ML18065A882

Responds to RAI Re Preliminary Thermal Annealing Rept & RAI Re Request for Approval of ASME Code Case N-557, In-Place Dry Annealing of PWR Nuclear Reactor Vessel (Section 11, Div 1)

ML18065A882
Person / Time
Site:	Palisades
Issue date:	08/26/1996
From:	Bordine T CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9609030282
Download: ML18065A882 (124)

Text

{{#Wiki_filter:consumers Power l'OWERIN& llllUllGAN"S l'IUllilfUJ Palisades Nuclear Plant: 27780 Blue Star Memorial High~ay, Covert, Ml 49043 August 26, 1996 U S Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT Thomas C. Bordlne Manager. Licensing

• RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PRELIMINARY THERMAL ANNEALING REPORT On June 27, 1996, the NRC requested additional information regarding the Preliminary Palisades Thermal Annealing Report which was submitted by letters dated October 12, 1995, December 1, 1_995, December 12, 1995, January 12, 1996(2), February 2, 1996, February 5, 1996, March 27, 1996 and April 29, 1996. Additional information was also requested concerning.our April 3, 1996 request for approval of the American Society of Mechanical Engineers (ASME) Code Case N-557, "In-Place Dry Annealing of a PWR Nuclear Reactor Vessel (Section XI, Division I)". The attachment to this letter provides the requested additional information.

SUMMARY

OF COMMITMENTS This letter contains no new commitments and no revisions to existing commitments. ~ -?*s=s;:;:? d - d4' Thomas C. Sardine Manager, Licensing CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector - Palisades Attachment A CMS' ENERGY COMPANY ~O\\ \\ '

i . -.... *-~. ' -* -..

• \\.,

... *. :~ '."'~- ATTACHMENT 1 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION DATED 06/27/96 REGARDING PRELIMINARY THERMAL ANNEALING REPORT (TAR) 122 Pages

Preliminary Thermal Annealing Report (TAR) Request for Additional Information Thermal and Stress Analvsis (Sections 1.3, 1.4, 1.5, 1. 7, 1,8, 2.1, 2.2, and 2.3) and ASME Code Case N-577

1.

By letter dated April 3, 1996, Consumers Power requested approval to use American Society of Mechanical Engineer (ASME) Code Case N-557, "In-Place Dry Annealing of a PWR Nuclear Reactor Vessel (Section XI, Division 1 ). " Figure 1 of Code Case N-557 contains stress categories and allowable stress limits for the thermal anneal. These stress categories and allowable limits are similar to those contained in Section Ill Subsection NB for design. The allowable. limit for secondary stre~ses is the same 3Sm limit contained in Section Ill. However, Section Ill requires the 3Sm limit to be met for the range of secondary stresses. Code Case N-557 is silent in this regard. Since the anneal will involve a heat up followed by a coo/down, the staff believes that it is appropriate to conside.r the full range of stresses in the evaluation of secondary stresses. Address the application of Code Case N-557 with regard to the full r,ange of secondary stress produced by the thermal anneal.

Response

Code Inquiry #IN-96-17 was passed by Section XI at the ASME Code meeting in Portland in August 1996. This Code Inquiry stated that the computed stress intensity value ihstead of the computed:stress intensity" rang*e be us*ed as the limit. Because the answe~ to the inquiry was terse, Mr. John P. Houstrop, Structural Consultant, Honorary Lifetime Member of the ASME Subcommittee XI, and one of the principal authors of the Code Case was requested to discuss the philosophy of the 3Sm limit aS, used in Code Case N-557. It must be emphasized that Mr. Houstrop's discussion is unofficial as far as the Code is concerned, and therefore is not a Code Consensus response. The RAI correctly states that the limit for the primary plus secondary stress intensities in the ASME B&PV Code, Section Ill, Subsection NB is 3Sm and applies to the range of the stress intensity, while the limit in Code Case N-557, also 3Sm, is silent on its application to the range of stress intensity. This difference was intentional. The task group charged with preparing the Code Case attempted to use as much as possible of the philosophy of the Section Ill Design Code, consistent with the recognized differences between an anneal construction-type activity and the design of a new component. The basis for the stress limits in Section Ill 1

were first revisited and then adapted as discussed below. In the development of th~ criteria for Section Ill, three types of stress limits were used. Primary stresses that relied on the experience proven "design by rule" philosophy to provide enough strength to resist burst from the primary or "real" loadings, Total stresses to resist cyclic damage from fatigue, Primary plus secondary stresses to validate the fatigue analysis based on elastic computations. This last criteria is the sole purpose of the 3Sm limit on primary plus secondary stress intensities and is discussed in detail in the ASME B&PV Code Section Ill criteria document. <1> In the development of the criteria for c.ode Case N-557, it was recogni;;rnd that none of the above criteria were directly applicable to an in-place anneal of.a reactor vessel. Discussions during the. Code Case development included the following:.* The fatigue analysis was not applicable. The. anneal is completely analogous to a one-time construction.activity. Just as with construction post-weld heat treatments, welding, forming, lifting, etc. that are not factored into fatigue, neither. shoulc::I ~ q11e-tirne aon.eal. ~ Th~.annea.I is. performed, ot'}.a non-operational, empty vessel. The only primary stresses are those resulting from dead weight that are essentially negligible but were included because they are simple to include and add generality to: the Code Case. Since the limits in Section Ill on the primary plus secondary stress range are to validate the fatigue analysis, they were not applicable, but it was recognized that something was necessary. After considerable discussion .it was unanimously concluded that a 3Sm limit on the stresses themselves, not the range, was conservative and appropriate. Section Ill has no specific limits for these construction stresses, but the 3Sm limit here will, in an elastic calculation, limit the total strains to twice the yield strain, or "Criteria of Section Ill of the ASME Boiler and Pressure Vessel Code for Nuclear Vessels," ASME, 1964. 2

approximately 0.4% and the maximum plastic strain to 0.2%, a very small value. The range was not included because, assuming that a cooldown is the mirror image of a heat up, the strains will simply reverse, but not increase, and cyclic plasticity is not a consideration. The conservatism in this rationale is derived from the fact that for all reactor vessel base materials included in the Code Case, Sm is derived from (1/3)Su, not (2/3)Sy, so that actual strains will be smaller. Additionally any strain hardening reduces them further. Thus, Figure 1 of the Code Case N-557, while similar in appearance to the hopper diagrams of Section Ill, uses a different but still conservative philosophy. Note 7 of this same figure applies additional restrictions to any nozzle-piping transitions that are affected by the anneal. Currently the way the Code Case is written it can be inferred that here the range of stresses is limited to 3Sm. Inclusion of the range was not intentional and this point is currently being discussed by the ASME Code Committees. It should be noted that the calculations described in Section 1. 7 and Appendix B to Section 1.7 of the TAR and used to determine compliance with Code Case N-557 were performed using 3-D finite element elastic-plastic analysis methods that can confuse the elastically derived limits of the Code Case. However, since all computed plastic strains are very, small, specifically in the base metal covered by the Code Case, the conelusions are still valid. It should also be recognized that all stress intensity values presented are maximum stress intensities, and comparison of these with the primary plus secondary limits can be quite conservative. (This is discussed in Section 1.7.D.4 of the TAR) Thermal and Stress Analysis (Sections 1.3, 1.4, 1:5,.. 1. 7, 1.81 2.1, 2.2, and 2.3) and ASME Code Case N-577 (Continued)

2.

Section 1. 3. D. 6 of the TAR indicates that the cold leg pipes are cold sprung to reduce system stresses resulting from vertical thermal growth of the reactor vessel (RV) nozzles. Describe how this cold spring was accounted for in the three-dimensional (3-0) elastic-plastic finite element model described in Section

1. 7 of the TAR.

Response

As described in the TAR Section 1.3.D.6, the cold spring effect is applied by raising and shimming the pump and support frame vertically one-half of the expected thermal growth during normal operation. This process reduces the system stresses resulting from vertical thermal growth of the reactor vessel nozzles during normal operation. 3

The Palisades 30 finite element analyses, as described in Section 1.7 of the TAR, were used to establish bounding thermal and stress results in order to assure that no harmful structural integrity effects to the reactor vessel and PCS piping will occur during an annealing operation. To obtain bounding thermal and stress results a number of conservative assumptions were made, among them conservatively excluding the cold spring effects. That is the centerline of the cold leg at the pump end was assumed to be at the same elevation as the centerline of the cold leg (inlet) nozzles at the start of the annealing operation. The result of this assumption is that the maximum vertical growth differential between the reactor vessel nozzles and the cold leg pump ends as calculated and reported in TAR Section 1. 7 is greater than it would have been if the cold spring effect had been included in the analysis. A larger vertical growth differential causes larger stresses at both the reactor vessel nozzles and the cold leg pump ends: Therefore, by excluding the cold spring effect, the nozzle and pipe

• stresses reported in the TAR Section.1.7 bound (are higher than) those that c~n be.

expected d~ring the actual anneal. Thermal and Stress Analvsis (Sections 1.3, 1.4, 1.5, 1. 7, 1.8, 2.1, 2.2, and 2.3) and ASME Code Case N-577 (Continued).

3.

Section 1. 5 of the TAR contains a description of the instrumentation to be used during the anneal. Provide a comparison of the instrumentation tq be used during the Palisades anneal with.the instrumentation used at the Marble Hill demonstration anneal. Describe-the differences between Palisades plant and the Marble Hill plant that would affect the measurements at comparable instrument locations. Also describe any differences between the analytical models for Palisades and Marble Hill that woµld. affect.the measurem~n_t

• : compafisons~at comparable instrument locations: Provide. a comparison
• between the predicted and measured results at the Marble Hill demonstration
• anneal. Discuss the implication of the Marble Hill.comparison with respect to the
• adequacy of the Palisades analytical results.

Response

A comparison of the *instrumentation to be used during the Palisades anneal with the instrumentation used at the Marble Hill demonstration anneal is given as Attachment A,. µltimately the document given in Attachment A is intended to lead to the development of a justification for the Palisades instrumentation monitoring approach and to changes

• in *this approach, if warranted.

The diffe.rences between Palisades and Marble Hill that would affect the measurements at comparable instrument locations are as follows: 4

Marble Hill Palisades Measurement Effect at Comparable Locations Nozzles 8 6 Marble Hill heat losses in the nozzle region, above Insulation 3" 4" and below the annealing Thickness zone, and through the Flange to Pipe . 85" 75.5" insulation are greater. This leads to larger C. L. Distance thermal gradients. Bottom Yes No Mounted Instrumentation Heat Exchanger Only for the Insulation at Higher temperatures would Insulation down comer the top and be expected for the Marble pipes and bottom guard Hill anneal in the region the bottom zone and the above the anneal zone. guard downcomer zone. pipes. The Marble;Hill anarytical model has.higher reactor vessel -support heat losses, which leads to cooler primary coolant pipe temperatures than for the Palisaqe,s analytical model. The other differences in the models are based on the physical differences given in the table above. In terms of the comparison between the predicted and measured results at the.Marble. Hill demonstration anneal, the anneal data is still in the process of being analyzed after the anneal was completed in early July 1996. The final report providing this comparison is scheduled to be provided to the ASME-DOE/SNL-lndustry Steering Committee in November 1996 for release and distribution. The rylarble Hill comparison will be used to ca lib.rate the analytical models for Palisades to achieve a "realistic" or more accurate representation of the temperature and stress distribution in the Palisades reactor vessel and attached PCS piping. The intent of this calibration is to establish more realistic structural integrity margins than established in the bounding analyses documented in Section 1.7 of the TAR and to determine displacements for use as administrative limits in Section 1.4 of the TAR. 5

The Marble Hill demonstration project is to compare the actual monitoring and dimensional data against the predicted values established in the Marble Hill thermal and stress model. Changes between the predicted and actual values will then be explained with respect to the Marble Hill project's thermal and stress model characteristics and factored into a revised model. The Palisades project will assess the Marble Hill project's model changes and make appropriate adjustments in the* Palisades' models. A revised Palisades thermal and stress analysis will be conducted with the model changes and incorporated into the TAR, where necessary. This process is expected to occur prior to the submittal of the final TAR Thermal and Stress Analvsis (Sections 1.3, 1.4, 1.5, 1. 7, 1.8, 2.1, 2.2, and 2.3) and ASME Code Case N-577 (Continued)

4.

Section 1. 5. C. 1. 3 of the TAR discusses the measurement bias uncertainty for the vessel internaUemperature sensor probes: The report indicates that correction factors developed for the probes will be verified at Marble Hill. Describe the method that will be used at Marble Hill to verify the correction factors used for Palisades..

Response

It is the intent of the Palisades annealing project to use the data from the annealing demonstration at Marble Hill as a means of verifying the Palisades correction factors: It should be noted, however, that this verification was not intended to be performed concurrent with Marble Hill, since this would unnecessarily complicate (and potentially compromise) the* objectives of that project. -lnstead,.a statistically significant numqer of Palisades prototype sensors were installed, and.the raw (uncorrected) data from these sensors was collected.along with the Marble Hill Annealing Demonstration Project (ADP) data. As of this date, the final correction factors for Palisades have not been established. Data from the Marble Hill ADP will be issued shortly as part of the deliverables for that project. At that point, data reduction and a complete correction factor analysis performed as part of the Palisades detailed design phase will begin. This effort is c.urrently scheduled to begin in September 1996 under the Palisades Phase II schedule. In addition to data from the twenty-four prototype sensors installed during the Marble Hill ADP, data for the establishment of correction factors for use at Palisades is available from the original prototype tests (four sensors and three tests). These tests, performed in a test stand constructed around a section of reactor vessel on loan from EPRI, provided valuable data and were instrumental in finalizing the design of the 6

prototype sensors used at Marble Hill. Based on preliminary investigations of the prototype test data, and informal observations made during the Marble Hill ADP, the performance of the sensor probes has been as anticipated. At this time, the Palisades probes are expected to be almost identical to the prototypes used at Marble Hill, and the estimated correction factors are expected to be as little as 25-35°F during heating and 5-10°F duririg soak. These factors are preliminary and are subject to change. There are several approaches that can be used to determine an appropriate correction factor for the Palisades sensor probes, and at this time no approach has been rejected from consideration. Potential approaches include the following: A temperature dependent factor based on a statistical curve fit of the' test data. A factor resulting from a complex surface fit of the test data based on several independent temperature measurements. A factor based on an analytical model that has been correlated to the test data. A simple fixed value correction factor during the annealing soak period may be appropriate. No correctic;m factor may be necessary if a slightly larger measurement uncertainty can be tolerated without interfering with the feasibility of the a.nnealing. The basic development of the correction factor will be based on the data available from the prototype tests. Once the best correction facfor candidates are established, the

• approaches will then be fine tuned on the Marble Hill data. Final verification that the
• correction factor is satisfactory will be based on the performance during a simulated annealing, using data from the Marble Hill ADP. A description of th.e. selected correction technique and the* final correction factors for Palisades will be included as part _of the,overallReactor Vessel--lnternal Temperature Measurement lnstrum1?nt.~ti.on.

Uncertainty Analysis. Formal issuance of th.is document is currently scheduled for" Janifary 1997 under the Palisades Phase II schedule. Thermal and Stress Analvsis (Sections 1.3, 1.4, 1.5, 1. 7, 1.8, 2.1, 2.2, and 2.3) and ASME Code Case N-577 (Continued)

5.

Section 1. 7. D. 1 of the TAR discusses the results of thermal and stress analyses using a two-dimensional (2-0) finite element model.* Clearly identify how these results were used to demonstrate compliance with ASME Code Case N-557.

Response

The 2D analysis was not used to directly demonstrate compliance with the ASME Code Case N-557. Compliance with the ASME Code Case N-557 is shown in the 3D analysis _where it was demonstrated that for the 3D annealing cases (Cases T1.1, 7

T1.1 P, T1.1 F, T1.2, T2, and T3, as described in Section 1.7 of the TAR) the reactor vessel base metal and adjacent piping undergo no plastic strain and the stresses are within the 3Sm limits. The origins of the 20 finite element analysis predate both the development and issuance of the ASME Code Case N-557. The 20 analysis was performed using elastic-plastic-creep methodology as a cost effective way to establish two important preliminary results: first that the creep and inelastic deformations of the reactor vessel are negligibly small for the times and temperatures of the anneal as required by Section 1.7 of Draft Regulatory Guide DG-1027 (now Regulatory Guide 1.162), and second to provide qualitative comparisons of parametric studies of various configurations and parameters to provide efficient input to the subsequent three dimensional analysis. The results of the 20 analysis demons'trate clearly that dimensional stability of the reactor vessel will be maintained (as shown in previously referenced analyses in Section 1.7 of the TAR). The stress and temperature requirements of the Palisades anneal also fall within the limitations of Table 1 of Code Case N-557 that control creep deformations and of Figure 1 of the Case that limits plastic deformations, providing additional verification of the Case, although verification was not an analysis objedi.ve. The parametric studies performed provided the basis for the design of the heat exchanger, realistic thermal emissivities. for the reactor vessel and the heat exchanger, the effects of RV wall thickness differences, demonstration that the chosen transient rates were reasonable, that equipmen_t boundary conditions were realistic, the effect of vessel claqding, and provided the starting place for the 30 analysis that justifies the anneal to Code Case N-557.

• Thermal and Stress Analvsis (Sections 1.3, 1.4; 1.5, 1. 7, 1.8, 2.1; 2.2, and 2:3) and ASME Code Case N-577 (Continued)

6.

Section 1. 7. D. 3 of the TAR contains a discussion of the evaluation of the surveillance capsule holder assemblies. Describe the evaluation to determine the acceptability of the welds to the RV wall and the fillet welds between the box beam and the bracket. Provide the acceptance criteria used during these evaluations. Describe any measures that will be taken during the anneal to limit the distortion of the surveillance capsule holder assemblies. ResOOnse: A thermal network analysis was made using the equivalent circuit of a heat exchanger fin. The network included the integrated effect of the conduction along the capsule holder assembly and the radiation/convection transfer from the box beam to the 8

surroundings. The following assumptions used in the thermal analysis permitted a conservative, yet reasonable estimate of the capsule holder assembly temperatures.

1)

Convection is neglected as compared to radiation.

2)

Emissivities of all surfaces have a value of 0.55.

3)

No net radiation occurs between the inside of the box beam walls and the sidewall of the box beam (e.g., surface perpendicular to the vessel).

4)

Radiation between the capsule holder assembly, the heat exchanger, and the reactor vessel wall is proportional to the respective view factors.

5)

The temperature distribution in the upper portion of the box beam is one dimensional with distance from the support, and can be calculated by the fin equation.

6)

Radiation heat transfer can be linearized with the reference temperature being the average temperature along the length of the box beam.

7)

Temperatures of the heat exchanger and of the vessel wall are not disturbed by the presence of the capsule holder assembly. The fin includes a 1 /4-section of the box beam, half of the support bracket and half the distance between the support brackets. The analysis is made conservative by only using the top and side portion of the box beam to conduct heat axially along the box b_eam. Heat is also conducted from the box beam to the RV wall by the support brackets. Temperatures at various locations on the surveillance capsule holder were obtained for three times in the heat-up period of the anneal when the vessel wall temperatures are 200, 500, and 800°F and at one time for the portion of the box beam located in -the upper guard zone. The average temperature difference between the box beam and the wall is greatest at the start of heating: The thermal analysis resultswere used as input to the capsule holder stress analysis. The capsule holder stress analysis was performed using a simple ANSYS beam model 'of the surveillance capsule holder assemblies, including the box beam and the nine pairs of support brackets anchored to the vessel wall. The model was loaded by applying the temperature distributions obtained from the thermal analyses. Displacements from the 20 vessel finite element analysis results were imposed at the nodes representing the support brackeURV wall interface. An elastic analysis was performed using the thermal conditions from the thermal analfses. Based on the elastic analysis, the welds to the RV wall were found to have a maximum weld stress intensity value of less than 3Sm. The 3Sm limit is based on the Code Case N-557 criteria. It should be noted that the RV weld stresses reported above are conservative since they are based on elastic analysis in which the box beam stresses were over 2 times the yield stresses. This is not realistic since the box beam 9

would undergo some plastic deformation before reaching such a high stress and therefore the RV weld stresses will be at least 40-50% lower than those qualified using the criteria above. - An inelastic analysis confirmed that the box beam would undergo plastic deformation that would significantly reduce the loads on the RV welds, and on the fillet welds connecting the box beam to the support brackets. The ASME Section Ill, Subsection NF Level D stress limits were used to qualify the fillet welds since these are not pressure boundary welds. Table 3552(b)-1 and Appendix F, paragraph F-1334.2, limits shear stresses to the lesser of 0. 72 Sy and 0.42 Sui where Sy and su* are the material yield and ultimate tensile strengths, respectively. The shear stresses were optionally furth~r limited to the Level C shear stress limits of 0.6 Sy, to provide added assurance against shear yield. The inelastic analysis fillet weld shear Stresses were found to be less than 0.6 Sy and therefore acceptable. The box beams:were evaluated for buckling. (overall and local). It was determined that the box beams will yield before they will buckle. The support bracket to RV wall welds and the fillet welds attaching the support brackets to the box beam were shown to retain their integrity, however the tube may yield. The measures to be taken during the anneaJ to limit distortion of the surveillance capsule holder assemblies are currently being *examined: These measures include *

• reducir;ig the thermal difference between the RV wall and the capsule hplder tube by -.

shielding or insulation. Additionally a refined thermal and stress analysis approach usihg a 30 finite element elastic-plastic methodology and various thermal distributions - (including a slowerheatup) throughout the annealing cycle to quantify maximum stresses and ~esid.ual qeformations -for comparison against fitup toleranc_es,is beir:ig examine.d. - Thermal and Stress Analvsis (Sections 1.3, 1.4, 1.5, 1. 7, 1.8, 2.1, 2.2, and 2.3) and ASME Code Case N-577 (Continued)

7.

