - - - - _ _ _ - _ _ . , _ ~ _ _ - - - - - _ , _ _ . . _ _ _ . . . . . , .,,._ - - ---- _ -__-_-..--- - -- _ -
RGvision: 1 Paga 50 of 52 RESULTS REPORT ISAP VI.a (Cont'd)
. Table 3 Critical Spaces List
- 1. Pipe Bumpers (Crush-pipe and Honeycomb types) .
- 2. Moment Limiting Components
- 3. U-Bar Type Pipe Whip Restraints
- 4. Metallic Expansion Joints (Bellows)
- 5. Expanding Equipment (slotted footings for equipment experiencing thermal growth)
- 6. RPVRI/ Shield Wall annular area
- 7. Containment Spray Sump Screens
- 8. Tank Diaphragas
- 9. Dry Pipe
- 10. Reactor Blowdown Tubes
- 11. Electrical Panels with ventilation requirements
- 12. Mechanical Penetrations
- 13. Critical Floor Drains
, 14. Fire Dampers
- 15. Platform Beans
- 16. Seismic gap at tops of secondary walls O
i ** 17. Polar crane gaps
- 18. Seismic gap between Containment Liner and interior structure
- 19. Tornado Vents
- 20. Guide Tube Pipe Supports
- 21. Rotating Platform Pivot Bearing
- 22. Pipe Vents
- 23. Pneumatic Instruments with one side vented to atmosphere 24 Steam Generator, Reactor Coolant Pumps, and Pressurizer Restraints
- 25. Building Separation Gaps
- Item transferred to ISAP II.c. " Maintenance of Air Gap Between Concrete Structures", inspection scope.
- Item transferred to ISAP VI.b " Polar Crane Shimming".
i
- O
R3 vision: 1 Page 51 of 52 RESULTS REPORT 3 ISAP VI.a y (Cont'd)
Table 4 Critical Space Items from Table 3 Exempt From Inspection item Reason For Exemption Moment Limiting Components Gaps verified by HFT U-Bar Type Pipe Whip Restraints Gaps verified by HFT RPVRI/ Shield Wall Annular Area Westinghouse contracted to remove debris. Area now protected by re-installing seal ring.
Containment Spray Sump Screens Readily inspectable by routine housekeeping activities (CPSES instruction QAI-001)
Electrical Panels Panels verified clear of debris during turnover process.
Operations procedures STA-607 and STA-612 ensure continued cleanliness Fire Dampers Outer perimeter of damper verified clear of debris as part of resolution to SDAR-132. Inner area verified clear by execution of start-up test procedure XCP-ME-3, and is maintained by operability tests required by Tech. Spec.
3/4.7.12.
Platform Beams Slotted holes to allow for thermal growth protected from significant accumulation of debris by washers.
Rotating Platform Pivot Bearing Cowling over bearing prevents debris intrusion (should not have been included on list)
Seismic Gaps at tops cf Addressed within ISAP II.c Secondary Walls Containment Wall / Internal Addressed within ISAP II.c Structure Polar Crane Addressed within ISAP VI.b Pipe Vents Not applicable at CPSES tshould not have been included on list)
Rsvision: 1 Paga 52 of 52 RESULTS REPORT ISAP VI.a (Cont'd)
Table 5 Sununary of Related DIRs RELATED REPORT DIR # TITLE CLASSIFICATION SECTIONS E-0268 Contact Between Reactor Unclassified 5.1,5.2.5.5 Insulation and Wall Deviation 5. 7, 5.8 E-0666 Impact of Non-Safety Unsubstantiated 5.2, 5.3, Related Design Changes 5.4 E-0985 Gap Between RPV Insulation Unsubstantiated 5.1 and Shield Wall E-1273 Reactor Pressure Vessel Unsubstantiated 5.4.2 Reflective Insulation D-0001 Control of Design Input Unsubstantiated N/A D-0002 Superseded Reference in Observation 5.1 NNS Calculation D-0016 Gibbs & Hill / Westinghouse Observation 5.1, 5.2 Interface D-0053 Incorrect Insulation Deviation 5.2, 5.3 Thickness use in Piping 5.4 Analysis D-2372 Analysis Consideration of Discrepancy
- 5.4.2 Piping Insulation C-0200 SWEC Piping Analysis Study -* 5.4.2 C-0201 SWEC Piping Analysis Study -* 5.4.2 O
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- Transferred to DSAP IX