Section 1. 7. D. 4 of the TAR contains a description of the evaluation results for the 3-D finite element model. The discussion of analysis case T 4 contains an evaluation of the strain to determine acceptability for Code Case N-557. Provide* the basis in Code Case N-557 that allows for this evaluation procedure.

Response

The Sm value in the ASME Section Ill code tables can be thought of as being less than 2/3 of the "design yield strength" for the material in evaluating the primary and secondary stresses. The draft basis document for Code Case N-557 states that "By 10

meeting the 3Smlimits as required by Figure 1 of the proposed Code Case, minimal residual strain should occur." This statement reflects that the use of the 3Sm limit on Code Case N-557 is quite different than the 3Sm limit on' primary plus secondary stress range in Section Ill. In Section Ill, it is a limit on the stress range to insure the applicability of the fatigue analysis. In Code Case N-557 the purpose of the 3Sm stress limit is to limit the total strain (elastic and plastic) to less than twice the yield strain to limit distorti_ons to acceptable levels. Thermal and Stress Analysis (Sections 1.3, 1.4, 1.5, 1.7, 1.8, 2.1, 2.2, and 2.3) and ASME Code Case N-577 (Continued)

8.

Section 1. 7. D. 4 of the TAR contains a discussion of the results of the evaluation of the RV flow skirt for analysis case T3. Figure 3-32 of Appendix 1. 7.8 contains a comparison of the stress-strain relationship used in the analysis for the. SB-168 material with test data. The comparison indicates a significant difference between the stress-strain relationship assumed in the analysis and the stress-strain relationship obtained from the actual test data. Describe the potential impact of the test data on the strain calculated in the analysis. Also describe the

• design criteria that is applicable to the flow skirt.

Response

The potential impact of the SB168 test data curve on the strain calculated in the analysis is small since the strain distributiori is really dependent on the overall structural response of the Palisades reactor vessel to the large axial thermal gradients*

at the* vessel ends (Case T;3).. -The flow skirt is a non-structural component designed to facilitat~ the primary coolant flow into the reactor core. The flo_w skirt is not designed'to carry significant loads and its stiffness is smal I compared to the-reactor vessel. It is acceptable to conclude that the SB168 test data curve would have a minimal impact on the reactor vessel strains.

As discussed in Section 1.2.E.2.2 of the TAR, the main effect of the thermal annealing operations on the flow skirt are expected to be stresses at the attachments resulting from thermal displacements. The displacements at these attachment points, the interface of the bottom head region and the flow skirt lugs, are mostly dependent on the SA302B base metal material and not on the SB168material.

• The deformations at this interface lead to secondary stresses in the flow skirt; that is stresses developed by a self-constrained (fixed displacement) structure. With the exception of small localized areas, these flow skirt secondary stresses are well within the elastic region and would not be impacted by the use of the test data curve. The region of localized plasticity shown in the attached figures would be impacted. The 11

stresses obtained using the test data would be smaller than those reported in Section 1.7.D.4 of the TAR and the strains would be somewhat larger but not much larger since the flow skirt is self-constrained. Since the strains being reported in Section 1.7.D.4 of the TAR are small (peak strains of about 2 percent), it was felt that the implementation of the test data would not change the overall conclusions of the analysis. The analysis demonstrated the acceptability of the flow skirt and the attachment points by predicting acceptably low strains during the anneal and acceptably small residual deformations after the anneal, such that the flow distribution function of the flow skirt would not be altered. 12

r-----

• ~I I l I I I

I I I I I I I I I I I I

r.

I I --":""----+/--,,,..__ J I. Region. with *l9c~li~~<l"""_ ~-~ .. -N__

_pl~s-~~i~~'.----

~

~ ~ -="'l""---:-1'----...;;:::--:__::;;;*:;;;;;:;: --- - --j I t ~ I I I I I I --+-----------+ - - ~ ~ - I I J I I / / ~ ,~ ........,,. ~ ""!: ~ I fsee 1 Figlire *2Y ' 1* ~ -~ ~-,. I i .! I I C~se T3 - L;;irge Axial Gradients at Vesse.l Ends Figure 1 for RAI #8 of.Therma! and Stress Analysis and ASME C_ode C~s(t __ N~S~!.. GJ 0

- ANSYS 5. 1 ~' JUt 17 1996. '11 :46:42 i\\NODAL SOL UT I ON,: I 5l-EP=1 I SUB *=1 TiME=33.2 SINT (AVG) DMX =1.562 SMIN =571.134 ~--............ SMX =52347 I_ A =3448 B =9200 I [ =14953 I D =20706 . E =26459 f =32212 G =37-965 I H =43718: I _=:4947_1. Localized plasticity due* , to bending of RV wall.

---

~""'-"--

.... -T--------, I I I I I I I I I I I I I I I I J I I I I I I I I Figure 2 for RAI #8 of Therm_ii! ~nd Stress Analysis and ASME Code Case N-557 ~ ,--i \\ 14 I I I

Thermal and Stress Analvsis (Sections 1.3, 1.4, 1.5, 1. 7, 1.8, 2.1, 2.2, and 2.3) and ASME Code Case N-577 (Continued)

9.

Appendix 1. 7. B of the TAR contains a description of the 3-0 finite element model used for the thermal and stress analysis of the reactor and primary coolant system for the Palisades anneal. The stress analysis used elastic-plastic calculational methods to calculate stresses. The TAR indicates that the vessel base metal did not experience plastic deformation. Since the criterion used in ASME Code Case N-557 is based on elastic analysis, describe the rationale for using the elastic-plastic calculational method to compute stresses. Provide a list of all elements in the model that exceeded the yield limit and describe the impact of the yielding on the results for the overall model that would have been obtained with an elastic analysis.

Response

The elastic-plastic calculational method was used fo( the Palisades stress analysis because it was expected that the cladding and the flow skirt would have regions* in which yield was exceeded and the use of the elastic-plastic method would best determine the effect of these non-structural components exceeding elastic limits on.the overall stresses of the gross deformed shape of the overall structure. Figure 5-67 of Appendix 1. TB of the TAR is a contour plot of the plastic strain intensity at time = 204 hours, *excluding.the cladding. Figure 5-67 shows that there is no plastic strain in the base metal (contour lin'es are only on the flow skirt flange). Plasticity is confined to very localized areas in the top flange of the flow skirt and in the cladding. The elements in the model that exceeded the yield limit tor the design basis ca~e are as follows: 640 elements in the reactor vessel cladding shown in the attached Figure 1.

• 8 elements in the flow skirt shown in attached Figure 2..

It should be noted that the calculations presented in Appendix 1. 7. B of the TAR were performed using finite element elastic-pl~stic methods. The Code Case N-557 allowables are implied for elastic analysis methods, however since all computed plastic strains are very small (less than 0.2% for the design base case 1.1 F), and no plasticity is present in the base metal covered by the Code Case, the conclusions are still valid. It should be recognized that all stress intensity values presented here are, because of the finite element calculation method used, maximum stress intensities, and comparison with primary plus secondary limits can be quite conservative. An elastic analysis of the vessel would have resulted in much higher stresses in the non-structural components of the vessel which locally yielded, the cladding and the 15

flow skirt top flange. The overall model deformations would be slightly smaller. The stresses in the base metal would be essentially the same, and the overall conclusion that the base metal stress intensities are less than 3Sm would remain unchanged. 16

-.~----- 1 ANSYS 5.1 JUL 19 1996 10 ::03: 48 NODAL SOLUTION STEP=3 I SUB =1 TIME=204 EPPLINT <AVG) OMX =1.919 SMX =0.001559 ! A =0.866E-04 . B =0. 260E-03 E =0.433E-03 El =0.606E-G3 E =0.779E-G3 lF =0.953E-03 G =0.001126 H - =0.001-299 I =0.001472 -- --------- ---- ---------- -- --- -~-1 ~ Plastic strains in ~ I ~ the qlaqding. There are 64<:felements in*the: -, Palisades structural. model. I I I _ l PALISADES 30 STRESS MODEL - Case T1.1F . *- -- Figure 1 for RAI #9 The~~~~~-~ Stress An~lys;s. ~n~~-~~E Code C-as_e_N __ 57 ________ _ I

17 I

\\ ______ _:

1* I*. 1 Maximum shown in

• Fjguie* '5-67 of Appendix.
• 1J7.B I

I ~-- 1 I I I I I I I I I I J I I I J J I I I f I I 1 r r I I I I I I -~------- I I I I l I

---

~

PALISADES 3D STRESS MOD~L r~-C-ase_jl.1F

~.-

Figure 2 for RAI #9 Thermal and Stress 1Analysis and ASME Code Case N-557

Thermal Annealing Operation (Section 1. 5)

1.

Provide analysis documentation to support the claim that the loss of any one sensor location can be adequately compensated for by interpolation between two or more adjacent sensors. Identify the location and effect (i.e., on temperature control and/or critical data accumulation) of the sensor position for which the greatest level of uncertainty would result from the interpolation procedure during each of the five annealing phases as defined in subsection 1. 5. D. 1.

Response

The heat exchanger is designed to emit circumferentially uniform radiant heat onto the reactor vessel. Both the Marble Hill and Palisades analysis results show that with exception of the nozzle region (due to the nozzles and the two variations of nozzle sizes), the temperature distribution is circumferentially uniform. The temperature "measurement sensors in Palisades are arranged in a series of rows and *columns as shown in Figures 1.5.8-1 and 1.5.8-2 of the TAR Section 1.5. Sensors in the same row are at the same elevation and sensors in the same column are at the same azimuthal

• location. For the nozzle region, additional temperature sensors have been added both above and below the nozzles to better define the temper:ature distribution. If both the primary and backup sensor fails, but the sensors in ttw rows above and below, and in the adjacent columns are working, enough redundancy would exist that an interpolation may *not be necessary. The sensor failure rates during the Marble Hill demonstration anneal was extremely low which also suggests that interpolation may not be necessary.

It is noted that the interpolation procedure was intended only for the temperature sensors although Section 1.5.B of the TAR did not explicitly state it. The final TAR will define the temperature sensors as the sensors being used in the interpolation procedure. If during or after the anneal a temperature at a certain location is judged to b_e needed due to a loss of sensor or for information purposes, the interpolation used would depend on the previously calculated finite element results at the annealing phase in question. For instance, during soak the value could be based on a linear interpolation between two sensors, based on finite element results showing linear gradients and circumferential uniformity, while during heatup and transitioning to soak it may require a higher order interpolation between two or more adjacent sensors, based on finite element results showing nonlinear gradients and circumferential variations. This approach is considered to represent an acceptable approximation. The actual interpolation approach will be established following the review and analysis of the

• Marble Hill annealing data. This interpolation approach will be a part of the justification for the Palisades instrumentation monitoring approach defined in Section 2.1 of the TAR This analysis will be provided to the NRC with the final TAR The sensor position for which the greatest level of uncertainty exists in terrns of 19

The sensor position for which the greatest level of uncertainty exists in terms of temperature using the interpolation procedure is where the largest thermal gradients with regard to structural integrity are anticipated to occur. For Palisades, these include the flow skirt region and across the vessel nozzles. The uncertainties in temperature will be larger during the time when the largest thermal gradients are occurring (e.g., during the heating of the reactor vessel, transitioning into the RV annealing soak band, and the RV cooldown). Heating the heat exchanger up to 700°F and the soak period are not as critical. These uncertainties, however, will be reduced due to redundant sensors at symmetrical locations that have a smaller tolerance in placement which can be used in the interpolation procedure. The effect of these temperature uncertainties will be incorporated into the procedural limits as defined on Section 2.1.A of the TAR Thermal Annealing Operation (Section 1.5) (Continued)

2.

Confirm that the instrumentation of the external temperature measurement sensor array described in subsection 1.5.B.2 conforms to the single location failure stipulation as presented in question #1. Re~ponse: The same approach of using finite element results to establish an appropriate interpolation for a given time and location as described in RAI #1, Thermal Annealing Operation will be used for the external sensors. Thermal Annealing Operation (Section 1. 5) (Continued)

3.

Section 1.5.D, subsection B, notes that all potential liquid flow paths to the reactor vessel will b~ either drained or tagged out of service. Specify how each of the flow paths will be isolated and confirmed to prevent a water intrusion during the annealing process.

Response

The plans for isolation of water from the reactor vessel during annealing are in a very early stage. Attachment C provides our current thinking, including a Reactor Annealing Water Isolation Draining/Tagging Plan, which is provided in a draft stage for information. Plant conditions and configuration change with time and the isolation plan will have to fit the exact plant configuratiori at the time of the actual anneal. Therefore, we can only commit to the general principles to be applied such as those listed in TAR Section 1.5.D. The isolation plan will be reviewed and revised as necessary by members of the Plant Operations Department prior to the implementation of the reactor vessel anneal. 20

Department independently reviewing the isolation plan. The actual valve positioning and equipment protective tagging (yellow caution tags) will be subjected to independent verification. Monitoring of tank levels and leakage/drain paths will be conducted to check for possible sources of water. The details of these monitoring points will be determined at a later date. 21

Materials!Fluence (Section 1. 1) Section 1.1.E discusses the surveillance program (also discussed in Sections 1.2 and 3. 0). With regard to the recent revisions to the reactor pressure vessel (RPV) f/uence estimates, what are the impacts on the thermal annealing submittal:

Response

If the revised fluence analysis is accepted by the NRC the TAR will not be implemented as currently written. The extent of the revision will depend on the results of the NRC's review. Materials!Fluence (Section 1. 1) (Continued)

1.

The revised fluence estimates were performed differently than the fluences estimated for the specimens that provide the bases for the Regulatory Guide (RG) 1.99, Rev. 2 embrittlement trend curves.* Since Consumers Power's embrittlement estimate was projected using the chemistry factors in Table 1 of the pressurized thermal shock (PTSY, rule, you should estimate the effect of using your.neutron f/uence methodology on the chemistry factor of its limiting RPV beltline material.

Response

As described in CPCo's June 21, 1996 response to NRC fluence submittal questions, the use of ENDF\\B-VI based cross-sections was the only portion of Consumers Power's , current fluence submittal that is outside the basis of RG 1.99 embrittlement trend curves.* This change has a minimal impact on the results of the measured capsule fluence values. Therefore, there is no reason that the chemistry factor should be expected to change or that some other value woul? be more appropriate. 22

Materials/Fluence (Section 1. 1) (Continued) a) The neutron fluence for each capsule has been reduced as a result of the change in neutron fluence methodology. How will the changes in the neutron fluence methodology impact the neutron fluence of a typical (generic) pressurized-water reactor (PWR) RPV wall capsule? Identify the changes in the neutron fluence methodology that are applicable only to Palisades. How much do each of the factors affect the change in neutron fluence for the capsules? _What is the !mpact of changing the neutron fluence methodology on the amount of embrittlement to the Palisades RPV?

Response

As stated above and in Consumers Power Company's (CPCo) June 21, 1996 submittal, the majority of the changes CPCo made in its fluence methodology are consistent with industry practices. The use of ENDF\\B-VI based cross-sections is a more recent change that is not part of the basis of the trend curves in RG1.99. However this change does not have a significant impaqt on the measured fluence values, +/-3%. The* implementation of ENDF\\B-VI based cross-sections may have a slightly smaller impact on the neutron fluence of Palisades compared to a typical (generic) pressurized-water reactor (PWR) RPV wall capsule because of the absence of_ a thermal shield in the Palisades vessel. Assuming the fluence levels as submitted are approved, the projected status of the Palisade~ reactor vessel beltlir:ie materials on March 14, 2007 are listed on the following table. The.beltline materials are projected to have reference temperatures below their respective PTS screening criterion and upper shelf energies greater than. 50 ft-lbs._ 23

• Material ID Fluence RTPTS

'XT Fluence CvUSE ldent'ification (1019 n/cm2 ) (oF) (1019 n/cm2 ) (ft-lbs) D-3803-1 1.86 205 1.10 68 D-3803-2 1.86 191 1.10 58 D-3803-3 1.86 219 1.10 60 D-3804-1 1.86 185 1.10 51 D-3804-2 1.86 157 1.10 54 D-3804-3 1.86 105 1.10 57 W5214 1.38 263 0.82 78 348009 1.38 249 0.82 .76 27204 1.86' 278 1.10 63 Materials/Fluence (Section 1. 1) (Continued) b) (Jompare the chemistry factors from the surveillance capsules materials to the chemistry factors in the PTS rule. Does this comparison indicate that the

• chemistry factors in the PTS ruie' are applicable to the Palisades RPV?
• Response:

The chemistry factor (CF}, as determined from the copper and nickel concentrations, is limited* to plate material. 1 OCFR50.61 does not list values for CF for welds with nickel concentrations in excess of 1.20% or for heat affected zone (HAZ) material. The following table lists projected shifts and measured shifts for the Palisades reactor vessel l)urveillance plate material.* The measured 30 ft-lb transition temperature (LlTT30} for the L-T plate material irradiated to 0.924 x 1019 ri/cm2 exceeds the projected LlRTNoT by more than 17°F and therefore does not meet the credibility criterion specified for base metal in 1 OCFR50.61 (c)(2)(i)(C). Similar results were obtained in*the past using fluence values as shown in TAR Tables 1.1.E-2 through 1.1.E-6. 24

Material Fluence CF Projected Measured (1019 n/cm2) (Cu, Ni) ~RTNDT ~TT30 C-1279-3 0.924 152 176 L-T 1.59 155 175 179 3.84 209 205 C-1279-3 0.924 155 152 158 T-L 3.84 209 194 The CF may be determined from measured shifts using equation 5 of 1 OCFR50.61. The following table lists projected shifts using this method and measured shifts for the Palisades reactor vessel surveillance materials. Again, measured ~TT30 for the L-T plate material irradiated to 0.924 x 1019 n/cm2 exceeds the projected ~RT Nor by more. than 17°F and therefore does not meet the credibility criterion specified for base metal in 1 OCFR50.61 (c)(2)(i)(C). Material Fluence CF Projected Measured (1019 n/cm2 ) ~RTNDT ~TT30

• C-1279-3 0.924 157 176 L-T 1.59 161 182 179

~ ' - 3.84 217 205 C-1279-3 0.924 150 147 158 T-L 3.84 ,*202 194 0.924 262 286 3277 1.59 268 ' 302 305 3.84. 361 342 0.924 197 194 HAZ 1.59. 201 227 226.. 3.84. 271 ~73 Using measured shifts of both L-T and T-L data results in a combined CF for the surveillance plate of 157°F. Measured ~TT30 for the L-T plate material irradiated to 0.924 x 1019 n/cm2 exceeds the projected ~RTNor by more than 17°F and measured ~TT30 for the T-L plate material irradiated to 3.84 x 1019 n/cm2 falls below the projected ~RT Nor by more than 17°F and therefore the combined plate material also does not meet the credibility criterion specified for base metal in 1 OCFR50.61 (c)(2)(i)(C). 25

Material Fluence CF Projected Measured (1019 n/cm2 ) .6.RTt-.1nT -.6.TT~n C-1279-3 0.924 154 176 L-T 1.59 177 180 3.84 157 212 205 C-1279-3 0.924 154 158 T-L 3.84 212 194 Due to its high nickel content, the chemistry factor for the surveillance weld must be determined using equation 5 instead of Table 1 of 1 OCFR50.61. The revised fluence analysis provides greater confidence in _the results from the A-240 surveillance capsule irradiated to 3.84 x 1019 n/cm2

• Revised fluence levels and inclusion of the A-240 measured shift results in the CF being revised from 270 to 268°F for the 3277 weld.

Based on the above, comparisons indicate that the chemistry factors in the PTS rule are applicable to the Palisades RPV, except perhaps for the L-T surveillance plate results at a fluence of 0.924 x 1019 n/cm2. However, recent studies have shown that the standard deviation for plate material in embrittlement correlations is adually >25°F. This suggests that although the L-T surveillance plate data is outside of the 1 OCFR50.61 criterion, the spread in the plate qata may be reasonable. Materials/Fluence (Section 1.1) (Continued)

2.

Sections 1. 1. E. 2. 3 and 1. 1. E. 3. 4 discussed the standard reference material (SRM) in the Palisades surveillance capsules. Considering the revised fluence evaluation, do the SRM results still indicate no anomalies in the Palisades operating conditions?

Response

NUREG/CR-4947 identifies the chemistry factor for standard reference material . HSST01 to be 136.1.°F. The only measurement for HSST01 material from the Palisades reactor vessel surveillance program is from capsule W-110. The calculated transition temperature shift using the revised fluence level of 1.59 x 1019 n/cm2 in equation 3 of 10CFR50.61 results in.6.RTNor = 154°F. The measured shift for this material as reported in Table 1.1.E-6is154°F. There appear to be no anomalies due to Palisades reactor vessel operating conditions. 26

Materials!Fluence (Section 1. 1) (Continued)

3.

Section 1.1.E.1.1 discusses weld repairs made to the suNeillance weld. What were the locations and extent of these repairs? Were mechanical test specimens removed from these areas?

Response

The stick electrode E8018 rod was available for initial weld fit-up and potential weld repairs. There were some weld repairs reported as discussed below: Location and extent: There were two records of weld repairs to the surveillance program weldments. [Note: Two weld seams were deposited using weld wire heat 3277, one for weld metal specimens termed the surveillance weld, and one for heat-affected specimens termed the HAZ weld.] The first was to repair defects in the base metal on the 'weld prep' surface of both plates used to fabricate the surveillance weld.* The grind-outs were 2 to 8 inches long, 318 to 314 inches deep, and were located on the

• OD portion of the weld prep.

The second repair was done after the welds were deposited and radiography was performed. The records indicate that only one of the welds required repair, but it is unclear whether it was the surveillance or HAZ seam. Neither the location nor size of the repair is indicated in the records examined. Proximit¥ of mechanical specimens to repair: It is very likely that some of the surveillance program weld mechanical specimens will include some of the. base metal repair material. The weld Charpy specimens would be "pure" 3277 weld metal around the notch with repair weld and/or base metal extending away from the notch. It i~ uncertain whether or not any mechanical specimens were removed from the region. of the second repair. Weld metal and HAZ mechanical specimens were removed from locations through the thickness excluding the outer half inch and the weld root. Therefore, a shallow repair should not affect either weld or HAZ specimens. However, if removed from the repaired region, the weld Charpy specimens would be repair weld material around the notch as opposed'to the 'pure' intended weld material (3277). E?<amination of the weld metal Charpy results for both unirradiated and irradiated conditions reveals consistent data (i.e., not highly scattered) with a large amount of. transition temperature shift due to the levels of copper and nickel in the 3277 weld metal. Had E8018 weld metal been included in the weld metal sections of the surveillance weld specimens, the unirradiated results should be scattered to reflect two distinctly different weld metals, and some of the irradiated data would not be shifted as far for the low copper E8018 stick weld. 27

Materials/Fluence (Section 1. 1) (Continued)

4.

Table 1.1.E-5 presents Charpy V-notch results for the Palisades surveillance program heat-affected zone (HAZ). Upper-shelf energy (USE) decreases of 37 and 35 are listed. Are these absolute decreased values or deltas? If absolute, since these values were so low, were the fracture surfaces ex1mined for failure mode?

Response

Consistent with the data presented in the other tables in Section 1.1. E *summarizing the sL,Jrveillance results, the values for decrease in upper shelf energy (USE) for the HAZ (Table 1.1.E-5) are.differences (deltas) between the unirradiated and irradiated conditions. The absolute values for USE are shown in the column to the left of the delta decrease. None of the absolute decreased*values fall near SO.ft-lb; therefore, no microscopic fracture surface examination was warranted. Materials/Fluence (Section 1. 2) (Continued)

5.

.In Section 1.2. C. 1.1 it is stated that there are no records of weld repairs to the beltline plates. Does this include only Combustion Engineering re.cords or plate fabricator (Lukens, U.S. Steel, etc.) records also? It would seem.unusual for .such large Q& T plates to have. no surface weld repairs. Section 1. 2. C. 2. 3 acknowledges the possibility of plate weld repairs that may not have been documented. This section also notes that plate repairs in the *upper shell course were only later discovered during vessel inlet nozzle weld inspections..

Response

Welding of repairs by the plate (or forging)* manufacturer was not allowed

• by CE specification and therefore, there will be no records on file. Welding of-repairs by CE would have been recorded only if the size of the repair exceeded the A$ME Code criteria (the lesser of 10% bf base metal thickness or 3/8-inch).

The plate was ordered oversized to allow for inside and outside surface machining; this machining could remove shallow surface defects on the as-received plate and those introduced during forming and heat treatment.. Repair of surface defects less than 3/8-inch found after machining was done in accprdance with procedures but the repair location was not recorded. A typical surface defect repair would have been grindouts for removal of temporary attachments such as tie straps. The repairs referred to in the upper shell plate were laminations that were acceptable per the ultrasonic criteria during the initial plate inspection. When the noz.Zle openings 28

in the shell plate were cut, the laminations apparently were not exposed to the surface nor close enough for the magnetic particle inspection to detect them as subsurface. The nozzle to shell weld radius extends outward onto the plate. The radiography of the nozzle to shell weld resulted in the detection of a plate lamination, and it was repaired during the standard repair cycle for the nozzle to shell weld. The following is a list of weld repairs to the reactor vessel plates:

• 1.

Upper shell outside diameter - repair of surface irregularities

2.

Upper shell plate 03802 repair of edge cracks

3.

Upper shell plate 03802 repair of weld prep. area

4.

Intermediate shell plate 03803 removal of embedded mill scale and repair (weld buildup) prior to cladding

5.

Bottom head torus segment 03807 repair of weld prep. area Of the five recorded repairs, pla~e 03803-1 is the. only beltline region plate. . Materials!Fluence (Section 1. 2) (Continued)

6.

Section 1. 2. C. 2. 1 discusses three separate stress relief heat treatments that were performed during the RPV fabrication sequence. The temperatures for these treatments ranged from 1100 °F to 1175 °F with a total cumulative time of 16 to 18 hours at temperature. Was accelerated cooling from the SRHT [stress-relieved heat treatments] used to prec~ude potential metallurgical degradation associated with impurity segregation (e.g., temper embrittlement)?

Response

Accelerated cooling, per se, was not allowed by the ASME' Code. The 100 °F/hr limit on the cooldown in the temperature range from 1100-1175 °F was established based on prior experience with the fabrication and heat treatment of large pressure vessel components. It is likely that prevention of metallurgical degradation during cooldown was a consideration when setting the upper limits on the cooldown rate, but it cannot be established that it was a prime reason fofallowing up to 100 °F/hr 'for the cooldown rate. 29

Materials/Fluence (Section 1.2) (Continued)

7.

Section 1. 2. C. 3 summarizes the fluence analysis for the RPV. The reactor vessel neutron fluence analysis has been revised per the submittal to NRG dated April 4, 1996. Assess the impact of the revised fluence analysis on the surveillance results per the previous questions on Section 1. 1.

Response

The responses to RAI #1.b and #2, MaterialslFluence (Section 1.1) assess the impact of the revised fluence analysis on the surveillance results as related to the 30 ft-lb transition temperature and PTS. The following table lists projected and measured values o.f Charpy upper shelf energy for the Palisades reactor vessel surveillance materials. The measured upper shelf energies for the plate materials remain higher. than the projected values. The measured upper shelf energies for the surveillance weld are lower than the projeded values, but remain above 50 ft-lbs even out to a fluence level of 3.84 x 1019 nlcm2. Regulatory Guide 1.99, Rev. 2 does not provide

• guidance on estimating the decrease in upper shelf energy for HAZ material.

Material Fluence Projected Measured (1019 n/cm2) CvUSE CvUSE C-1279-3 0.924 106 112 L-T 1.59 99 103 3.84 86 93 C-1279-3 b.924 . 70 84 T-L 3.84 57 68 0.924 73 63 . 3277 1.59 '67 58 3.84 57 51 0.924 79 HAZ 1.59 81 3.84 . 59 ' HSST01 1.59 95 99 30

Materials/Fluence (Section 1. 2) (Continu~d)

8.

Section 1.2.E.2 discusses annealing effects on internal attachments. Are there any effects of the anneal on the bottom-mounted instrument penetrations in the lower head?

Response

The Palisades reactor vessel design does not have bottom-mounted instrument penetrations. This is indicated in Figure 1.Z.D-1 in Section 1.2 of the TAR Materials/Fluence (Section 1.2) (Continued)

9.

Sections 1.2.E.2.1and1.2.E.2.2 discuss annealing effects on the Core Support Lugs, Core Stabilizing Lugs and the Flow Skirt. The potential for sensitization of the IN-600 material is acknowledged. It is stated that the sensitized material is not a concern in the normal PWR environment. Is the sensitization potentially a problem when coupled with enhancement from the irradiation environment?;

Response

The end of life fluence on the alloy 600 material at the core support lugs, core stabilizing lugs and the flow skirt is less than 1018 nlcm2 (E > 1.0 MeV). No metallurgical changes are expected at fluences less than 1018 nlcm2. <2> Therefore no enhancement of the sensitization from the irradiation 'environment is expected. 2 R. Bajay, W. J. Mills, M. R. Lebo, B. Z. Hyatt, M. G. Burke, "Irradiation Assisted Stress Corrosion Cracking of HTH Alloy X-750 and Alloy 625," Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, August 1995, p. 1093 (NACE). 31

Materials (Section 1.3) (Continued)

10.

Comment - As a general obseNation much of this section relies on the predictions from the thermal models as to the maximum temperatures that will be achieved in various components. The predictive capabilities of these types of models will be evaluated in the Marble Hill anneal. If the Marble Hill measurements show the models* to be deficient, the conclusions of this section will obviously have to be re-evaluated. * .. Response: The statements made in the comment are true, however, the thermal and stress models as described in Section 1.7 o(the TAR used to establish the conclusions stated in Section 1.3 of the TAR are considered to be bounding to what is considered to be the "realistic" portayal of *the actual process. Our intent for the Palisades Annealing Program is to obtain a "realistic" or more accurate representation of the temperature and stress distribution in the Palisades reactor vessel and attached piping using knowledge gained from the Marble Hill process. The purpose of this new analysis is to

• establish a more representative understanding of the structural integrity margins. If this analysis indicates that the conclusions of Section 1.3 are incorrect, which is not expected, such conclusions will be re-evaluated and the. appropriate changes will be
• made to the TAR Currently this new analysis wi!!' be completed prior to issuance of the final TAR Materials (Section 1.3) (Continued)

11.

Sectif?ns 1.3.B.4 and 1.3.B.4 discuss heating effects on the RV supports. The assumption js that the supports will not be heated beyond their design temperature of 650°F. Section 1.3.8.4 acknowledges that the lubricant will degrade at temperatures above 650 °F. There is also the possibility for metallurgical degradation of the bronze alloy at temperatures above 650 °F. If the prediction of 650 °F is not verified, degradation of the support assembly will have to be addressed.

Response

In Section 1.7, Table 1.7.0-6 it is indicated that the highest predicted temperature of the RV nozzle support paqs is 563 °F, well below the 650 °F criteria defined in Section 1.3. This temperature prediction was based on a bounding thermal case T2 which was created to give hotter temperatures than expected. To achieve this, the analysis assumed a higher ambient air temperature, RV insulation properties 33% better, and no internal PCS piping convection to the steam generators and primary coolant piping. Our intent for the Palisades Annealing Program is to obtain a "realistic" or more 32

accurate representation of the temperature and stress distribution in the Palisades reactor vessel and attached piping_ using knowledge gained from the Marble Hill process. However it is expected that this more accurate portrayal of the RV nozzle support pad temperature will be less than that reported in Table 1.7.D-6. This is

• anticipated because the preliminary findings from Marble Hill indicated that the measured temperatures in the RV nozzle support pad area were far less than the 650°F design temperature.

The actual temperatures on the RV support pads during the anneal will be determined by analysis using the temperature sensors above and below the RV nozzles and those located. near the PCS piping interface. Since confidence is very.high that the actual RV *

• support pad temperature will be maintained much less that 650°F, no further emphasis

'is being placed on the subject. If the actual temperatures in the.RV support pads are found to be.higher than 650 °F then the temperature degradation effects to the lubricant and bronze alloy will be addressed.

• • Materials (Section 1.3) (Continued)
• .*12.

In Sections 1.3.A and 1.3.B, insulation properties for various components are

• addressed. For th_e nonreflective types *of insulation, does a possibility for generation of potentially tpxic gases from. the insulation exist at the elevated
• temperatures of the* cavity? Cah such gases be generated from other materials in tt}e cavity due to the elevated temperatures? *

' Response: * :, '

• i The nonrefl.ective insulation*within the cavity consists of austenitic stainless steeL sheet.
• metal, aluminum foil, mineral wool, and/or Unibestos. None of these materials should

.'.release potentially toxic gases. The mineral wool and Unibestos can withstand . temperatures up to 1000 °F and 1200 °F, respectively, as reported in Section 1.3~8.8 of . the TAR, without deterioration. With the supplemental cooling in the cavity as,described in Section 1.6.A.7.. 1 of the . TAR, temperatures outside of the reflective insuJation on the vessel are not much different than normal operation thus other materials in the cavity should not release potentially toxic gases. The coating of the reactor vessel outer diameter surface (Gilman St~ck Paint Lot 33-E-102) will be at a higher temperature during the anneal than at normal operating temperatures. This coating; however, has a heat resistance of 350°F (continuous) and 500°F (intermittent) and therefore any release of fumes would probably have occurred during the first and subsequent normal operational cycles. 33

Materials (Section 1. 3) (Continued)

13.

The concept of "sacrificial" concrete is presented in Sections 1.3.C.2 and 1.3.C. 7. Section 1.3.C. 7 references the Bechtel Palisades Plant design criteria when discussing the "sacrificial" concrete concept. Provide the Bechtel reference discussing "sacrificial" concrete. What is the basis for the "sacrificial" thickness being set at 10 inches?

Response

Early in the design of the Palisades Biological Shield cooling system, the thermal analysts recognized that embedding cooling coils directly in the concrete would induce steep temperature gradients in the concrete. <3> The low conductivity of concrete requires a steep gradient to drive significant heat to the coils. The coils have low heat flux. The water velocity in the 314-inch schedule 80 coil tubes is about 1 foot per

• second and the coils are 23 feet long. The design temperature rise in the coils is approximately 3°F. At design conditions, this low heat flux depresses the concrete temperatures about 70°F over the 3-inch distance between the concrete surface and the nearest coil.

Structural engineers recognized that over time this steep gradient could cause degradation in the concrete such that the concrete integrity cannot be ensured. Their solution was to classify the concrete in the immediate vicinity of the coils as "sacrificial'.' concrete and consider it nonstructura1.<4> To decouple the sacrificial concrete from the structural concrete,. the conceptual design called for the sacrificial concrete around the coils to be poured ahead of the main biological shield pours. The inner form for these pours was the reactor cavity liner plate which was to be reinforced with a moveable spider in the inner diameter as the pour advanced upward. The outer diameter form . was to be plywood through-bolted to the liner plate. <5> This 10-inch thick circle of concrete has significant buckling resistance and was to be used as the inner diameter form for the main pour. The plywood form at the 10-inch radial distance from the liner plate was left in place. The location of this plywood at 10 inches defines the limit of the 3 4 Bechtel Letter dated 8/4/67 from D. J. Olver to W. C. Cooper on Biological Shield Liner Plate [Attachment B1] H. S. Tsang Calculation dated 11/7/67 on the Radial Deflection of Biological Shield Due to 1 OOOK Load, 1 page excerpt only [Attachment B2]

• Bechtel Memo dated 8/1/67 from D. J. Olver to S. Pohtos on Biological Shield Liner Plate, without Dwg. C-153 [Attachment B3]

34

sacrificial concrete. It is shown on Palisades Plant Drawings C-153 and M-72. cs>.m Construction progress photographs show that pours for the sacrificial and the structural concrete were actually advanced approximately together so that the through-bolting (shown on Attachment 84) was not required. These photographs clearly show that the plywood was still inserted at the 10-inch radial location and that the moveable spider at the inner diameter of the liner plate was used. It is likely that the existence of large amounts of reactor vessel supporting structural steel penetrating the plywood caused significant leakage problems to develop in the outer diameter plywood boundary and resulted in the change in the construction strategy to advancing the pours together. Although alternative designs were known to have been considered, it appears they were rejected and the designs indicated in Attachments 81 through 85 were used. Because a great deal of latitude was allowed in construction practice at the time of construction some of the details in this material might not be totally accurate, however the design concepts involved are believed to be the actual ones used. Materials (Section 1.3) (Continued)

14.

Section 1. 3. C. 5 discusses the enhanced cooling capabilities of the Bio-Shield Cooling System (SGS) in regions adjacent to the RPV.supports. However, 'the Bio-Shield thermal model (described in Section 1.3.C.9) does not appefJr to account for thermal conduction down the steel embedments for the RPV supports. Why was this not included in the model? What are the expected temperatures from conduction down the embedments? Can the SGS maintain adequate cooling in these regions during the anneal?

Response

During the anneal, the maximum predicted RPV nozzle support pad temperature is 563°F with the bottom of the nozzle support pad at approximately 500°F (Table 1.7.D-6 of Section 1. 7 of the TAR). This is below the nozzle design temperature of 650°F (Section 1.2.E.1.3 of the TAR). The thermal response of the RPV nozzle supports is expected to be bounded by normal plant operation based on the lower than design temperatures of the nozzle and the results of the annealing cavity thermal model which demonstrate lower cavity liner temperatures. The thermal model assumes that the nozzle support at the cavity liner 6 Palisades Plant Drawing C-153, Reactor Containment Biological Shield Liner, Rev. 6 dated 8/8/89 [Attachment 84, Detail 4] Palisades Plant Drawing M-72, Primary Shield Cooling Coils Sections & Details, Rev. 1 [Attachment 85] 35

• J~ f

interface to be less than or equal to 250°F. Therefore a detailed model of heat transfer down the steel was not constructed. The reactor cavity.will be cooled by heat transfer through the primary shield, the SCS, and by a temporary air forced cooling system during the anneal. The analyses described in Section 1.3.C.9 of the TAR determined that the primary heat removal mechanism is through the temporary forced cooling system. The insulation thermal properties were characterized based on the temperature measurement data taken following the 1995 refueling outage. Sensitivity studies were performed when doing this characterization which conservatively predict heat loss from the reactor vessel during t.he anneal. The thermal analysis for the reactor cavity includes the heat removal mechanism through the SCS, heat transfer through the concrete to containment, and the temporary forced cooling system. The cavity heat load is calculated using conservative in.sulation thermal properties. The calculated cavity temperatures remain below the temperature limit for the Palisades biological. shield wall concrete inner. surface (250 °F). Materials (Section 1.4) (Continued) 15: Section 1.4.B proposes 800 ~F and 48 hours as the minimum temperatureand time for annealing. Given the actual annealing response for the Palisades materials in the subr(littardated April 2, 1996, has there been any thought given to revising the minimum conditions?.

Response

Even 'though the results from the annealing response for the Palisades materials in the submittal dated April 2, 1996 appear to be lower than expected (especially for the plate* upper shelf energy), there are insufficient data available upon which to base a challenge to the time-temperature trends suggested by the NUREG/CR-6327 model. The CPCo experiments did not evaluate the parametric effects of time and temperature, but there is no reason to expect significantly different trends than that suggested by the model. Therefore, at this time CPCo does not intend to revise the minimum time at temperature specified in the TAR CPCo is currently conducting additional tests on surveillance materials to further quantify a*nd/or possibly explain the lower than

• expected values provided in the April 1996 submittal.

36

Materials (Section 1.4) (Continued)

16.

Section 1.4.B discusses temper embrittlement (TE) but does not discuss the Palisades plans for examination of TE in HAZ materials. The section should be revised to include the plans.

Response

The plans for examination of TE in HAZ materials were supplied December 18, 1995 and are further discussed in response to RAI # 21 and #22, Materials. The final version of the TAR will include discussion of the examination plans and will include information provided in responses to RAI #21 and #22, Materials (Section 3.0). Materials (Section 1. 5) (Continued) *

17.

Section 1. 58 notes that strains will not be measured at Palisades but will be inferred based.on temperature measurements and the stress analysis.* This will be highly dependent on how well the stress analysis for Marble Hill is benchmarked by the experimental measurements made during the anneal. What is the current projection for when the Marble Hill annealing information will be submitted to NRG?

Response

The cu.rrent Marble Hill Annealing Demonstration Project (~DP) schedule shows that the project documentation will be provided to the ASME-DOE/SNL-lndustry Steering Committee in November 1996. This committee has control ofthe final project documentation release and distribution. The final Marble Hill ADP documentation is scheduled for release prior to the end of the first quarter of 1997. At which time the NRC will be provided a copy. This project documentation will address how well the stress analysis for Marble Hill is benchmarked by the experimental measurements made during the anneal. 37

Materials (Section 1. 8) (Continued)

18.

Sections 1.8.B and 1.8.C discuss time and temperature limits for the anneal. Time and temperature limitations on the anneal could also possibly result if TE is found to be an issue with the Palisades HAZ materials. These sections should ackr:10wledge the possibility of limitations associated with TE.

Response

CPCo does not believe that temper embrittlement is a problem for the Palisades vessel annealing plan. CPCo is planning to further investigate any intergranular fracture that

• could result due to temper embrittlement in the heat-affected-zone (HAZ) region of the surveillance weld, but the data and industry experience to date for U.S. quality reactor vessel steels do not suggest any real potential for this type of material damage mechanism. At this time, there is no need to acknowledge temper embrittlement as a problem in Sections 1.8.B and 1.8.C.

Materials (Section 1. 8) (Continued)

19.

Section 1.7.D.3 discussed the possibility of high stresses in the surveillance. capsule holder assemblie~ during the annealing heatup. There exists a possibility that these stresses could cause cracking in the fillet welds joining the wall attachment bracket to the box be~m. Table 2.2.A-1 requires a VT-1 inspection_ oftf;lese areas. VT-1 may not be sufficient to find the cracking if it is tight. An "enhanced" VT-1 procedure capable of resolving a 112-mil diameter wire sho.uld be performed on}hese areas.

Response

Per the arfalysis results given in Section 1.7 of the TAR no cracking is expected in the fillet welds joining the wall attachment bracket to the box beam. However since the 'purpose of Section 2.2 of the TAR is to affirm that the annealing operation has not damaged such attachments it is recognized that tight cracks may be difficult to detect with conventional ASME Code Section XI VT-1 procedures demonstrated only to resolve a 1/32-inch black line on an 18% neutral gray card. A demonstrated resolution of a 1-mil diameter wire should be sufficient for this purpose and equipment exists to conduct this type of enhanced VT-1 examination within the vessel environment. A 112:- mil diameter wire resolution capability while possibly achievable with special equipment would require qualification for use in a PWR vessel environment. Present plans are to use a demonstrated resolution of a 1-mil diameter wire. 38

Materials (Continued)

20.

Section 2.2.A.3 discusses the possibility of iacking" loop components during the anneal. What criteria will be used to determine if jacking is necessary? Will the evaluation be performed on-line during the. heatup, or in advance of the anneal?

Response

Jacking (horizontally) of the steam generators is a contingency plan possibility to facilitate reassembly of the head onto the reactor vessel. Heating of the open reactor vessel will m.ove the two steam generators radially on their sliding bases. After

• cooldown, the residual friction acting on the steam generator sliding bases may hold the "reduced stiffness" RV in an out-of-round position. The "reduced stiffness" of the RV occurs because the stiffness normally offered by the closure head and the stiffening.

effect of internal pressure are both absent during this one*time thermal transient. '. The smallest clearance in the flange area is the radial cleara-nce between the head and the vessel, which i~ 0.045-inch minimurn. The Sl!lall head-to-vessel clearance could be reduced to zero if the steam generator sliding base friction does not allow the generator to return to ~he pre-anneal position. If this were to occur, the vessel will still be well in the elastic range. Jacking of the steam generators would only-be needed if there was some question as to whether the RV head can be re-installed. This assessment can be made "on line" by evaluation' of steam generator movement during the heatup and cooldown, and verified by the post-anneal measurements of the head seating counterbore. Materials (Section 3.0). (Continued) . *The staff issued a request for additional information on this section previously (November 1995) and received a response in December 1995. The following are follow-up questions based on information obtained since that time:

21.

If intergranular fracture is observed in irradiated and annealed HAZ material, the licensee indicates that its surveillance program will only include weld and plate material but will not include HAZ material. Provide the basis for concluding that HAZ material need not be Charpy impact tested to determine the effect of intergranular fracture on irradiated and annealed HAZ material. If additional testing of Palisades HAZ materials is indicated, can the testing be completed to allow time for NRG review before the anneal (May 1998)?.

Response

If intergranular fracture is observed in irradiated and annealed HAZ material, broken plate and weld specimens previously annealed and tested will be examined for 39

intergranular fracture. If intergranular fracture is observed in plate or weld material, then the existing surveillance program is already evaluating material demonstrating this phenomenon. If intergranular fracture is not observed in plate and weld material, then HAZ material will be added to the surveillance program because it is the only material exhibiting the phenomenon. Additional testing, if required, will be completed before November 1997 to allow for NRC review prior to the anneal. Materials (Section 3.0) (Continued)

22.

What is the neutron fluence received by the HAZ materials that are being fractured? How much phosphorus is in the HAZ materials? Will the neutron fluence and phosphorus of the HAZ material that are being fractured represent the neutron f/uence andphosphorus of the HAZ materials in the Palisades RPV? Explain.

Response

HAZ material will be tested using surveillance plate HAZ material that has received neutron fluence levels of 1.59 x 1019 and 3.84 x 1019 n/cm2. The phosphorus content in the HAZ is assume~ to be the same as that of the plate material D-3803-3. The phosphorus contents of the six beltline plates are estimated to be: Beltline Plate Phosphorus(%) D-3803-1 <0.010 D-3803-2 0.010. D-3803-3

• 0.011 D-3804-1 0.016 o:.3804-2 0.015 D-3804-3 0.010
• The neutron fluence of the surveillance plate HAZ material is greater than the fluence accumulated by the reactor vessel. The phosphorus content of this material is directly representative of one beltline material and generally representative of the remaining materials.

40

Equipment Qualification (Sections 1.3 and 1.6)

1.

Describe in detail those measures that will be taken to assure that safety-related components (including electric cables) will not be adversely affected by

• temperature and radiation effects during and following the annealing process (recognizing that there will be some degradation of the RV insulation materials).

Also, discuss any provisions that will be implemented for independent oversight of the evolution as it relates to surrounding plant equipment and also address the following specific questions pertaining to Sections 1. 3 and 1. 6 of the Preliminary TAR.

Response

The principal equipment, components and structures which will be potentially exposed to higher than normal temperatures during the annealing evolution are those in areas surrounding the reactor vessel itself(e.g., the biological shield wall, neutron detectors and cabling}, and the permanent equipment and s~ructures over which the heat exchanger inlet and outlet ducting will pass. The temperature of the concrete biological shield wall and other critical structures and components surrounding the reactor vessel will be monitored during the annealing evolution. To ensure that the biological shield wall does not exceed the steady-state maximum temperature limit of 250°F, an active biological shield supplemental cooling system will be provided which directs airflow tjownward within the reactor cavity annulus. Cooling airflow provided by this RV Biological Shield Supplemental Cooling System will protect the temperature sensitive

• equipment, components, or structures outboard of the reactor vessel. As described in the response to RAI #2, Equipment Qualification, the neutron detectors' temperature limits will be bounded by the biological shield wall temperature.

The heat exchanger inlet and outlet ducting, which enters containment via the equiplT)ent hatch, traverses over to and past the temporary RV internals shielding

• structure, and then is directed downward to its connection to the top of the heat exchanger, will be ins*u1ated and supported such that no permanent structure or component will be subjected to unacceptable tem*peratures. (See also the response to RAI #4, Equipment Qualification below). During the Marble Hill Annealing Demonstration, casual contact with the ducting insulation was only warm to the touch and posed no concern to personnel or plant equipment.

Regarding potential radiation effects on plant structures and equipment, the radiation levels to equipment surrounding the reactor vessel during the annealing evolution will be less than those existent during normal plant operation or refueling. The responses to RAI #3 and #5, Equipment Qualification, given below, provide additional information regarding higher radiation levels and their potential effects. Regarding provisions for independent oversight of the annealing evolution as it relates to surrounding plant equipment, frequent plant walkdowns are planned (see response 41

to RAI #7, Equipment Qualification below) to ensure that equipment is performing as designed and intended. The walkdowns will be controlled procedurally and the results will be documented. Additionally CPCo and their vendor annealing personnel not directly involved in the conduct of the annealing operation will participate in the walkdowns and observations and provide a measure of independent oversight. Equipment Qualification (Sections 1.3 and 1.6) (Continued)

2.

How will temperature vs. time be established? What about electrical wiring associated with the detectors? (1.3.E.1)

Response

Temperature sensors located on the biological shield liner will provide the temperature versus time profile input as recorded by the data acquisition system for determining the temperature effects to the neutron detector instrumentation. This data will be used to determine the environment to which the plant equipment such as the detectors and, associated electrical wiring wer~ subjected during the annealing process. The electrical w.iring supplied with the detectors can withstand steady. state operating temperatures up to 300°F. The detector leads are routed inside the detector wells within the biological shield down to the 590-foot elevation in containment. The limiting temperature, therefore, is the 250°F maximum steady state temperature for the biological shield wall. Equipment Qualification (Sections 1.3 and 1.6) (Continued)

3.

What effects will radiation have on electrical cables? What assurance is there

• that the actual positioning of cables will be adequate to avoid adverse effects?

(1.3.E.4)

Response

As discussed in the response to RAI #1, Equipment Qualification, the radiation levels in the area above the reactor vessel during the annealing evolution are less than those existent during normal plant operation. The CSB and UGS will be shielded for purposes of protecting personnel who will be performing the annealing operations. The resulting general area radiation levels in the refueling cavity (maximum of 500 mrem/hr) and at the operating floor level (60 mrem/hr) will not cause radiation degradation of electrical cabling and equipment during the two to three week period of the actual annealing. It is normal for any refueling outage to remove cabling from the refueling cavity (e.g., 42

for reactor instrumentation, control rod drive and position indication, etc.). No special positioning of cables is anticipated for avoiding adverse effects from heat during the anneal. Also based on the discussion above, positioning of cables is not required to avoid adverse effects from radiation, except perhaps for the UGS dry lift as discussed in response to RAI #5, Equipment Qualification below. Since the fuel transfer tilt machine is inside the dry RV internals shielding enclosure, attention to the associated equipment hoses and electrical switches and wiring is required. An evaluation of the materials will be performed after identification and analysis of the equipment dose within the internals shielding system design. The equipment will be reviewed to determine its ability to function following radiation exposure of this magnitude. If failure is considered likely, the affected piece of equipment will be shielded or removed. By maintaining some minimal water level in the tilt pit it may be possible to attenuate the dose to bound the effect as equivalent to the underwater fuel handling. The final TAR will include a discussion of the radiation

• effects to the fuel transfer tilt machine.

Equipment Qualification (Sections 1.3 and 1.6) (Continued)

4.

How will the temperature profile outside the ducts be established? (1.3.E.6)

Response

Assessing the effects of heat load from the heating system ducting on adjacent plant structures depends in part on determining the heat flow from the insulated ducts themselves. This heat flow is usually determined using very standard, one-dimensional

• steady state. heat transfer techniques, such as are contained in the ASTM "Proposed Procedure for Calculating Heat Losses Through Furnace Walls", prepared by the C-8 Subcommittee on Heat Transfer. At Palisades, heat flows to containment from the ducts will be determined by this method, using conservative assumptions regarding the state of the ducting itself. For example, although the di.Jct metal (hot face) temperature varies throughout the annealing cycle, assessments are generally made using the maximum anticipated duct temperature. Although the duct metal temperature is generally slightly lower than the temperature of the hot air inside, the duct is assumed to be at the fluid temperature. Similarly, conditions for the duct surroundings (ambient temperature, wind speed, etc.) are chosen conservatively. Based on the resulting duct insulation cold face temperature and heat flow through the insulation, assessments on
• the surrounding structures can be made.

43

Equipment Qualification (Sections 1. 3 and 1. 6) (Continued)

5.

What about effects of radiation levels/exposure of plant equipment? (1.3.F.511.3.F.6)

Response

In general, there is not a significant amount of equipment in the area of containment exposed to the unshielded upper guide structure (UGS). Mechanical equipment and electrical cabling in this area are expected to receive exposure levels during normal operation substantially greater than the exposure received during the transfer of the UGS. The high dose rates may have a negative effect on some of the instrumentation and controls in the area (e.g., the refueling machine control panel, refu_eling area monitors, humidity sensors, etc.). This equipment will be reviewed to determine its ability to function following radiation exposure of this magnitude. If instrument failure is considered likely, the affected piece of equipment will be removed.or shielded prior to performing the UGS transfer.. Equipment -Qualification (Sections 1. 3 and 1. 6) (Continued) 6. Through what areas of the plant and in the vicinity of whflt plant equipment does the annealing system ductwork pass? What measures will be taken to assure that the heat exchanger ducting will remain intact and adequately supported and will not impact safety-related equipment throughout the course of the annealing operation? Should a failure of the ducting occur, identify any plant structures and components that could be affected by the failure, including the effects of any fire that might result, such that the consequences are made worse (e.g., create a challenge for the spent fuel pool cooling system, result in an evacuation of the control r<;Jom, cause a radiological release, etc.) and identify any special measures that will be taken to eliminate or minimize the vulnerabilities that exist. Provide a.summary of the analysis for the most safety significant postulated ductwork rupture event. (1.6.A.1.2)

Response

The burners will be located on the platform to be erected on the east side of the spent fuel building. From the burners, the hot air inlet ducting will pass through the east wall of the spent fuel building near the containment. The ducting would run west to the containment equipment hatch, then through the hatch into the containment building. From the equipment hatch, the ductwork would be routed across the operating deck toward the reactor cavity, over the temporary RV internals shield structure, then down to its connection to the heat exchanger inside the reactor vessel. The exhaust ductwork follows the same path through containment, up from the heat exchanger and across the operating deck to the equipment hatch. Once through the hatch, the 44

e~haust ductwork turns away from the direction of the *burners and exits through the west wall of the spent tu.el building. The attached Figure 1 is a sketch of the ductwork layout. Structural integrity of the hot air ductwork is assured by utilizing design and construction criteria generally in accordance with ASME 831.3. A thermal and structural analysis will be performed, taking into account the operating temperatures of the ducting, deadweight of the ducting and insulation, flow induced vibration loads, duct support reaction loads, and thermal expansion loads. Although the routing of the ductwork may yet be modified, one of the primary considerations in the location of the ducting is to minimize the heat load on adjacent equipment and structures, and for the exhaust ducting, to minimize impact at plant air intake locations. The impact on plant equipment and structures will be controlled

• through the utilization.of additional protective measures such as insulation and increased standoff distance, as necessary. Combustibles on the operating floor and other affected areas will be controlled utilizing normal administrative control p~ocedures for hot work permits and by performing frequent walkdowns as described in RAl'#7, Equipment Qualification below:

Minimizing the_ routing of the hot air ducting through plant areas greatly reduces the possibility of impact that a duet failure would h~ve on plant equipment and strudu,res. In the containment, structures potentially 'affected by a ductfailure would *be the interior concrete and the equipment hatch, neither of which would likely experi~nce adverse effects. In the spent fuel building, since the routing takes the ducting only across the south erid of the building at the operating deck elevation, there is again minimal possibility of affecting safety related equipment.* The main control room, emergency diesel generators, radwaste processing and holdup systems, and the spentfuel pool

• cooling system equipment are all physically removed from the operating deck area of the spent fuel building, as well as protected by concrete structures. The only structure susceptible to a duct failure would be the spent fuel pool itself.* The insulation applied to the ductwork that*runs through the spent fuel building will be securely attached to the ducting to' preclude it becoming loose debris in the spent fuel pool. Considering the routing of the ducting, the existence of intervening structures along a portion of its path, and the distance of the pool from the ductwork, ~he possibility of duct failure affecting the spent fuel pool or the stored fuel would be remote. Even so,. ~he postulated rupture of a hot air.duct has been evaluat~d.

Each of the five hot air ducts running from the burners to the heat exchanger carry ~uperheated aide one of the five zones of the heat exchanger. Each zone has a different heat demand based on the heat load and temperature to be maintained in that particular portion of the reactor vessel. Thus, the air temperature in each duct may be different, depending on it.s particular heat exchanger zone requirement.. For this evaluation, conservative air flow and temperature conditions were used. The air temperature was assumed to be 1800°F (nominal air temperature is about 1400°F). 45

While a flow rate of about 2,000 SCFM is not expected to be exceeded during burner operation, an upper bound value of 3,000 SCFM was evaluated. This represents a heat load of approximately 6.9 million BTU/hour. An evaluation has been performed to determine the effect on the spent fuel pool water conditions in the event of a break in the hottest duct. It was assumed that the duct fails and is positioned such that the superheated air is discharged into the spent fuel pool water. The initial spent fuel pool water temperature is assumed to be 150°F, which is the maximum allowed by Plant Technical Specifications. All spent fuel assemblies are in the spent fuel pool, plus the full core offload. To maintain the integrity of the spent fuel and assure that cooling of the fuel is not affected, two spent fuel pool parameters are considered. Sufficient water inventory must be maintained such that continued operation of the spent fuel pool' cooling system is not jeopardized. Also, the water temperature should be maintained sufficiently below the boiling point so that heat transfer from the fuel is not impaired by local boiling in the fuel assemblies. Two cases were evaluated. The first case conservatively assumed that all available heat from the superheated air is transferred to the pool water. It is reasonable to assume that such a break in the hot air system would be discovered . quite quickly and the burners shut down, halting further heat input to the spent fuel pool. The evaluation conservatively. assumed that the discharge would continue for 30 minutes before the burners are shut down. After the.30-minute discharge into the spent fuel pool, the temperature of the water, assuming a homogeneous distribution of the heat, increases by only 3°F to about 153°F. Clearly the heat distribution would not be so uniform; however, the hot water would tend to remain toward the top of the pool, with the water at the fuel elevation remaining closer to the initial pool conditions. With the inlet to the spent fuel pool cooling system located near the top of the pool, the hotter* water would tend to be drawn into the spent fuel pool cooling system, rather than migrate or be forced down toward th~ fuel racks. The second case assume,d that dry air is discharged into the water and saturated air exits the pool, maximizing the amount of water that can be removed by the hot air. Again, it was assumed that the broken duct discharges into the spent fuel pool for 30 minutes before the burners are shut down. To maximize the water inventory

• removed, this.case also assumes that no heat is transferred to the water remaining in the pool. The results of this analysis indicate that the loss of inventory in the spent fuel pool would be approximately 181 gallons of water. This represents a decrease in the level of;water in the spent fuel pool of 0.56 inch.

The conditions following such a hot air duct failure would likely fall somewhere between the results of the two cases considered. The water would experience a small increase in temperature, accompanied by an inconsequential loss of water inventory. In summary, the two cases evaluated indicate that there would be little effect on the spent fuel pool water conditions, and thus no adverse effect on the fuel stored in the spent fuel pool. 46

I I~ I~ <C c ; ..a. O'... m .c c -* "Cl 3 CD

i -

0 c !! 3i n m 0

i tn

r

~.

i

<C tn A CD - n

r 0

c.

c n ~ A m 0 c - SECTION A-A ,n...n......,..... ~ - - - I I I \\ .1 I I I PALISADES PLANT CONSUMERS PO~ER COMPANY J DUCTING PLAN @ EL. 649'.-0' J

---

---- ___ ___:._J

Equipment Qualification (Sections 1.3 and 1.6) (Continued)

7.

How often will the 'job walks" be performed and what disciplines will be involved? Also, will there be an initial walkdown of the installation and if so, what will it entail and what disciplines will be involved? Discuss any training that will be provided, procedures that will be used, documentation that will be required, and how discrepancies and problems will be highlighted and resolved. (1.6.A.1.9.2)

Response

The frequency of walkdowns performed during the annealing can vary depending on the item being checked. For example, the fuel level may be checked only once every twelve hours, since it changes slowly over time, while checks on the duct supports may be required every four. Generally, most chec~s are performed at intervals of two to tow hours. All walkdowns of the heating system performed during the annealing are the responsibility of qualified.heat treatment technicians. Regarding an initial walkdown of the installation prior to startup, the purpose of the walkdowns as described in TAR Section 1.6.A.1.9.2 is to ensure that as the annealing process proceeds, the heating system and surroundings are responding as expected or

• required. Nonetheless, an initial walkdown prior to startup will be performed by the lead operator and Cooperheat shifrengineer, as a minimum. Note that *this walkdown is generally for performing last minute visual checks. More formal verification of the integrity of the installation is made 'throµgh the QC hold points and checks that are included in the. equipment installation procedures.

.Requirements for the periodic walkdowns will be contained in the approved written procedures which will govern the performance of the annealing. These procedures will explicitly define the required frequency of all checks, what minimum specific checks are required, and what documentation is required. All heat treating.technicians, operators, and engineers responsible for carrying out the operations contained in that procedure will receive training on that procedure, as described in TAR Section 1.6.A.1.9.1. At a minimum, walkdowns will be documented in the job log book which will be submitted to CPCo along with the completed procedure at the conclusion of the jab. If deemed necessary, certain checks may be documented usihg checklists included with the procedure. Documentation will include any problems enGountered and corrective action taken, or a statement that all is normal. If problems are encountered at any time during the annealing, including during walkdowns, the individual(s) identifying the problem will be required to inform the lead operator. Solutions to most common problems (e.g., open thermocouples, routine equipment failure, etc.) are contained in rectification instructions that are part of the main procedure. It is the responsibility of the lead operator to ensure that corrective action is taken consistent with these procedures. For problems not identified in these 48

procedures, the lead operator has the authority to take corrective action if the solution to the problem is self-evident or obvious. For more complex problems, the annealing may be placed in a hold status until corrective action can be decided upon, in conjunction with the Cooperheat shift engineer or Cooperheat shift supervisor if required. Depending on the severity of the problem, the lead operator may be required to inform the Westinghouse shift supervisor and/ or utility representatives, even if the problem was rectified per the rectification instructions. For any problem where the solution is outside the bounds of written procedures or rectification instructions, it is the lead operator's responsibility to inform the Westinghouse Shift Supervisor and CPCo representative promptly. 49

Instrumentation

1.

In its letter dated January 12, 1996, Consumers Power Company stated in Section 1. 3. E. 1 that the source range, wide, and power range detectors can withstand operating temperatures up to 300 °F. The temperature limits for the detectors will not be exceeded during the annealing, and Consumers Power prefers to keep the instrument in place if the detector temperature can be maintained below the specified maximum. Additionally, in Consumers Power's Jetter dated February 5, 1996, it is stated that the target inner diameter annealing temperature will be 850 °F to 900 °F for 168 hours and the maximum annealing temperature will not exceed 940 °F. Clarification is needed to determine how detector temperature will be controlled to remain below the instruments' temperature limits/capabilities.

Response

The nuclear detectors are installed in the detector wells in the biological shield. The reactor cavity annulus will be cooled by heat transfer through the concrete to containment, the Biological Shield Cooling System (SCS}, and by a temporary forced air cooling system during the anneal. The analyses described in TAR Section 1.3.C.9 determined that t.he primary heat removal mechanism is through the tempo~ary forced air cooling system. The calculated reactor cavity annulus temperatures noted in TAR Section 1.3.C.9 will remain below the annealing temperature limit for the Palisades biqlogical shield wall concrete inner surface.(250°F). Maintenance of the Palisades biological shield wall concrete in.ner surface at 250°F or less will bound the nuclear detector operating limit of 3d0°F. 50

~..* Fire Protection

1.

Specify the editions of the National Fire Protection Association (NFPA) Codes and Standards (NFPA 54, 58, 70, 79, etc.) referenced in Section 1.6 of the preliminary TAR, that the installation and operation of the gas delivery system, control train, and support equipment will comply with. Identify any deviations from the referenced codes and standards.

Response

Specific Editions of the codes and standards referenced in. TAR Section 1.6.A.1.3 are as follows: NFPA-54, "National Fuel Gas Code": 1992 Edition (NFPA-54-1992) NFPA-58, "Standard for the Storage and Handling of Liquefied Petroleum Gases": 1995 Edition (NFPA-58-1995) NFPA-70, "National Electrical Code": 1993 Edition (NFPA-70-1993) NFPA-79, "Electrical Strandard for Industrial Machinery": 1991 Edition *(NFPA \\ 1991) Note that during.the detailed design phase and the equipment construction and qualification phase of the Palisades project, the requirements of more curr~nt versions

• of the above codes and standards may be employed.

The portions of the above referenced codes and standards.Which apply to the combustion equipment used at Palisades will be defined for the equipment specific to the Palisades project. This information will be formalized during the detailed combustion and control equipment design phase of the project, *currently scheduled to begin in September 1 ~96 under the Palisades Phase II schedule.

• Fire Protection (Continued)'

2. Specify the applicable safety requirements from the American Gas Association Code that the gas control train is in compliance with.

Response

The portions of the codes and standards referenced in TAR Section 1.6 which apply to the combustion equipment used at Palisades will be defined for the equipment specific to the Palisades project. This information will be formalized during the detailed combustion and control equipment design phase of the project, currently scheduled to begin in September 1996 under the Palisades Phase II schedule. 51

Fire Protection (Continued)

3. Specify the maximum quantity of liquefied petroleum gas (LPG) that will be permitted on site at one time. Also provide specific details regarding the location and orientation of the docking area and LPG tankers relative to safety-related structures, including the actual separation distances and elevation differences.

Response

The initial estimates for total propane needed for the on-site tests and the annealing operation is expected to be between 75,000 and 100,000 gallons. Following the Marble Hill Annealing Demonstration Project a refined estimate will be developed. The tanker docking arrangement will be such that as one tanker is providing fuel to the system, a second tanker cari be connected to the propane supply systern piping. The changeover can then be made without any stoppage in the fuel supply. LPG road tankers, which will supply the propane for the annealing operation, come in various.sizes up to a maximum capacity of 19,000 gallons. The size of the road tanker used will be a compromise between the vqlume of gas stored.on site and the number of tanker moves onto site, and the maximum weight of the tanker, given road weight

• restrictions. Preferably, the largest available tankers will be used, since this will minimize the number of changeovers required between tankers. In the evaluation of
• the consequences of potential propane release events.associated with the docking station, a total* quantity of propane equaling. 18,810 gallons was used~ This value considers that the tanker connected to the system at switch over will be partly filled and the second road tanker to be connected is essentially full. Thus, the total quantity of propane at the docking station in the two tankers would be 18,810 gallons. To assure

. that the annealing process is not impacted by a shortage of fuel potentially caused by unforeseen delays in the arrival of the second tanker, the second tanker will be on the owner-controlled site well before connection to the docking station. Thus, the total quantity of propane on the site will be no more than 38,000 gallons; however, this second tanker will be located in an area that will not impact safety-related equipment and structures, in case of a propane release. Several locations for the road tanker docking* station are Under consideration, including

• east of the containment.and the spent fuel buildings. The location of the docking station will be consisterit with the results of the hazardous materials evaluation which

. will include the requirements of NFPA-58, described in TAR Section 1.6.A.1.7.1. The long axis of the tanker will be oriented at right angles to the direction to the containment and the spent fuel building, unless the docking station location and/or the surrounding terrain makes this unnecessary. Detailed information regarding the storage location on site will be available when the locations are finalized during the detailed design phase of the project. This information is currently scheduled to be finalized in February 1997. 52

Fire Protection (Continued)

4. Specify the fire-suppression equipment that will be provided for the docking area, gas delivery system, and control train. Describe the access routes available for mobile fire apparatus to respond to a fire incident involving the docking area considering potential environmental conditions.

Response

Hose with portable deluge nozzles will be provided for the tanker docking area. 'A fire-suppression system would not be used for the gas delivery system or the control trains should a fire occur. These items would first be isolated from their gas supply and the remaining gas in the line would be left to burn out on its own. This action would prevent gas vapor from escaping and gathering in some undesirable location. Fire streams would be available to control the spread of any fire to adjacent buildings or equipment during the burn out stage of a gas delivery system or control train fire. Fire brigade training on propane hazards has begun with an emphasis on road tanker docking operations. This included hands on application of fire suppression tactics and leak repair under emergency conditions. Additional live training is planned to be conducted in the spring of 1997. The location for tanker docking is still being evaluated as described in the response to the previous RAi: The final location chosen will have normal plant access roads available for mobile fire apparatus to respond to a fire incident. Fire department access to the plant is available at all times and at least annual drills are held on site for the local fire department. Fire Protection (Continued)

5. Specify the adminstrative controls on smoking in the vicinity of the docking area, gas delivery system, and control trains that will be in place when LPG is on site.

Response

Administrative controls on smoking in the vicinity of the fuel docking area, gas delivery system, and control trains (in fact, any restrictions on lighted materials or open flames described in TAR Section 1.6.A.1.7) will be strictly enforced. As a minimum, these controls will consist of physical barriers (such as ropes with signs, sawhorses, etc. as required) around the areas in question. These measures will supplement the limited smoking areas already in* effect at the Palisades site. 53

Fire Protection (Continued)

6. Specify the listing or approval that has been obtained for the components of the gas delivery system, and control train.

Response

AGA Certification (the "Blue Star" seal) is given to specific models of an appliance or device. This certification entails testing by AGA of the applicance or device to some nationally recognized standard (often an ANSI standard) or a specially prepared standard (an "AGA Requirement") in the absence of a suitable existing standard. Also, inspection of the manufacturer's Quality Assurance program and manufacturing facility is required. Although designed to meet the applicable safety and perfomance requirements of the codes and standards described in TAR Section 1.6.A.1.3, no listing or approval has been obtained for the Cooperheat Gas Control Train, since such listing or approval is not required in the normal course of Cooperheat's business. Note that this equipment is designed and built for Cooperheat use in contract heat treatment services only, and is not used for permanent installations. Other major components of the gas delivery system (vaporizers, tanks, regulator, hoses) may or may not have specific listings or approvals, since such listing or approval is not required in the normal course of Cooperheat's business. At this time, specific model numbers for those components are not identified. Until specific model numbers for these components are identified, such listings or approvals can not be specified. Specific model numbers for these components will be identified during the equipment *construction and qualification phase of the Palisades project, currently scheduled to begin in 1997 under the Palisades Phase Ill schedule. 54

Radiation Protection (Sections 1.3, 1.5, 1.6, and 1.9)

1. Section 1.3.F.2 of the Thermal Annealing Operating Plan states that the "dose rate target for the shielding was established as a calculated dose rate of 500 mR/hr or less in the refueling cavity work areas and 60 mR/hr or less at the 649 foot elevation."

1) Discuss why you did not design the shielding for the reactor internals to obtain a lower calculated dose rate in the "refueling cavity work areas.

2) Discuss what major work evolutions will be conducted in the refueling cavity work areas following draindown of the water in the refueling cavity prior to annealing.

• Provide the estimated person-rem doses associated with each of these work evolutions.

3) Provide layout maps of the major plant areas where annealing operation activities will take place. Include on these layout maps the estimated dose rate.

zones that will be present during the annealing operation (i.e., once the reactor ' internals have been stored. in the refueling cavity and the water in the refueling cavity has been drained down). _Response: .The dose receptor points described in Section 1.3.F.2.and as summarized above were estab,lished to limit*personnel exposure ALARA in work areas where frequent or extensive annealing - related activities will be conducted (e.g. at the 649 ft. elevation) and to allow higher doses in areas where annealing related work will be minimal or where physical restriction prevents the provision of additional gamma shielding (e.g. in the refueling ca'vity), Also factored into the establishment of these dose receptor points is an economic factor of $9,000/person-rem. Because of space restrictions in the reactor vessel flange area and due to the fact that minimal work will be perfprmed in the refueling cavity area, the dose receptor value for t8is area is 500 mrem/hr. Most annealing related activities in the refueling cavity area such as thermal barrier installation, internals shield installation, etc. will be performed with the refueling cavity flooded to minimize personnel dose. Only limited, short-term activities such as heat exchanger control thermocouple actuation (switchblades) and he.at exchanger ductwork interconneCtion will be performed by personnel located in the 500 mrem/hr refueling. cavity area. Efforts will be made in the currently ongoing detail

• design phase of the project to minimize activities and stay times in the relatively high 500 mrem/hr radiation area. The attached table provides:a*breakdown of the total personnel exposure estimated for each activity during the annealing evolution.

Referring to the table, the activities to be performed within the refueling cavity with the cavity drained are Activities 17, 19, 20, 21, 22, 26, 28, and 31. The attached table is a preliminary dose assessment for the total annealing evolution. The project will be 55

developing and issuing a comprehensive ALARA Plan during the currently ongoing Phase II Detail Design effort which will update.and expand significantly the preliminary assessment in the table. A preliminary issue of this ALARA Plan is currently scheduled to be released in November 1996. (See also, response to RAI #7 Radiation Protection below.) The preliminary ALARA Plan to be issued in November will include containment layout maps of the maj~r work areas where various annealing activities will take place. These maps will include the general area dose rates expected during the annealing evolution both with the refueling cavity filled and drained. These dose layout maps, in conjunction with the site activities schedule will be utilized to develop a more detailed and expanded person'nel exposure tabulation. 56

PALISADES THERMAL ANNEAL PROGRAM DOSE ASSESSMENT - PRELIMINARY Average Total ID Task Description Location Dose Rate Dose lmr/hr) (man-rem) 1 Transport and assemble equipment within Aux. Bldg. SFP 2 0.4 2 Install / setup Internals storage I shielding support equipment Cavity 30 1.0 3 Move refueling equipment out of c.ontainment I clear work areas Deck 5 1.3 4 Move equipment into containment and assemble Deck 5 1.2 5 Install heat exchanger ducting to refueling cavity Deck 2 0.3 6 Install /setup temporary ducting I support equipment for heat exchanger test Deck 10 '5.5 7 Perform heat exchanger containment test I remove temporary setup Deck 10 1.9 8 Remove surveillance capsules & CSB and place CSB into temporary storage stand Deck 60 6.0 9 Place temporary shield around cavity/move UGS to CSB Deck 60 1.9 10 Install shielding around internals, top cover & contamination cqntrol sys. Deck 90 4.0 11 Perform RV pre-anneal measurements and visual inspections Deck 30 2.9 12 Install external instrumentation around RV & perform visual inspections Annulus 2,000 30.0 13 Install biological shield supplemental cooling system Annulus 1,000. 10.0 14 Install external instrumentation on PCS piping & SIG base measurement Loops 1,000 20.0 15 Drain refueling cavity to RV flange I hookup contamination control system Deck 30 1.3 16 Install nozzle thermal barriers Deck 30 1.5 17 Remove cavity seal, stud hole plugs and detector well covers / Install RVTC pins Cavity 400 2.6 18 Lower heat exchanger assembly into RV and drain RV Deck 60 0.6 19. Seat heat exchanger assembly and remove lift rig Cavity 500 2.0.. 20 Complete heat exchanger ducting I Hookup RV Purge Cavity 500 4.0 21 Remove RV Guide Pins and sleeves Cavity 500 0.3 .22. Actuate R"'. internal instrumentation I System walkdown Cavity 500 0.5 23 Secure all water sources to RV Loops Later Later 24 System heatup for RV drying and raise to temperature Deck 15 0.7 25 Anneal RV Deck 15 2.5 26 Retract convection barrier /-hookup forced air cooling system Cavity 500 0.3 27 Cool down RV Deck 15 2.2 28 Rem9ve heat exchanger ducting I Prepare heat exchanger for lift Cavity 500 4.0 29 Restore water sources to RV Loops Later Later 30 Pull heat exchanger ( Fill RV Deck 15 0.4 31. Remove RVTC pins, install seal, stud plugs; guide pins/sleeves & detector well covers Cavity 400 4.4 32 Remove nozzle thermal barriers Deck 30 1.0 33 Fill refueling cavity I Remove refueling cavity contamination control system Deck 30 1.3 34 Perform RV post-anneal measurements and visual inspections Deck 30 2.9

35. Perform post-anneal ISi Deck Later Later 36 Remove internals shielding top cover I Move UGS to storage stand Deck 60 1.4 37 Remove internals shielding Deck 90 4.0 38 Install CSB and surveillance capsules into RV Deck 60 7.0 39 Remove biological shield supplemental cooling system Annulus 1,000,*

10.0 40 Remove external instrumentation around RV & perform visual inspections Annulus 2,000 30.0 41 Remove external instrumentation on PCS piping & SIG base measurement Loops 1,000 20.0 42 Disassemble equipr:nent and move out of containment Deck 5 5.0 43 Move refueling equipm.ent into containment I restore areas Deck 5 1.2 44 Transport and disassemble equipment within Aux. Bldg. SFP 2 0.4 45 Remove Internals support equipment Cavity 30 1.0 46 Health Physics (10 %) 20 47 CPCo Engineering I Management oversight 2.0 Total Dose Assessment ILaters not included in total\\ 220.5 Table 1 for RAI #1, Radiation Protection (Sections 1.3, 1.5,1.6, and 1.9) 57

Radiation Protection (Sections 1.3, 1.5, 1.6, and 1.9) (Continued)

2. Section 1.3.F.5 discusses the process of removing the upper guide structure (UGS) and core support barrel (CSB) from the RV and placing these components on

• storage pads in the west end of the refueling cavity. Provide plant layout drawings (including sectional cut drawings) showing the proposed storage location of these components in the refueling cavity. Include on these drawings a depiction of the shielding structure and horizontal top cap that will be constructed around the core internals in the refueling cavity.

Response

Conceptual drawing WGC 9005 for which details are shown in the attached Figures 1 and 2 of this RAI depicts the concept of how the CSB and USG will be stacked in the west end of the refueling cavity on temporary pads and how the reactor internals shield will enclose and shield the stacked internals. Recently, the project decided to change to reinforced concrete for much of the internals shield structure, however, the concept depicted in the subject.drawing is still accurate. The QAD shielding analysis and structural framing detaHs are currently under project review. The final TAR will reflect the change in design. 58

---

~-

I ( ' a* \\ ENLARGED VIE\\{\\ rDI COC: FU D.. 61*"-6" SE:CTION A-A ~:*:... Figure 1 for RAI #2, Radiation Protection (Sections 1.3, 1.5, 1.6 and 1.9) showing a sectional view of the intern~ls shielding arrangement (from WGC 9005) 59 \\ i

Figure 2 for RAI #2, Radiation Protection (Sections--1;3~ 1.5, 1.6 and 1.9) showing a top view of the internals shielding arrangement (from WGC 9005) i 601 \\ _______ :

Radiation Protection (Sections 1.3, 1.5, 1.6, and 1.9) (Continued)

3. Sections 1.3.F.5 and 1.3.F.6 discuss the use of temporary shielding on the refueling deck to minimize personnel exposures during the insertion and withdrawal of the UGS from the CSB. Describe where this temporary shielding will be erected and who it will be used to provide protection for.

Response

Radiation dose to personnel during the dry lift of the UGS will be minimized by the use of remotely operated cameras for the positioning of the UGS into the CSB and a water shield with plexiglass viewing port for the crane signalman and heavy loads supervisor. The location for this shielded viewing port may be on top of _the steam generator secondary platform (north). The steam generator blockhouse will provide additional shielding from the direct shine imparted from the lower section of the UGS. The crane operator will have the steam gen.erator blockhouse (and temporary shielding if needed) for radiation protection during the critical lift. Non-essential personnel will be cleared

• from containment.

Radiation Protection (Sections 1.3,. 1.5, 1.6, and 1.9) (Continued)

4. Section 1.5.D states that there will be provisions to control radioactive contamination

.. during the annealing operation. In addition to the provisions to control airborne.

• contamination currently in use at Palisades, describe any additional *measures that.

.will be taken to control airborne *contamination created by the annealing process. *In particular, discuss measures to control airborne contamination (which may result from dryout of the core internals) inside the constructed shield structure. Also discuss the venfilation system that will be installed to filter and exhaust the contaminated air in the RV and reactor vessel annulus areas during the drying and* annealing process.

Response

. The RV internals and refueling cavity airborne contamination control systems will be used'to create an airflow into the RV internals shielded enclosure and refueling cavity to minimize the escape of airborne radioactivity into the containment atmosphere. Both of these systems will be operated prior to, during and after the annealing. In addition to a high efficiency particulate air (HEPA) filter stage, other filter stages, including a prefilter and/or moisture separator will be used to maximize system efficiency as required. Discharge from the systems will be directed in a manner that minimizes the creation of airborne radioactivity in adjacent areas. The RVTC cover plate insulation interface with the RV flange and RVTC penetrations insulating packing will provide a barrier against convective air flow from the reactor 61

vessel and thereby minimizing the escape of airborne radioactivity into the containment atmosphere. After the reactor vessel is drained, the RV drainage/purge system piping integrated into the heat exchanger assembly will be configured to allow connection to a filtered ventilation system. The reactor vessel will then be heated to apprpximately 250°F and held to dry. Any steam from residual moisture in the reactor vessel will be removed by the RV purge system. Provisions for cooling and condensing the hot air to assure system performance shall be provided as appropriate. The exiting air will be filtered during the drying portion of the heatup and during the cool down phase. During the annealing phase, the RV purge SY,stem may be used in a low flow mode to maintain the RV internal pressure at atmospheric. This could be required to minimize airborne contamination in the reactor vessel from escaping during the annealing without significantly cooling the reactor vessel. The reactor vessel may be vented in this manner throughout the annealing process. The RV purge system and RVTC airborne contamination control.provisions are also supported by the airflow into the reactor cavity annulus created by the biological shield suppl"emental cooling system and also by the RV internals and refueling cavity airborne contamination control systems which draw air in from the areas surrounding the RVTC. The biological shield supplemental cooling system will be used to create a cooling airflow down into the reactor cavity annulus area and to minimize the escape of.

• airborne radioactivity into the containment*atmosphere. The air flow from the reactor cavity access tube will be filtered and directed in a manner that minimizes the creation of airborne radioactivity in adjacent areas.

Radiation Protection (Sections 1.3, 1.5, 1.6, and 1.9) (Continued)

5. Section 1.6.A.5. 1 c;Jescribes the RV top cover (RVTC) that will be installed over the dry RV to support the heat exchanger assembly, provide shielding, and control airborne contamination.

1) It is unclear from the description in this section whether or not the RVTC is an integral part of the heat exchanger. If the RVTC is not an integral part of the heat exchanger, describe the procedure for lowering the RVTC over the heat exchanger ducting following insertion of the heat exchanger into the RV.

2) Describe how the two 6-foot semicircular cylindrical shielding segments will be installed around the heat exchanger ducting on top of the RVTC (will this be performed remotely?). Discuss the potential for streaming of radiation up through the heat exchanger ducting and out the top of these cylindrical shielding segments and how you plan to protect plant personnel from this potential source of radiation.

3) Provide an estimate of the dose rate on top of the RVTC once it has been 62

installed on top of the reactor vessel. Describe any work functions that will be performed in this area in preparation for the annealing operation.

Response

The Reactor Vessel Top Cover (RVTC) is an integral part of the heat exchanger assembly. Prior to installation into the reactor vessel the RVTC will be assembled over the heat exchanger and stub ducting on the operating floor. The cylindrical shielding segments are an integral part of the RVTC.. The cylinder is assembled on the lift frame ring girder. The installation will be performed as part of the RVTC and heat exchanger assembly work on the operating floor prior to insertion irito the reactor vessel. Pers.onnel performing fitup of t~e heat exchanger ducts, actuating the control thermocouples (switchblades), etc. will be located on the RVTC. The RVTC is designed in combination with the RV internal shield structure to attenuate radiation sources such that the general area dbse in the refueling cavity area will be 500 mrem/hr or less. The aforementioned cylindrical shield in combination with adminstrative controls (e.g., HP coverage, procedure controls) will protect personnel working in the refueling cavity from the streaming source which results from the-ducting penetrating the RVTC. After landing the heat exchanger assembly onto the RV flange it will be necessary to*complete.ttie installation and set-up of the heat exchanger assembly ducting and support equipment. The use of temp.orary shadow shielding or remote tooling will be employed as appropriate during the removal and reinstallation of guide pins and sleeves, RVTC lift rig, RV Drainage/Purge system, actuat.ion of temperature probe delivery mechanisms and convection.barriers, etc. as de~cribed in response to RAI #1, Radiation Protection above. Radiation Protection (Sections 1.3, 1.5, 1.6, and 1.9) (Continued)

• 6. Section 1.9.A.3 states that the projected.dose for the annealing projectis approximately 200 person:rem, excluding other outage scope tasks and support.

1) Provide a job dose breakdown for each of the tasks associated with the annealing project and provide the most recent total job dose estimate for the annealing project..

Response

The table attached to RAl#1, Radiation 'Protection provides a preliminary activity dose breakdown for each major task associated with the annealing evolution. As stated in. the responses to RAI #1 and to #7, Radiation Protection, a much more detailed dose breakdown table will be provided in the preliminary ALARA plan.currently scheduled for 63

issue in November 1996. A copy of this plan will be provided to NRC. Radiation Protection (Sections 1.3, 1.5, 1.6, and 1.9) (Continued)

2) Discuss how you plan to integrate the annealing project with the other planned outage scope tasks so that the other scheduled tasks will not interfere with the annealing project. Discuss your plans to limit personnel entries into the containment once the reactor internals have been removed from the RV and the water in the refueling cavity has been drained down.

Response

A prelimina*ry schedule of the annealing activities will be drafted by the annealing project personnel considering generic Palisades refueling activities that occur every refueling outage. This draft schedule will be used by annealing project personnel and Palisades Outage Management personnel to integrate the annealing project into the annealing/refueling schedule to minimize outage durationwhile considering safety and ALARA issues. Personn_el access to the 649-foot level of.con"tainment will be limited during the period of the Upper Guide Structure dry air lift into the Core Support Barrel through annealing. and return.of the UGS to its normal storage location. Generally, the refueling outage schedule itself will limit personnel on the 649-foot level of containment during this period simply due to the fact there is little room left to perform other activities. The final schedule will consider ALARA when scheduling other activities during annealing on the 649-foot level of cc;mtainment. The allowed access during this period to the 649-foot level of containment will include personnel as required to support the annealing project

• and others as required to support non-annealing activities in support of the. refueling
• . outage schedule.. Access to containment for work on lower levels will. generally be through the personnel hatch to avoid unnecessary radiation dose due to walking through the 649-foot level of containment.

Radiation Protection (Sections 1.3, 1.5, 1.6, and 1.9) (Continued)

7. Provide a copy of the annealing as low as is reasonably achievable (ALARA) plan mentioned.in Sec_tion 1. 9.A. 3 of the Thermal Annealing Operating Plan.

Response

As stated in the responses to RAI #1 and #6, Radiation Protection, the project will issue a preliminary ALARA Plan in November, 1996. This plan will also be provided to the NRC. This plan will include containment layout maps defining general area dose rates for key activity areas. The plan will also inclu.de an activity dose breakdown for each 64

~* annealing work activity and the expected number of personnel performing each activity, the duration and the total dose expected to be received. This preliminary ALARA plan will be updated as the design of the equipment is completed. The final outage schedule will also be evaluated for impacting the ALARA plan. Radiation Protection (Sections 1: 3, 1. 5, 1. 6, and 1. 9) (Continued)

8. Section 1.9.C.4 states that the experience gained from the installation of thermocouples in the reactor cavity annulus during the 1995 refueling outage at Palisades was beneficial in minimizing personnel dose. Since a majority of the estimated dose from.the annealing project will come from installation of instrumentation (in the reactor cavity annulus, primary coolant system piping loop areas, and other high dose rate areas), you should expand Section 1.9.C.4 of the Thermal Annealing Operating Plan to describe some of the dose reduction features and practices used to minimize personnel dose during instrumentation installation.

Response

The experience gained and techniques developed for thermocouple installation in the reactor cavity annulus during the 1995 Palisades refueling outage were beneficial in minimizing personnel dose. Maximum advantage will also be taken of the Marble Hill annealing demonstration project to find additional ways to enhance the equipment, processes, and procedures. Lessons learned during the Marble Hill annealing demonstration project will be incorporated into the Palisades annealing project, as, applicable. Since much of the same instrumentation equipment is planned to be used, any upgrades to the systems which could reduce radiation exposure will be considered. for incorporation. Some of the dose reductio11. features and practices used to minimize personnel dose during external instrument installation are described below. These approaches will be utilized in the annealing instrumentation design and preparations for installation* in the reactor cavity annulus, primary coolant system piping loop areas and ot~er high dose areas. For certain annealing project tasks, such as reactor cavity annulus instrumentation installation, personnel assigned will undergo training, as required, on full size representative mock-ups while simulating the working area field conditions and protective clothing restrictions expected to be encountered. This mock-up training is . extremely critical in reducing the amount of time required to perform a task, thus saving worker exposure. Where practical, the actual annealing project equipment and simulated environments will be used in order to minimize risk of error, facilitat~ tool development, familiarize the crew with equipment features to enhance task proficiency, and to develop ALARA work practices. Remote long handled tooling will be used to install some external temperature measurement devices. The tooling and equipment to be utilized during the Palisades 65

annealing effort will be designed to facilitate efficient assembly, disassembly and repair, thus precluding unnecessary maintenance time in high radiation areas. The setup of equipment and the performance of the annealing will utilize personnel who have experience in design, maintenance and repair operations in a radioactive work environment which can be drawn upon to minimize radiation exposure. The final TAR will include wording similar to the above. Radiation Protection (Sections 1.3, 1.5, 1.6, and 1.9) (Continued)

9. Section 1. 5. C. 3 describes the use of sensors for external displacement measurements. Therefore the first word in Section 1. 5. C. 3. 4 should b~. changed * *

. from "temperature" to "external displaceme(lt".

Response

. The statement made in the question is correct. As a result in the final version of the. TAR the first word "Temperature" in Section 1.5.C.3.4 will be changed to "External displacement". Radiation Protection.. (Sections 1.3, 1.5, 1..6, and 1.9) (Continued)

• -10.

Section 1. 9 of the Thermal Annealing Operating Plan. is entitled "ALARA Considerations" and it discusses some of the methodologies and procedures thC!t Consumers Power Company will use to ensure that occupational doses will be *. maif!fained ALARA during the annealing process. Regulatory Guide 8. 8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Is Reasonably Achievable"contains . guidelines (such as equipment selection and design, proper use of shielding,

• ALARA training, and the establishment of a good radiation protection program) on how to maintain radiation*dosesALARA. In Section 1.9 of the operating plan, verify that you utilized the guidance contained in Regulatory Guide 8. 8 to ensure

. that radiation doses will be maintained ALARA during the annealing project. .Response: In addition to the specific issues highlighted in Section 1.9 of the TAR, all operations associated with the annealing project will. be performed in accordance with the Palisades Radiation Safety Program as implemented by the Chemical and Radiological Services Department procedures. This program requires that the guidelines set forth in Regulatory Guide 8.8 be followed throughout the annealing procedure. All aspects of the equipment design and annealing procedures will be subjected to design reviews including ALARA considerations. Pre-and post-job ALARA reviews of all phases of 66

the operation will be conducted. The Palisades Chemical and Radiological Services Department will be utilized throughout the operation to provide job follow, dose monitoring, and radiation protection oversight for all procedures that take place within the Radiation Control Area. The ALARA plan as mentioned in Section 1.9 of the TAR is intended to meet the requirements of Regulatory Guide 8.8. 67

Radiation Protection (Section 1. 10)

1. Section 1. 10. E of the Thermal Annealing Operating Plan provides a summary of the planned sequence for the annealing operation. This summary states that the temporary shielding around the reactor internals will be removed prior to removal of the UGS from the CSB. Since it would appear that dose rates in the vicinity of the temporary shielding would be lower once the UGS was removed from the CSB, justify why the temporary shielding will be dismantled prior to removal of the UGS.

from the CSB.

Response

In order to remove th~ UGS from the CSB the top cover of the temporary shielding will require disassembly and removal. The UGS top cover shielc;tirig will be removed. This

• will allow the lift rig to be installed and the UGS to be lifted out of the CSB. The UGS will be lifted out of the water'and moved east to its storage stand and lowered into the refueling cavity water. This operation is expected to take less than 15 minutes. Once the UGS is removed from the CSB and returned to its* normal storage position, disassembly and removal of the temporary shielding walls around the CSB can continue. The final TAR will clarify this sequence both in S~ctions 1.5. D and 1.1 o*. E.

Radiation Protection (Section 1. 10) (Continued)

2. Section 1. 10. 1 states that industry experience gained from the annealing demonstration project at Marble Hill will be applied to Palisades. Discuss any lessons learned from the Marble Hill annealing project that will result in estimated dose savings during the upcoming Palisades annealing project.

• Response:

There are several lessons learned stemming from the Marble Hill Annealing Demonstration Program. Examples of some of the major ALARA lessons learned are as follows:

a. Enhance the anchor points/mechanism for the vertical ductwork extending upward from the heat exchanger.

b. Investigate enhancement of the ductwork coupling methodology particularly in the high dose areas (i.e.: mechanical joints in lieu of welding).

c. Re-design the Reactor Vessel Top Cover (RVTC) alignment pins to allow for greater flexibility for installation and removal.

d. Modify the existing holes in the RVTC to allow for increased flexibility in the removal 68

and subsequent installation of the RV guide pins.

e. Integrate known refueling cavity temperatures from the Marble Hill ADP into the design efforts for the Refueling Cavity Airborne Contamination Control System.

f.

Refine the vessel external instrumentation delivery system to allow for quicker connects and disconnects.

g. Utilize shadow shielding and cameras for insertion and removal of the heat exchanger.

h. Incorporate RV flange insulation on the RVTC.

i. -Investigate alternate RVTC/Heat Exchanger leveling methodology:

j.

• Review the possibility of reduCing both pre/post RV measurements based on results I

from Marble Hill annealing effort.

k. Utilize communications head sets for high dose efforts.

I. Pre-assemble as many components as possible prior to transportation into containment. All of the lessons iearned from the Marble Hill Annealing Demonstration Project will be. - reviewed and incorporated as applicable into the Palisades Annealing.efforts. Per the current schedule, the lessons learned document will be issued with the rest of the final Marble Hill Annealing Demonstration Project report to the ASME-DOE/S_NL-lndustry Steering -comm.itte~ in* November 1996 for releas_e and distribution. 69

CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 ATTACHMENT A RAI #3, THERMAL AND STRESS ANALYSIS AND ASMECODECASEN~5~ PALISADES THERMAL ANNEALING PROJECT ADP/PALISADES INSTRUMENTATION MONITORING POINTS 22 Pages

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points ' PALISADES THERMAL ANNEALING PROJECT Rev. 1 July 1996 ADP/PALISADES INSTRUMENTATION MONITORING POINTS

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996 1.0 Purpose and Background The purpose of this document is to identify the Marble Hill Annealing Demonstration Project (ADP) instrumentation monitoring points which will be used to compare against the Palisades Thermal Annealing instrumentation approach as outlined in the Palisades Thermal Annealing Report (TAR) Sections 1.5 *and 2.1 (References 1 and 2). For the Monitoring Contingency Approach as described in TAR Section 2.1.B a single subset of sensors which meets the intent of Tables 2.1.A-3 and 2.1.A-4 of TAR Section 2.1 will be addressed. This comparison ~s intended to lead to the development of a justification for the Palisades instrumentation 'monitoring approach and to changes 'in this approach, if warranted. The Palisades monitoring approach is intended to satisfy four objectives as provided in .. Reference 4. These include:

1.

Make sufficient measurements for difect comparison with limiting parameter values to demonstrate that the annealing process was successful.

2.

Make sufficient measurements for direct comparison with limiting parameter values to demonstrate that the annealing process did not adversely affect the integrity :_of the reactor vessel and adjacent structures. '3. Make sufficient measurements for input to the analysis model for determining the temperature, stress, and strain profiles for the vessel and adjacent structures.

4.

Make* sufficient measurements for direct comparison with the output from the analysis model to illustrate that the assumptions made for the computer analysis were sufficient to yield realistic results. These objectives.are consistent with the ADP-Marble Hill instrumentation functions, however because of the noncontarninated environment at Marble Hill and the differences in the geometries between Marble Hill and Palisades there are differences in instrumentation placement and in instrumentation quantity and type. Thus not all of the ADP instrumentation is directly applicable to justifying the Palisades monitoring approach as described in References 1 and.2. Additionally because of the differences in the geometries some sensors in the Palisades monitoring approach are unique to Palisades and thus are not reflected in the Marble Hill ADP instrument plan. However even the remainder of the ADP instrumentation is important to validate the need or. lack of need for additional sensors at other locations for Palisades, as described in References 1 and 2. The following information will be provided in the futu,re as a revision to this document:

1.

Criteria for evaluating the monitoring data from the ADP. 7/8/96 2 insloc.pal

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996

2.

Definition of a single subset of sensors which meet the intent of Tables 2.1.A-3 and 2.1.A-4 of TAR Section 2.1.

3.

Justification logic for using the ADP monitoring data to justify the Palisades monitoring approach.

4.

Revision of Palisades monitoring approach, if necessary.

5.

Justification for using the Palisades monitoring approach. 2.0 References

1.

Palisades Plant Thermal Annealing Report, Section 1: Thermal Annealing Operating Plan; Section 1.5 - Annealing Method, Instrumentation and Procedures, January 12, 1996.

2.

Palisades Plant Thermal Annealing Report, Section 2: Requalification, -Inspection. ano

• .. Test Program, Section 2.1 - Monitoring the Annealing Process, February 5, 1996.

3.

Westinghouse Document, Reactor Vessel Annealing Demonstration Project at Marble* Hill Unit 1, Task 3 - Instrllmentation Locations, January 29, 1996, Revision 2~

4.

Palisades Reactor Vessel Annealing Project, Protocol for External Measurements,

• Parsons Power Document, TIN-95-1288.[DRAFT].

3.0 Identification of Instrumentation Locations

• The instrumentation.locations and types identified in References 1 *and 2 were compared.*

against the instrumentation locations and types in Reference 3, Instrumentation with a* common function and location between the two projects are identified in Table 1. Figures 1 and 2 show the instrumentation location. approach for Palisades and show the relative locat~ons of the ADP monitoring points identifieq in Table 1. These will be directly used to justify the Palisades monitoring approach. Zones G and J sensors identified for Palisades in Table 1 have no corresponding locations in the Marble Hill ADP because of different vessel support and different reactor cavity cooling features, respectively. Instrumentation identified in Table 2 are additional Marble Hill ADP sensors which do not correspond directly with the Palisades monitoring approach sensors. Figures 3 and 4 show the Table 2 sensor locations relative to the Palisades plant configuration. The results from these sensors at the Marble Hill ADP will need to b~ reviewed to determine whether additional sensors and/or different sensor locations are required. 7/8/96 3 insloc.pal

P-es Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Rev.1. July 1996 Table 1: Marble Hill ADP/Palisades Equivalent Instrumentation Monitoring Points Zone Type of Sensor ADP Palisades Comments

• (Number of_

(Number of Sensor Sen8or Locations)/ Locations )<1> (Sensor Identification) A temperature 6 3 External monitoring location on the RV bottom head. T_V9_05 T_V9_06 l T_V9_07 . T_V9_08 T_ V9_09 T_V9 10 - A displacement 1 1 . One location is listed, however three sensors are at one location representing three orthogonal directions. The loc_;ation for Palisades has only a vertical displacement sensor. External monitoring location on RV bottom head. D_ V9_01 D_V9_02 7/8/96 4

des Thermal Annealing Project . ADP/Palisades Instrumentation Monitoring Points

• i Rev.1-July 1996 Table 1:

Marble Hill ADP/Palisades Equivalent Instrumentation Monitoring Points Zone Type of Sensor ADP Palisades Comments . (Number of _(Number of Sensor. Sensor Locations)/ Locationsi<1> * (Sensor Identification) D_V9_03 B temperature 4 3 External monitoring location at bottom of RV annealing zone. T~V9_01 T_V9_02 T_V9_03 T_V9_04 c temperature 4 '3. External rn:onjtoring location at middle of RV annealing zone. T_V8_01 T_ V8_02 T_,V8_03 T_V8_04 m~ s

des Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996 Table 1: Marble Hill' ADP/Palisades Equivalent Instrumentation -Monitoring Points Zone Type of Sensor ADP Palisades Comments (Num~er of (Number of Sensor Sensor Locations)/ Locations )<1> (Sensor Identification) D temperature 4 3 External monitoring location at top of RV annealing zone. T_V7_05 T_V7_07 T_ V7_08 T_ V7_10 E temperature 5 6 The ADP ineasurements are taken on the external bottom surface of the safe end to pipe weld. The Palisades measurements are taken on the internal bottom surface near the PCS piping interface. T_P1A_03 T_P1B_03 T_P2A~05 T_P3A_05 T_P4B_05 m~ 6

des Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996 Table 1: Marble Hill ADP/Palisades Equivalent Instrumentation Monitoring Points Zone Type of Sensor ADP Palisades Comments (Number of (Number of Sensor Sensor Locations)/. Locations )<1> (Sensor Identification) F temperature 5 6 The ADP measurements are taken on the external top surf ace of the safe end to pipe weld. The Palisades measurements are taken on the internal top surface near the PCS piping interface. T_P1A_02 T_P1B_02 T_P2A_04 T_P3A_04 T_P4B_04-G temperature 0 1 External monitoring location on the RV support structure coincident with RV annealing zone. H temperature 4 2 External monitoring location on the RV cavity liner coincident with the middle and lower sections of the RV annealin$ zone. T_Cl_Ol 7/8/96 7

a.. Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points

• Rev.I
• July 1996 Table 1:

Marble Hill ADP/Palisades Equivalent Instrumentation Monitoring Points Zone Type of Sensor ADP Palisades Comments (Number of (Number of Sensor -, Sensor Locations)/ Locations )<1> (Sensor Identification) T_C1_02 T_C1_03 T_C1_04 I displacement 5 6 Five external monitoring locations outside of the biological shield wall and on the primary piping, representing five piping runs, are listed for the ADP, however three sensors are at each location representing three orthogonal directions. The locations for Palisades are similarly configured, however six piping runs are represented. D_P2A_Ol D_P2A_02 D_P2A_03 D_P2B_Ol D_P2B.:_02 m~ s

Ptlltdes Thermal Annealing Project ADP/Palisades IllStrumentation Monitoring Points Rev. 1 July 1996 Table 1:.,. Marble Hill ADP/Palisades Equivalent Instrumentation Monitoring Points Zone Type of Sensor ADP Palisades Comments (Number of (Number* of Sensor Sensor Locations)/ Locations}°> (Sensor Identification) D_P2B_03 D_P3A_Ol D_P3A_02 D_P3A_03 D_P3B_Ol D_P3B_02 D_P3B_03 D_P4B_Ol D_P4B_02 D_P4B_03 J temperature 0 2 External monitoring location on the RV cavity liner coincident with the upper shell region of vessel. 7/8/96 9

-des Thermal Annealing hoject ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996 Table 1: Marble Hill ADP/Palisades Equivalent Instrumentation Monitoring Points Zone Type of Sensor ADP. Palisades Comments (Number of (Number of Sensor Sensor Locations)/ Locations )<1> (Sensor

• Identification)

K temperature 22 30 Internal monitoring location on the RV shell in the RV annealing zone: T_V4_01 T_ V4_02 T_ V4_03 T_ V4:_04

• T_V4_05 T_V4_06 T~Vl_Ol T_V1_02 T_V1_03 T_V1_04 T_V1_05 7/8/96 10

-des Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Poinjs Rev.l. July 1996 Table 1: Marble Hill ADP/Palisades EquivaleµtJnstrumentation Monitoring Points Zone Type of Sensor ADP Palisades Comments (Number of (Number of Sensor. Sensor Locations)/ Locations )<1> (Sensor Identification) T_V1_06 T_V1_07 T_V1_08 T_V1_09 T_Vl - 10 T:...Vl - 11 T_Vl - 12 T_Vl - 13 T_Vl - 14 T_Vl - 15 T_Vl - 16 L temperature s* 12 Internal monitoring location on the RV shell above the nozzles. 7/8/96 11

-des Thermal Anneali~g Project ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996 Table 1: Marble lijll ADP/Palisades Equivalent Instrumentation Monitoring Points Zone Type of Sensor ADP Palisades Comments (Number of

• (Number of Sensor Sensor Locations)/

Lo~ations )0> (Sensor Identification) T_V5_01 T_V5_02 T_V5_03 T_ V5_04 T_V5_05 T_ V5_:06 T_ V5_:_07 T_V5_08 M temperature 7 12 Internal monitoring location on the RV shell above the RV annealing zone but below the nozzles.

• T_ V4_07 T_ V4_08 T_V4_09 7/8/96 12

des Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Table 1:

• Marble Hill ADP/Palisades Equivalent Instrumentation Monitoring Points Zone Type of Sensor ADP Palisades Comments (Number of.

(Number of Sensor Sensor Locations)/ Locations )<1> (Sensor* Identification) T_V4_10 T_Y4_11 T_V4_12 T~V4_13 Rev. I. July 1996 N temperature 4 6 Internal *monitoring location on the RV shell below the RV annealing zone. T_V2_01 T_ V2_02 T_V2_03 T_V2_04 Note: 1.

• Palisades sensor identification has not* been established at this time.

7/8/96 13

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996 Table 2: Other ADP Instrumentation Monitoring Points Not Directly Aligned With Palisades Instrumentation Monitoring Approach Points Type of Sensor ADP Comments (Number of Sensor Locations)/ (Sensor Identification) temperature 11 External monitoring points in the upper vessel area. T_V6_01 below the vessel flange T_V6_02 below the vessel flange T_V6_03 above vessel nozzle T_V6_04 above vessel nozzle T_V6_05 above. vessel nozzle T_V6_06 above vessel nozzle T_V6_07 between vessel nozzles T_V6_08 between vessel nozzles T_V6_09 between vessel nozzles ) T_V6_10 between vessel nozzles T_V6_11 between vessel nozzles temperature 6 External monitoring points in the upper vessel area below the vessel nozzles. T_V7_01 below vessel nozzle T_V7_02 below vessel nozzle T_V7_03 below vessel nozzle

• T_ V7_04 below vessel nozzle T_V7_06 top of anneal zone under vessel nozzle 7/8/96 14

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996 Table 2: Other ADP Instrumentation Monitoring Points Not Directly Aligned With Palisades Instrumentation Monitoring Appro_ach Points Type of Sensor ADP Comments (Number of Sensor Locations)/ (Sensor Identification) T_V7_09 top of anneal zone under vessel nozzle temperature 1 ExternaI monitoring point in the lower vessel region. T_V9_11 bottom of lower head displacement

4.

External monitoring points at. vessel support paqs in the radial direction. D_V7_01 D_V7_02 D_V7_03* D_V7_04 temperature 18 External mo'nitoring points on

• the primary.coolant system piping o~tside of the biological shield wall.

T_PlA_Ol T_PlB_Ol T_P2A_Ol T_P2A_02 T_P2A_03 T_P2B_Ol T_P2B_02 T_P2B_03 T_P3A_Ol 7/8/96 15

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996 Table 2: Other ADP Instrumentation Monitoring Points Not Directly Aligned With Palisades Instrumentation Monitoring Approach Points Type of Sensor ADP Comments (Number of Sensor Locations)/ (Sensor Identification) . T_P3A_02 T_P3A_03 T_P3B_Ol 1* T_P3B_02 T_P3B_03 T_P4A_Ol T_P4B_Ol T_P4B_02 T_P4B_03 temperature 2 External monitoring points on biological shield wall concrete, above primary loop piping in the pipe cavity. T_C3_01. T_C3_02 strain 4 External monitoring points on primary coolant pipe near steam generator inlet elbow, top and bottom. S_P1A_03 S_P1A_04 S_P2A_06 S_P2A_07 temperature 4 External monitoring points on top of vessel flange. 7/8/96 16

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996 Table 2i Other ADP Instrumentation Monitoring Points Not Directly Aligned With Palisades Instrumentation Monitoring Approach Points Type of Sensor ADP Comments (Number of Sensor Locations)/ * (Sensor Identification) _ T_V6_12 T_ V6_13 T_ V6_14 T_ V6_15 strain 10 External monitoring points near weld of safe-end to pipe on top

• and bottom.

S_PlA_Ol S_P1A_02 S_PlB_Ol S_P1B_02 S_P2A_Ol S_P2A_02 S_P3A_Ol S_P3A_02 S_P4B_Ol S_4BA_02 temperature 6 External monitoring points on

• the vessel support structure.

T_S2A_Ol T_S2A_02 T_S2A_03 T_SlB_Ol 7/8/96 17

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points Rev. 1 July 1996 Table 2: Other ADP Instrumentation Monitoring Points Not Directly Aligned With Palisades Instrumentation Monitoring Approach Points Type of Sensor A.DP Comments (Number of Sensor Locations)/ (Sensor Identification) T_S1B_02 T_S1B_03 temperature 4 Internal monitoring points in the vessel lower head region. T_V3_01 T_V3_02 T_V3_03 T_V3_04 temperature 3 Internal monitoring points in the vesser upper region between vessel nozzles. T_V4_14 T_ V4_15 T_V4_16 7/8/96 18

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points REACTOR VESSEL DTYP 3PLACES TYP 4 places ,!>Mq.oput -~ T TYP &A.I 4 places a: C TYP l97*ap0n. 8 3 PLACES t-u a: BTYP ,~ .....L tllDZi* 3 PLACES* TYPICAL SECTION NOTE: RV BOTTOM HEAD WILL BE INSULATED DURING ANNEALING Rev. l July 1996 TYP

Splacet, 3-dlrectlom rw~~~~~i INSllATED ACCESS PANas OP DURING ANNEALING ADP Palisades

~ TEMPERATURE ~* DISPLACEMENT NOTE: ADP Semor LocaUons Approximated RelaUve to Palisades Geometry Figure 1: 7/8/96 Equivalent External Monitoring Points of ADP and Palisades Monitoring Approaches (Table 1) 19

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points I * )I( )I( I I ~ ~-~+~..f ~* I *i. ~+I* I I )I( I s ~-~+-14mi-I I*! I I I I )I( )I( )I( I I ~~-~t~**I+

• I I.

~~-i+tf©t-1. ~ .1 )I( )I( I I ~~-~+H-<+>>* I*! I I I t_llltaS\\ u.

• . I I.,

I I I I I I I I )I( * )I( I I I I

• ==::.

I

• I I

I I I _J I ~ I CSlu...1-I '51 I I "UJ(.!)--t DJ>- ~;::'.~ I u J I* I i )I( ' I I I I I I I I 1::::::1 I I i I I I I ~;:::::l I I I )I( * * .1.~ § H '* r t ~~3NOZ~ t 1 3NOZ u...!'.J w!'.J

f:

z wir wir w w Palisades Zones ZU> ZU> z z ~~ o* 0 0 N~ N N Palisades ADP Rev. 1 July 1996 LU ~iS o-Lii-O< ~8 ~_, I- )8( NOTE: ADP Sensor Locations Approximated Relative to Palisades Geometry Figure 2: 7/8/96 Equivalent Internal Monitoring Points of ADP and Palisades Monitoring Approaches (Table 1) 20

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points 1YP REACTOR VESSEL 2 locatlom batlma,SIG CTYP 3 PLACES a: 0... (.) _l_' TYPICAL SECTION NOTE: RV BOTTOM HEAD WILL BE INSULATED DURING ANNEALING Rev. 1

• July 1996 a locatlom w;~~~~~~lYP

!locaUom on pipe top A bottam H ADP Palisades l8I TEMPERATURE

• . :J.t DISPLACEMENT STRAIN Figure 3:

ADP Monitoring Points Not Directly Aligned With The Palisades Monitoring Approach Points (Table 2) 7/8/96 21

Palisades Thermal Annealing Project ADP/Palisades Instrumentation Monitoring Points <SI lO.,, I-

• wlSl z! ~-

I-

• W~

f N N...

I- * )Ii( ADP Zones I )I( )I( 1

• r r@* Ii===,I I

I I I I I I I I )I( )I( )I( I I .J I ~ I ISli..,.1- )I( * )I( I ISi I -Nw<!l>-----1-1---...

• "'11 I

I')>~ ~ :E-l-

)

0 N.. I-

• wlSl

.JN-:- z - I ,....~~ I U I I I I' I I

1.

I )I( I I I I I ' I I I I I I I I I I

== ::l )I( )I( I..:

• 7..:

I::::::: I-

• wlSl-

_. lO z I- )I( )I( I I I I I I I I I I * * ~ s -- l..J--,--. l:gl it I i( '8' I I lf1!!!5'1!.,.... 1-5 1 3NOZ --1 Pailsades Zones ' _i....~ ~~ ~~ f.. t UJ~

E:

UJ~ UJ 5~ 5 N~ N ~~3NOZ~ NOTE: ADP Sensor Locations Approl<lmated Relative to Palisades Geometry t z UJ z 0 N Palisades ADP Rev. 1 July 1996 U.J ~~ o~ l.Jl-i8 ~--' I- )Ii( Figure 4: ADP Internal Monitoring Points Not Directly Aligned With The Palisades Monitoring Approach Points (fable 2) 7/8/96 22

CONSUMERS POWER COMPANY PALISADES PLANT . DOCKET 50-255 ATTACHMENT B RAI #13, MATERIALS REFERENCED DOCUMENTS

CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 ATTACHMENT B1 . BECHTEL LETTER, D. J. OLVER TO w.c~ COOPER ON BIOLOGICAL SHIELD LINER PLATE DATED 8/4/67 2 Pages

t,*+*9: ~ -:. . /.;:

• <, *:.t.~.:~ __ ;_._\\~-:.'*>*:.;-} <*<.-.:*.

. *._ *.~:-* . -**.. :-~'-~-~. *:. ". .. :.::... ~:.:- ::*~!:);"(":/1,:**.. . **.*.* llE (fr1 EL Co ~~R<1R /\\llt't'J :llf .,N if,:.,£ ms: cd~i;~=]p~~ ~.. ." <:., 0 nvo Tv~* ~ NTY u.u.?1-1 sTR ~Fr:~/*.**- <. :~:i">'riJ'..Ef4Ar:~c 1s_c:;;_9;eAL,,IF: ~411\\;;,:::'.:;i).:/~~ i~:is::Y~ <z~&J,4s@~~*,;:-'.f?9~r,t]J!*/_ \\,~ ';.;.*-\\.~*.:.-<...

• ~**; *<'* ?*

,_ c.. **** -.*..P *. ~*r. Com*.. n '.li.:'rfi;;;~'*:f<.;:lhf;~7tA~:~s~;;.

• "T

'Si~:~~~*?,t~&¥~!~:~~~ 0 ~~:.i..

~....

• • • • • • *..-.)(£k~;nh~rVi;'.: -***tlir.:~ w~:c*;*-'CO:op~r\\

  • :. *-*~_:J'.IE~> '..;._.**.** *

~ <.)~*. "~:: ';**;., *. *:.. ~ ~ ;*sui,)je(;t:: _CqDsµry1e::*r'~:*;P:c)~~r Cd~~~~i.

.~y

'i.'

>\\*'" *

• palisacles PT~nt::
• *' ~
  • \\. -~~,_:l:,~/;;_,:;;:,, *.
  • ~I~!~*~

1 ~il:~"~i~~*~"9g~5,~~i;;f.i41........ *.*... ******* . *-~

• .. Gei1tler:*::::n::.

~:;.eq\\lis.i.tion 59 35-C-{7..., 'Li.nef* Pia ~s ari'd**'.*_~c: _1 1 1 ~c_ 1 .ec-****e**l-'.:-.d~s_* .* 91Y ~teel fcir.. *a&olocical _ .. * *:.=:~ **~:~--.* *. :_:. :*_. -.;.. ~..., braw"ing' S:~JS'-C-153: - Biolpgi..B<:li: :~1ii21Ci** £I.rie.r* 'J:1is liner-. serves.aE?* a.. cc:m*cret~:.for::, and a'*,;;,-upport: fo£the*** !l~~! ~;;i~~j!!i!!~i~~i!~iltl~ilt;!!li:~li~l!f jii~;;~;;~i~ir::~ othe::::- b:i;O.s, i ~~;~ci!t!~~~~.~;"~;~;,~~*;~~f '.;~~~~~~.~r~;f~i :{.,3'~~~{~~*~~~9 * *. ~- -:~-- [1870]

... *<: ~:~.** -*:.-*:.* '.1967.

f?

!.*::-j*"

.,/-~*;'

1tl)t.... ~~~:::: '::.*:*. -7. S<t.:/...

• .Enc le sL;::;::s d:: : : _;_.

i*~r:. v.c. Hai'l w/~nc *(J:) David.J~<*'C~ve:: Project. Er:;,; 0i:-:2'>::r Bechte*1** CC.;;9:2--,/ ~.. ._. :-~

-...... ~,.~

'.;:~~:-*--~:*,... ~-...........

• • • -~-~-*::":";: (:: <"-(.: !"**

CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 .ATTACHMENT 82 H. S. TSANG CALCULATION ON , THE RADIAL DEFLECTION OF BIOLO.GICAL SHIELD DUE TO 1000K LOAD DATED 11/7/67

• 1 Page

,'_:_\\,:** ~*-*-

CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 ATTACHMENT BJ BECHTEL MEMO FROM D. J. OLVER TO S.. POHTOS ON BIOLOGICAL SHIELD LINER PLATE DATED 8/1167. 5 Pages

To* .*~op.ies to '/;...::<.

.-* *~*?.*.'..*.

'.**...

• o~e:
• Au~~t: l; 196.1*~

.~.. >>~ .~~~.. *::~~~~r In~µs,~r+/-al.J*~~~ .. ** : *. ~,*' : ,.. -~ :*,. _. *At : ' * * *ci4?;:~~~>'.; 116n.tgomery *st~~.~#g;

• • • • • • • • • • • • • • • ~~~,****11~~!!~~1~::~**

1 . *--~ .*..... ;:-... ~:*. ":* plate w-ill be. c;olU1ected to ~h:¢ i*J:lf=-l~n~d' J,e,g; byy: ~ans bf a ~hinrter.plate!, .(:s~y J'lo. 9.gage.slj~¢:t.-riietaf}; whic:h is,-welded to the liner ?late arid holt:E!.<r: tO. *a::.v~rti~:a+:*:µ\\e-JDb:er. at:tach,E!.d: to: th.e inclined leg. Thii; de~~'i*l wi,1,1 b.e >ii:dqed* la:ter to Drawings C~l5.3 and c... 157 (ReactQr* Ye:sseL;s_i:ippoits:)~ .HS'r:fub End:os.~J:es ~:. D.J *

• Qclyei:-~

---

~*_:_::.::-*

e:~~i!:?ici 1 i~!:Sl wiJt ~ncq_:r:.pdf:aiei ._I:fl~.F ;~cfiha\\1:e.; ., ~-* .. *---... *.:*_x ~~-*-- .{. ... *.:{ '.*.x U: t ": ~

• y*:*.. :*;*

_.f.*~J**

12

• .**,~o'D

• . *:B:*'.**,'.*.s

.. *Eu**N***.**,*~. J.;**;:_*.::.*.2.* *.* 1*.*,**_-_* ~-

• . :;'-: :~

u ~--. /; ' *;c,.:.* f :- --~ ;... ~ ' **..

. -: -,:~-:~ \\~?:.:~:~:_.~-~-::.~~~~~f ~~If!::~~t~r}~~7;(~-~~:}~~~f1~*;;~ 7 ~7fff ::*;¥:~~~;:-rf:*[~?:~~7;/~;-~~:*f~::~:::~~~~~~ft'.7;~:

.. ~ ~ *......

/' --._,.,,..... _-.. _.,,_......,.. ~........ '1'.-:-.-~...... ~':'."_...... w:';"'*...,_.,,_ j.... 1**1*. "*I '~* l~ . l-1EN8£L IQ *Pt.At.Go.. A:E.Tt:

• e.*~A/e,.. -S-e'(.c4J: 614.~l...:
• ->$* ~ec"7.

)C_t"~"(;ye*:: f;t,*'&>G* ~;,,~E_'i.. -?1.l;.i:i:i~ ~-:~z.; * > *-J.;.7;. :; 1 .. > * ~/ ~,_.:.* ... ~........ :* 1*.: I ~*.* 'C.OA/.~i?- 'T--e IJ Gt.Dl(J ' 'Z*2' !. 2-t

• i5 ' ~-S,';"

. i '. \\ -._

• * ;.\\

\\_:' :* : -~ [ N ** Clz-g;4J~ 1'.-;. -.* -=-:.::;: ;:, - --::=-:::= T =:_~-:~-:: ~ ~-= -/':;:>:.; < :.

~ J.

1 ".

• rot?..

OP£ N 1 /'/ c r,,. (v L1NeR. tf .1 Y"-"".

F

~-'--~-

-*~** _ _;.._....._.:

,{,\\ "\\\\

.{'...... \\.l""* . \\ .**.*.=.

)

<*/'.* '/ . II I 'fl. f_i?.f._<1. t; '~~lk'~t !,_ 9,, ' I. 1:~ SEfi Dwr. C-15f(TyPJ / i. I .. *--.~*-.----~~--=:==..:= :==t= *. '.

• ... /

?I l*- 1 o*****av. f i5&'t. T ~INER n. *' T~ "'. Y' - * . $;UPP&1l?...T sr~~~t :...* ".. -' /

• '.*--.__::,;;~MA.J~

elJ~e. '.,i-1JM~f.:), '.J ... * '**"\\)*::"'"' *'(.~ll":'l". FL_* wvc;.

• ~"'-'* f":.

. -3"*.,; _. """ * "' '='_,, fl) '. *:t1)1" _,..:;.;.:,,_" ~- t's_ -';:;,/'/ R"i:y ji A_ ):: 7'~'..P '.. : :/1 :. .-. :: i.. ....,. Jil-f'C,,.. .._J -.'j~~- l'f* u... I I - /(.,. Co"-C* 1?e-iC;41 tro6'r9**,.s *

• 1

. ~ *,:* _....,)f/Ff . *.,~,-,) ;"i,** {~~:.::;~i:f ~x~-;:~,2;~,~ *. ~~;i;f1~:~~;:it~~-~ -.,:J*-"* ~..

• • ~.... *.*.

Pt. -_-. .-*r:*

• • • 'i

~ *'

l:::* -~.-

'. ***-; -*,-;-*-~~. J.,4Gf:t.,*.:,v1,*7A;::.. +/-:: _*-~*>: ~ ., l

j.

• -~~-.:..!.......,;..__.: *--**-*--**** :*-"

_\\T.~l?-L ~~-.. !. -~-~*.....

.~-**

,.. * *

• X-; ~?~ f~%&:~T.f.f-'~'Y*=f *'_

. ~ ~-~. *'.. *-**... . -~*-. **..

• .:.t

.,..,l:!:"f.AJ.i, ~?.

~.@;~~;-;,.. ~/{~'..;:;:

r j ~

---

~---- *--~:-~~~-: ______ _;_ ~----~--------- --*-*

....... -*. *- *-..,.... ~ -,

• '.:;*. -* ~.*.:. :-,..,'.:::*.*::.:.*. -.

[1869]

CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 ATTACHMENT 84 PALISADES PLANT DRAWING C-153, REACTOR CONTAINMENT. BIOLOGICAL SHIELD LINER, REV: 6 DATED 8/8/89. 2 Pages

G . ~* D T c f -- A ,/....

• ,,*. "*t u,...;~,
• *. \\/lf:"W:. 4CUIJ ~.:.

~;:;*.. _____...:. "'-;....,------...... :::::.~ 1-----~.. J.:._ c: ;;. ~ T : p :-.' A 1~

• ar*...: 1 ".l."I"**:

.U."**'*;*4 . 0=1..o.1L.@ ,..,..o*

• j

. 1. ti ~

=TAI!..~

'1* '" O' £. 'i 1* 'I I ' I I;* I j J .-.. l..:... :=*

i.

~

~

.,--:\\ "V"'., .-_*~*~*~*_!.:._.+=.~-=*-=-~-~-~-=-=~~-,,_,....c='-~,-~~*~.:...~',,,.*_.. _,~:.. _-=*.::.~:*_=**~:'-"~'~* *=~~~-~-~--~-'-=-=-=-=-~~~~~=....,.-~.,..,,,.,.

• t" 1..::v-*.c

--~r *. *.*.: 4* ...;, \\,,'..... ;- ;.:~,*../ ' I .. ~:Ii.' !1Qlli.

• .L**I* "lb.1D co...-OIW 1DllST* onc..... 1~

.1.--.tiu.l1ftfll10co*ro*1110*""ICK_,..,.

• • M7:"

-'":::x~ro:. -:o: :*,~=-'--":_:;~:,. "*°"°.... 1111004 *OU"'lD.

• • r~**.,

,.,... T I ~- ~ I C**I' C:.*U4 _, *.,,L"':. G A

I I I I. I l I ~I¢ f:,Ol.7S J( 11- 0 I 1...0, W.//.../U7S1 W4S;.le~ E.J 2." /h'O, .I ~* ~PLYWOOD ~M I~ I / 0. - __.,, /I -= I~~..... L

CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 ATTACHMENT 85 PALISADES PLANT DRAWING M-72, PRIMARY SHIELD COOLING. COILS SECTIONS & DETAILS, REV. 1

• 2 Pages

iH G D c A ~ '!J ZONE NO. 1-j-* -.--.C-- 1; ~ ~ ~ ). ~ SIDI ZONE NO. 3-t--.:- ~ 9 ~ i . *-.='-~-,-==-~r* ~~:;,*:~j..,~.j=~~:~~~ NQ*-1---:--~- 1 'lL B i l -~ I I

• l\\l.!..!,,j 1!

~:~~;.:~: __ J--~-'t7:-~ ;--....... - I / / 1-:

• /

7 I i I .'.'I i / JI.. lll!o:,_*,......u... ... ~-~- .....,.-..,*,o4*AC.TGllt--- --... -. t *

!;!;!;~H;

~;::t!;;W,;;+H;::!;:::lf=:::::::l 4U..Ul'*D (lLUIT*'>

tu*'°':'" cn.10&-cr t~.:=-~::. t""*'°'*** Cb..Hl'*r* I "l...,11~* I c'i..*'°'*** ~,.. I, ft.. *oo*... (t'"I ii L~!1:'.0'\\\\l:::!I....., HHl-tt--tt-tt-r .1-4t.1..ffl'*D ,_,,_.._,._-+--'>,~I l~:~:'.:: ~.'t;'k~"" ,,l\\.Hl~r I ( ,,1,._,......

=t'.::t::t::f:ftin:l'.::t::t::f:fj~ftitn:~~~rtl~~OJ::f:X:+/-:t:ta~

I' ~=;==-~~=* -~'"" 11 I I I 1!1:. I' I j Ii 1-' 111.. """*o* lb..'t*1*-a.*

• • *,,*.c,*TC) 4 tl[..t(.T~-\\r-OE VELOPEQ ELE VllT!ON Of COIL sc.e.... t.,,.,r.o*

OE V£LOPED ELEVATION Of COIL ZONES Ntl I 3 ANO 4 ac.-.1..* "11**'-cr ttL.'HT'*c.* f::U.Ut:*o:.. rull:.~~ f-(llttw)llllL' [""'.,

---

u---**

~~--- "'I I . 1=,_ K .'-t tL.It %!-N,L@ SE,~JIQ.~. ~-E sE~it-~~ *.r.-r See ne.'1-..\\- pd5 c. ~r e"\\~<5e""e."t PAllUDH PUNT Cl MSUMllS row11 COMPANY PIPING DRAWING Plitt.>, f?Y' 'StetElO COOLING COILS

ECTI0

6 I. OE T~lS M--TZ

1.

,t si_.;i4'-o I 3ip:-u~~i~~~:~;-.:;~. . ~ :OL.;,i;-;;.~* t r~;~_7:;~:;'.'.!r;~,.;c;i. c i I. l C::L.GIO'-G'" r j. LttJER~ ~ -!. EL.'391'-0 ~E.L~0J-;, r I I I / DETAILS OF HANGER D\\14 C -IS3 .. -50 BEND!TYP.) / ---~ ~;&J!:f'l:!'..!QIF!HD8, t,,~PPLV(.Qt-OiiOR.

• r-s*

TYP SECT. RISERS SCALE. /1.,":t'-0'

CONSUMERS POWER COMPANY PALISADES PLANT DOCKET. 50-255 ATTACHMENT C RAI #3 THERMAL ANNEALING OPERATION (SECTION 1.5) 11 Pages

General Rules For Isolation of Water From the Reactor Vessel During Annealing In general, all directly connected sources of water should be isolated by twc;> (2) valves with an open drain between them to take care of any through seat leakage. In cases where this is not possible the upstream system should be isolated from water sources and drained. A closed valve or open drain should exist in the system between the water source and the*reactor vessel. Systems required for emergency response during annealing or to provide necessary support to the annealing function can be handled by justified exemption from this rule or by operating instructions to.be included in the annealing procedure. For emergency core cool.ing and radiation protection reasons the plant was designed to resist being drained, therefore, exceptions exist to these simple rules. ALARA considerations must be weighed against.the actual risk of flooding ir:wolved in each instance. With that in mind situations that might justify less than two valve isolation were examined. Four categories were considered as groupings of common.conditions:

1. The water source is normally and continuously pressurized as by head tank or pneumatic accumulator. *For this condition a single isolation valve is acceptable if it can be leak checked and there is a path for leakage to escape befqre entering the vessel.

2. The water source can be pressurized by automatic pump start. In this condition a single isolation valve is acceptable if it is leak checked and the pumping source can be tagged and locked out.

3. The water source is only pressurized or activated by manu.al action. For this condition a single valve is acceptable if an alarm will sound before significant flow toward the ves~el can occur or if water movement is only by local*

manual action such as by hose or by drain funnel.

4. The line does not normally contain water but can become wet due to flooding.

runoff or by water backup within a pipe. In this condition the degree of_ isolation is based on degree of risk as determined by operating experience and judgement of the ability of the operating staff to become aware of water movement before it reaches the reactor vessel.. Water sources that enter the refueling cavity are generally considered the same as directly connected to the reactor vessel because the water can run out the refueling cavity seal opening and quench the vessel outside diameter (OD) (and inner diameter (ID) if flow is voluminous or forceful enough). Water sources that enter the tilt pits can 1

utilize the volume of the tilt pit to gain reaction time if there is a timely method of alerting the Operators such as alarms, remote monitored indicators or periodic rounds. I Water sources that can flood the containment lower levels are considered if they have the potential of overcoming the normal two (2) inch gravity drain from the containment sump. Small lines that are capable of holding only a small amount of water that can potentially drain to the reactor vessel will remain unisolated. These are mostly% inch instrument lines entering very large Primary Coolant System (PCS) loops far enough from the vessel that leakage will evaporate in the loops as the water flows slowly toward the vessel along the gradually warming piping. Some lines do riot exactly meet any of the rules. They will be considered on a case by case basis much like condition #4 above with plant operating experience being heavily weighted in making isolation decisions. Additionally during the actual annealing period, the Outage Scheduling Group will schedule other refueling related activities in a manner to minimize valving evolut!ons and open maintenance orders that would affect water isolation from. the reactor vess~I. The current draft of the valving arrangement, including valve numbers and' positions, is* available in the following "Reactor Annealing Water Isolation Draining/T_agging Plan". 2

REACTOR ANNEALING Water Isolation Draining/Tagging Plan

1. DIRECT ROUTES INTO PRIMARY COOLANT SYSTEM::

A SAFETY INJECTION 12" (SI) TANK LINES - Only isolation.available are the MOV's M0-3052 (T-820) M0-3049 (T-82C) M0-3045 (T-828) M0-3041 (T-82A) DRAFT 8/21/96 Page 1 - Empty and vent SI tanks. Leave MOV's open. This will be done as a part of R0-105. Leave SI tanks empty and vented until after annealing.

8. PRESSURIZER SURGE LINE

- Can not be isolated - See "Pressurizer" C. SHUTDOWN COOLING DROP LINE - Close MOV-3015 and MOV-3016 & open breakers - Look into using RV-0401 as possible between the valves drain - Do not want to qrain downstream of M0-3016 so that alternate Spent FL1el Pool (SFP) cooling via Shutdown Cooling (SOC) remains available: D. CHARGING LINE - Close charging pump discharge manual valves MV-CVC2106 (P-55C) MV-CVC4100 (P-558) MV-CVC2094 (P-55A) MV-CVC108 (P-558 & C common di$chargej MV-CVC2089 (P-55A suction) Gives double isolation for P-55A - Open breakers for charging pumps - Vent & 'drain shell side of Regenative Heat Exchanger and all of charging line between isolation valves & PCS (Note: Charging stop valves are air to close, therefore, not used to isolate) E. eves LETDOWN LINE - Close manual isolation valves downstream of orifices (MV-CVC2227 & MV-CVC2007) - Vent & drain letdown line between PCS & isolation valves via valves MV-CVC2288 through 2291 to Primary System Drain Tank T-74.

REACTOR ANNEALING Water Isolation Draining/Tagging Plan F. PRIMARY COOLANT SYSTEM (PCS) LOOP DRAIN LINES - Open all loop drain valves to drain any in-leakage: - Place spectacle flanges to open position. DRAFT 8/21/96 Page2 - or do the above, but leave one valve closed (to prevent back leakage) which

• can be manually opened if required.

NOTE: T-74, Primary System Drain Tank (PSDT) automatically pumps down & is alarmed at C-40 panel on high level. Sufficient time should be available to manually drain T-74 to containment sump if required. . G. PCS HOT LEG INJECTION - Close manual isolation valve MV-PC1094A - Close and rack out breakers for M0-3083 and M0-3082

• - Close and caution tag drains to PSDT (T-74) CV-3084 & CV-3085.

- Open manual valves MV-ES3091, MV-ES3091A, and remove the pipe cap on the end of the line to provide an intermediate drain path. H. PCS INSTRUMENT PRESSURE & LEVEL TAPS - No isolation required. Low Volume. Should largely drain with PCS. - Can double isolate where possible & where easily accessible & open equalizing lines where applicable on sensing equipment if a large volume of water is available to be held up by vacuum. I. REACTOR VESSEL FLANGE LEAKOFF AND 0-RING LEAKOFF LINES - Close inner leakoff isolation valves MV-PC1095A and MV-PC10958 (See Item 4G). - No isolation of other leakoff line since there is.no water source available to it. J. PCS SAMPLE POINTS - Close local sample" isolation valves MV-PC1023A & MV-PC10238 (SX-1023) MV-PC1012 & MV-PC1012A (SX-1012) MV-PC1034A & MV-P C1034B (SX-1034). - Close and caution tag hand switches to CV-.1910 & CV-1911, CV-1902. K. PRIMARY COOLANT PUMP (PCP) SEAL INJECTION AND LEAKOFF (and Seal Cooling via Component Cooling Water (CCW) Lines) - Isolate local CCW supply and return isolations. MV-CC113 MV-CC115 MV-CC109 MV-CC111 MV-CC114 MV-CC116 MV-CC110 MV-CC112.

REACTOR ANNEALING Water Isolation Draining/Tagging Plan - Double isolate seal flush lines to PCP's coming.from shutdown cooling MV-PC1147, 1148, 1149, 1150 & MV-PC1151, 1152, 1153, 1154. DRAFT 8/21/96 Page 3 - Seal leakage to T-74 (PSDT) protected only by a check valve. No ability to further isolate (unless PSDT is pumped qown and the drain to containment sump left open - Not Recommended). L. STEAM GENERATOR TUBE LEAKS - Per Sop 7, Steam generators are not to be.lowered to <30%. -Any compensatory actions (eg, plugging leaking tubes) should be. done prior to annealing.

2. ROUTES ENTERING VIA THE PRESSURIZER A. NORMAL SPRAY

- Part of PCS. Should drain as PCS drains.via spray valves bypasses (MV-PC 1056 ~ 1058).. B. AUXILIARY SPRAYS - Close and caution tag CV-2117 hand switch. CVCS Header up stream will be drained per Item 1 D (Charging). C. PORV'S - Isolate Primary Makeup Water (PMW) supply to T-73 Quench Tank: MV-PC1134 & CV-0155 (air to open) - Isolate sample lines from Quench Tank. -(Note MV-PC1046'& 1047 for SX-0116 to be left open to allow monitoring of Quench Tank level.) -MV-PC1049 for SX-1049. -MV-PC1053 for SX-1053 -Close & Caution Tag CV-1910 & 1911. - Drain Quench Tank via CV-0148. Maintain closed, but monitor Quench Tank level & drain again if needed. - Isolate sample line SX-1045 via MV-PC1045B & MV-PC1045C - Close and caution tag CV-1910 & 1911. D. CODE RELIEF VALVES - Same as PORV's.

REACTOR ANNEALING Water Isolation Draining/Tagging Plan E. ATMOSPHERIC VENT LINES (HIGH POINT VENTS) - No Ingress Path (Source) of water. DRAFT 8/21/96 Page4 - During PCS drain, open Primary Relief Valves (PRV's) temporarily to ensure line drains. F. HEAD VENT LINE - No ingress path (source) of water. - During PCS drain, open PRV's temporarily to ensure lines are drained. G. PRESSURE AND LEVEL INSTRUMENT LINES - Should largely get drained with PCS, check with l&C if necessary to isolate due to environmental (heat) conditions during anneal. - Can locally isolate if needed. H. SAMPLE POINTS ~ See item "C" above.

3. ROUTES ENTERING VIA SAFETY INJECTION:

A PMW VIA eves - N/A if charging downstream of P-55A/B/C is isolated & drained. B. CONCENTRATED BORIC ACID VIA eves - N/A if charging is isolated & drained. . c~ SAFETY INJECTION & REFUELING WATER (SIRW) VIA SI. eves. VENT & RECIRCULATION LINES - CVCS is N/A if charging lines are isolated & drained.. - High Pressure Safety Injection (HPSI) - Close & deenergize MOV's (Caution-Tag) M0-3007 M0-3011 M0-3009 M0-3013. ~Close manual isolation valves upstream of MOV's MV,.ES3007, 3009, 3011, 3013. - Uncap and open in-between test connection to drain: MV-ES3007 A, 3009A, 3011 A, 3013A. - Redundant HPSI - Close & deenergize MOV's (C8.bltion Tag) M0-3068 M0-3064 M0-3066 M0-3062.

REACTOR ANNEALING Water Isolation Draining/Tagging Plan -Also; close HPSI pump discharge manual valves MV-ES3178 (P-60B) MV-ES3186 (P-66A). - Open & caution tag P-66A & B breakers and hand switches. - Low Pressure Safety Injection (LPSI) -Close & deenergize MOV's (caution tag) M0-3008 M0-3010 M0-3012 M0-3014. -Uncap & open test connection valves downstream to drain MV-ES3008 -ES3010

• -ES3012

-ES3014 MV-ES3008A -ES301 OA -ES2012A -ES3014A. - Close and caution tag LPSI manual disch valves

• MV-ES3202 (P-67 A)

. *MV-ES3193 (P-67B). -*Open & caution tag P-67 A & B breakers & hand switches.. DRAFT 8/21/96 Page 5

• NOTE: If use of alternate SFP cooling is required or needs to remain available, CV-3025 and CV-3006 can be hand-jacked closed instead. This may be preferable.

- Recirculation - Isolation for HPSI & LPSI will also cover isolation for individual pump recirculation. lines. - Check closed and caution tag MV-ESS3234 and MV-ESS3234A to isolate the header recirculation path from the SI tanks. D. TUBE LEAKS IN SOC HEAT EXCHANGERS (E-60A&B) - Improbable source of in-leakage. - HPSI, LPSI, and Containment Spray will have at least one downstream isolation. - additional measures not required. E. TUBE LEAKS IN NON-REGEN HEAT EXCHANGER (ie, Letdown Heat Exchanger E-58) - Not an issue if letdown isolated upstream via item 1 E. F. BACKUP FROM GAS COLLECTION HEADER - Any backup from the Vent Gas Collection Header (or Waste Gas Surge Tank) would be isolated or monitored by actions performed to address in-leakage from the tanks which have access to the PCS. Some additional measures could be taken:

-Quench Tank REACTOR ANNEALING Water Isolation Draining/Tagging Plan DRAFT 8/21/96 Page6 -Close & caution tag CV-0152 vent to Containment Ventilation Header. - (may also want to open atmospheric vent MV-PC1054 and remove its blind flange). - SI Tanks - Need to keep vented. - Highly improbable that sufficient head could be developed to back water.up to the top of the SI bottles. Indications would be. detected elsewhere allowing time to close vent CV's. to SI tank prior to in-leakage. - Primary System Orain Tank (T-74) - Normally isolated/closed by PCV-1002A. - See item 1 F note for compensatory measures if required. -Clean Waste Receiver Tanks (CWRT's) (T-64A/B/C/D) - See item 5A G. VOLUME CONTROL TANK (VCT) T-54 -*Normal discharge path addressed under 1 D, Charging. - Controlled bleed off - Close 2 isolation valves & open intermediate drain ( & remove blind flange) as follows:* P-50A: close MV-PC1130 & MV-PC1130A open MV-PC11.35 P-508: Close MV:.PC1131 & MV-PC1131A open MV-PC1136 P-50C: Close MV-PC1132 ~ MV-PC1132Aopen MV-PC1137 P~SQD: Close MV-PC1133 & MV-PC1133A open MV-PC1138.

4. ROUTES ENTERING VIA CAVITY FLOOR:

A. CONTAINMENT SPRAY - Close & caution tag CV-3001 & CV-3002. NOTE: These are air-to-close valves but have nitrogen backup. - Close & caution tag MV-ES3258 & MV-ES3259. - Open test valves MV-ES3344 & MV-ES3346 for between the valves drain/vent. . - Drain headers inside containment via MV-ES3246A and MV-ES3244A. - Open and caution tag pump breakers.

REACTOR ANNEALING Water Isolation Draining/Tagging Plan B. FIRE FIGHTING WATER DRAFT 8/21/96 Page 7 - Fire water in containment is provided by Critical Service Water. This cannot and should not be made unavailable. The system consists of 2 fire hose stations with local valve isolation. Normally closed. Can caution tag "For Emergency Use Only" but this_should be the only time they would be used anyway. C. TRANSFER TUBE <FROM SFP) - SFP gate installed and leak-tight - Fuel transfer tube flange is installed - or Flange remains off & AO monitoring for leakage and pump down tilt pit (back to SFP) if leakage occurs. - or Close and caution tag Transfer Tube Isolation Valve MV-SFP141 (if deemed preferable to the above two options with respect to ALARA and the valve is functional and leak tight at the time of the anneal). D. SPENT FUEL POOL (SFP) COOLING SYSTEM - Close & caution tag fill, drain & recirculation valves MV-SFP120 MV-SFP117 MV-SFP121 MV-SFP118. E. TILT PIT DRAIN - See SFP Cooling above. - Other drain line goes to Containment Sump, should this backup, the reactor vessel annulus cavity would already be flooded. F. TILT PIT FILL LINE - See Items D & E above. G. DISCONNECTED HEAD VENT LINES IN THE CAVITY WALL - Isolation requirements already addressed under-"PORV's" item 2C and "Head Vent Lines" Item 2F. - Per Dwg M-201 Sh 1, Rev 64, the drain to the tilt pit is not functional (See M-201 Sh 1 Rev 64 Note 3) due to a temporary modification. H. VENTILATION SYSTEM CONDENSATION - Volume per unit time would be sufficiently small, if at all, that any condensation could be drained via normal means. I. FUEL TRANSFER EQUIPMENT HYDRAULICS - Finite volume is within limits of tilt pit capacity. Caution tag pump.

REACTOR ANNEALING Water Isolation Draining/Tagging Plan J. DECONTAMINATION HOSES (WALL WASHDOWN WATER) DRAFT 8/21/96 Page 8 - PMW water to containment isolated under "PORV's" Item 2C. (CV-0155 closed & caution tagged). - additionally, ensure closed & caution tag local PMW header valves MV-DMW751, 752, 753, 754.

5. REACTOR VESSEL OUTSIDE DIAMETER QUENCHING BY CONTAINMENT FLOODING A. CLEAN WASTE RECEIVER TANK (CWRT) OVERFLOW

- With proper water management, this should not be a risk factor. PCS cold volume is 84,000 gal. with system drainage (charging/letdown, etc) total additional volume from normal operations would be approximateiy 100,000 gal. Total CWRT volume (4 tanks) is >230,000 gal. With a concentrated pre.:.outage effort to process clean waste from the T-64's, CWRT overflow should not be a concern. B. CONTAINMENT SPRAY . - See "Containment Sp_ray" item 4A. .C. EMERGENCY CORE COOLING SYSTEM (ECCS) RECIRCULATION LINE BACKUP - Concern is the possible emptying of SIRW Tank to Containment Sump if check valves CK-ES3181 & 3166 back-leak & CV-3029/3030 lose air. - CV-3030 & CV-3029 are high pressure air with instrument air backup. Very low probability of loss of all air. - Caution tag hand switches in control room and open temporarily and drain te~t tap between check & CV tc:i ensure check valves are properly seated. D. FIRE FIGHTING WATER - Same as item 48. E. SERVICE WATER SYSTEM (SWS) LINE LEAK - Service water to containment cannot be isolated as a precautionary measure since it will be required to maintain containment temperatures during annealing. - Leakage would be detected by alarm EK-1347 for flows~ 300GPM or by Containment Sump levels (EK-1351 & 1350). - Isolation steps from ARP response or by ONP 6.1 would be performed. If isolation of all SWS to containment for a significant time period were required,

REACTOR ANNEALING - Water Isolation Draining/Tagging Plan DRAFT 8/21/96 Page 9 annealing wo"uld likely need to be aborted to avoid overheating containment, unless an evaluation on the acceptability of elevated containment temperatures on equipment had been performed as a precaution. F. SECONDARY SIDE WATER VIA CONDENSATE RECYCLE WITH VENTS OR DRAINS OPEN - Steam Generators will be in wet layup using the blowdown system. Associate_d vents and drains are closed and capped. - General Note All vc:ilves, breakers, and switches which have a required position for annealing should be caution tagged accordingly prior to vessel heat up above 210°F unless the valve is positioned by the annealing procedure which takes precedent over this general plan.}}